Similar Documents at Ginna |
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Category:CORRESPONDENCE-LETTERS
MONTHYEARML17250B3041999-10-20020 October 1999 Forwards Changes to Tech Specs Bases for Re Ginna Nuclear Power Plant.Change Bars Indicate Those Revs Which Have Been Incorporated IR 05000244/19990081999-10-14014 October 1999 Forwards Insp Rept 50-244/99-08 on 990809-0919.Severity Level IV Violation of NRC Requirements Occurred & Being Treated as non-cited Violation Consistent with App C of Enforcement Policy ML17265A7651999-10-0808 October 1999 Forwards Fifteen Relief Requests That Will Be Utilized for Ginna NPP Fourth Interval ISI Program That Will Start on Jan 1,2000.Attachment 1 Includes Summaries & Detailed Description of Each Relief Request ML17265A7631999-10-0505 October 1999 Requests Approval for Use of Relief Request Number 43 to Address Volumetric Examination Limitations (Less than 90%) Associated with a & B RHR Heat Exchanger Outlet Nozzle to Shell Welds.Approval Is Requested by Dec 31,2000 ML17265A7641999-10-0505 October 1999 Requests Approval for Use of Relief Request Number 42 to Address Volumetric Examinations Limitations (Less than 90%) Associated with Eight Class 1 Identified Welds or Areas of Reactor Pressure Vessel ML20212J3561999-09-30030 September 1999 Forwards Four Copies of Re Ginna NPP Training & Qualification Plan for Security Officers, Rev 7,dtd 990930. Synopsis of Changes,Encl.Encl Withheld Per 10CFR73.21 ML20212J3801999-09-30030 September 1999 Forwards Four Copies of Rev to Re Ginna NPP Security Plan.Rev Changes Contingency Weapons Available to Response Force to Those Most Effective in Current Defensive Strategy.Encl Withheld Per 10CFR73.21 IR 05000244/19990051999-09-24024 September 1999 Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs ML17265A7571999-09-24024 September 1999 Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs IR 05000244/19992011999-09-24024 September 1999 Forwards Insp Rept 50-244/99-201 (Operational Safeguards Response Evaluation) on 990621-24.No Violations Noted. Primary Purpose of Osre to Assess Licensee Ability to Respond to External Threat.Insp Rept Withheld ML17265A7461999-08-31031 August 1999 Submits Response to NRC Administrative Ltr 95-03,rev 2, Availability of Reactor Vessel Integrity Database,Version, Dtd 990726 ML17265A7401999-08-26026 August 1999 Requests Approval for Use of Relief Request Number 35 Re Use of ASME Section XI Code,1995 Edition,1996 Addenda.Code Will Be Used to Develop Plant Fourth 10-year Interval ISI Program on Class 1,2 & 3 Components ML17265A7411999-08-26026 August 1999 Forwards LER 99-004-01 Re Plant Being Outside Design Basis Due to Containment Recirculation Fan Moisture Separator Vanes Being Incorrectly Installed.Part 21 Notification of 990512 Is Being Rescinded ML17265A7451999-08-23023 August 1999 Forwards fitness-for-duty Performance Data Rept for Six Months Ending 990630,per 10CFR26.71(d) ML17250B3021999-08-23023 August 1999 Informs That Util & NRC Had Conference Call on 990816 to Review Approach in Responding to Questions,As Result of Questions Re Main Steam Check Valve Performance Included in Insp Rept 50-244/99-05,dtd 990806 ML17265A7271999-07-30030 July 1999 Forwards 10CFR21 Interim Rept Per Reporting of Defects & Noncompliance,Section 21 (a) (2).Interim Rept Prepared Because Evaluation Cannot Be Completed within 60 Days from Discovery of Deviation or Failure to Comply ML17265A7151999-07-23023 July 1999 Forwards LER 99-007-01 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Reactor Trip.Further Investigation of Event Identified Addl Corrective Actions ML17265A7061999-07-22022 July 1999 Forwards LER 98-003-02,re Actuations of CR Emergency Air Treatment Sys.All Creats Actuations,Including Those Originally Believed to Be Valid Actuations,Were,In Fact Invalid Actuations ML17265A7191999-07-21021 July 1999 Forwards Ginna Station ISI Rept for Refueling Outage Conducted in 1999 ML17265A7141999-07-21021 July 1999 Withdraws Relief Request 35 for Plant Inservice Insp Program Section XI Requirements,Submitted on 980806.Licensee Plans to Resubmit Relief Request,Which Includes Addl Level of Detail,In Near Future ML17265A7041999-07-16016 July 1999 Submits Info Re Specific Licensing Actions Which May Be Expected to Generate Complex Reviews,In Response to Administrative Ltr 99-02,dtd 990603 ML17265A6911999-06-30030 June 1999 Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Gl 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Readiness Disclosure Attached ML17265A6871999-06-22022 June 1999 Forwards Response to RAI Made During 990225 Telcon Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves. Calculation Encl. Encl ML17265A6841999-06-21021 June 1999 Informs That Util Wishes to Amend Extend of Alternate Exams Provided for Relief Request Re ISI Program ASME Section XI Require Exams for First 10-Yr Interval for Containment ML17265A6741999-06-15015 June 1999 Submits Annual ECCS Rept IAW 10CFR50.46(a)(3)(ii) Requirements.No Changes Have Been Made to Large Break LOCA PCT & Small Break LOCA PCT ML17265A6751999-06-11011 June 1999 Responds to NRC RAI Re Licensee GL 96-05 Program.Encl Info Verifies That Util Is Implementing Provisions of JOG Program on MOV Periodic Verification ML17309A6551999-06-0707 June 1999 Responds to NRC 990310 RAI Re Verification of Seismic Adequacy of Mechanical & Electrical Equipment ML17265A6671999-06-0101 June 1999 Requests Approval of Ginna QA Program for Radioactive Matl Packages,Form 311,approval Number 0019.Ginna QA Program for Station Operation, Was Most Recently Submitted to NRC by Ltr Dtd 981221 & Supplemented on 990301 ML17265A6641999-05-25025 May 1999 Forwards Addl Info on Use of GIP Method a for Re Ginna Nuclear Power Plant.Copy of Re Ginna Station USI A-46 Outlier Resolution Table as Requested ML17265A6611999-05-24024 May 1999 Forwards LER 99-007-00 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Rt.Util Is Planning to Submit Suppl to LER by 990730 ML20206U0921999-05-13013 May 1999 Forwards Four Copies of Rev R to Gnpp Security Plan,Per Provisions of 10CFR50.54(p).Rev Clarifies Armed Response Team Assignments & Does Not Degrade Physical Security Effectiveness.Rev Withheld,Per 10CFR73.21 ML17265A6461999-05-12012 May 1999 Forwards 1998 Annual Radioactive Effluent Release Rept & 1998 Annual Radiological Environ Operating Rept, for Re Ginna NPP ML17265A6411999-05-12012 May 1999 Forwards LER 99-004-00 IAW 10CFR50.73 & 10CFR21.Further Assessment Will Be Provided in Suppl to LER by 990630 ML20206E7221999-04-29029 April 1999 Forwards Four Copies of Rev Q to Re Ginna Nuclear Power Plant Security Plan,Per 10CFR50.54(p).Changes Do Not Degrade Physical Security Effectiveness.Encl Withheld,Per 10CFR73.21 ML20206H6911999-04-22022 April 1999 Forwards Info Requested During Informal Telcon on 980408 Concerning Upcoming Osre at Ginna Station.Info Requested Listed.Without Encls ML17265A6271999-04-19019 April 1999 Forwards Rev 0 & Rev 1 to Colr,Cycle 28 for Re Ginna NPP, Per TS 5.6.5 ML17309A6501999-04-14014 April 1999 Forwards Revised Ginna Station EOPs & Procedures Index ML17265A6121999-03-29029 March 1999 Forwards Rept on Status of Decommissioning Funding for Re Ginna Npp,For Which Rg&E Is Sole Owner,Per 10CFR50.75. Data Presented Herein,Current as of 981231 ML17265A6041999-03-24024 March 1999 Forwards LER 99-001-00,IAW 10CFR50.73(a)(2)(ii)(B) & 10CFR21.Addl Analyses Are Being Performed to Support Future Cycle Operation & Supplemental LER Is Scheduled to Be Submitted by 990618 ML17265A5671999-03-0101 March 1999 Forwards Application for Amend to License DPR-18,to Revise TSs Battery Cell Parameters Limit for Specific Gravity (SR 3.8.6.3 & SR 3.8.6.6).Supporting Tss,Encl ML17265A5641999-03-0101 March 1999 Forwards Response to NRC 990217 RAI Concerning Changes to QA Program for Re Ginna Station Operation.Rg&E Is Modifying Changes Requested in 981221 Submittal.Modified QA Program,Encl ML17265A5551999-02-25025 February 1999 Informs That Util Is in Process of Revising fitness-for-duty Program,Developed in Accordance with 10CFR26.Util Will Continue to Use Dept of Health & Human Svcs Certified Test Facility for Majority of Tests During Yr ML17265A5561999-02-22022 February 1999 Forwards FFD Performance Data Rept for Six Months Ending 981231,per 10CFR26.71(d) ML17265A5451999-02-12012 February 1999 Forwards Simulator Four Year Certification Rept,Per 10CFR55.45(b)(5)(ii) ML17309A6491999-02-12012 February 1999 Forwards Ginna Station EOPs ML17265A5431999-02-0909 February 1999 Supplements 980806 Relief Request with Attached Table.Util Third 10-Yr ISI of Reactor Vessel Being Performed During 1999 Refueling Outage,Beginning on 990301 ML17265A5361999-02-0202 February 1999 Forwards Response to NRC 981203 RAI Re Resolution of Unresolved Safety Issue USI A-46.Util Does Not Agree with NRCs Interpretation.Detailed Bases,Encl ML17311A0691999-01-25025 January 1999 Forwards Revs to Ginna Station Emergency Plan Implementing Procedures (Epips).Previous Rev Had Incorrect Effective Date ML17311A0671999-01-14014 January 1999 Forwards Revised Emergency Operating Procedures for Re Ginna NPP ML17265A5131999-01-12012 January 1999 Forwards Revised Cover Page for Ginna Station Technical Requirements Manual (Trm), Rev 7,correcting Error in Effective Date 1999-09-30
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML17250B3041999-10-20020 October 1999 Forwards Changes to Tech Specs Bases for Re Ginna Nuclear Power Plant.Change Bars Indicate Those Revs Which Have Been Incorporated ML17265A7651999-10-0808 October 1999 Forwards Fifteen Relief Requests That Will Be Utilized for Ginna NPP Fourth Interval ISI Program That Will Start on Jan 1,2000.Attachment 1 Includes Summaries & Detailed Description of Each Relief Request ML17265A7631999-10-0505 October 1999 Requests Approval for Use of Relief Request Number 43 to Address Volumetric Examination Limitations (Less than 90%) Associated with a & B RHR Heat Exchanger Outlet Nozzle to Shell Welds.Approval Is Requested by Dec 31,2000 ML17265A7641999-10-0505 October 1999 Requests Approval for Use of Relief Request Number 42 to Address Volumetric Examinations Limitations (Less than 90%) Associated with Eight Class 1 Identified Welds or Areas of Reactor Pressure Vessel ML20212J3561999-09-30030 September 1999 Forwards Four Copies of Re Ginna NPP Training & Qualification Plan for Security Officers, Rev 7,dtd 990930. Synopsis of Changes,Encl.Encl Withheld Per 10CFR73.21 ML20212J3801999-09-30030 September 1999 Forwards Four Copies of Rev to Re Ginna NPP Security Plan.Rev Changes Contingency Weapons Available to Response Force to Those Most Effective in Current Defensive Strategy.Encl Withheld Per 10CFR73.21 IR 05000244/19990051999-09-24024 September 1999 Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs ML17265A7571999-09-24024 September 1999 Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs ML17265A7461999-08-31031 August 1999 Submits Response to NRC Administrative Ltr 95-03,rev 2, Availability of Reactor Vessel Integrity Database,Version, Dtd 990726 ML17265A7401999-08-26026 August 1999 Requests Approval for Use of Relief Request Number 35 Re Use of ASME Section XI Code,1995 Edition,1996 Addenda.Code Will Be Used to Develop Plant Fourth 10-year Interval ISI Program on Class 1,2 & 3 Components ML17265A7411999-08-26026 August 1999 Forwards LER 99-004-01 Re Plant Being Outside Design Basis Due to Containment Recirculation Fan Moisture Separator Vanes Being Incorrectly Installed.Part 21 Notification of 990512 Is Being Rescinded ML17250B3021999-08-23023 August 1999 Informs That Util & NRC Had Conference Call on 990816 to Review Approach in Responding to Questions,As Result of Questions Re Main Steam Check Valve Performance Included in Insp Rept 50-244/99-05,dtd 990806 ML17265A7451999-08-23023 August 1999 Forwards fitness-for-duty Performance Data Rept for Six Months Ending 990630,per 10CFR26.71(d) ML17265A7271999-07-30030 July 1999 Forwards 10CFR21 Interim Rept Per Reporting of Defects & Noncompliance,Section 21 (a) (2).Interim Rept Prepared Because Evaluation Cannot Be Completed within 60 Days from Discovery of Deviation or Failure to Comply ML17265A7151999-07-23023 July 1999 Forwards LER 99-007-01 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Reactor Trip.Further Investigation of Event Identified Addl Corrective Actions ML17265A7061999-07-22022 July 1999 Forwards LER 98-003-02,re Actuations of CR Emergency Air Treatment Sys.All Creats Actuations,Including Those Originally Believed to Be Valid Actuations,Were,In Fact Invalid Actuations ML17265A7191999-07-21021 July 1999 Forwards Ginna Station ISI Rept for Refueling Outage Conducted in 1999 ML17265A7141999-07-21021 July 1999 Withdraws Relief Request 35 for Plant Inservice Insp Program Section XI Requirements,Submitted on 980806.Licensee Plans to Resubmit Relief Request,Which Includes Addl Level of Detail,In Near Future ML17265A7041999-07-16016 July 1999 Submits Info Re Specific Licensing Actions Which May Be Expected to Generate Complex Reviews,In Response to Administrative Ltr 99-02,dtd 990603 ML17265A6911999-06-30030 June 1999 Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Gl 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Readiness Disclosure Attached ML17265A6871999-06-22022 June 1999 Forwards Response to RAI Made During 990225 Telcon Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves. Calculation Encl. Encl ML17265A6841999-06-21021 June 1999 Informs That Util Wishes to Amend Extend of Alternate Exams Provided for Relief Request Re ISI Program ASME Section XI Require Exams for First 10-Yr Interval for Containment ML17265A6741999-06-15015 June 1999 Submits Annual ECCS Rept IAW 10CFR50.46(a)(3)(ii) Requirements.No Changes Have Been Made to Large Break LOCA PCT & Small Break LOCA PCT ML17265A6751999-06-11011 June 1999 Responds to NRC RAI Re Licensee GL 96-05 Program.Encl Info Verifies That Util Is Implementing Provisions of JOG Program on MOV Periodic Verification ML17309A6551999-06-0707 June 1999 Responds to NRC 990310 RAI Re Verification of Seismic Adequacy of Mechanical & Electrical Equipment ML17265A6671999-06-0101 June 1999 Requests Approval of Ginna QA Program for Radioactive Matl Packages,Form 311,approval Number 0019.Ginna QA Program for Station Operation, Was Most Recently Submitted to NRC by Ltr Dtd 981221 & Supplemented on 990301 ML17265A6641999-05-25025 May 1999 Forwards Addl Info on Use of GIP Method a for Re Ginna Nuclear Power Plant.Copy of Re Ginna Station USI A-46 Outlier Resolution Table as Requested ML17265A6611999-05-24024 May 1999 Forwards LER 99-007-00 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Rt.Util Is Planning to Submit Suppl to LER by 990730 ML20206U0921999-05-13013 May 1999 Forwards Four Copies of Rev R to Gnpp Security Plan,Per Provisions of 10CFR50.54(p).Rev Clarifies Armed Response Team Assignments & Does Not Degrade Physical Security Effectiveness.Rev Withheld,Per 10CFR73.21 ML17265A6461999-05-12012 May 1999 Forwards 1998 Annual Radioactive Effluent Release Rept & 1998 Annual Radiological Environ Operating Rept, for Re Ginna NPP ML17265A6411999-05-12012 May 1999 Forwards LER 99-004-00 IAW 10CFR50.73 & 10CFR21.Further Assessment Will Be Provided in Suppl to LER by 990630 ML20206E7221999-04-29029 April 1999 Forwards Four Copies of Rev Q to Re Ginna Nuclear Power Plant Security Plan,Per 10CFR50.54(p).Changes Do Not Degrade Physical Security Effectiveness.Encl Withheld,Per 10CFR73.21 ML20206H6911999-04-22022 April 1999 Forwards Info Requested During Informal Telcon on 980408 Concerning Upcoming Osre at Ginna Station.Info Requested Listed.Without Encls ML17265A6271999-04-19019 April 1999 Forwards Rev 0 & Rev 1 to Colr,Cycle 28 for Re Ginna NPP, Per TS 5.6.5 ML17309A6501999-04-14014 April 1999 Forwards Revised Ginna Station EOPs & Procedures Index ML17265A6121999-03-29029 March 1999 Forwards Rept on Status of Decommissioning Funding for Re Ginna Npp,For Which Rg&E Is Sole Owner,Per 10CFR50.75. Data Presented Herein,Current as of 981231 ML17265A6041999-03-24024 March 1999 Forwards LER 99-001-00,IAW 10CFR50.73(a)(2)(ii)(B) & 10CFR21.Addl Analyses Are Being Performed to Support Future Cycle Operation & Supplemental LER Is Scheduled to Be Submitted by 990618 ML17265A5641999-03-0101 March 1999 Forwards Response to NRC 990217 RAI Concerning Changes to QA Program for Re Ginna Station Operation.Rg&E Is Modifying Changes Requested in 981221 Submittal.Modified QA Program,Encl ML17265A5671999-03-0101 March 1999 Forwards Application for Amend to License DPR-18,to Revise TSs Battery Cell Parameters Limit for Specific Gravity (SR 3.8.6.3 & SR 3.8.6.6).Supporting Tss,Encl ML17265A5551999-02-25025 February 1999 Informs That Util Is in Process of Revising fitness-for-duty Program,Developed in Accordance with 10CFR26.Util Will Continue to Use Dept of Health & Human Svcs Certified Test Facility for Majority of Tests During Yr ML17265A5561999-02-22022 February 1999 Forwards FFD Performance Data Rept for Six Months Ending 981231,per 10CFR26.71(d) ML17265A5451999-02-12012 February 1999 Forwards Simulator Four Year Certification Rept,Per 10CFR55.45(b)(5)(ii) ML17309A6491999-02-12012 February 1999 Forwards Ginna Station EOPs ML17265A5431999-02-0909 February 1999 Supplements 980806 Relief Request with Attached Table.Util Third 10-Yr ISI of Reactor Vessel Being Performed During 1999 Refueling Outage,Beginning on 990301 ML17265A5361999-02-0202 February 1999 Forwards Response to NRC 981203 RAI Re Resolution of Unresolved Safety Issue USI A-46.Util Does Not Agree with NRCs Interpretation.Detailed Bases,Encl ML17311A0691999-01-25025 January 1999 Forwards Revs to Ginna Station Emergency Plan Implementing Procedures (Epips).Previous Rev Had Incorrect Effective Date ML17311A0671999-01-14014 January 1999 Forwards Revised Emergency Operating Procedures for Re Ginna NPP ML17265A5131999-01-12012 January 1999 Forwards Revised Cover Page for Ginna Station Technical Requirements Manual (Trm), Rev 7,correcting Error in Effective Date ML17265A5111999-01-11011 January 1999 Requests Relief Per 10CFR50.55a(a)(3)(ii) from Certain Requirements of Section XI of ASME Bp&V Code for ISI Program.Relief Requests 37,38 & 39 Encl ML17265A5101999-01-11011 January 1999 Requests Relief Per to 10CFR50.55a(a)(3)(ii) from Certain Requirements of Section XI of ASME B&PV Code for ISI Program.Relief Request 40 Encl 1999-09-30
[Table view] Category:UTILITY TO NRC
MONTHYEARML20059L4481990-09-18018 September 1990 Forwards Responses to Rer 900709-13 Team Visit Findings, Per .Responses Withheld (Ref 10CFR73.21) ML17309A4491990-09-13013 September 1990 Forwards Slides Presented by Facility & Westinghouse to NRC in 900724 Meeting.Encl Withheld ML17262A1331990-09-11011 September 1990 Responds to Violations & Several Unresolved Items Noted in SSFI Rept 50-244/89-81.Update of Appropriate Unresolved Items Encl.Specific Actions Re All NRC Unresolved Items Being Tracked to Completion ML17261B1511990-08-29029 August 1990 Forwards Semiannual Radioactive Effluent Release Rept Jan- June 1990 & Rev 3 to Process Control Program for Ginna Station, Per Tech Specs 6.9.1.4 & 6.16,respectively ML17261B1481990-08-28028 August 1990 Lists Understanding of Issues Util Planning to Address Re Containment Integrity,Per 900718 Telcon.Any Concerns or Action Items Different from Listed Submittal Should Be Provided to Util Prior to NRC 900905 & 06 Visit to Plant IR 05000244/19880261990-08-20020 August 1990 Submits Update on Util Plans to Provide Appropriate Ventilation to Intermediate Bldg Clean Side,Per Violation Noted in Insp Rept 50-244/88-26.Extensive Program to Manually Close All External Openings Implemented ML17261B1501990-08-20020 August 1990 Submits Update on Util Plans to Provide Appropriate Ventilation to Intermediate Bldg Clean Side,Per Violation Noted in Insp Rept 50-244/88-26.Extensive Program to Manually Close All External Openings Implemented ML17261B1371990-08-17017 August 1990 Forwards Rev 0 to, Inservice Insp Rept for Third Interval (1990-1999),First Period,First Outage (1990) at Re Ginna Nuclear Power Plant. ML17309A4481990-08-16016 August 1990 Responds to NRC 900717 Ltr Re Violations Noted in Insp Rept 50-244/90-80.Corrective Actions:Westinghouse Drawing E-2508 Approved & Issued to Central Records ML17261B1351990-08-15015 August 1990 Provides Info & Assessment on Integrity of Connection of Containment Bldg to Foundation,Per 900808 Telcon.Util Believes That Existing Condition Not Safety Concern & No Reason Exists to Suspect Joint Will Not Perform Function ML17261B1361990-08-14014 August 1990 Responds to Commitment Tracking Concerns Noted in Insp Rept 50-244/90-09 & Planned Corrective Actions.Util Confirms Commitments Dates for Implementation of Effective Shelf Life Program & Comprehensive Preventive Maint Program for Parts ML17261B1331990-08-13013 August 1990 Recommits to Performing Enhanced primary-to-secondary Leak Rate Monitoring,Per NRC Bulletin 88-002 ML17261B1261990-07-30030 July 1990 Clarifies Commitment Made in 900316 RO Re Restoring Inoperable Fire Damper I-411-21-P.Util Plans to Design Removable Track Which Will Allow Charcoal Drawer to Be Manipulated.Definitive Schedule Will Be Provided in 60 Days ML20055H9291990-07-23023 July 1990 Forwards Revised Page 6:4 of Plant Security Plan.Page Withheld Per 10CFR73.21 ML17261B1091990-07-20020 July 1990 Advises That Structural Evaluations of Containment Sys Being Performed in Response to Containment Integrity Insp ML17309A4471990-07-17017 July 1990 Forwards Decommissioning Rept, for Plant,Per 10CFR50.33(k) & 50.75(b) ML17261B0881990-07-13013 July 1990 Provides Update to Util 860616 Ltr Re Implementation of NUREG-0737,Item 6.2,Suppl 1, Emergency Response Capability. ML17261B0921990-07-11011 July 1990 Responds to 900709 Request for Addl Info Re Containment Integrity Insp.Util Will Provide Preliminary Results by 900716 Re Where Groundwater Entering Annular Access Area. Meeting Proposed for Wk of 900723 ML20044B0481990-07-10010 July 1990 Discusses Review of Station Blackout Documentation,Per 10CFR50.63.Licensee Will Complete Enhancements to Station Blackout Documentation Identified in Attachment 1 as Indicated & Other Items Will Be Completed within 2 Yrs ML17309A4461990-07-0909 July 1990 Responds to Generic Ltr 90-04 Re Status of Licensee Implementation of Generic Safety Issues Resolved W/ Imposition of Requirements or Corrective Actions ML17261B0931990-07-0909 July 1990 Responds to Generic Ltr 88-14, Instrument Air Supply Sys Problems Affecting Safety-Related Equipment. Util Completed All Actions Requested in Generic Ltr & Will Retain Documentation Verification for Min of 2 Yrs ML17261B0901990-07-0909 July 1990 Advises That Licensee Will Install Containment Isolation Signal Going to Valve AOV 745 by End of 1992 Refueling Outage,Per Util 900608 Ltr Notifying of Condition Outside Design Basis of Plant Under 10CFR50.72 Criterion ML17250B2151990-06-29029 June 1990 Forwards Application for Amend to License DPR-18, Reformatting Auxiliary Electrical Sys Tech Specs & Action Statements for Offsite & Onsite Power Sources Available for Plant Auxiliaries ML17261B0851990-06-28028 June 1990 Forwards LER 89-016-02 Re 891117 Failure of Safety Injection Block/Unblock Switch Which Could Render Both Trains of Safety Injection Sys Inoperable.Also Reported Per Part 21 ML17250B1991990-06-0808 June 1990 Responds to NRC 900509 Ltr Re Violations Noted in Insp Rept 50-244/89-81.Corrective Actions:Improved Battery Load Profile Developed Incorporating Calculational Improvements Contained in Current Industry Std IEEE 485-1983 ML17250B1951990-06-0505 June 1990 Responds to Generic Ltr 89-08, Erosion/Corrosion Induced Pipe Wall Thinning. Util Developed Erosion/Corrosion Program for Single & Two Phase Sys Consistent W/Requirements of NUREG-1344 & NUMARC 870611 Rept ML17250B1851990-06-0101 June 1990 Forwards Application for Amend to License DPR-18,providing Guidance for Action Statements Associated W/Power Distribution Limit Specs ML17261B0691990-06-0101 June 1990 Discusses Testing Frequency for Insp of Incore Neutron Monitoring Sys Thimble Tubes,Per NRC Bulletin 88-009.Thimble Tube Indicating Greatest Wear Recently Repositioned in Effort to Minimize Future Wear ML17250B1811990-05-31031 May 1990 Forwards Addl Info Re Response to Notice of Violation from Insp Rept 50-244/90-04.Info Withheld (Ref 10CFR73.21) ML17250B1801990-05-29029 May 1990 Responds to NRC Bulletin 90-001, Loss of Fill-Oil in Transmitters Mfg by Rosemount. Util Does Not Need to Develop Enhanced Surveillance Program to Monitor Currently Installed Transmitter Based Upon Limited Installed Quantity ML17250B1841990-05-29029 May 1990 Responds to NRC Bulletin 89-002, Stress Corrosion Cracking of High Hardness Type 410 Stainless Steel (SS) Internal Pre-Load Bolting.... Valves Disassembled & 410 SS Studs Removed & Visually & Liquid Penetrant Examined ML17261B0961990-05-29029 May 1990 Responds to NRC 900427 Ltr Re Violations Noted in Insp Rept 50-244/90-07.Corrective Actions:Worker Records Immediately Corrected Indicating That Worker Received No Significant Exposure for Time While Error Occurred ML17250B1721990-05-18018 May 1990 Forwards Rev 1 to Summary Exam Rept for 1990 Steam Generator Eddy Current Insp. Rept Summarizes Observations & Corrective Actions Resulting from Insps Performed During 1990 Outage ML20042G8931990-05-0808 May 1990 Forwards Revised Emergency Operating Procedures,Including Rev 3 to AP-RHR.2,Rev 1 to ES-0.3 & Rev 11 to ES-1.3 ML17261B0611990-04-26026 April 1990 Advises of Changes to 890201 Commitments Made Under Programmed Enhancement Response to Generic Ltr 88-17.Changes Will Not Reduce Capability to Operate Safely in Reduced Inventory Condition or Scope of Programmed Enhancements ML17261B0591990-04-17017 April 1990 Advises That No Interlocks Required for RHR motor-operated Valves 701 & 720 Based on Present Arrangement.Listed Failures Would Have to Occur in Order for Potential Overpressurization of RHR Sys to Occur ML17261B0461990-04-12012 April 1990 Responds to Issues Discussed During 900323 Telcon Re Inservice Testing Program Status & Relief Request.Current Test Methodology for Seat Leakage Acceptable Based on Application of Direct Measurement Sys ML17262A1431990-04-12012 April 1990 Responds to NRC 900202 Ltr Re Weaknesses Noted in Insp Rept 50-244/89-80.Corrective Actions:Procedure ES-0.3 Modified to Provide Guidance for Rapid Cooldown & Depressurization W/ & W/O Reactor & Vessel Instrumentation Sys ML17250B1451990-04-0606 April 1990 Discusses Impact of SER Issuance for Inservice Insp & Inservice Testing Programs & Advises That Timing Will Not Affect Util Implementation Plans for Programs,Except as Listed ML17250B1441990-04-0505 April 1990 Responds to NRC 900309 Ltr Re Violations Noted in Insp Rept 50-244/90-02.Corrective Actions:Positive Indicator Will Be Installed for All Check Valves to Enhance Visible Positive Position Verification Ability & to Avoid Confusion ML17261B0401990-03-30030 March 1990 Advises That Final Results of Station Blackout Documentation Review Will Be Submitted to NRC on or About 900501 ML17261B0321990-03-28028 March 1990 Forwards Annual Rept of ECCS Model Revs as Applicable to Facility,Per 10CFR50.46.Mods to Model for Small Break LOCA Do Not Affect Calculated Peak Clad Temp ML17261B0331990-03-28028 March 1990 Forwards Info Re Reactor Vessel Issues,Per 900305 Telcon Concerning Change of License Expiration Date ML17261B0231990-03-26026 March 1990 Responds to NRC 890222 Ltr Re Violations Noted in Insp Rept 50-244/89-17.Corrective Actions:Personnel Verified Safety Injection Block/Unblock Switch in Proper Position & Operator Procedure 0-1.1 Changed as Indicated ML17261B0281990-03-23023 March 1990 Forwards Addl Info Re Proposed Tech Spec Amend Concerning Use of Reconstituted Fuel,Per 900315 Telcon.During Fuel Assembly Reconstitution,Failed Fuel Rods Will Be Placed W/ Filler Rods ML17261B0221990-03-22022 March 1990 Provides Revised Test Schedule for motor-operated Valve Diagnostic Test Program,Per IE Bulletin 85-003.NRC Notification of Changes to Valve Operability Program Required by Generic Ltr 89-10,dtd 890828 ML17261B0411990-03-20020 March 1990 Forwards Summary of Onsite Property Damage Coverage Currently in Force at Plant,Per 10CFR50.54(w)(4) ML17261B0151990-03-19019 March 1990 Responds to Generic Ltr 89-19, Safety Implication of Control Sys in LWR Nuclear Plants (USI A-47). Overfill Protection Provided Through Trip Bistables in Reactor Protection Racks Powered from 120-volt Instrument Buses ML17261B0101990-03-15015 March 1990 Forwards Application for Amend to License DPR-18,allowing Use of Reconstituted Fuel Assemblies ML20012D2831990-03-14014 March 1990 Responds to Issues Discussed During 900307 Telcon W/Util & Eg&G Re Inservice Testing Program Status & Relief Request. Valves 5960A & 5960B Will Be Disassembled to Verify Forward Flow.Relief Requests PR-8 & VR-25 Encl 1990-09-18
[Table view] |
Text
REGULATORY INFORMATION DISTRIBUTION SYS H (RIDS)
ACCESSION NBR'800815(108 DOC DATE! 84/08/08 NOTARIZED: NO DOCKET. 0
~
FACIL:50-240 Robert)Emmet Ginna Nuclea'r Planti Unit li Rochester G 05000240 AUTH'BYNAME AUTHOR AFFILIATION ROBERtRel4e Rochester Gas 5 Electric Carp.
'EC IP, NAME RECIPIENT AFFILIATION PAULSONiWeA ~ Operating Reactors Branch 5.
SUBJECT:
Discusses encl, results of analysis or Reactor Vessel Surveillance Capsule T~supporting 821208 application to License DPR-18.Fluence data 8 analysis presented in for'mend YCAP 10086 verified, DISTRIBUTION CODE: A001S COPIES RECEIVED:LTR ENCL / SIZE:
TITLE: OR Submittal: General Distr ibution NOTES:NRR/DL/SEP 1cy, 05000244 OL+ 09/19/69 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL NRR ORB5 BC 01 7 7 INTERNALS ADH LFMB 0 ELD/HOSP 0 NRR/DE/MTEB 1 NRR/DL DIR 1 NRR/DL/ORAB 1- 0 N HETB 1 NRR/DSI/RAB 1 EG F E 04 l.
RGN1 1 1 EXTERNALS ACRS 6 LPDR 03 NRC PDR 02 1 NSIC 05 NOTES'9 NTIS 1 1
1 1
TOTAL NUMBER OF COPIES REQUIRED: LTTR 27 ENCL 24
II' o; l C
t
'II IIIIIIIIIg ROGER W. KOBER VICE PRESIDENT III" XWn'P::::
ZZZZZS ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER, N.K 14649-0001 TELEPHONE 1am< Owl ~
- ELECTRIC & STEAM PRODUCTION AREA CODE Tld 546-2700 August 8, 1984 Director of Nuclear Reactor Regulation Attention: Mr. Walter A. Paulson, Acting Chief Operating Reactors Branch No. 5 U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Subject:
Reactor Vessel Surveillance Capsule T R. E. Ginna Nuclear Power Plant Docket No. 50-244
Dear Mr. Paulson:
On December 8, 1982 an Application for Amendment to Operating License was submitted to the NRC incorporating the results of the analysis of reactor vessel surveillance capsule T. This analysis was performed by Westinghouse and published as WCAP-10086.
The results of the neutron transport calculations presented in WCAP-10086 were obtained using a statistically based generic 2-loop core power distribution and cross-sections employing a Pl expansion of the scattering matrix. In spite of the good agreement between measurements and calculations at the surveillance capsule locations, both of these assumptions were questioned by the NRC staff in their review of WCAP-10086. In particular, the consideration of plant specific core power distributions and the use of a P expansion of the neutron scattering cross-sections were requested.
In order to address these two issues, neutron transport cal-culations in R II geometry were carried out using the Dl@Idiscrete ordinates code and the SAILOR cross-section library . The SAILOR library is a 47 group, ENDF-BIV based data set produced specifically for light water reactor applications. Anisotropic neutron scattering is treated with a P 3 expansion of the cross-sections.
Two sets of transport calculations were carried out to provide the appropriate data to address the issues at hand. The first, a single computation in the forward mode, utilized the P cross-sections in conjunction with the generic 2-loop power distribution to provide a direct P3/Pl comparison. The second set of calculations consisted of a serxes of seven adjoint analyses relating power distributions within the core to the fast neutron flux at four locations on the pressure vessel as well as at the 8408150108 g QQoP 05000244, 840808'DR ADOCK P PDR
e ROCHESTER GAS AND ELECTRIC CORP. SHEET NO.
DATE August 8 1984 g
Mr. Walter A. Paulson center of each surveillance capsule. These adjoint importance functions were then used along with actual core burnup information to generate plant specific fluence levels on a cycle by cycle basis for the R. E. Ginna reactor. A P3 scattering cross-section expansion was used for all of the adjoint analyses.
The transport methodology, both forward and adjoint, using the SAILOR cross-~potion library has been benchmarked against the ORNL PCA facility as well as against the Westinghouse power reactor surveillance capsule data base. The results of these benchmarking studies will be available in a report to be issued by Westinghouse by the end of 1984. However, as indicated in our letter of October 10, 1983, the benchmarking studies indicate that the use of SAILOR cross-sections and generic design basis power distributions produces flux levels that tend to be conservative by about 7-22%. This conservatism is to be expected since the generic power distributions include a 2o statistical uncertainty based on plant-to-plant and cycle-to-cycle variations in peripheral power. It is unlikely that a single reactor would have a power distribution at the nominal +2o level for a large number of fuel cycles.
When plant specific power distributions are used with the adjoint importance functions, the benchmarking studies show that
.fluence predictions are within +15% of'easured values at sur-veillance capsule locations. Further, these predictions also tend to be in good agreement with the P calculations based on the generic power distribution. That ls, the 2o uncertainty added to the peripheral power tends to offset deficiencies in the P 1 analy-sis relative to the P3 approach.
Results of Anal sis Results of the plant specific fluence evaluations for the R. E. Ginna reactor are 0 summarized jn Tables 1 0 through 7. Data are presented for the 0 < 14.5 , 30 , and 44.5 azimuthal locations at the pressure vessel inner radius in Tables 1 through 4, respectively. Corresponding data applicable to tge center of surveillance capsules positioned at 13 , 23 , and 33 are given in Tables 5 through 7. In all cases, plant specific fast neutron flux and fluence levels for each operating cycle are listed along with predictions using a constant flux level based on P and P
1 cross-sections and the generic design basis core power dis(ri-butions.
In regard to the data contained in Tables 1 through 7, the plant specific core power distributions used for cycles lA through 13 were derived from actual burnup data. The cycle 14 power distribution was taken from the nuclear design data given in WCAP-10505. As with the power distribution data, the full power irradiation times for cycles lA through 13 were obtained from the cycle specific burnup data. The irradiation time for cycle 14 was computed from the design burnup of 8800 MWD/MTU specified in WCAP-10505. Cycle 14 began in Spring 1984.
I ROCHESTER GAS AND ELECTRIC CORP. SHEET NO. 3 DATE August 8, 1984 Mr. Walter A. Paulson Where appropriate, measured fluence values from surveillance capsules V, R, and T have been compared with calculated values.
These analytical/experimental comparisons are shown in Tables 5 and 6. It should be noted that the measured values represent the average fluence derived from all neutron sensors contained in each surveillance capsule. Further, in keeping with the NRC request, the fluence levels are based on spectrum average cross-sections obtained from the SAILOR P3 analysis. Therefore, measured fluence values listed in Tables' and 6 differ slightly (from 1% to 8%)
from the corresponding values given in Table 6-12 of WCAP-10086.
These variations in the experimental values are well within the uncertainty in the measurements themselves.
The following conclusions can be drawn from an examination of Tables 1 through 7.
- l. Over the long term the plant specific SAILOR P3 analysis is in excellent agreement with the GAMBIT P analysis.
At the projected end of cycle 14 the P3/Pl luence ratios are as follows:
SAILOR P3/GAMBIT Pl 0
0 vessel 0.98 0 vessel 14.5 1. 01 30 vessel 1.04 44.5 vessel 1.14 13 0 capsule 0.92 23 capsule 0.95 33 0 capsule 0.92
- 2. For surveillance capsules, V, R, and T the fluence predictions based on plant specific SAILOR P3 analysis are in excellent agreement with dosimetry data.
Analytical/experimental comparisons are as follows:
SAILOR P3/MEASUREMENT Capsule V 0.99 Capsule R 0.95 Capsule T 0.92
- 3. Analysis using SAILOR P 3 cross-sections and the generic design basis core power distribution produces con-servative results relative to the other two analytical approaches as well as to the surveillance capsule measurements.
ROCHESTER GAS AND ELECTRIC CORP. SHEET NO. 4 D4TE August 8, 1984 Mr.,Walter A. Paulson Based on these conclusions, the fluence data and analysis presented by WCAP-10086 is verified and supports our application of December 8, 1982.
er truly yours,
~ul Roger W. Kober
ROCHESTER GAS AND ELECTRIC CORP. SHEET NO. 5 AUgUst 8 g 1 984 Mr. Walter A. Paulson
References:
Soltesz, R. G., R. K. Disney, J. Jedruch, and S. L. Zeigler, "Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation Volume 5 Two-Dimensional Discrete Ordinates Transport Technique", WANL-PR-(LL)-034, Vol. 5, August 1970.
- 2. SAILOR RSIC DATA LIBRARY COLLECTION DLC-76, "Coupled, Self-shielded, 47 Neutron, 20 Gamma-ray, P3, Cross-section Library for Light Water Reactors".
- 3. McElroy, W. N., "LWR Pressure Vessel Surveillance Dosimetry Improvement Program: PCA Experiments and Blind Test",
NUREG/CR-1861, June 1981.
4 ~ Letter J. E. Maier to D. M. Crutchfield, October 10, 1983.
SD-RSA-84/2153 Table 1 FAST NEUTRON E > 1.0 MeV EXPOSURE. AT THE PRESSURE VESSEL INNER RADIUS 'O'ZIMUTHAL ANGLE CUMULATIVE FLUENCE (n/cm2)
Irradiation Cycle Avg.
Cycl e Time Fj,ux Plant Specific Generic, Generic No. (EFPS) (n/cm~-sec) P3 P3 1A 2.09 x 10 3.88 x 10 8.10 x 10 8.26 x 10 1.01 x 10 18 2.38 x 10 4.16 x 10 1.80 1p18 1.77 x 10 2.15 x 10 2 0.71 x 10 4.22 x 10 2.10 x 1018 2,05 x 10 2.50 x 10 3 2.88 x 10 3.62 x 10 3.14 x 10 3.18 x 10 3.88 x 10 2.17 x 10 3.60 x 10 3.93 x 1018 4.04 x 10 4.93 x 10 5 1.84 x 10 4.63 x 10 4.78 x 10 4.77 x 10 5.82 x 10 6 2.42 x 10 3.93 x 10 5.73 x 1018 5.72 x 10 6.98 x 10 7 2.35 x 10 3.74 x 10 6.61 x 1018 6.65 x 10 8.12 x 10 8 2.20 x 10 4.40 x 10 7.58 x 1P18 7.52 x 10 9.18 x 10 9 2,62 x 10 4.13 x 10 8.66 x 1018 8.55 x 10 1.04 x 10 19 10 2.51 x 10 3.97 x 10 9.65 x 1018 9.55 x 10 1.16 x 10 19 11 1.83 x 10 3.86 x 10 1.04 x 10" 1.03 x 10 1.25 x 10 19 12 2.30 x 10 4.53 x 10 1.14 x 10 1.12 x 10 1.37 x 10 13 2.32 x 10 3.11 x 10 1.21 x 10 1.21 x 10 1.48 x 10 1p19 19 14 2.27 x 10 3.05 x 10 1.28 1.30 x 10 1.59 x 10
-
P 10 2 Notes: 1 9 avg = 3.95 x 10 n/cm -sec.
1p
= 4.82 x 10 n/cm -sec.
'0 n
p1
SD-RSA-84/2153 Table 2 FAST NEUTRON E > 1.0 MeV . EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 14.5'ZIMUTHAL ANGLE CUMULATIVE FLUENCE (n/cm2)
Irradiation Cycle Avg. :- 1 Cycl e Time Fj,ux Plant Specific Generic , Generic No. (EFPS) (n/cm~-sec) P3 P1 P3 1A 2.09 x 10 2.34 x 10 4.89 x 10 4.97 x 10 6.11 x 10, 1B 2.38 x 10 2.48 x 10 1.08 x 10 1.06 x 10 1.31 x 10 2 0.71 x 10 2.57 x 10 1.26 x 1018 1.23 x 10 1.52 x 10 3 2.88 x 10 2.22 x 10 1.90 x 10 1.92 x 10 2.37 x 10 4 2.17 x 10 2.38 x 10 2.42 x 10 2.44 x 10 3.01 x 10 5 1.84 x 10 2.81 x 10 2.94 x 10 2.87 x 10 3.55 x 10 6 2.42 x 10 2.47 x 10 3.53 x 10 3.45 x 10 4.27 x 10 7 2.35 x 10 2.48 x 10 4.12 x 10 4 01 x 10 4.96 x 10 8 2.20 x 10 2.67 x 10 4.70 x 10 4.53 x 10 5.61 x 10 9 2.62 x 10 2.59 x 10 5.38 x 10 5.16 x 10 6.38 x 10 10 2.51 x 10 2.46 x 10 6.00 x 10 5.75 x 10 7.12 x 10 11 1.83 x 10 2.39 x 1010 6.44 x 10 6.19 x 10 7.66 x 10 12 2 '0 x 10 2.61 x 10 7.04 x 10 6.74 x 10 8.34 x 10 13 2.32 x 10 1.93 x 10 7.49 x 10 7.29 x 10 9.03 x 10 14 2.27 x 10 1.97 x 10 7.94 x 10 7.83 x 10 9.70 x 10
- 10 2 Notes: 1 5 = 2 38 x 10 n/cm -sec.
10 2
= 2.95 x 10 n/cm -sec.
~l SD-RSA-84/2153 Table 3 FAST NEUTRON E > 1.0 MeV EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 30 AZIMUTHAL ANGLE CUMULATIVE FLUENCE (n/cm2)
Irradiation Cycle Avg.
Cycle Time F/ux 'Plant Specific Generic , Generic
'No. (EFPS) (n/cm -sec) P3 Pl -
P3 lA 2.09 x 10 '.50 x 10 3.14 x 10 3.24 x 10 .-
4.12 x 10 .
1B 2.38 x 10 1.53 x 10 6.78 x,10 6.93 x 10 8.81 x 10 2 0.71 x 10 1.75 x 10 . 8.02 x 10 8.03 x 10 1.02 x 10 3 2.88 x 10 1.46 x 10 1.22 x 10 1.,25 x 10 1.59 x 10 4 2.17 x 10 1.80 x 10 1.61 x 10 1.59 x 10 2.02 x 10 5 1.84 x 10 1.71 x 10 1.93 x 10 1.87 x 10 2.38 x 10 6 2;42 x 10 1.62 x 10 2.32 x 10 2.25 x 10 2.85 x 10 7 2.35 x 10 1.71 x 10 2.72 x 10 2.61 x 10 3.32 x 10 8 2.20 x 10 1.76 x 10 3.11 x 10 . 2.95 x 10 3.75 x 10 9 2.62 x 10 1.84 x 10 3.59 x 10 3.36 x 10 4.27 x 10 10 2.51 x 10 1.78 x 10 4.04 x 10 3.75 x 10 4.76 x 10 ll l.83 x 10 1 ~ 48 x 10 4.31 x 10 4 03 x 10 5.12 x 10 12 2.30 x 10 1.43 x 10 '4.63 x 10 4.39 x 10 5.58 x 10 13 2.32 x 10 1.27 x 10 4.92 x 10 4.74 x 10 6 03 x 10 14 2.27 x 10 1.54 x 10 5.27 x 10 5.09 x 10 6.48 x 10 Notes: 1 - 5 ayg = 1.55 x 10 10 n/cm 2 -sec.
10 Pay= 1 .97x1 0n/cm-sec P
'
ill
- 4
SD-RSA-84/2153 Table 4 FAST NEUTRON E > 1.0 MeV EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 44.5'ZIHUTHAL'ANGLE CUMULATIVE'FLUENCE (n/cm2)
Irradiation Cycle Avg.
Cycl e Time Fiux Plant Specific Generic Generic No. (EFPS) (n/cm~-sec) P3 Pl P3 1A 2.09 x 10 1.29 x 10 2.70 x 10 , 2.65 x ] 017 3.64 x 10 1B 2.38 x 10 1.31 x 10 5.81 x .10 5.68 x 1017 7.78 x 1017 2 0.71 x 10 1.60 x 10 6.95 x 10 '.58 x 10" 9. Ol x 10 3 2.88 x 10 1.30 x 10 1.'07 x 10 1.02 x 1018 1.40 x 1018 4 2.17 x 10 1 ~ 62 x 10 1.42 x 10 1.30 x 1018 1.78 x 10" 5 1.84 x 10 1.36 x.l0- 1.67 x 10 1.53 x 018 2.10 x 10 6 2.42 x 10 1.30 x 10 1.99' 10 1.84 x 1'0" 2.52 x 1018 7 2.35 x 10 1.38 x 10 2.31 x 10 2.14 x 10 2.93 x 1018 8 2.20 x 10 1.58 x 10 2.66 x 10 2.42 x 1018 3.31 x 1018 9 2.62 x 10 1.68 x 10 3.10 x 10 2.75 x 1018 3. 77 x 1018 8
10 2.51 x 10 1.63 x 10 3.51 x 10 3.07 x 10 4. 21 x 1018 11 1'.83 x 10 1.45'x 10 3.77 x 10 3.30 x 1018 4.52 x 10" 2.30 x 10 1.39 x 10 4.09 x 10 3.59 x 018 4.92 x 1018 12 .
13 2.32 x 10 1.05 x 10 4.34 x 10 3.84 x 1018 , 5.33 x 10 2.27 x 10 1.54 x 10 4.69 x 10 4.13 x 1018 5.72 1018 14
- 10 2 Notes: 1 9 = 1.27 x 10 n/cm -sec.
10 2 2 - p = 1.74 x 10 n/cm -sec.
FSD-RSA-84/2153 Table 5 FAST NEUTRON E > 1.0 MeY EXPOSURE AT THE 13'URVEILLANCE CAPSULE .CENTER CUMULATIVE FLUENCE (n/cm )
Irradiation Cycle Avg.
Cycl e Time Pjux '1ant 'Specific'n/cm Capsule Generic Pl Generic P3 No. (EFPS) -sec.) P Measurement
,
1A 2.09 x 10 1.19 x 10 2.49 x 10 2.'78 x'1018 3.07 x 10 7 18 5. 95 x 1018 6.57 x 10 1B 2 38 x 10 1.27 x 10 5.51 x 10 1.30 x 10 6.44 x 10 6.53 x 10 6. 89 1018 7.61 x 10 2 0.71 x 10 3 2.88 x 10 1.13 x 10 9.69 x 10 . 1.02 x 10 1.07 x 10 1.18 x 10 1.22 x 10 1.36 1019 1.50 x 10 4 2.17 x 10 1.17 x 10 19 5 1.84 x 10 1 ~ 43 x 10 1.49 x 10 1.61 x 10 1.77 x 10 6 2.42 x 10 1.24 x 10 1.79 x 10 1.93 x 10" 2.13 x 10 19 7 2.35 x 10 1.23 x 10 2.08 x 10 2.24 x 10 2.48 x 10 2.37 2.53 1019 2.80 x 10 8 2.20 x 10 1.36 x 10 x 10 9 2.62 x 10 1.29 x 10 2.71 x 10 2.88 x 10 3.18 x 10 10 2.51 x 10 1.23 x 10 3.02 x 10 3. x 10 3.55 x 10 2-1'.46 3.24 x 10 1019 3.82 x 10 11 1.83 x 10 ~
1.22 x 10 3.56 3.76 1019 4.16 x 10 12 2 30 x 10 1.36 x 10 x 10 13 2.32 x 10 9.43 x 10 3.78 x 10 4.07 x 10 4.50 x 10 19 4.37 x 10 4.83 x 10 14 227 x10 9.49 x 10 4.00 x 10 10 2 Notes: 1 - P
= 1.33 x 10 n/cm -sec.
11 2 2 - 5 = 1.47 x 10 n/cm -sec.
'V
'l
'
~ II
~ ~ p C II
~f'
FSD-RSA-84/2153 Table 6 FAST NEUTRON E ) 1.0 MeV EXPOSURE AT THE 23'URVEILLANCE CAPSULE CENTER CUMULATIVE FLUENCE (n/cm )
Cycl e Irradiation Time Cyc1e Ayg. ~:
Generic 1 ~
beneric 2
, F)ux P1ant Specific No. (EFPS)'n/cm -sec.) P3 Capsu1e Pl P3 Measurement lA 2.09 x 10 6.86 x 10 1.43 x 10 1.60 x 10 1.86 x 1018.
1B 2.38 x 10 7.14 x 10 3.13 x 10 3;42 x 10 , 3.98 x 1018 0.71 7.80 18'.
2 x 10 x 10 3.69 x 10 3.97 x 10 >4.61 x 10 3 2.88 x 10 6.61 x 10 5.59 x 10 6.18 x 10 7;18 1018 4 2.17 x 10 7.82 x 10 7.29 x 10 7.84 x 10 ,9.11 x 1018 5 1.84 x 10 8.14 x 10 8.79 x 10 9.25 x 10 1.08 x 10 2.42 7.50 19 10" 6 x 10 x 10 1.06 x 10 1.11 x 10 1.29 x 7 2.35 x 10 7.82 x 10 1.24 x 10 1.29 x 10 1.50 x 10 2.20 'x 7.95 8 10 x 10 1.42 x 10 1.46 x 10 1.70 x 10 9 2.62 x 10 8.07 x 10 1.63 x 10 1.78 x 10 1.66 x 10 1.93 1019 10 2.51 x 10 7.74 x 10 1.82 19 x 10 1.85 x 10 2.15 x 10 ll 1.83 x 10 7.18 x 10 1.96 x 10 1.99 x 10 2.32 x 10" 12 2.30 x 10 6.85 x 10 2.11 x 10 2.17 x- 10 '2.52 1019 13 2.32 x 10 5.95 x 10 2.25 x 10 2.35 x- 10 2.73 x 10" 14 2.27 x 10 6.55 x 10 2.40 x 10 2.52 x 10 2.93 x 10 Notes: - 10 2 1
Hay 7 66 x 10 n/cm sec.
2 pay=8.9 1x10n/ cm-sec 10
wc ll
~f ~
D-RSA-84/2153 Table 7 FAST NEUTRON E > 1.0 MeV EXPOSURE AT THE 33'URVEILLANCE CAPSULE CENTER CUMULATIVE FLUENCE (n/cm2)
Irradiation 'yc1e Avg.
Cyc1e Time Flux Plant Specific Generic Generic No. (EFPS) (n/cm~-sec) P3 Pl P3 18 1A '2.09 x 10 6.06 1010 1.27 x 10 1.48 x 10 1.68 x 10 7 18 1B 2.38 x 10 6.14 x 10 2.73 x 10 3.16 x 10 3.60 x 10 18 2 0.71 x 10 7.24 1010 3.24 x 10 3.66 x 10 4.18 x 10 3 2.88 x 10 5.95 x 10 4.96 x 10 5.69 x 10 6.50 x 10 4 2.17 x 10 7.40 x 10" 6,.56 x 10 7.22 x 10 8.25 x 10 5 1.84 x 10 6.75 x 1010 7.80 x 10 8.52 x 10 9.73 x 10 6 2.42 x 10 6.45 x 10" 9 37 x 10 1.02 x 10 1.I7 x 10 10 1.19 x 10 1.36 x 10 7 2.35 x 10 6.86 x 10 1.10 x 10 8 2.20 x 10 7.22 1010 1.26 x 10 1.34 x 10 1.53 x 10 9 2.62 x 10 7.61 x 10" 1.46 x 10 1.53 x 10 1.75 x 10 10 2.51 x 10 7.39 x 10 1.64 x 10 1.71 x 10 1.95 x 10 ll 1.83 x 10 6.03 x 10 1.75 x 10 1.84 x 10 19 2.10 x 10 12 2.30 x 10 5.78 1010 1.88 x 10 2.00 x 10 2.28 x 10 10 19 13 2.32 x 10 5.00 x 10 2.00 x 10 2.16 x 10 2.47 x 10, 19 14, 2.27 x 10 6.38 x 10 2.14 x 10 2.32 x 10 2.65 x 10 10 2 Notes: 1 9 = 7.08 x, 10 'n/cm -sec.
- 10 2 2 9 avg = 8.06 x 10 n/cm -sec.
0