Letter Sequence Other |
---|
|
|
MONTHYEARML20039B3991981-12-18018 December 1981 Application to Amend Ol,Increasing Storage Capacity of Spent Fuel Pool.Class III Fee Encl Project stage: Request ML20039B4031981-12-18018 December 1981 Proposed Tech Spec Section 5.6 Re Fuel Storage Project stage: Other ML20040E7181982-02-0101 February 1982 Forwards Amend 1 to Spent Fuel Pool Mod Project stage: Other ML20040E7201982-02-28028 February 1982 Amend 1 to Spent Fuel Pool Mod Project stage: Other ML20049J8701982-03-19019 March 1982 Forwards Amend 2 to Spent Fuel Pool Mod, Responding to Questions in Encl 1 of NRC ,Correcting Design Stds & Typographical Errors & Clarifying Tech Spec Safety Evaluation.Insertion Instructions & Safety Evaluation Encl Project stage: Other ML20049J8721982-03-31031 March 1982 Amend 2 to Spent Fuel Pool Mod Project stage: Other ML20054C6211982-04-0505 April 1982 Forwards Amend 3 to 811218 Spent Fuel Pool Mod in Response to 820304 Questions Project stage: Other ML20054E1571982-04-21021 April 1982 Forwards Corrected Page III-7 to Spent Fuel Pool Mod Amend 3,re High Density Rack Installation Project stage: Other ML20054K4071982-06-23023 June 1982 EIS Supporting Amend 14 to License NPF-8 Project stage: Other 1982-03-19
[Table View] |
|
---|
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 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 L-99-031, Forwards Info Requested in Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Exams1999-09-13013 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 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 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 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 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 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 L-99-027, 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 Guidelines1999-07-22022 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 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 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 ML20196J7471999-07-0202 July 1999 Forwards RAI Re Cycle 16 Extension Request.Response Requested within 30 Days of Date of Ltr 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 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 L-99-024, 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 ML20196J3591999-06-30030 June 1999 Forwards SE of TR WCAP-14750, RCS Flow Verification Using Elbow Taps at Westinghouse 3-Loop Pwrs 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 ML20196D1931999-06-22022 June 1999 Discusses Requesting Approval & Issuance of Plant Units 1 & 2 ITS by 990930.New Target Date Agrees with Requested Date ML20196A3401999-06-10010 June 1999 Forwards Insp Repts 50-348/99-03 & 50-364/99-03 on 990404-0515.No Violations Noted ML20196H9801999-06-10010 June 1999 Submits Two RAI Re ITS Section 4.0 That Were Never Sent. Reply to RAI Via e-mail L-99-021, 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-99-022, 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 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 ML20212C2351999-09-16016 September 1999 Submits Corrected Info Concerning Snoc Response to NRC GL 99-02, Lab Testing of Nuclear-Grade Activated Charcoal L-99-031, Forwards Info Requested in Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Exams1999-09-13013 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 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 ML20210G4901999-07-30030 July 1999 Responds to GL 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal, Issued 990603.Ltr Contains NRC License Commitment to Utilize ASTM D3803-1989 with Efficiency Acceptance Criteria Utilizing Safety Factor of 2 L-99-028, Responds to NRC 990730 RAI Re 990423 OL Change Request to Allow for Risk Informed Approach for Evaluation of SG Tube Structural Integrity as Described by NEI 97-06, SG Program Guidelines1999-07-30030 July 1999 Responds to NRC 990730 RAI Re 990423 OL Change Request to Allow for Risk Informed Approach for Evaluation of SG Tube Structural Integrity as Described by NEI 97-06, SG Program Guidelines L-99-027, 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 Guidelines1999-07-22022 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 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, Submits Correction to Errors Contained in to NRC Re TS Changes Re Control Room,Penetration Room & Containment Purge Filtration Systems & Radiation Monitoring Instrumentation.Errors Do Not Require Rev of SA1999-06-30030 June 1999 Submits Correction to Errors Contained in to NRC Re TS Changes Re Control Room,Penetration Room & Containment Purge Filtration Systems & Radiation Monitoring Instrumentation.Errors Do Not Require Rev of SA L-99-249, Submits Correction to Errors Contained in to NRC Re TS Changes Re Control Room,Penetration Room & Containment Purge Filtration Systems & Radiation Monitoring Instrumentation.Errors Do Not Require Rev of SA1999-06-30030 June 1999 Submits Correction to Errors Contained in to NRC Re TS Changes Re Control Room,Penetration Room & Containment Purge Filtration Systems & Radiation Monitoring Instrumentation.Errors Do Not Require Rev of SA L-99-025, 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-99-225, 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 ML20195E9581999-06-0707 June 1999 Responds to GL 98-01, Yr 2000 Readiness of Computer Sys at Nuclear Power Plants L-99-022, 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 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-99-021, 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-99-203, 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 L-99-020, Forwards Response to NRC RAI Re GL 96-05 for Farley Nuclear Plant.Farley Is Committing to Implement Phase 3 of JOG Program1999-05-28028 May 1999 Forwards Response to NRC RAI Re GL 96-05 for Farley Nuclear Plant.Farley Is Committing to Implement Phase 3 of JOG Program ML20195F2101999-05-24024 May 1999 Requests That Farley Nuclear Plant Proprietary Responses to NRC RAI Re W WCAP-14750, RCS Flow Verification Using Elbow Taps at W 3-Loop Pwrs, Be Withheld from Public Disclosure Per 10CFR2.790 L-99-180, Forwards Responses to NRC RAI Questions for Chapter 3.8 of Ts.Proposed Revs to TS Previously Submitted with LAR Related to RAI1999-04-30030 April 1999 Forwards Responses to NRC RAI Questions for Chapter 3.8 of Ts.Proposed Revs to TS Previously Submitted with LAR Related to RAI ML20206F4321999-04-30030 April 1999 Forwards Responses to NRC RAI Questions for Chapter 3.8 of Ts.Proposed Revs to TS Previously Submitted with LAR Related to RAI 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 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-99-172, 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 L-99-012, Forwards 10CFR50.46 Annual Rept for 1998,re Effects of ECCS Evaluation Model Mod on Peak Cladding Temp Results Since 1997 Annual Rept & Most Recent PCT Error Rept Submitted 9809101999-03-19019 March 1999 Forwards 10CFR50.46 Annual Rept for 1998,re Effects of ECCS Evaluation Model Mod on Peak Cladding Temp Results Since 1997 Annual Rept & Most Recent PCT Error Rept Submitted 980910 ML20205A2871999-03-19019 March 1999 Forwards Rev 0 to W Rept WCAP-15171, Analysis of Capsule Z from Alabama Power Co Jm Farley Unit 2 Reactor Vessel Radiation Surveillance Program, Presenting Surveillance Capsule Test Results from Capsule Z L-99-125, Forwards Rev 0 to W Rept WCAP-15171, Analysis of Capsule Z from Alabama Power Co Jm Farley Unit 2 Reactor Vessel Radiation Surveillance Program, Presenting Surveillance Capsule Test Results from Capsule Z1999-03-19019 March 1999 Forwards Rev 0 to W Rept WCAP-15171, Analysis of Capsule Z from Alabama Power Co Jm Farley Unit 2 Reactor Vessel Radiation Surveillance Program, Presenting Surveillance Capsule Test Results from Capsule Z ML20205A1531999-03-19019 March 1999 Forwards Corrected Typed & marked-up Current TS Pages for Replacing Previous Pages Submitted on 990222,re CR, Penetration Room & Containment Purge Filtration Sys & Radiation Monitoring Instrumentation L-99-010, Forwards ISI Refueling 15,Interval 2,Period 3,Outage 3 for Jfnp,Unit 1, & Vols 1 & 2 to ISI Refueling 15,Interval 3, Period 1,Outage 1 for Jfnp,Unit 1. Summary of Results May Be Found in Tab B of Encl 21999-03-18018 March 1999 Forwards ISI Refueling 15,Interval 2,Period 3,Outage 3 for Jfnp,Unit 1, & Vols 1 & 2 to ISI Refueling 15,Interval 3, Period 1,Outage 1 for Jfnp,Unit 1. Summary of Results May Be Found in Tab B of Encl 2 ML20205A7611999-03-18018 March 1999 Forwards Annual DG Reliability Data Rept for 1998,per Plant TS 6.9.1.12 & 10CFR50.36.Rept Provides Number of Tests (Valid or Invalid) & Number of Failures for DGs at Jm Farley Nuclear Plant.Ltr Contains No New Commitments ML20205H2741999-03-18018 March 1999 Forwards Info on Status of Decommissioning Funding for Jm Farley Nuclear Plant,Units 1 & 2,IAW 10CFR50.75(f)(i) ML20204D4281999-03-16016 March 1999 Forwards SG-99-03-001, Farley Unit-1 1999 Voltage-Based Repair Criteria 90-Day Rept, Per GL 95-05.Required Rept for Fall 1998 SG Insp Is Included in Rept ML20204E5841999-03-15015 March 1999 Submits Info on Current Levels & Sources of Insurance on Jm Farley Nuclear Plant,Units 1 & 2 ML20207H6781999-03-0202 March 1999 Forwards Revised Epips,Including Rev 3 to FNP-0-EIP-8.1, Rev 95 to FNP-0-EIP-8.0,rev 1 to FNP-0-EIP-8.2 & Rev 4 to FNP-0-EIP-8.3 & Rev 32 to EP FNP-0-EP-0.0 ML20207D4261999-02-25025 February 1999 Forwards Relief Request RR-40 for Units 1 & 2 for SG Primary Nozzles Inside Radius.Util Requests NRC Approval of Proposed Relief Request by Mar 4,2000 to Support Unit 1 SG Replacement Outage in Spring of 2000 & Unit 2 SG in 2001 ML20207C2271999-02-22022 February 1999 Forwards Complete Replacement for 970630 Submittal Re TS Amend to Clarify SR Refs to ANSI N510 Sections 10,12 & 13 to ASME N510-1989,with Errata Dtd Jan 1991 & to Add Footnote Which Ref FNP FSAR for Relevant Testing Details ML20207B8591999-02-22022 February 1999 Forwards Proprietary 1998 Form NRC 5 Annual Rept, of Results of Individual Monitoring Carried Out at Plant.Rept Includes Record of Each Individual for Whom Monitoring Was Either Required by 10CFR20.1502 or Not.Encl Withheld 1999-09-16
[Table view] Category:UTILITY TO NRC
MONTHYEARML20064A7131990-09-17017 September 1990 Advises That Due to Reassignment,Jj Clark No Longer Needs to Maintain Senior Reactor Operator Licenses ML20059J2811990-09-14014 September 1990 Forwards List of Key Radiation Monitors Which Will Be Used as Inputs to Top Level Radioactivity Status Bar Re Spds.List Identifies Monitors Which Would Provide Concise & Meaningful Info About Radioactivity During Accidents ML20059J1661990-09-13013 September 1990 Forwards Monthly Operating Rept for Aug 1990 for Jm Farley Nuclear Plant & Rev 10 to ODCM ML20065D5961990-09-13013 September 1990 Responds to Violations Noted in Insp Repts 50-348/90-19 & 50-364/90-19.Response Withheld ML20065D6621990-09-12012 September 1990 Forwards NPDES Permit AL0024619 Effective 900901.Limits for Temp & Residual Chlorine Appealed & Stayed ML20059J2891990-09-12012 September 1990 Confirms Rescheduling of Response to Fitness for Duty Program Notice of Violation 90-18-02,per 900907 Telcon 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 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 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 ML20064A3431990-08-28028 August 1990 Forwards Corrected Insertion Instructions to Rev 8 to Updated FSAR for Jm Farley Nuclear Plant ML20059D4711990-08-22022 August 1990 Forwards Fitness for Duty Performance Data for Jan-June 1990 ML20059B5101990-08-22022 August 1990 Forwards Semiannual Radioactive Effluent Release Rept for Jan-June 1990.No Changes to Process Control Program for First Semiannual Period of 1990 Exists ML20056B2751990-08-20020 August 1990 Forwards Relief Requests from Second 10-yr Interval Inservice Testing Program for Class 1,2 & 3 Pumps & Valves. Request Incorporates Commitments in 891222 Response to Notice of Violation ML20056B2741990-08-20020 August 1990 Forwards Rev 2 to Unit Inservice Testing Program,For Review & Approval.Rev Incorporates Commitments Addressed in Util 891222 Response to Notice of Violation & Other Editorial & Technical Changes 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 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 ML20055G7701990-07-18018 July 1990 Updates 900713 Response to NRC Bulletin 90-001, Loss of Fill-Oil in Transmitters Mfg by Rosemount ML20055F7411990-07-11011 July 1990 Forwards Monthly Operating Rept for June 1990 & Corrected Monthly Operating Repts for Nov 1989 Through May 1990.Repts Revised to Correct Typo on Value of Cumulative Number of Hours Reactor Critical ML20055F3781990-07-10010 July 1990 Submits Final Response to Generic Ltr 83-28,Items 4.2.3 & 4.2.4.Util Position That Procedures Currently Utilized by Plant Constitute Acceptable Ongoing Life Testing Program for Reactor Trip Breakers & Components ML20055D4861990-07-0202 July 1990 Requests Authorization to Use Encl ASME Boiler & Pressure Vessel Code Case N-395 Re Laser Welding for Sleeving Process Described by Oct 1990,per 10CFR50.55a,footnote 6 ML20044A6191990-06-26026 June 1990 Suppls 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 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 ML20034C9971990-06-0101 June 1990 Advises That Implementation of Vital Area Door Alarms Previously Scheduled for Completion on 900608 Will Now Be Completed on 900716 Due to Other Plant Mods ML20043C1851990-05-29029 May 1990 Forwards Proposed Schedules for Submission & Requested Approval of Licensing Items ML20043B5941990-05-25025 May 1990 Provides Rept of Unsatisfactory Performance Testing,Per 10CFR26,App A.Error Caused by Olympus Analyzer Which Allowed Same Barcode to Be Assigned to Two Different Samples. Smithkline Taken Action to Prevent Recurrence of Scan Error ML20042G7461990-05-10010 May 1990 Certifies That Plant Licensed Operator Requalification Program Accredited & Based Upon Sys Approach to Training,Per Generic Ltr 87-07.Program in Effect Since 890109 ML20042F0831990-05-0101 May 1990 Forwards Rev 18 to Security Plan.Rev Withheld ML20042G3081990-04-25025 April 1990 Forwards Alabama Power Co Annual Rept 1989, Unaudited Financial Statements for Quarter Ending 900331 & Cash Flow Projections for 1990 ML20034B5791990-04-23023 April 1990 Responds to Violations Noted in Insp Rept 50-364/90-08. Corrective Actions:Deficiencies on Feedwater Flow Instruments Repaired & Power Range Nuclear Instruments Adjusted to Read within 2% of Actual Power ML20034A3001990-04-12012 April 1990 Responds to NRC 890803 Request for Addl Info Re Inservice Insp Program.Util Will Submit Rev 1 to Program Upon Receipt of Confirmation by NRC That Issues Satisfactorily Resolved ML20042E4121990-04-12012 April 1990 Provides Addl Info Re Review of Second 10-yr Inservice Insp Program,Per NRC 890803 Request.Relief Request RR-30 Requested Reduced Holding Time for Hydrostatically Testing Steam Generator Secondary Side ML20012E9571990-03-27027 March 1990 Forwards Annual Diesel Generator Reliability Data Rept,Per Tech Spec 6.9.1.12.Rept Provides Number of Tests (Valid or Invalid),Number of Failures for Each Diesel Generator at Plant for 1989 & Info Identified in Reg Guide 1.108 ML20012D9661990-03-22022 March 1990 Forwards Annual ECCS Evaluation Model Changes Rept,Per Revised 10CFR50.46.Info Includes Effect of ECCS Evaluation Model Mods on Peak Cladding Temp Results & Summary of Plant Change Safety Evaluations ML20012D8901990-03-20020 March 1990 Clarifies 891130 Response to Generic Ltr 83-28,Item 2.2.1 Re Use of Q-List at Plant,Per NRC Request.Fnpims Data Base Utilized as Aid for Procurement,Maint,Operations & Daily Planning ML20012C4701990-03-15015 March 1990 Responds to NRC 900201 Ltr Re Emergency Planning Weaknesses Identified in Insp Repts 50-348/89-32 & 50-364/89-21. Corrective Actions:Cited Procedures Revised.Direct Line Network Notification to State Agencies Being Implemented ML20012C6241990-03-14014 March 1990 Informs of Resolution of USI A-47,per Generic Ltr 89-19 ML20012C4651990-03-13013 March 1990 Provides Verification of Nuclear Insurance Reporting Requirements Specified in 10CFR50.54 w(2) ML20033F3501990-03-0808 March 1990 Emphasizes Belief That Info Contained in Util Re Station Blackout Responsive to 10CFR50.63 NL-90-1454, Forwards SPDS Critical Function Status Trees,Per G West Request During 900206 SPDS Audit at Plant.W/O Encl1990-03-0505 March 1990 Forwards SPDS Critical Function Status Trees,Per G West Request During 900206 SPDS Audit at Plant.W/O Encl ML20012A1621990-03-0202 March 1990 Forwards Addl Info Inadvertently Omitted from Jul-Dec 1989 Semiannual Radioactive Effluent Release Rept,Including Changes to Process Control Program ML20012A1301990-03-0101 March 1990 Responds to Generic Ltr 90-01 Re Request for Voluntary Participation in NRC Regulatory Impact Survey.Completed Questionnaire Encl ML20043A7481990-02-0202 February 1990 Forwards Util Exam Rept for Licensed Operator Requalification Written Exams on 900131 ML20006D2311990-01-31031 January 1990 Responds to NRC Bulletin 89-003 Re Potential Loss of Required Shutdown Margin During Refueling Operations. Refueling Procedures Will Be Revised to Incorporate Guidance That Will Preclude Inadvertent Loss of Shutdown ML20006A9091990-01-23023 January 1990 Forwards Response to Generic Ltr 89-13 Re Svc Water Sys Problems Affecting safety-related Equipment.Util Has Program to Perform Visual Insps & Cleanings of Plant Svc Water Intake Structure by Means of Scuba Divers ML20005E3681989-12-28028 December 1989 Responds to Violations Noted in Insp Repts 50-348/89-28 & 50-364/89-28 on 891002-06.Corrective Actions:All Piping Preparation for Inservice Insp Work in Containment Stopped & All Participants Assembled to Gather Facts on Incident ML20005E4931989-12-28028 December 1989 Provides Certification That fitness-for-duty Program Meets 10CFR26 Requirements.Testing Panel & cut-off Levels in Program Listed in Encl ML20005E1971989-12-27027 December 1989 Responds to Violations Noted in Insp Repts 50-348/89-22 & 50-364/89-22 on 890911-1010.Corrective Actions:Steam Generator Atmospheric Relief Valve Closed & Core Operations Suspended.Shift Supervisor Involved in Event Counseled ML20011D5041989-12-22022 December 1989 Responds to Violations Noted in Insp Repts 50-348/89-26 & 50-364/89-26.Corrective Actions:Personnel Involved in Preparation of Inservice Test Procedures Counseled. Violation B Re Opening of Pressurizer PORV Denied ML19332F1241989-12-0707 December 1989 Forwards Response to NRC 890803 Request for Addl Info Re Review of Second 10-yr Inservice Insp Program,Per 1990-09-17
[Table view] |
Text
f Miihng Address 9
Af arwna Power Company 600 thrth 18th Street Po',t Of fice [lon 2041
[hrmingham, Alabama 35291 1
T e le pt'one 205 783 0081 F. L. Clayton, Jr.
b
?""QCC"'
Alabama Power n-wun, u ae w>.m April 5, 1982 Docket No. 50-364 g,j
\\r Director, Nuclear Reactor Regulation
.8 U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Attention:
Mr. S. A. Varga Joseph M. Farley Nuclear Plant - Unit 2
~
Spent Fuel Pool Modification Amendment 3 Gentlemen:
In response to your letter of March 4,1982, enclosed is Amendment No. 3 to Alabama Power Company's Spent Fuel Pool Modification submitted on December 18, 1981 and amended in February 1982 and March 1982.
This amendment responds to the questions identified in Enclosures 2, 3 and 4 of your March 4,1982 letter.
In addition, page III-7 was revised to incorporate the response to Question #5 of Enclosure 4.
If you have any questions, please advise.
Yours very truly, j-. 3
'qj/ /p/g, Aw g,
O L
3 F. L. Clayton,<r.
FLCJ r/ J AR :j c -D7 Enclosure cc:
Mr. R. A. Thomas SWORNTOANDjUBSCRIBE EF.0 E Mr. G. F. Trowbridge ME THIS /A.
DAY OF Mr. J. P. O'Reilly 1982.
Mr. E. A. Reeves Mr. W. H. Bradford
. /v W
/ /
Notafy Public My commission expires:
0
/O-22-75 8204210437 820405 DR ADOCK 05000364 PDR J
r-e e
FARLEY NUCLEAR PLANT UNIT 2 SPENT FUEL POOL MODIFICATION AMENDMENT 3 REVISION INSERTION INSTRUCTIONS April 5, 1982 Section Page Instruction' III III-7 Replace Tab Response to Add NRC Questions Q
Q-ll Add Q
Q-12 Add Q
Q-13 Add Q
Q-14 Add Q
Q-15 Add Q
Q-16 Add Q
Q-17 Add Q
Q-18 Add Q
Q-19 Add t
l l
to the spent fuel pool 10 days after the emergency shutdown.
3.
105'E component cooling water temperature.
Results:
No. of Total Spent Fuel Cooling Trains Heat Load Pool Bulk Operation (los Stu /hr)
Temeerature (*F) 1 30.384 158 2
30.384 131 The computer code BPCOL is used to analyse the natural circulation cooling of the spent fuel in the event of a loss of all external means of cooling for the spent fuel pool.
BPOOL is a proprietary program of NAI.
The code is based on the assumption that boiling takes place near the top of the fuel channel.
BPOOL evaluates the saturation properties of the coolant on the basis of the static pressure at the top of the storage racks.
These properties include water density, temperature, and steam density.-
The steam is assumed to separate and flow out of the pool.
The water at the saturation temperature corresponding to the pressure at the top of the racks flows downward to the inlet of the storage racks.
The static pressure at this location is higher than the pressure at the top of the storage racks and as a result the fluid is subcooled as it enters the fuel assembly.
The fluid becomes less dense as it passes up the fuel channel.
Near the top of the fuel channel the fluid reaches saturation conditions and net boiling occurs.
The computer cock, BPOOL, assumes a loss of all external means of cooling, but it should be noted that the Earley spent fuel pool cooling system is redundant and single failure-proof.
Under normal conditions, voiding between fuel assemblies is highly unlikely because these spaces are not sealed to keep out water.
Holes are provided at the top and bottom of each inner can to permit a definitive flowpath for circulation of water in these spaces.
III.l.5.(4)
Potential Fuel and Rack Handling Accidents The high-density poison racks are of a free-standing design, utilizing bottom support pads, resting on the floor of the spent fuel pool.
The installation of the high-density racks I
will include removal of the existing 13-in. center storage racks.
The high-density racks will be installed wet since there is spent fuel in the storage pool.
l The following is a sequence of events for installing the high-density poison racks.
Phase I Install and test a temporary crane for handling the existing racks and the high-density racks.
The AMEND 3 4/82 l
l III-7 l
ENCLOSURE-2 OF NRC LETTER DATED MARCH 4, 1982 NRC Question 1:
In section III.I. 2 (2) of your fuel pool modification report you state that the size of the load that can be handled over the spent fuel pool when fuel is in the pool is limited to 3,000 lb by Farley Unit 2 Technical Specifications.
In Section IV. (1).b of the report you state that the spent fuel racks are designed to withstand a fuel bundle drop from 42 in. under various conditions and a 9-in. gate drop.
Provide the following information to assure your Technical Specifications are sufficient to cover all cases since your Technical Specifications do not impose a height restriction.
a.
Verify that the load drops identified in Section IV. (1).b of your report do not have a higher possible kinetic energy than that assumed in your accident analyses used in Meveloping the Technical Specifications.
b.
Verify that when considering a load drop the weight of the handling fixtures was used in your analyses.
c.
For loads lighter than one fuel assembly verify that their lifting height will not result in a higher kinetic energy if dropped than the maximum used in your accident analyses.
Also specify how the height of lighter loads will be limited to an acceptable elevation.
APC Response:
a.
The Parley Unit 2 Technical Specification, Section 3.9.7.1, prohibits loads in excess of 3,000 lb from travel over fuel assemblies in the storage pools.
This load bounds the spent fuel assembly with control rod and handling fixture (1992 lb).
The fuel assembly drop is analyzed at a drop height of 42 in., which bounds the actual drop of 39.5 in.
The drop height is limited by a high level limit switch on the lift in combination with a long handling tool.
Administrative controls prohibit a gate (3600 lb) from being carried over spent fuel.
However, the racks are designed to withstand a gate drop from 9 in., which bounds the actual drop of 6 in.
The drop height is limited by a physical limitation in lifting capability.
The fuel assembly drop assuming a load of 3,000 lb at 42 in.
is the worst impact load condition, but is very conservative considering an actual drop of 1992 lb at 39.5 in.
Therefore, the load drops identified in Section IV. (1) b of the report do not have a higher kinetic energy than that assumed in the accident analysis.
Q-11 AMEND 3 4/82 I
b.
The loads associated with a fuel assembly with control rod (1650 lb) and handling fixture (376 lb) are bounded by the 3000 lb load specification, which was used in the drop accident analysis.
The handling fixture is not used with the 3600-lb gate.
The gate is designed with a clevis at top center, which the crane hook fits into for lifting and' relocating.
c.
As discussed in response to Question 1(a), the analyzed fuel assembly drop load of 3000 lb at 42 in. represents the worst fuel rack impact load condition; i.e.,
highest kinetic energy; of all loads that could be moved over the spent fuel pool.
The kinetic energy of loads lighter than one fuel assembly will be less than that calculated for the analyzed fuel assembly drop load descr3 bed above even if the lighter loads are released from the maximum possible elevation.
Therefore, limitations c:. lift height of these lighter loads are not required.
G AMEND 3 4/82 Q-12
ENCLOSURE-3 OF NRC LETTER DATED MARCH 4, 1982 NRC Question 281-1:
The February 1, 1982 Amendment request does not indicate any proposed modification of the spent fuel pool cooling and cleanup system (SFPCCS) in conjunction with the installation of high density poison spent fuel storage racks.
Describe what changes, if any, will be made to the SFPCCS to maintain the level of pool water purity with respect to visual clarity and activated corrosion and fission product buildup the same as for the original spent fuel storage capacity.
Assume that the number of defective fuel assemblies increases in proportion to the increased spent fuel storage capacity.
If no changes to the SFPCCS are to be made, indicate how the same level of pool water purity will be maintained.
APC Response:
No modifications to the spent fuel pool cooling and cleanup system will be made as a result of the installation of high density poison spent fuel storage racks.
The Farley spent fuel pool cleanup system is described in FSAR subsection 9.1.3.
The system's demineralizers and filters are designed to provide adequate purification to permit unrestricted access for plant personnel to the spent fuel storage area and maintain optical clarity of the spent fuel pool water.
The optical clarity of the spent fuel pool water surface is maintained by use of the system's skimmers, strainer, and skimmer filter.
Operating plant experience has indicated that crud release occurs immediately after refueling with the fuel pool water returning to normal conditions within a few days.
The increase in spent fuel assembly capacity of the spent fuel pool will not affect the optical clarity of the spent fuel pool water.
NRC Question 281-2:
Describe the samples and instrument readings and their frequency of measurement that will be performed to monitor the spent fuel pool water purity and need for demineralizer resin and filter replacement.
State the chemical and radiochemical limits to be used in monitoring the spent fuel pool water and initiating corrective action.
Provide the basis for establishing these limits.
Your response should consider variables such as:
boron, gross gamma and iodine activity, demineralizer and or filter differential pressure, demineralizer decontamination factor pH, and crud level.
APC Response:
A maximum radiation level and minimum decontamination factor (DF) will be established by procedure for the spent fuel pool (SFP) demineralizer.
Accordingly the demineralizer resins may be ex-changed at a more frequent interval than would be required for the original rack configuration.
The SFP filters may also be exchanged Q-13 AMEND 3 4/82
at a more frequent rate according to aP limitations in table Q-2.
The SFPCCS monitoring parameters, sampling frequency, limits to be maintained and their bases are shown in table Q-2.
I i
1 I
I i
i 4
Q-14 AMEND 3 4/82
TABLE Q-2 (SHEET 1 OF 2)
PURITY MONITORING A.
Samples Parameter Frequency Limit Basis Boron Weekly 2000-3000 ppm The limitations on re-activity conditions during refueling ensure that:
- 1. The reactor will remain suberitical during core alternations, and 2.
a uniform boron concentration is main-tained for reactivity control in the water volume having direct access to the reactor vessel.
These limi-tations are consistent with the initial con-ditions assumed for the boron dilution incident in the accident analyses.
pH Weekly 4.0-4.9 To be consistent with the boron concentration.
Cl-Weekly 50.15 ppm Consistent with the acceptable limit for stainless steel systems.
~
F Weekly 50.15 ppm Consistent with the acceptable limit for stainless steel systems.
Demineralizer DF Monthly ")
I}
I 210 Iodine specific DF to ensure the removal of 99 percent of the assumed 10-percent iodine gap activity released from the rupture of an irradi-ated fuel assembly.
B.
Instrument Parameter Frequency Limit Basis 1
Area Monitor RE-5 Continuous 15 mrem /hr N/A i
Filter AP Once per 20 psid Manufacturer's 8 hr recommendation.
Q-15 AMEND 3 4/82
TABLE Q-2 (SHEET 2 OF 2)
(a)
This will be done daily during forced oxygenation clean up prior to refueling.
(b)
If iodine inventory in the spent fuel pool is low; i.e.,
I-131
-4 s 1 x 10 pCi/g'; the demineralizer may be allowed to stay in service longer.
This may be at the discretion of the C&HP supervisor or C&HP section supervisor.
4 I
i I
l l
l Q-16 AMEND 3 4/82
ENCLOSURE-4 OF NRC LETTER DATED MARCH 4, 1982 NRC Question 1:
What is the calculated nominal effective multiplication factor for the high density storage racks in their design configuration?
APC Response:
The nominal reactivity of the new Farley high density spent fuel pool (HDSFP) is determined using the Monte Carlo Code KENO-IV to be 0.9217*0.0044 with the 95-percent confidence interval ranging from 0.9127 to 0.9305.
The results have been adjusted to include the effects due to the presence of U-234 and Inconel spacer grids.
NRC Question 2:
What are the values of the calculated bias, calculational uncertainty, and mechanical uncertainty which are applied to the nominal effective multiplication factor?
APC Response:
The following summarizes the reactivity inventory of the Farley HDSFP.
Nominal K 0.9217*0.0044 eff Calculational bias
+0.0027
' Mechanical uncertainty (worst case geometry) +0.0159 Final K 0.9403*0.0044 eff c>
95-percent confidence interval 0.9315-0.9491 NRC Question 3:
What value of water temperature yields the maximum reactivity?
What value was used in the calculations?
APC Response:
A temperature reactivity sensitivity study covering the range from 68 F to 212 F shows that the effective multiplication factog has the highest value at the lowest reasonable pool temperature (68 F).
The nominal effective multipfication factor reported in response to Question 1 is for the 68 F condition.
Q-17 AMEND 3 4/82
NRC Question 4:
What organization performed the criticality calculations for the high density storage racks?
Has the organization benchmarked the versions of the KENO-IV and PDQ-7 codes which are operating on their computer system against critical experiments or does the reference to benchmarking refer to some other organization's efforts on a different computer system?
APC Response:
The criticality calculations are performed by Nuclear Associates International (NAI), a consulting service of the Control Data Corporation.
Both the KENO-IV and PDQ-7 code packages have been installed and in use on the CDC CYBER systems for a number of years.
The AMPX/ KENO-IV model, which is the basic code package used in the calculation of the effective multiplication factor of the infinite rack array, has been benchmarked by NAI against earlier ORNL critical experiments measured by Dr.
E.
Johnson, at La Crosse BWR measured cold critical;'and the more recent critical experiments conducted by S.
R.
Bierman, et al, at the Batelle Pacific Northwest Laboratory as reported in PNL-2438.
Although the PDQ-7 model is mainly used to provide relative reactivity results for the several sensitivity studies conducted for the Farley analysis, it is also used to calculate the infinite array reactivity of the nominal configuration in part to cross check and verify the results of the Monte Carlo KENO-IV calculations mentioned above.
The PDQ-7 model has been in use at NAI for core physics and fuel management work since the late 1960's when the code was first released for industrial use.
This model has been used to calculate cold critical measured at a dozen pressurized water reactor and boiling water reactor power plants in the United States.
Furthermore, it has also been used in calculating the La Crosse BWR cold critical and the ORNL/ Johnson critical experiments.
NRC Question 5:
The thermal hydraulic analysis of the cooling of the spent fuel states that voiding between fuel assenblies is not possible because these spaces contain poison plates.
This is not obvious from figure II-4.
1s there no coolant between fuel assemblies?
APC Response:
Under normal conditions, voiding between fuel assemblies is highly unlikely because these spaces are not sealed to keep out water.
Holes are provided at the top and bottom of each inner can to permit a definitive flowpath for circulation of water in these spaces.
AMEND 3 4/82 Q-18
O Voiding in the space between fuel channels can occur only when there is boiling in the fuel cavity itself, as might occur if all external means of cooling were lost.
Under these conditions, the voids in the volume between the fuel channels are always less than those in the fuel cavities and would not increase the reactivity of the
~
storage arrangement.
When the fuel cavities have no voids, the temperature is low enough to preclude voiding in the space between fuel channels.
6 i
Q-19 AMEND 3 4/82