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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 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 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 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 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 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 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-99-027, 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 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 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 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 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 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 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 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 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 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 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 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 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 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-99-027, 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 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 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 ML20195E9581999-06-0707 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 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 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 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 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 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 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 ML20206B4391999-04-21021 April 1999 Forwards Corrected ITS Markup Pages to Replace Pages in 981201 License Amend Requests for SG Replacement 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 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 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 L-99-153, 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 Error1999-04-13013 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:RESEARCH INSTITUTION/LABORATORY TO NRC
MONTHYEARML17261A8541989-01-13013 January 1989 Discusses 890110-11 Meetings Re Nozzle Sizing Study.Most Accurate Sizing Obtained by Collecting Data of Edge Diffracted Waves from Geometric Extremities of Flaw ML19324A1111988-05-19019 May 1988 Forwards Memo Describing Results of Testing of Plant Bolting Matls Under FIN A-3866,Task Assignment 12 ML19324B2701988-03-0202 March 1988 Forwards Rept on Metallurgical Evaluation of Five Bolts Obtained from Farley Plant.Bolts Found Acceptable ML20207R7571987-03-10010 March 1987 Forwards Technical Evaluation Rept Input for South Texas Initial Plant Test Program Through FSAR Amend 56,Feb 1987 Sser & & Beaver Valley Unit 2 Initial Plant Test Program Through FSAR Amend 15 & Nov 1986 Sser ML20207S5001987-03-0606 March 1987 Forwards Technical Evaluation Repts for Domestic Mark III Plants (Grand Gulf,Clinton,River Bend & Perry) & Gessar Ii. Inserts to Be Included in Section 6.2.1.8 of Draft Sser 2 for Clinton,River Bend & Perry Plants Also Encl ML20214K1391986-11-12012 November 1986 Forwards Comments on Emergency Preparedness Exercise Scenario Review.Amount & Quality of Plant Parameter Data,In Addition to Lack of Controller & Contingency Messages,Will Not Support Meaningful Exercise ML20195F8651986-10-31031 October 1986 Forwards Request for Addl Info Re MSIV Operability at Facility,Based on Initial Review of Util Rept Entitled, Final Rept 10CFR50.55(e) MSIV Actuators, Forwarded by Util ML20206H3211986-09-10010 September 1986 Forwards Summary Repts from 860811-14 Visit to Kewaunee Nuclear Station & 851002-03 Visit to Cooper Nuclear Station Re Generic Issue 83 on Control Room Habitability ML20206J6001986-08-0808 August 1986 Advises That NUREG-0956 Support Calculations Using ORNL Trends Code to Evaluate Influence of Containment Chemistry or Retention of Hi Can Also Provide Useful Info for General NUREG-1150 Issue Paper,Per Telcon Discussion ML20206H1131986-06-23023 June 1986 Submits DHEAT2 Parametric Calculations for Plant,Per 860620 Discussions.Calculations Run Assuming Any Steam Spike Would Develop Too Slowly to Contribute to DCH Peak Pressure ML20214E1721986-02-25025 February 1986 Ack Receipt of 860210 & 21 Ltrs Accepting Task Orders 3 & 4 Under FIN A-3552 & Task Order 5,respectively.Orders Include Work at Palisades & LaSalle Stations & Mod Review at Davis-Besse.Modified Task Order 4 Under FIN A-3550 Encl ML20206J2521986-01-24024 January 1986 Forwards Early Draft of Executive Summary from Analysis of Station Blackout Accidents for Bellefonte PWR, Per Request. Initial Draft of Rept Scheduled to Be Completed by 860315 ML20206J3681986-01-15015 January 1986 Discusses Latest Calculations of Molecular Iodine (I2) & Organic Iodide (CH3I) Scrubbing During TC1 Sequence from Draft Plant Rept ML20136F4771985-12-31031 December 1985 Forwards PNL-5718, Review of Tdi Diesel Generator Owners Group Engine Requalification Program,Final Rept, Technical Evaluation Rept ML20206J4031985-12-13013 December 1985 Summarizes Preliminary Scoping Calculations for Molecular Iodine (I2) Scrubbing in BWR Pressure Suppression Pools. Calculations Made for Most Recent TC1 Sequence in Draft Plant Rept ML20133A3141985-09-27027 September 1985 Forwards Review of Section 4.7 of Technical Evaluation Rept PNL-5600, Review of Resolution of Known Problems in Engine Components for Tdi Emergency Diesel Generators, Reflecting Views Re Crankshafts for 16-cylinder Engines ML20128H7371985-06-27027 June 1985 Forwards Rev 1 to Review of Engine Base & Bearing Caps for Tdi DSRV-12,DSRV-16 & DSRV-20 Diesel Engines, Technical Evaluation Rept ML20141N2981985-05-29029 May 1985 Requests Listed Addl Info Re S&W 1982 Rept, Ultimate Pressure Capacity of Shoreham Primary Containment ML20126E8721985-05-24024 May 1985 Forwards PNL-5200-3, Review of Emergency Diesel Generator Engine & Auxiliary Module Wiring & Terminations, Dtd May 1985 ML20133N7721985-03-27027 March 1985 Forwards Info Telecopied on 850325 Re Locational Distribution of Three Fission Product Species for Surry V Sequence & Time Dependent Release for Fission Product Groups for Peach Bottom ML20214F5101985-03-13013 March 1985 Forwards List of Questions Re Plant Pra,Per 850512 Telcon. Statement of Work, Review of PRA for Seabrook Nuclear Power Plant, Encl ML20133N3461985-02-20020 February 1985 Discusses Review of March Results for Surry & Peach Bottom Sequences,In Order to Quantify Expected Noble Gas Releases. March Model Appropriate for Behavior of Noble Gases ML20134A1011985-02-20020 February 1985 Submits Results of Noble Gas Release Sequences Using March Code for Facilities ML20128P3401984-12-0505 December 1984 Submits Rept for Task 1 Per M Silberberg 841116 Memo Re Concerns Re Lanthanum Releases in BMI-2104.Discusses Discrepancies in Corcon Calculations,Presents Revised Tables 6.14 & 7.16 & Recommends Reanalyses ML20138N6971984-09-18018 September 1984 Forwards Repts Re Irradiation,Decontamination & DBA Testing, Per Request of Y Korobov of Carboline Co ML20134E1591984-08-0202 August 1984 Forwards Brief Summary Rept on Leakage Characteristics of Nuclear Containment Hatches During Severe Accident. Draft of Complete Rept Will Be Mailed Later in Aug ML20127B9221984-07-20020 July 1984 Forwards PNL-5201, Review & Evaluation of Tdi Diesel Engine Reliability & Operability - Grand Gulf Nuclear Station, Unit 1 ML20127B8881984-05-21021 May 1984 Comments on May 1984 Draft Tdi Diesel Generator Owners Group Program Plan.Full Insp of One Engine to Owners Group Spec Recommended ML20093G4821984-05-0202 May 1984 Provides Summary of Battelle Subcontract W/Tdi to Evaluate Special Silicon carbide-impregnated Cylinder Liners & Piston Rings.Task Does Not Involve Dependability Tests of Tdi Engines.No Apparent Conflict Found ML20113A0201984-04-0404 April 1984 Forwards Requests for Addl Info Developed During Review of FSAR Chapter 14 Re Initial Plant Test Program.Review Conducted Through Amend 4.Listing of Items Requiring Resolution Encl ML20117C9421984-04-0303 April 1984 Forwards Proposed Radiological Source Term Input for Facility.Encl Amended to Reflect NRC Suggested Changes ML20093C5281984-03-30030 March 1984 Forwards Audit of Susquehanna Unit 2 Tech Specs, Technical Evaluation Rept ML17320A9711984-03-0606 March 1984 Forwards First Round Questions Following Evaluation of Exxon Rept, Steam Tube Rupture Incident at Prairie Island Unit 1, PTSPWR2 Vs Data,Preliminary Benchmark Analysis. ML17320A9741984-03-0505 March 1984 Forwards First Round Questions on Exxon Methodology Rept for PTSPWR2.Rept Lacks Specific Details Re Biases in Initial Conditions & Boundary Conditions ML20128N6571984-02-22022 February 1984 Comments on Draft Vols IV-VI of BMI-2104,presented at Peer Review 840126 & 27 Meetings.Comments Concern Completed Calculations for Sequoyah Ice Condenser Plant,Recalculated Surry Results & Completed Calculations for Zion Plant ML19306A0151984-02-10010 February 1984 Summarizes 840206-07 Visit to Nevada Test Site Hydrogen Burn Facility to Inspect Condition of Equipment & Cable/Splice Samples After Series of Hydrogen Burn Tests Conducted by Epri.No Indication of External Damage to Equipment Noted ML17320A9731983-11-22022 November 1983 Forwards First Round Questions on Mods to Exxon Draft Repts, PTSPWR2 Mods for St Lucie Unit 1 & Description of Exxon Plant Transient Simulation Model for Pwrs. ML17320A9721983-09-30030 September 1983 Forwards First Round Questions on Exxon Plant Transient Code,Based on Review of Proprietary Rept, Description of Exxon Nuclear Plant Transient Simulation Model for Pwrs. ML20080B5631983-08-12012 August 1983 Forwards Draft Preliminary Review of Limerick Generating Station Severe Accident Risk Assessment,Vol I:Core Melt Frequency. Rept Satisfies Milestone for Task 1 of Project 3 Under FIN A-3393 ML20211D5971983-04-11011 April 1983 Forwards Summary of Independent Development of Finite Element Models & Determination of Natural Frequencies for Piping Problems in Containment Spray Discharge Line & Accumulator Loop 4 ML20132B4961983-02-11011 February 1983 Forwards Evaluation of Generic Key Indicators for Project Engineering/Design Activities at Facilities ML20072L7101982-12-20020 December 1982 Forwards BNL 821213 Memeo Re Degraded Core Accidents at Facilities,Per Task III.3 Defined in Design Basis for Hydrogen Control Sys in CP & OL Applications. No-cost Extension to Contract Requested ML20079J5641982-12-16016 December 1982 Summarizes Findings of 821207-08 Plant Tour Re Use of PORVs or Depressurization Scheme for Removing Decay Heat.Addl Info Requested Includes General Arrangement & Isometric Drawings of Piping Near San Onofre Pressurizer ML20027D1751982-10-18018 October 1982 Responds to 820806 Request to Evaluate Rapid Depressurization & DHR Sys for C-E Plants W/O Porvs, Discussing Role of Task Action Plan A-45, Shutdown Heat Removal Requirements, in Supporting Evaluation of Facility ML20126F4791982-10-0101 October 1982 Requests Permission to Observe Upcoming Types A,B & C Tests at Listed Facilities,Per FIN B-0489, Containment Leak Rate Testing ML20072L6851982-09-21021 September 1982 Forwards Analysis of Full Core Meltdown Accidents in Grand Gulf Reactor Plant, Draft Informal Rept.Rept Satisfies Preliminary Rept Milestone for Task I Defined in Safety Evaluation of Core Melt Accidents:Cp,Ml & OL Applicants ML17276B0731982-02-10010 February 1982 Comments on Const of Two Power Reactors by Skagit/Hanford Nuclear Project at Hanford.Purchasing & Finishing Terminated Wppss Plants at Hanford & Satsop Should Be Considered as One Alternative to Proposed Action ML19343A2031980-09-0404 September 1980 Forwards Monthly Progress Rept for Aug 1980 Under NRC Licensing Case Review Assistance Program ML20098C8521976-03-26026 March 1976 Forwards Revised Concept Calculations Requested by Jc Petersen for Facilities & Results of Calculations ML20098D3701975-06-13013 June 1975 Forwards Revised Concept Calculations for WPPSS Projects 1 & 4,in Response to Jc Petersen Request 1989-01-13
[Table view] |
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- BROOKHAVEN NATIONAL LABORATORY 0 ' ' ASSOCIATED; UNIVERSITIES,ilNC.":
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Department of Nucleor Energy; FTS 666' 7005
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7 March 2, 1988
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Mr." James Conway; i
U.S. Nuclear Regulatory Commission -
Vendor Programs Branch Mail'Stop 9D4 Washington, DC- 20555 ay Reft Testing of Farley Bolting Materials, Fin A-3866, Task Asuignment 9 *
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! . . Dear Jim L h n. ' '
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Enclosed is four copies of a report on the metallurgical evaluation of five
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bolts that had been obtained from the Farley plant and tested by C.J. Czajkowski of BNL. These five bolts (S0-18 through S0-21'and S0-23) had originally been 'out a y , of specification after the' original tensile and chemical testing. The results of ,
the testing are
>
S0-18 Exceeded chromium level and had below maximum carbon level. (Notes insuf ficient carbon level was inadvert.ently not identified in the pre-vious report.). This bolt is considered " suitable for service" after reevaluation and examination. t
'
' SO-19 : Exceeded chromium levels. The.s bolts are consi<iered "suiuble for
~ Through- service" after reevaluation and examination.
.
S0-21 S0-23: Exceeded maximum hardaess. This bolt is considered acceptable af ter retesting in accordance with ASTM A370-77. t
'7 If there are any questions, please feel free to call. j,r.. ;
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'WStaf , John H. Taylor, Group Leader
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Encl. Plant Systems & Equipment Analysis j.
cc:
E. Baker, NR C. Czajkowski ;
B. Grenier, NRC R. Hall W. Kato W.1 Shier J. Stone, NRC
' File '
8911030029 891030 PDR FOIA g/r/
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BROOKHAVEN, NATIONAL ~ LABORATORY. ,
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FROM: C.J.'CZajkowski(FTS 666-4420) " ,:
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. suancT Retesting' of Bolts S0-19 through S0-2 and S0-23 for USNRC I, l& -
Pursuant with Task Order No. 9 under FIN A-3866, please find attached copies of metallurgical evaluations for bolts identified as S0-18 through S0-21 and S0-23. These bolts had previously been found (nty memo to you 10/20/87) to be out of specification after the original tensile and chemical i testing.
The-results of the retesting are:
S0-18: Exceeded chromium level and had below maximum carbon level.
(Note: insufficient carbon level was inadvertently not identified in 10/30/87 report.) This bolt is considered " suitable for service" i t after reevaluation and examination.
S0-19 thrugh S0-21: ' Exceeded chromium levels. These bolts are con-sidered " suitable for service" after reevaluation and examination. r S0-23: . Exceeded maximum hardness. T ,is bolt is considered accept-able after retesting in accordance with ASTM A370-77. h;?
This completer Task 9.under FIN ' A-3866. Four additional copies of the If there are any quest 9ons, d'
. report are attached for transmittal to the NRC. N please contact me at the above number.
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O # BOLT IDENTIFICATION: .50-13l .80LT SPECIFICATION:'; A193-87 L ".,,'1
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TENSILE STRENGTH:
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I 1 e Actual Required by Specification '
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Not Required:
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CHEMICAi. ' ANALYSIS:
Actual w/o Required by Specification w/o t
0.37 - 0.49
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Carb'on 0.34 (Note 2)
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1' ~ Manganese 0.96 0.65 - 1.10 i '4 "
.; Phosphorus 0.12 0.035 max l.i . Sulfur 0.006 0.040 reax ;
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Silicon 0.28 0.15 - 0.35 Chromium 1, 1.55 (fbte 1) 0.75 - 1.20
' Nickel 0.06 - E Molybdenum 0.19 0.15 - 0.25 (0
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C0i9 TENTS: 1) Chromium value exceeds specification requirements even factoring
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L after factoring permissible variation. Note: inadvertently not identified on original 10/30/87 report.
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METALLURGICAL EVALUATION n
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, BOL7 IDENTIFICATION: S0-18 BOLT SPECIFICATION: ' A193-B7 .f.
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- 1. A section was cut from the boh, mounted in epoxy, then metallurgically l ground, polished and etched (2% Nial). The section showed (Figure 1) a 3,!
tempered martensite microstructure consistent with this type of material. il r 2. Half _ of the fracture face (after tensile testing) was- examined under the scanning election microscope (SEM) (Figure 2). The resulting fractograph A,.
showed a _ dimpled rupture appearance which indicated good ductility in the "i <'
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Conclusions:
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Ti : bolt is considered to be suitable for service for the following reasons: ~
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- 1. The polished section showed a microstructure consistent for this grade of bolt. The SEM examination revealed that the fracture (after tensile
- testing) was' ductile in nature. This coupled with the fact that the bolt -j 4 met the tensile requirements of A193-B7 material and was only 0.01% below
- the minimum carbon level and 0.35% above the maximum chromium level (all other chemical requirements were met) leads one to believe that the ten- W '
sile requirements will not be a problem for this bolt. The only other y major consideration would be if the bo?t could fail in a brittle (as .
l opposed to ductile) manner in service (Notch Toughness) due to these chem- .@f ical composition variances, .de
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L The Metals Handbook, 8th Eoition. Vol. 1. Properties and Selection of PI%
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...the ability of a metal to yield plastically under high ?
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c from 0.15 to 0.80% in normalized plain carbon steels, the notch tonghness L decreases- (at room temperature). This lowering of energy absorption is s. y
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accompanied by a subsequent raising of the transition tamperature (6*F per 0.01% increase in carbon content above 0,30% carbon). The net effect of ds r ~
this lower amount of carbon on this particular bolt would then be to i increase the amount of energy absorbed and decrease the transition temper-ature as much as 6'F. Both of which should be beneficial to the bolt's
" suitability for service."
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BOLT IDENTIFICATION:' S0-19 . BOLT SPECIFICATION:, A193,87 q y
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SOLT SIZE:,
.
K. .;
-
1" -. 8 UNC L. ,
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a
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TENSILE-STRENGTH: 'T
" ' .:!.
-
Actual Required by Specification
~ @'
.u ;
143.03 ksi 125 ksi (min.). '
t Failure Location - Shoulder ;
.-
. , i r
,
HARDNESS: ..
'
-
t Actual Required by Specification
,
l
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27.6 R Not Required C ;j i
; F
. . CHEMICAL APALYSIS:
,E.
,
.
Actual w/o Required by Specification w/o
.
o' Carbon. -0.40 9.37 - 0.49 ?
'
Hanganese- 0.98 0.65 - 1.10 $,
Phosphorus' O.005 ' O.035 max Sulfur. 0.016 0.040 max W_ -
Silicon
~
0.21 0.15 - 0.35' Ch romium 1.80 (No,te 1) 0.75 - 1,90 !'
. NIcr el
-
0.34 r.
- Molybdenum 0.19 0.15 - 0.25 ii ;
Vanadium <0.05 -
$
l Columbium + Tantalum s0.05 -
OT '
sg4 .~,
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< = Less than 'N4-( .u se gg .
- a. <
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a COPMENTS: 1) Chromium value exceeds specification requirements even factoring 3,,fA l -'
n ' i in permissible variations (0.05% over) '
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BOLT. IDENTIF.ICATION: S0-19 BrLT SPECIFICATION: A193-B7 ,
A +,
.$
BOLT SIIE: 1"- 8 UNC y l:p Methodology: Y ,
- 1. A section was cut from the bolt, mounted in epoxy, then metallurgically '.
ground, polished and etched (2% Nital). The section showed (Figure 3) a p+
+
tempered martensite microstructure consistent with this type of material.
- 2. Half of the fracture face (after tensile testing) was examined under the Z scanning election microscope (SEM) (Figure'1). The resulting fractograph showed a dimpled rupture appearance which indicated good ductility in the fracture.
Conclusions:
The bolt is considered to'be suitable for. service for the following reasons:
- 1. The polished section showcd a microstructure consistent for this grade of bolt. The SEM examination revealed that the fracture t'after tensile .
1 testing) was ductile in nature. These observations plus the fact that the bolt met the tensile requirements and all of the chemical requirements
'
.t (except chromium) of the specification leads one to examine the ability of the bolt to resist rapid failure (notch toughness).
,
.&
Chromium Content
,;-,
The Metals Handbook, 8th Edition, Vol. 1, Properties and Selection of Metals, states: .*.
" Chromium has slight effect on tran ition temperature..." g!
.,
Since the impact properties should not be significantly impaired due to the higher chromium content, the bolt is considered " suitable for s
.. service." y
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- B0LT IDENTIFICATION: { $02 20 , '80LTSPECIFICATI0m/A'193-87:3. I
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BOLT SIZE: 3/8" - 16 UNC 9,
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TENSILE STRENSTH: i WG Actual Required by Specification' M- ;..
151.74 ksi 125 ksi (min.) :. l 4
. :.", e Fa11'ure Location - Threads h 7
*?
.
HARDNESS: g Actual Required'by Specification tg 23.83 R Not Required c
CHEMICAL ANALYSIS: ..
I f Actual w/o Required by Specification w/o ,n
.
!
- 1 Carbon 0.41 0.37 - 0.49 l- ' Manganese 0.97 0.65 - 1.10 .(f),
Phosphorus <0.005 0.035 max ,
v; Sulfur 0.008 0.040 max nl Silicon 0.27 0.15 - 0.35 N-ll, Chromium 1.51 (Note 1) 075 1S
- 1. -
, ,
Nickel- 0.05 -- QI.il Mt j Molybdenum 0.22 0.15 - 0.25 c>;
Vanadium <0.05 -
.A -:
I Columbium +' Tantalum <0.05 - i
; 4
<e. . , -
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-t C0tWENTS: 1) Chromium value exceeds specification requirements even factoring r, 1 g.
in permissible variations (0.05% over) fi,
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BOLT IDEKIIFICATiON: S0-20
- BOLT SPECIFICATION: 'A193-B7 N'
; e
,
'
4-B0LT SIZE: . 3/8" - 16 UNC a;
Methodolony:
,
- 1. A.section was cut from the bolt, mounted in epoxy, then metallurgically ground, polished and etched (2% Nital). The section showed (Figure 5) a tempered martensite microstructure consistent ~with this type of material.
'
- 2. Half of the fracture face (after ten:ile testing) was examined under tht scanning election microscope (SEM) (Figure 6). The resulting fractograph showed a dimpled rupture appearance which indicated good ductility in-the '(
fracture, t
Conclusions:
.
The bolt is considered to be suitable for service for the following reascns:
- 1. The polished section showed a microstructure consistent for this gra:le of '
bol t. The SEM examination revealed that the fracture (after tensile
( testing)wasductileinnature. These observations plus the fact that the ~
bolt met the tensile requirements and all of the chemical requirements (except chromium) of the specification leads one to examine the ability of the bolt to . dst rapid failure (notch toughness). >
Chromium Content
'
- . ,
'
L The Metals Handbook, 8th Edition, Vol. 1, Properties and Selection of '
!
Metals, states:
" Chromium has slight effect on transition temperature..." !
,
- l. Since the impact properties should not be significantly impaired due to E
'
l the higher chromium content, the bolt is considered " suitable for c
.. service."
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.EBOLT IDENTIFICATION: s50-21,).'
;r BOLT'.SPCCIFICATION: JA193,87) > . , .e.
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- e . BOLT SIZE: 1/4" - 20 UNC ,u r ,. ,
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..fg k.
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m.
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TENSILE STRENGTHt
- '
m,,
'
Actual Required by Specification 'it-
, .
,
- k.
,,
'
d A
fc 158.80 ksi 125 ksi (min.)' <
* ,
,
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'
, , Failure Location ~- Threads
- y. .
HARDNESS:
- , - , .
t
"
Actual Required by Specification 62.'6 R Not Required A
,
CHEMICAL' ANALYSIS:
'
Actual w/o Required by Specification w/o 4
>
Carbon 0.41 0.37 - 0.49
. Manganese 0.94 0.65 - 1.10 g 7 Phosphorus 0.020 0.035 max la Sulfur 0.024 0.040 max % "
Silicon 0.27 0.15 - 0.35 l Chromium 'e 1.4C (f;t,tt.1) 0.75 - 1.20 '
M<
.
l Nickel ' O.47 -
l Molybdenum- 0.23 0.15 - 0.25 f, Vanadium <0.05 - 'i
4 Columbium + Tantalum <0.05 -
E>
'
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.. .-
- d.
s9
-
< = Less than 7 .'-
l,
- 1) Chromium value exceeds specification requirements even factoring
, o C0t91ENTS:
dk (
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in permissible variations (0.05% over)
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METALLURGICAL EVALUATION .n
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( ,
r DOLT IDENTIFICATION:. S0-21 BOLT SPECIFICATION: 'A193-B7 A
-
, &r_
~
liQ S0LT SIZE: 1/4"-20UNC Y
.
M: :f.
>
Methodology:
f... ;
- 1. A section'was cut from the bolt, mounted in epoxy, then metallurgically 7
ground, polished and etched (2% Nital). The section showed (Figure 7) a .;j tempered martensite microstructure consistent with this type of material. ~
L 2. Half'of the fracture face (after tensile testing) was examined under the '
scanning election microscope (SEM) (Figure 8). The resultin0 fractograph
,
showed a dimpled rupture appearance which indicated good ductility in the
-
fracture.
I
Conclusions:
,
'
The bolt is considered to be suitable for service for the following reasons:
I 1. The polished section showed a microstructure consistent for this grade of bolt. The SEM examination revealed that the fracture (after tensile
{ testing)wasductileinnature. These observations plus the fact that the bolt met' the tensile requirements and all of the chemical requirements ,
(except chromium) of the specification leads one to examine the ability of ,"
the bolt to resist rapid failure (notch toughness).
4?gC
,
The Metals Handbook, 8th Edition, Vol. 1, Properties and Selection of ,'
l Metals, states: ,
-
A .a4 "
1: " Chromium has slight effect on transition temperature..."
Ls L Since the impact properties should not be significantly impaired due to L the higher chromium content, the bolt is considered " suitable for 4;. :
E
. service." g
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- n. 'B0LT IDENTIFICATION: $0-23 BOLT SPECIFICATION: , , ,A 193-88 i g .. ' .gjf
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30LT SIZE: 5/8" - 11 UNC ti th .v I
TENSILE STRENGTH: .
'
h Actual Required by Specification
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89.65 ksi 75 ksi (min.)
- .
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Failure location - Threads . ~ l n ,
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[ ., i HARDNESS: '
,
1
.
Actual Required by Specification
'
k 63.6 RA (Note 1) 223HB(max.) 4
(equates to 262 HB) 1
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CHEMICAL ANALYSIS: i 1
.
Actual w/o Required by Specification w/o q Carbon 0.06 Manganese 1.75 0.08 max 2.00 max 3
Ja l Phoschorus
- <
Sulfur-0.050 (Note 2) 0.021 0.045 max 0.030 max tf1 V7 '
2.,, Silicon 0.72 1.00 mn f
-
Cnromium 19 0 18.00'- 20.00 9+
. Nickel 9.0 8.00 - 10.50 -
Molybdenum M..
L 0.40 -
Vanadium Columbium + Tantalum
<0.05
<0.05
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< ' Less than 'i[ .n a
,f.' l k
%~, a 1 COMENTS: 1_) Although specification allows a maximum hardness of 241 HB %' ,
1 (A193). this bolt exceeds hardness maximum. 2) Permissible variation for l phosphorus (0.010% over) by specification allows acceptance of this value. A <
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ETALLURGICAL EVALUATION ^ o, , ,.
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BOLT IDENTIFICATION: S0-23 BOLT SPECIFICATION: A193-88 %,'
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BOLT SIZE: '5/8" - 11 UNC 4
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' Methodology:
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- 1. A section was cut from the bolt, mounted in epoxy, then metallurgically 3
- ground, polished and etched (electrolytic oxalic acid). The section dgy, showed (Figure 9) an austenitic microstructure consistent with this type
,
of material.
y-
- 2. Half of the fracture face (after tensile testing) was examined under the "
SEM (Figure 10). The resultant fractograph showed a dimpled rupture '
appearance which indicates good ductility. f,
'
- 3. Consistent with the requirements of ASTM A370-77, a transverse section through the bolt was cut and six hardness ' readings taken along the axial :
. length.
E
Conclusions:
M.'
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The bolt is considered to meet specification (ASTM A193-81a) requirements
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, after hardness retesting. The specification requirements allow a maximum . ':
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hardness of 223HB (with a maximurr hardness of 241HB allowed for 3/4" diameter g
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and smaller bolts), the six hardness retests showed the following hardnesses:
L RB 82, 83, 82, 85, 79, 88.5 j RB. 89.5, equates to 18,1 HB ,
- y All of which are below the specification maximum. *
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