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MONTHYEARML20082C5461983-10-31031 October 1983 Analysis of Capsule U from Alabama Power Co Jm Farley,Unit 2 Reactor Vessel Radiation Surveillance Program Project stage: Other ML20082C5361983-11-10010 November 1983 Forwards WCAP-10425,Analysis of Capsule U from Alabama Power Co,Jm Farley,Unit 2 Reactor Vessel Radiation Surveillance Program Project stage: Other ML20080K6461984-02-10010 February 1984 Proposed Tech Specs Re RCS Pressure & Temp Requirements Project stage: Other ML20080K6361984-02-10010 February 1984 Application for Amend to License NPF-8,changing Tech Specs Re RCS Pressure & Temp Requirements Project stage: Request ML20092G8961984-04-30030 April 1984 Fracture & NDE Evaluations for Closure Flange Regions of Comanche Peak Units 1 & 2 Project stage: Other ML20092G8931984-06-18018 June 1984 Forwards Info Clarifying 831110 Reactor Vessel Surveillance Capsule Rept & 840210 Associated Tech Spec Change Request, Per 840503 & 14 Requests Project stage: Other ML20107M1501984-11-0606 November 1984 Forwards Updated Status of 1984 Licensing Items Identified in .Approval of Item 28 Re Deletion of Requirement to Remove Reactor Vessel Capsule Crucial Prior to Scheduled Startup.Decision on Item Needed by 850104 Project stage: Other 1984-11-06
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Category:CORRESPONDENCE-LETTERS
MONTHYEARTXX-9924, Forwards Responses to Questions by NRC Re Application for Amends to Licenses NPF-87 & NPF-89,by Incorporating Changes Increasing RWST low-level Setpoint from Greater than But Equal to 40% to Greater than But Equal to 45% of Span1999-10-22022 October 1999 Forwards Responses to Questions by NRC Re Application for Amends to Licenses NPF-87 & NPF-89,by Incorporating Changes Increasing RWST low-level Setpoint from Greater than But Equal to 40% to Greater than But Equal to 45% of Span ML20217M5711999-10-20020 October 1999 Forwards Insp Repts 50-445/99-15 & 50-446/99-15 on 990822- 1002.Two Severity Level IV Violations of NRC Requirements Identified & Being Treated as non-cited Violations Consistent with App C of Enforcement Policy L-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 TXX-9923, Forwards Monthly Operating Repts for Sept 1999 for CPSES, Units 1 & 2,per Plant TS 5.6.4.No Failures of Challenges to PORVs of SV for Units Occurred1999-10-15015 October 1999 Forwards Monthly Operating Repts for Sept 1999 for CPSES, Units 1 & 2,per Plant TS 5.6.4.No Failures of Challenges to PORVs of SV for Units Occurred ML20217E7951999-10-12012 October 1999 Forwards COLR for Unit 1,Cycle 8,per TS 5.6.5 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 ML20212L2891999-10-0101 October 1999 Discusses Closeout of GL 97-06, Degradation of Steam Generator Internals. Purpose of GL Was to Obtain Info That Would Enable NRC to Verify That Condition of Licensee SG Internals Comply with Current Licensing Bases TXX-9922, Forwards Revised COLR, for Cycle 5 for Unit 21999-10-0101 October 1999 Forwards Revised COLR, for Cycle 5 for Unit 2 ML20216J5571999-10-0101 October 1999 Provides Final Response to GL 98-01,suppl 1, Y2K Readiness of Computer Sys at Npps 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 ML20212F7481999-09-24024 September 1999 Forwards SER Authorizing Relief from Exam Requirement of 1986 Edition ASME Code,Section XI Pursuant to 10CFR50.55a(a)(3)(ii) for Relief Request A-3 & 10CFR50.55a(g)(6)(i) for Relief Requests B15,16,17 & C-4 ML20212G0721999-09-24024 September 1999 Forwards Rev 4 to Augmented Inservice Insp Plan for CPSES, Unit 1. Future Changes & Revs to Unit 1 Augmented Inservice Insp Plan Will Be Available on Site ML20212H0461999-09-24024 September 1999 Forwards Rev 6 to CPSES Glen Rose,Tx ASME Section XI ISI Program Plan for 1st Interval on 990820 ML20212F1041999-09-23023 September 1999 Requests That NRC Be Informed of Any Changes in Scope of Y2K System Deficiencies Listed or Util Projected Completion Schedule for Comanche Peak Steam Electric Station,Units 1 & 2 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 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 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 ML20212F1111999-09-21021 September 1999 Discusses Closeout of GL 97-06, Degradation of Steam Generator Internals ML20212E6661999-09-21021 September 1999 Advises That Info Contained in Application & Affidavit, (CAW-99-1342) Re WCAP-15009,Rev 0, Comache Peak Unit 1 Evaluation for Tube Vibration Induced Fatigue, Will Be Withheld from Public Disclosure ML20212D9111999-09-16016 September 1999 Informs That on 990818,NRC Completed Midcycle PPR of CPSES & Did Not Identify Any Areas in Which Performance Warranted Insp Beyond Core Insp Program.Core Insp Plan at Facility Over Next 7 Months.Insp Plan Through March 2000 Encl 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 ML20212A7601999-09-14014 September 1999 Forwards Insp Repts 50-445/99-14 & 50-446/99-14 on 990707-0821.Four Violations Occurred & Being Treated as Ncvs.Conduct of Activities Was Generally Characterized by safety-conscious Operations & Sound Radiological Controls 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 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 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 TXX-9921, Suppls 981221 LAR 98-010 to Licenses NPF-87 & NPF-89, Clarfying Conditions of Use Re Analytical Methods Used to Determine Core Operating Limits,Per Telcon with NRC1999-09-10010 September 1999 Suppls 981221 LAR 98-010 to Licenses NPF-87 & NPF-89, Clarfying Conditions of Use Re Analytical Methods Used to Determine Core Operating Limits,Per Telcon with NRC 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 ML20211P3761999-09-0707 September 1999 Ack Receipt of Ltr Dtd 990615,transmitting Rev 30 to Physical Security Plan,Per 10CFR50.54(p).No NRC Approval Is Required TXX-9915, Responds to 990701 & 0825 RAI Telcons Re Spent Fuel Pool Temp,Per LAR 98-008,which Requested Increase in Spent Fuel Storage capacity.Marked-up Page 4-1 of CPSES Fuel Storage Licensing Rept, Encl1999-09-0303 September 1999 Responds to 990701 & 0825 RAI Telcons Re Spent Fuel Pool Temp,Per LAR 98-008,which Requested Increase in Spent Fuel Storage capacity.Marked-up Page 4-1 of CPSES Fuel Storage Licensing Rept, Encl ML20211L9871999-09-0303 September 1999 Forwards Rev 31 to Technical Requirements Manual. All Changes Applicable to Plants Have Been Reviewed Under Util 10CFR50.59 Process & Found Not to Include Any USQs 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 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 ML20211K2231999-08-31031 August 1999 Forwards Txu Electric Comments of Rvid,Version 2 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 ML20211J3801999-08-27027 August 1999 Forwards Corrected TS Page 3.8-26 to Amend 66 to Licenses NPF-87 & NPF-89,respectively.Footnote on TS Page 3.8-26 Incorrectly Deleted ML20211G7301999-08-26026 August 1999 Forwards Revs 29 & 30 to CPSES Technical Requirements Manual (Trm). Attachments 1 & 2 Contain Description of Changes for Revs 29 & 30 Respectively ML20211G1081999-08-26026 August 1999 Responds to NRR Staff RAI Re Util Mar 1999 Submittal for NRC Review & Approval of Changes to CPSES Emergency Classification Procedure 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 ML20211G3441999-08-25025 August 1999 Forwards Response to NRC RAI on LAR 98-010 for Cpses,Units 1 & 2.Communication Contains No New Licensing Commitments Re Cpses,Units 1 & 2 1999-09-09
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARTXX-9924, Forwards Responses to Questions by NRC Re Application for Amends to Licenses NPF-87 & NPF-89,by Incorporating Changes Increasing RWST low-level Setpoint from Greater than But Equal to 40% to Greater than But Equal to 45% of Span1999-10-22022 October 1999 Forwards Responses to Questions by NRC Re Application for Amends to Licenses NPF-87 & NPF-89,by Incorporating Changes Increasing RWST low-level Setpoint from Greater than But Equal to 40% to Greater than But Equal to 45% of Span L-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 TXX-9923, Forwards Monthly Operating Repts for Sept 1999 for CPSES, Units 1 & 2,per Plant TS 5.6.4.No Failures of Challenges to PORVs of SV for Units Occurred1999-10-15015 October 1999 Forwards Monthly Operating Repts for Sept 1999 for CPSES, Units 1 & 2,per Plant TS 5.6.4.No Failures of Challenges to PORVs of SV for Units Occurred ML20217E7951999-10-12012 October 1999 Forwards COLR for Unit 1,Cycle 8,per TS 5.6.5 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 ML20216J5571999-10-0101 October 1999 Provides Final Response to GL 98-01,suppl 1, Y2K Readiness of Computer Sys at Npps TXX-9922, Forwards Revised COLR, for Cycle 5 for Unit 21999-10-0101 October 1999 Forwards Revised COLR, for Cycle 5 for Unit 2 ML20212H0461999-09-24024 September 1999 Forwards Rev 6 to CPSES Glen Rose,Tx ASME Section XI ISI Program Plan for 1st Interval on 990820 ML20212G0721999-09-24024 September 1999 Forwards Rev 4 to Augmented Inservice Insp Plan for CPSES, Unit 1. Future Changes & Revs to Unit 1 Augmented Inservice Insp Plan Will Be Available on Site 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 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 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 TXX-9921, Suppls 981221 LAR 98-010 to Licenses NPF-87 & NPF-89, Clarfying Conditions of Use Re Analytical Methods Used to Determine Core Operating Limits,Per Telcon with NRC1999-09-10010 September 1999 Suppls 981221 LAR 98-010 to Licenses NPF-87 & NPF-89, Clarfying Conditions of Use Re Analytical Methods Used to Determine Core Operating Limits,Per Telcon with NRC TXX-9915, Responds to 990701 & 0825 RAI Telcons Re Spent Fuel Pool Temp,Per LAR 98-008,which Requested Increase in Spent Fuel Storage capacity.Marked-up Page 4-1 of CPSES Fuel Storage Licensing Rept, Encl1999-09-0303 September 1999 Responds to 990701 & 0825 RAI Telcons Re Spent Fuel Pool Temp,Per LAR 98-008,which Requested Increase in Spent Fuel Storage capacity.Marked-up Page 4-1 of CPSES Fuel Storage Licensing Rept, Encl ML20211L9871999-09-0303 September 1999 Forwards Rev 31 to Technical Requirements Manual. All Changes Applicable to Plants Have Been Reviewed Under Util 10CFR50.59 Process & Found Not to Include Any USQs ML20211K2231999-08-31031 August 1999 Forwards Txu Electric Comments of Rvid,Version 2 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 ML20211G7301999-08-26026 August 1999 Forwards Revs 29 & 30 to CPSES Technical Requirements Manual (Trm). Attachments 1 & 2 Contain Description of Changes for Revs 29 & 30 Respectively ML20211G1081999-08-26026 August 1999 Responds to NRR Staff RAI Re Util Mar 1999 Submittal for NRC Review & Approval of Changes to CPSES Emergency Classification Procedure ML20211G3441999-08-25025 August 1999 Forwards Response to NRC RAI on LAR 98-010 for Cpses,Units 1 & 2.Communication Contains No New Licensing Commitments Re Cpses,Units 1 & 2 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 ML20211C0991999-08-17017 August 1999 Forwards Rev 3 to ASME Section XI ISI Program Plan,Unit 2 - 1st Interval, Replacing Rev 2 in Entirety ML20210U3981999-08-17017 August 1999 Forwards Monthly Operating Repts for July 1999 for CPSES, Units 1 & 2,per TS 6.9.1.5.No Failures or Challenges to PORVs or SVs for Plant Occurred TXX-9919, Forwards Relief Request A-3,Rev 1 to Unit 1 ISI Program,Per Conversations Between NRC & Txu Electric on 9908021999-08-16016 August 1999 Forwards Relief Request A-3,Rev 1 to Unit 1 ISI Program,Per Conversations Between NRC & Txu Electric on 990802 ML20210R6561999-08-13013 August 1999 Forwards Response to NRR 990805 Telcon RAI Re License Amend Request 98-010,to Increase Power for Operation of CPSES Unit 2 to 3445 Mwth & Incorporating Addl Changes Into Units 1 & 2 TS ML20210S6411999-08-12012 August 1999 Informs That Wg Guldemond,License SOP-43780,is No Longer Performing Licensed Duties.Discontinuation of License Is Requested 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 ML20210N1101999-08-0404 August 1999 Provides Supplemental Info to Util 990623 License Amend Request 99-005 Re Bypassing DG Trips.Info Replaces Info Contained in Subject Submittal in Attachment 2,Section II, Description of TS Change Request ML20210J2301999-08-0202 August 1999 Forwards Amend 96 to CPSES Ufsar.Replacement of FSAR Figures with Plant Process Flow Diagrams Meets Intent & Requirements of NRC Reg Guide 1.70,Rev 2 ML20210J6071999-08-0202 August 1999 Forwards line-by-line Descriptions of Changes in Amend 96 to CPSES UFSAR Transmitted by Util Ltr TXX-99166,dtd 990802. Replacment of FSAR Figures with Plant Process Flow Diagrams Meets Intent & Requirements of NRC Reg Guide 1.70,rev 2 TXX-9916, Notifies NRC That CPSES Units 1 & 2,improved TS Implemented on 9907271999-08-0202 August 1999 Notifies NRC That CPSES Units 1 & 2,improved TS Implemented on 990727 TXX-9918, Forwards CPSES 10CFR50.59 Evaluation Summary Rept 0008,for 970802-990201 & CPSES Commitment Matl Change Evaluation Rept 0003,for 970802-9906301999-08-0202 August 1999 Forwards CPSES 10CFR50.59 Evaluation Summary Rept 0008,for 970802-990201 & CPSES Commitment Matl Change Evaluation Rept 0003,for 970802-990630 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 ML20210G5861999-07-29029 July 1999 Forwards fitness-for-duty Program Performance Data for Six Month Period of Jan-June 1999 ML20210J0121999-07-27027 July 1999 Forwards Summary of Methodology for Determination of NDE Measurement Uncertainty,In Response to Recent Discussions with NRC Re LAR 98-006 Concerning Rev to SG Tube Plugging Criteria 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 TXX-9917, Provides Info Re Augmented Inservice Insp Plan,Which Requires Periodic Insp of Rv Head & Internals Lifting Devices at CPSES1999-07-26026 July 1999 Provides Info Re Augmented Inservice Insp Plan,Which Requires Periodic Insp of Rv Head & Internals Lifting Devices at CPSES ML20210F3121999-07-26026 July 1999 Responds to GL 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal, 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 1999-09-03
[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 ML20059K6141990-09-17017 September 1990 Submits Documentation of 900906 Telcon Re Addl Util Candidate for Licensed Operator Generic Fundamentals Exam Section 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 ML20065D6621990-09-12012 September 1990 Forwards NPDES Permit AL0024619 Effective 900901.Limits for Temp & Residual Chlorine Appealed & Stayed 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 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 ML20059J2891990-09-12012 September 1990 Confirms Rescheduling of Response to Fitness for Duty Program Notice of Violation 90-18-02,per 900907 Telcon ML20059G8761990-09-10010 September 1990 Forwards Rev 8 to Physical Security Plan.Rev Withheld (Ref 10CFR73.21) ML20059F8461990-09-0404 September 1990 Responds to NRC Re Violations Noted in Insp Repts 50-445/90-22 & 50-446/90-22 on 900606-0705.Corrective action:OPT-467A Changed to Provide Addl Prerequisite Info Re Solid State Protection Sys Switch Lineup During Testing ML20059E4981990-08-31031 August 1990 Forwards Revised Radial Peaking Factor Limit Rept for Cycle 1 Per Tech Specs Through Amend 1 & ML20059E4201990-08-30030 August 1990 Forwards Objectives & Guidelines for 1990 Emergency Preparedness Exercise Scheduled for 901113 ML20028G8281990-08-29029 August 1990 Forwards Comanche Peak Unit 1 Semiannual Radioactive Effluent Release Rept,900403-0630 & ODCM ML20064A3431990-08-28028 August 1990 Forwards Corrected Insertion Instructions to Rev 8 to Updated FSAR for Jm Farley Nuclear Plant TXX-9030, Forwards Endorsements 40 & 4 to Nelia Policy NF-274 & Maelu Policy M-90,respectively & Endorsements 10,11,12,13,14 & 15 to Maelu Policy MF-1311990-08-27027 August 1990 Forwards Endorsements 40 & 4 to Nelia Policy NF-274 & Maelu Policy M-90,respectively & Endorsements 10,11,12,13,14 & 15 to Maelu Policy MF-131 ML20059F4381990-08-23023 August 1990 Forwards Public Version of Revised Emergency Plan & Corporate Emergency Response Procedures,Consisting of Rev 1 to EPP-313,Rev 7 to CERP-101,Rev 7 to CERP-102 & Revised Forms.W/Dh Grimsley 900906 Release Memo 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 TXX-9029, Responds to NRC Re Violations Noted in Insp Repts 50-445/89-57 & 50-446/89-57.Corrective Actions:Storage Facilities for Instrumentation & Control Work Packages Improved to Provide Acceptable Protection for Documents1990-08-20020 August 1990 Responds to NRC Re Violations Noted in Insp Repts 50-445/89-57 & 50-446/89-57.Corrective Actions:Storage Facilities for Instrumentation & Control Work Packages Improved to Provide Acceptable Protection for Documents 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 ML20058P0571990-08-13013 August 1990 Forwards, Comanche Peak Steam Electric Station Unit 1 Control Room Simulator 10CFR55 Certification Initial Rept ML20058N2311990-08-0909 August 1990 Forwards First Half 1990 fitness-for-duty Program Performance Data,Per 10CFR26.71(d).Point of Contact for Personnel,Random Testing Program Results & Confirmed Positive Tests Summary & Mgt Summary Included in Data ML20059A5831990-08-0404 August 1990 Responds to Info Request Made During 900802 Meeting Re Inadvertent Safety Injections on 900726 & 30 & Results of 100% Power Plateau Testing TXX-9024, Provides Status of Exam Activity & Revised Completion Schedule for Portion of Exam.Util Deferred Preoperational Testing of Facility Spent Fuel Pool Cooling & Cleanup Sys Until Prior to First Off Loading of Spent Fuel1990-08-0303 August 1990 Provides Status of Exam Activity & Revised Completion Schedule for Portion of Exam.Util Deferred Preoperational Testing of Facility Spent Fuel Pool Cooling & Cleanup Sys Until Prior to First Off Loading of Spent Fuel TXX-9022, Forwards Detailed Descriptions of Changes in Amend 79 to FSAR for Comanche Peak Steam Electric Station1990-07-31031 July 1990 Forwards Detailed Descriptions of Changes in Amend 79 to FSAR for Comanche Peak Steam Electric Station TXX-9027, Responds to NRC Re Violations Noted in Insp Repts 50-445/90-19 & 50-446/90-19.Corrective Action:Specific Activity within Tech Spec Limits.No Further Corrective Action Required1990-07-30030 July 1990 Responds to NRC Re Violations Noted in Insp Repts 50-445/90-19 & 50-446/90-19.Corrective Action:Specific Activity within Tech Spec Limits.No Further Corrective Action Required ML20055J1571990-07-27027 July 1990 Submits Info Re Replacement of Borg-Warner/Intl Pump,Inc Check Valve Swing Arms,Per .Swing Arms in ASME Code Class 2 Check Valves Will Be Replaced During First Refueling Outage ML20056A0461990-07-24024 July 1990 Requests Regional Waiver of Compliance for Containment Pressure Transmitter 1-PT-934 During Coming Outage ML20055G7701990-07-18018 July 1990 Updates 900713 Response to NRC Bulletin 90-001, Loss of Fill-Oil in Transmitters Mfg by Rosemount TXX-9023, Responds to NRC Bulletin 90-001, Loss of Fill-Oil in Transmitters Mfg by Rosemount. No Subj Transmitters in Existence at Facility1990-07-18018 July 1990 Responds to NRC Bulletin 90-001, Loss of Fill-Oil in Transmitters Mfg by Rosemount. No Subj Transmitters in Existence at Facility ML20055G6351990-07-17017 July 1990 Submits Results of Remote Shutdown Panel Environ Survey ML20055F9391990-07-13013 July 1990 Advises That Operators Have Gained Significant Experience in Operating Plant Under Wide Range of Conditions & Have Demonstrated Necessary Proficiency to Safely Operate Unit ML20055F9371990-07-13013 July 1990 Submits Info Re Revised Acceptance Criteria for THERMO-LAG Fire Barrier,Per 900705 Telcon.On Exposure to Heat Flux at Surface of Barrier,Listed Mechanisms Activated.Fire Testing Demonstrates That Panels Qualified W/Variation in Thickness 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 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 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 ML20044A7001990-06-26026 June 1990 Responds to Generic Ltr 90-04, Request for Info on Status of Licensee Implementation of Generic Safety Issues Resolved W/Imposition of Requirements or Corrective Actions. Util Will Implement for Unit 2 Activities Performed for Unit 1 ML20044A1411990-06-20020 June 1990 Forwards Endorsements 4 & 5 to Nelia Certificate N-90, Endorsements 5 & 6 to Maelu Certificate M-90 & Endorsements 34,35,36,37,38 & 39 to Nelia Policy NF-274 ML20043H7031990-06-18018 June 1990 Discusses Mod/Rework to Auxiliary Feedwater Sys Check Valves Mfg by Bw/Ip Intl,Inc.All But One Affected Check Valve modified.Motor-driven Auxiliary Feedwater Pump Discharge Valve Check Valve Body Replaced ML20043H2181990-06-15015 June 1990 Forwards Response to 900515 Request for Addl Info on Emergency Preparedness.Ongoing Emergency Preparedness Programs Address Concerns Raised by Citizens for Fair Util Regulation ML20043G3131990-06-15015 June 1990 Forwards Addl Info Re Pressurizer Surge Line Thermal Stratification & leak-before-break Evaluation & Nonproprietary Suppl 3 to WCAP-12247, Supplementary Assessment... & Proprietary Suppl 3 to WCAP-12248 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 ML20043E7131990-06-0101 June 1990 Responds to NRC 900517 Notice of Violation & Imposition of Civil Penalty in Amount of $25,000.Corrective Action:Qc Mgt Counseled Level III Inspector & QC Supervisor on Importance of Good Communications W/Receipt Inspectors 1990-09-04
[Table view] |
Text
-- _
gamm'g Address
- -s Alabama Power Company su((G P *, ,~ 600 North 18th Street Post Office Box 2641 Telephone 205 783-6090 R. P. Mcdonald Senior Vice President-Mmeenetnscene Flintndge Building Alabama Power June 18, 1984 " " " " " ' * ~
Docket Nos. 50-364 Director, Nuclear Reactor Regulation U. S. Nuclear Regulatory Comission Washington, D.C. 20555 Attention: Mr. S. A. Varga Joseph M. Farley Nuclear Plant - Unit 2 Response to NRC Questions - Reactor Yessel Surveillance Capsule Report and Associated Technical Specification Change Request Gentleme$:
Pursuant to NRC Staff request for information of May 3,1984 and NRC letter dated May 14, 1984, Alabama Power Company provides the enclosed response.
This response serves to clarify both the Reactor Vessel Surveillance Capsule Report, which was transmitted to the NRC by Alabama Power Company letter dated November 10, 1983, and the associated technical specification change request dated February 10, 1984.
If there are any questions, please advise.
Yours very truly J >
R. P. Mcdonald b
RPM /CJS:ddr-09 Attachment cc: Mr. L. B. Long Mr. J. P. O'Reilly Mr. E. A. Reeves Mr. W. H. Bradford 8406250279 PDR 840618 ADOCK 05000364 P \
PDR \
-.2:*:
Enclosure NRC Staff Request for Information Heatup/Cooldown Curves Joseph M. Farley Nuclear Plant Unit 2
- 1) NRC Request Provide the nickel composition for all plate materials in the reactor vessel beltline.
Alabama Power Company Response The nickel composition of the plate material in the reactor vessel beltline is tabulated below:
Component Heat No. Plate No. Ni (wt %)
Inter Shell C6309-2 B7203-1 .60 Inter Shell C7466-1 B7212-1 .60 Lower Shell C6888-2 B7210-1 .56 Lower Shell C6293-1 G7210-2 .57
- 2) NRC Request Provide pressure temperature limit curves that comply with the explicit closure flange material temperature requirements of the amended (May 27,1983) Appendix G,10CFR50, or provide the information described in Item 3.
Alabama Power Company Response The pressure temperature limit curves submitted to the NRC by letter dated February 10, 1984 do not reflect the 120*F (normal operation) and 90*F (hydrostatic testing) requirements of Appendix G to 10CFR50 for the flange area since Appendix G also allows a lower temperature to be used, if properly justified. Alabama Power Company's response to question 3, below, provides the required justification.
- 3) NRC Request Provide the analysis that shows that the closure flange region is less limiting than the beltline region.
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4
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Page 2 t
Alabama Power Company Response !
Westinghouse has performed an analysis to demonstrate that the closure ,
flange region is less limiting than the beltline region. As stated in 10CFR50 Appendix G, "the margins of safety for those regions (i.e.,
closure flange regions) when they are controlling are equivalent to ,
those required for the beltline when it is controlling". The Westinghouse analysis was originally performed for the Comanche Peak ;
Nuclear P1 ant - Units 1 and 2. Westinghouse has evaluated the Comanche Peak analysis ( Attachment 2) and has determined that its methodology and results are conservatively bounding for the Farley Nuclear Plant -
Unit 2. Westinghouse has determined that the Farley closure flange i region is less ' limiting than the beltline region as demonstrated in ;
response to NRC questions 3a, 3b, 3c, 3d and 3e. ;
3a) NRC Request Include.as a minimum the following information: i a) A description of the finite element analysis used to determine the ;
- stresses within the closure flange region. i Alabama Power Company Response i
A two dimensional finite element model of a typical 4 loop reactor l vessel closure head flange and vessel flange geometry was used in the analysis. .The WECAN finite element program was used to develop the :
model . A discussion of why the 4 loop model can be applied to the 3 !
loop Farley Nuclear Plant - Unit 2 is provided in response to NRC t question 3b. The finite element model was used to obtain temperature !
and stress gradients induced by the heatup and cooldown transients.
Separate iterations of the finite element model were performed to !
determine the bolt-up, pressure and thennal stresses. Figure 1 of }
Attachment 2 shows the cross sections analyzed for the closure flange -
regions. Figure 2 of Attachment 2 shows the mechanical boundary 5 conditions and Figure 3 of Attachment 2 illustrates the thermal I boundary conditions of the finite element model. A summary description [
of how the model was developed and the rationale for the mechanical and <
thermal boundary conditions is provided in Sections 2.0, 2.1 and 2.2 of ;
Attachment 2.
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Enclosure Page 3 3b) NRC Request Include as a minimum the following information:
b) Indicate the peak bolt-up, pressure and thermal stresses determined by the finite element analysis at the inside and outside surface
> locations of the flange to head and flange to shell junctions. +
Alabama Power Company Response The peak bolt-up, pressure and thermal stresses determined by the finite element analysis are provided in Tables 1 through 4 of Attachment 2. Stress values are provided in these Tables for both the ,
inside and outside surface locations of the flange to head and flange i to vessel cross-sections. Included are both Heat-up and Cooldown transients and associated longitudinal and circumferential stress values. A suumary of the methodology for developing the stress values is provided in Section 3.0 of Attachment 2. It is noted that cross-sections 1 and 2 are for the flange to head junctions and cross-section 3 is for the flange to vessel junction as shown in Figure i 1 of Attachment 2. The tabulated values are considered representative '
for the Farley Nuclear Plant - Unit 2. Stress values are lower for Farley Unit 2 for the following reason:
In Reactor Vessel Design, the vessel wall, head and flange dimensions are sized in such a manner that the total mechanical stresses due to pressure and bolt-up are virtually identical. Thermal stresses are a function of the vessel wall thickness. The Comanche Peak (4 loop) vessel, which was used in the analysis ( Attachment 2), has a thicker reactor vessel wall than Farley - Unit 2. The Comanche Peak Vessel exhibits higher thermal inertia and thereby has higher thermal stresses than the thinner walled 3 long reactor vessels like Farley - Unit 2. :
3c) NRC Request Include as a minimum the following information:
c) Indicate how the bolt-up, pressure and thermal stresses were combined to determine the maximum applied stress intensity factors.
Alabama Power Company Response A safety factor of 2.0 was applied to the stress intensity factor for primary stresses (bolt-up and pressure stresses) as required by ASME Code Section III, Appendix G.
7 o., ..
Enclosure Page .4 The formula for the combination of primary and secondary (thermal) stress intensity factors (K1 ) is as follows:
(K1 ) total = 2 (K')1 primary.+ (K )1 secondary In order to conservatively neglect the reduction of either the primary or secondary stress intensity factors by a corresponding negative
' stress intensity factor value, negative stress intensity factors were assumed to be equal to zero.
3d) NRC Request Include as a minimum the following information:
d) Indicate the flaw geometry used to calculate the maximum applied stress intensity factors.
Alabama Power Company Response The-flaw assumed in the analysis is a 0.625 inch deep surface flaw with an aspect ratio of 1:6. The long dimension' of the flaw is assumed to be in the longitudinal direction for the calculation of stress intensity. factors for longitudinal flaws (both inside and outside surfaces) and to be in the circumferential direction for the calculation of stress intensity factors for circumferential flaws (both inside and outside surfaces). The methods of ASME. Code Section IX, Appendix A,1983 were used to generate the fracture analysis results.
3e) NRC Request Include as a minimum the following information:
e) Indicate the maximum applied stress intensity factors for the flange to head and flange to shell junctions.
Alabama Power Company Response The maximian stress intensity factors, for the three closure flange cross-sectional areas analyzed, are tabulated in Tables 5 through 12 of Attachment 2. The tabular maximisn stress intensity factors for the flange to head and flange to shell junctions are 52.79 ksi Iin and 64.74 ksi JTn respectively. The tabulated values are considered to be representative for Farley Unit 2. The stress intensity factors are
Enclosure Pago 5 lower for Farley Unit 2 for basically the same reasons stated in response to question 3c above. A dimensional analysis was performed by Westinghouse to verify that the Comanche Peak analysis is conservative for Farley - Unit 2.
Table 10 of Attachment 2 indicates that the maximum total stress intensity factor (KI) of 64.74 ksi dei occurred for a hypothetical outside surface circumferential flaw, at the flange to shell juncture during a cooldown transient. The thermal stresses at the Farley Unit 2 (3 loop) flange to shell juncture can be approximated by comparing thermal stresses of two cylinders; one cylinder with a thickness of 9.125 inches (3 loop) and the other cylinder with a thickness of 10.75 inches (4 loop). From Figures A.3-5 and A.3-6 of " Tentative Structural Design Basis for Reactor Pressure Vessels and Directly Associated Components (Pressurized, Water Cooled Systems)," U.S. Department of Commerce, December 1,1958 and February 27, 1959 Addenda, it can be shown that the thermal stresses for the thinner vessel are approximately 75 percent of the thicker vessel. For Farley Unit 2, the resulting maximum total stress intensity factor for primary and secondary stresses is 61.63 ksi JUi for the same hypothetical outside surface circumferential flaw, at the flange to shell juncture during a cooldown transient. This maximum stress intensity factor for Farley Unit 2 is considered to be relatively small.
3f) NRC Request Include as a minimum the following information:
f) Indicate the nondestructive examination methods that will be used during inservice examination to determine that the critical flaw size, which was used in determining the maximum applied stress intensity factors, is not within the flange to head and flange to shell junctions.
Alabama Power Company Response Nondestructive examinations currently used for inservice inspection of the reactor flange-to-vessel and the flange-to-head welds are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, Division 1 - Subsection IWB, " Requirements for Class 1 Components of
. Light-Water Cooled Power Plants" 1974 Edition, Summer 1975 Addenda and APCo's, " Augmented Reactor Vessel Examination Program" transmitted to the NRC by letter dated October 26, 1983. Table IWB-2600 requires volumetric examination of flange-to-shell welds and flange-to-dome welds. The 1974 Edition, Stsnmer 1975 Addenda of Section XI specifies
Enclosure Page 6 the boundaries for voltmetric examination to include the weld and adjacent base material for a distance equal to one-half the weld thickness on both sides of the weld. The areas to be inspected are graphically represented in the Attached Figures 1 and 2.
Volumetric coverage of the reactor vessel flange-to-upper shell weld and specified adjacent base material is accomplished by two ultrasonic scan routines. Coverage from the flange side of the weld involves use '
of angled longitudinal waves from the flange seal surface. Beam angles are selected based on their ability to provide coverage of the weld and specifled adjacent base material to the extent practical and provide t near normal incidence to the plane of the weld. Refracted beam angles in the range 0* to 16* are typically used for these examinations. <
Examinations from the shell side of the weld involve 0 , 45*, and 60 refracted angle beam coverage from the vessel inside diameter surface.
Angle beam scanning is performed in two directions parallel to the weld and perpendicular to the weld from the shell side. Access for the shell side examinations is limited to the Ten Year ISI outage when the ;
core barrel is removed from the reactor vessel.
Volumetric examination of the reactor flange-to-head weld and specified adjacent base material is accomplished by 0*, 45* and 60* refracted angle coverage from the head outside surface. Angle beam scanning is performed in two directions parallel to the weld and perpendicular to the weld from the dome side. ,
It is the judgement of Alabama Powbr Company that the inservice examination methodology and techniques will detect critical flaws for the areas examined.
3g) NRC Request Include as a minimum the following information:
g) Indicate whether the nondestructive examination methods identified in (f) have been evaluated to demonstrate that the examination methods are capable of locating and sizing flaws of the geometry used for calculating the maximum applied stress intensity factors.
Indicate the results of the evaluation.
F Enclosure .
Page 7 -
- Alabama Power Company Response a..
Flaws assumed for this analysis were 0.625 inch deep planar surface flaws with 1:6 aspect ratios. The flaw may be oriented circumfer-entially or axially with respect to the vessel or head and may lie on the OD or ID surface.
The fact that the postulated flaws are surface related is significant from a detection probability point of view. Incipient cracks starting at right angles to a given surface (0D or ID) provide favorable conditions for detection via ASME Code specified 45' shear wave ultrasonic examinations from the opposite surface. Circumferential flaws are oriented favorably for detection during axial scanning.
Axial flaws are oriented favorably for detection during circumferential scans. Circumferential1y oriented flaws in the vessel flange weld region also provide favorable conditions for detection during ultrasonic examinations from the flange seal surface.
It is noted that the maximum stress intensity factor occurs for a postulated outside surface circumfereritial flaw at the flange to vessel juncture during a cooldown transient. As mentioned aboyc, in response to NRC question 3f, the volunetric examinations of the shell side of the flange-to-vessel juncture are performed from the inside surface, thereby enhancing the probability of flaw detection. For the flange to head juncture the maximum stress intensity factor occurs for a postulated inside surface longitudinal flaw during a cooldown transient. The volumetric examinations of the reactor closure head flange is accomplished from the outside surface, thereby enhancing the probability of flaw detection. Beam angles selected for.these particular scans provide near normal incidence to the anticipated flaw plane thereby further enhancing the probability of detection.
Application of near surface exainination methods in the form of full mode 45' or shallow angle techniques significantly increases the probability of detecting flaws at the examination surface, i.e., the vessel inside and the head outside.
l While the qualitative assessment indicates that detection probabilities j' are reasonably good for flaws postulated in this analysis, certain unknown factors such as clad effects, defect roughness, orientation, and transparency due to high compressive stresses influence the ability i to detect and ultimately provide a realistic estimate of the flaw size
. with current techniques. Defect sizing by ultrasonic methods has been the subject of several recent studies. To date, no single method has-been identified which consistently provides precise sizing data.
Typically several different methods must be applied and the most
- conservative results used in any analysis that might be necessary.
4 l
Enclosure Page 8 No quantitative information concerning detection and sizing capabilities of the' techniques currently applied during examinations of closure flange junctions has been developed based upon qualification demonstrations, nor are such demonstrations specifically required by existing codes and standards. However, the above features of the examinations may be considered to establish that flaws of the type postulated in this analysis which fall within the volumes subject to examination are likely to be detected.-
The state-of-the-art of reactor vessel examination has improved over
-the past several years. Enhanced near-surface detection capabilities and tip-diffraction sizing methods are examples. Continued emphasis on NDE technique development. promises to provide further improvements and more quantitative data concerning detection and sizing accuracies.
- 4) NRC Request.
For each capsule in Table 4.4-5 of the Farley 2 Technical Specifications, provide the predicted neutron fluence (E>1MEV) to be received by the capsule at the time of .its withdrawal.
Alabama Power Company Response The predicted neutron fluence (E>1EV) to be received by each capsule at' the scheduled time of withdrawal, submitted in 'the February 10, 1984
. Proposed Technical Specification Change, are listed below:
Capsule ' Lead Removal Estimated Flygnce Factor' Time [a] n/cm' x 10 W U 3.12 Removed (1.1) .56 W 2.70 4 2.18 X -3.12 6 3.78[b]
Z 2.70 12 6.54[c]
V 3.12- 18 11.34 Y 2.70 Standby -
[a] Effective full power years from plant startup
[b] Approximates vessel end of life 1/4 thickness wall location fluence
[c] Approximates vessel end of life inner wall location fluence
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Enclosure Page 9
- 5) NRC Request (Per telephone conversation)
Provide a better quality copy of the proposed heat-up/cooldown curves. t l
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- t NOTE E ALL FLAWS ARE EXAGGERATED IN SIZE AND SCALE I FIGURE 2 HEAD TO FLANGE WELD JOINT w c--- , , , , - ~ , - - - - - - - - - - - - - - - - - - - , - , - - - - - - - - , - - - - ----,--,---~-e,e v-