Similar Documents at Salem |
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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217M2101999-10-19019 October 1999 Forwards NRC Rept Number 17, Requal Tracking Rept from Operator Licensing Tracking Sys.Rept Was Used by NRC to Schedule Requalification Exam for Operators & Record Requal Pass Dates ML18107A5601999-10-18018 October 1999 Submits 30-day Fuel Clad Temp Rept,Iaw 10CFR50.46.Rept Describes Changes to Calculated Peak Clad Temp (PCT) for Salem Generating Station Large & Small Break LOCA Analyses ML18107A5531999-10-0808 October 1999 Forwards Summary Rept of Plant Startup & Power Ascension Testing for Sgs,Unit 2 Cycle 11,per Requirements of TS 6.9.1.1 ML18107A5561999-10-0707 October 1999 Requests Relief Associated with Containment Examinations at Hope Creek & Salem Generating Stations.Attachment 1 Includes Proposed Alternatives & Supporting Justification for Relief Requests ML18107A5501999-10-0505 October 1999 Provides Current Status of Pse&G Actions Re GL 98-01, Y2K Readiness of Computer Sys at Npps, for Salem Nuclear Generating Station,Units 1 & 2 & Hope Creek Nuclear Generating Station ML18107A5521999-10-0505 October 1999 Encourages NRC to Support Abb Combustion Engineering Nuclear Power Request for Priority Review of Generic TR Re Crossflow Ultrasonic Flow Measurement Sys ML18107A5591999-10-0505 October 1999 Informs That Nj Dept of Environ Protection Has No Comments on License Change Request S99-07 for Sgs,Units 1 & 2 IR 05000272/19990071999-09-28028 September 1999 Forwards Insp Repts 50-272/99-07 & 50-306/99-07 on 990721- 0831.One Potentially Safety Significant Issue Identified Dealing with Control Room Special Ventilation System.Four Addl Issues of Low Safety Significance Identified ML18107A5341999-09-22022 September 1999 Provides Data Re Operator Licensing Exam for Salem & Hope Creek Station,In Response to NRC Form 536 (7-1999) ML18107A5421999-09-22022 September 1999 Forwards Discharge Monitoring Rept for Salem Generating Station for Aug 1999.Rept Is Required by & Prepared Specifically for EPA & Nj Dept of Environ Protection ML20212B3631999-09-14014 September 1999 Forwards Rev 13 to Salem - Hope Creek Security Plan,Iaw 10CFR50.54(p).Summary of Proposed Changes to Plan,Encl. Encl Withheld ML18107A5331999-09-13013 September 1999 Provides Notification That PSEG Intends to Utilize ASME Code Case N-481 During Second ISI Interval at Sgs Units 1 & 2 ML18107A5351999-09-13013 September 1999 Informs That NRC Has Reviewed Pse&G Request Proposing to Modify TS Which Allow EDG to Be Operated for 24 Surveillance Test During Any Mode,Iaw 10CFR50.91(b) & Has No Comments ML18107A5321999-09-13013 September 1999 Forwards Revised 10CFR50.92 Evaluation to Clarify Util Response to Question Number 1 Re Amend to Modify TS 3/4 8.1, AC Power Sources. ML18107A5221999-09-0808 September 1999 Requests Approval to Use ASME Code Case N546,which Provides Alternative Qualification Requirements That Allow Personnel Most Familiar with Walkdown of Plant Sys,Like License Operators to Perform VT-2 Examinations ML18107A5231999-09-0808 September 1999 Forwards Salem Generating Station Semiannual Radioactive Effluent Release Rept for Jan-June 1999. Rept Summarizes Liquid & Gaseous Releases & Solid Waste Shipments from Salem Generating Station for Period of Jan-June 1999 ML18107A5181999-08-26026 August 1999 Forwards Ninety Day Rept for ISI Activities Conducted at Sgs,Unit 2 During Ninth Extended Outage & Tenth Refueling Outage.List of Encl,Provided ML18107A5061999-08-26026 August 1999 Provides First Feedback from Observation of NRC Insp Under Pilot Nuclear Power Plant Insp Program.Attached Are Completed Insp Feedback Forms for Procedure 71111, Attachment 21 & Procedure 71151 ML18107A4981999-08-26026 August 1999 Forwards Response to NOV That Resulted from Predecisional Enforcement Conference Conducted on 990624.Corrective Actions:Communications to Supervisors Reinforced Employee Right & Duty to Raise Nuclear Safety Issues ML18107A5051999-08-24024 August 1999 Informs That Nj Dept of Environ Protection Bureau of Nuclear Engineering Has Reviewed Request IAW Requirements of 10CFR50.91(b) & Has No Comments Re Amend to FOL Change Request S99-02 to Modify TS Re Penetration Valves ML18107A4921999-08-23023 August 1999 Provides Suppl Info Re 971024 Amend Request to Modify TS 3/4 7.7, Auxiliary Bldg Exhaust Air Filtration Sys. Util Withdraws from Request All Proposed Changes Associated with Filter Testing,Per Issuance of GL 99-02 ML18107A4911999-08-20020 August 1999 Forwards Revised Plant Attribute Libraries for Salem & Hope Creek Generating Stations,Iaw 10CFR50,App E,Vi, Emergency Response Data Sys, 3.b.Changes Are Identified by Rev Bars ML18107A4861999-08-19019 August 1999 Forwards NPDES Discharge Monitoring Rept, for Salem Generating Station for Month of Jul 1999.Rept Required by & Prepared Specifically for EPA & Nj Dept of Environ Protection ML18107A4831999-08-18018 August 1999 Submits Licensee Comments on NRC 990730 Ltr Which Provided Notification of Close Out of TAC Numbers MA0567 & MA0568 Re GL 92-01,Rev 1,Suppl 1 ML18107A4801999-08-13013 August 1999 Requests That Pse&Gs Contact in NUREG-0383, Directory of Compliance for Radioactive Matl Packages, Be Changed ML18107A4751999-08-0505 August 1999 Forwards Fitness for Duty Performance Data Rept for Six Month Period Ending 990630 ML20210M7571999-08-0404 August 1999 Forwards Response to Requesting Addl Info Re Status of Decommissioning Funding for Lgs,Pbaps & Sngs. Attachment Provides Restatement of Questions Followed by Response ML18107A4531999-07-20020 July 1999 Forwards Discharge Monitoring Rept for Salem Generating Station, for June 1999.Rept Is Required by & Prepared for EPA & Nj Dept of Environ Protection ML18107A4431999-07-0606 July 1999 Informs That Nj Dept of Environ Protection Bureau of Nuclear Engineering Has Reviewed Pse&G License Change Request S98-17 Re Permissible Enrichment Values for New Fuel Storage & Has No Comments ML20196J6301999-07-0101 July 1999 Requests Addl Info Re Status of Decommissioning Funding for Limerick Generating Station,Units 1 & 2,Peach Bottom Atomic Power Station,Units 1,2 & 3 & Salem Nuclear Generating Station,Units 1 & 2 ML18107A4181999-06-30030 June 1999 Submits Response to NRC Request for Info Re Y2K Readiness at Npps,Per GL 98-01,suppl 1.Disclosure Encl ML20209B6751999-06-29029 June 1999 Ack Receipt of from Dr Powell in Response to NRC Re Fitness for Duty.Attachment 2 of Will Be Withheld from Public Disclosure,Per 10CFR2.790 ML18107A4131999-06-25025 June 1999 Provides Further Clarification of Licensing & Design Basis for 125 Vdc Battery Margins for Sgs & HCGS for Meeting Station SBO & Loca/Loop Loading Requirements,Per Util 990426 Ltr & Discussion with NRC ML18107A4091999-06-22022 June 1999 Forwards Discharge Monitoring Rept for May 1999,containing Info as Required by Permit NJ0005622.Rept Prepared Specifically for EPA & Nj Dept of Environ Protection ML18107A4061999-06-21021 June 1999 Provides Supplemental Info to Proposed License Amend, Modifying TS 3/4 7.7, Auxiliary Bldg Exhaust Air Filtration Sys. Info Re Acceptance Criterion Discussed ML18107A3691999-06-11011 June 1999 Forwards Corrected Monthly Operating Rept for Apr 1999 for Salem Generating Station,Unit 1.Original Submittal Contained Typo for year-to-date Value for Numbers of Hours Generator Was on Line (Service Hours) ML18107A3641999-06-0404 June 1999 Requests Enforcement Discretion for TS 3/4.6.2.3 Re Containment Cooling Sys for Salem Generating Station,Unit 1 ML18107A3561999-06-0303 June 1999 Informs That Nj Dept of Environ Protection Bureau of Nuclear Engineering Has Reviewed Pse&G License Change Request S99-05 & Has No Comments ML18107A3611999-05-27027 May 1999 Forwards Responses to NRC 990301 & 990323 RAIs for Salem & Hope Creek Generating Stations Relating to GL 96-05 ML18107A3301999-05-24024 May 1999 Forwards Suppl Info Re GL 95-07, Pressure Locking & Thermal Binding of Safety Related Power Operated Gate Valves. Encl Contains Methodology Used in Determination of Pressure Locking Susceptibility of PORVs Block Valves ML18107A3371999-05-21021 May 1999 Forwards NPDES Discharge Monitoring for Salem Generating Station for Apr 1999, Containing Info as Required by Permit NJ0005622 ML18107A3291999-05-20020 May 1999 Forwards Redacted Response to NRC 990322 RAI Re Notification of Licensed Operator That Tested Positive for Alcohol. Attachment 2 Withheld,Per 10CFR2.790(a)(6) ML18107A3031999-05-18018 May 1999 Provides Summary of Changes to NRC Commitments That Have Been Made But Not Reported by Other Means,Iaw with NEI Process for Managing NRC Commitments ML18107A2891999-05-13013 May 1999 Forwards Rev 36 to Pse&G Nuclear Business Unit Emergency Plan. Rev 36 Incorporates Changes to Section 1-3,6 & 7 & 9-17.Attached Copy Includes All Sections of EP for Completeness ML18107A2951999-05-12012 May 1999 Submits SG Tube Plugging Rept,Per Plant TS 4.4.6.5.a.Total of 47 Tubes Were Plugged During SG Tube Insps,Which Were Completed During Plant Tenth RFO ML18107A2861999-05-11011 May 1999 Forwards Rev 0 to NFS-0174, COLR for Salem Unit 2 Cycle 11. COLR Rept Was Received by Util as Part of Reload SE ML18107A2481999-04-29029 April 1999 Forwards 1998 Annual Radiological Environ Operating Rept for Salem & Hope Creek Generating Stations. Rept Summarizes Results of Radiological Environ Surveillance Program for 1998 ML18107A2511999-04-27027 April 1999 Submits 30-day Fuel Clad Temp Rept for Salem Generating Station,Units 1 & 2.Rept Describes Changes to Calculated Peak Clad Temp (PCT) for Plant Large & Small LOCA & Small Break LOCA Analyses ML18107A2631999-04-26026 April 1999 Provides Clarification on Licensing & Design Basis for 125 Vdc Battery Margins for Sgs & HCGS for Meeting SBO & Loca/ LOOP Loading Requirements ML18107A2371999-04-26026 April 1999 Forwards Corrected Response to NRC RAI Re Licensee Request for Change to TS Permissible Enrichment Values for New Fuel Storage.Incorrect Attachment Was Provided with Util 990412 Ltr to Nrc.Encl Supersedes 990412 Submittal 1999-09-08
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML18107A5601999-10-18018 October 1999 Submits 30-day Fuel Clad Temp Rept,Iaw 10CFR50.46.Rept Describes Changes to Calculated Peak Clad Temp (PCT) for Salem Generating Station Large & Small Break LOCA Analyses ML18107A5531999-10-0808 October 1999 Forwards Summary Rept of Plant Startup & Power Ascension Testing for Sgs,Unit 2 Cycle 11,per Requirements of TS 6.9.1.1 ML18107A5561999-10-0707 October 1999 Requests Relief Associated with Containment Examinations at Hope Creek & Salem Generating Stations.Attachment 1 Includes Proposed Alternatives & Supporting Justification for Relief Requests ML18107A5501999-10-0505 October 1999 Provides Current Status of Pse&G Actions Re GL 98-01, Y2K Readiness of Computer Sys at Npps, for Salem Nuclear Generating Station,Units 1 & 2 & Hope Creek Nuclear Generating Station ML18107A5521999-10-0505 October 1999 Encourages NRC to Support Abb Combustion Engineering Nuclear Power Request for Priority Review of Generic TR Re Crossflow Ultrasonic Flow Measurement Sys ML18107A5591999-10-0505 October 1999 Informs That Nj Dept of Environ Protection Has No Comments on License Change Request S99-07 for Sgs,Units 1 & 2 ML18107A5341999-09-22022 September 1999 Provides Data Re Operator Licensing Exam for Salem & Hope Creek Station,In Response to NRC Form 536 (7-1999) ML20212B3631999-09-14014 September 1999 Forwards Rev 13 to Salem - Hope Creek Security Plan,Iaw 10CFR50.54(p).Summary of Proposed Changes to Plan,Encl. Encl Withheld ML18107A5321999-09-13013 September 1999 Forwards Revised 10CFR50.92 Evaluation to Clarify Util Response to Question Number 1 Re Amend to Modify TS 3/4 8.1, AC Power Sources. ML18107A5331999-09-13013 September 1999 Provides Notification That PSEG Intends to Utilize ASME Code Case N-481 During Second ISI Interval at Sgs Units 1 & 2 ML18107A5351999-09-13013 September 1999 Informs That NRC Has Reviewed Pse&G Request Proposing to Modify TS Which Allow EDG to Be Operated for 24 Surveillance Test During Any Mode,Iaw 10CFR50.91(b) & Has No Comments ML18107A5231999-09-0808 September 1999 Forwards Salem Generating Station Semiannual Radioactive Effluent Release Rept for Jan-June 1999. Rept Summarizes Liquid & Gaseous Releases & Solid Waste Shipments from Salem Generating Station for Period of Jan-June 1999 ML18107A5221999-09-0808 September 1999 Requests Approval to Use ASME Code Case N546,which Provides Alternative Qualification Requirements That Allow Personnel Most Familiar with Walkdown of Plant Sys,Like License Operators to Perform VT-2 Examinations ML18107A4981999-08-26026 August 1999 Forwards Response to NOV That Resulted from Predecisional Enforcement Conference Conducted on 990624.Corrective Actions:Communications to Supervisors Reinforced Employee Right & Duty to Raise Nuclear Safety Issues ML18107A5181999-08-26026 August 1999 Forwards Ninety Day Rept for ISI Activities Conducted at Sgs,Unit 2 During Ninth Extended Outage & Tenth Refueling Outage.List of Encl,Provided ML18107A5061999-08-26026 August 1999 Provides First Feedback from Observation of NRC Insp Under Pilot Nuclear Power Plant Insp Program.Attached Are Completed Insp Feedback Forms for Procedure 71111, Attachment 21 & Procedure 71151 ML18107A5051999-08-24024 August 1999 Informs That Nj Dept of Environ Protection Bureau of Nuclear Engineering Has Reviewed Request IAW Requirements of 10CFR50.91(b) & Has No Comments Re Amend to FOL Change Request S99-02 to Modify TS Re Penetration Valves ML18107A4921999-08-23023 August 1999 Provides Suppl Info Re 971024 Amend Request to Modify TS 3/4 7.7, Auxiliary Bldg Exhaust Air Filtration Sys. Util Withdraws from Request All Proposed Changes Associated with Filter Testing,Per Issuance of GL 99-02 ML18107A4911999-08-20020 August 1999 Forwards Revised Plant Attribute Libraries for Salem & Hope Creek Generating Stations,Iaw 10CFR50,App E,Vi, Emergency Response Data Sys, 3.b.Changes Are Identified by Rev Bars ML18107A4831999-08-18018 August 1999 Submits Licensee Comments on NRC 990730 Ltr Which Provided Notification of Close Out of TAC Numbers MA0567 & MA0568 Re GL 92-01,Rev 1,Suppl 1 ML18107A4801999-08-13013 August 1999 Requests That Pse&Gs Contact in NUREG-0383, Directory of Compliance for Radioactive Matl Packages, Be Changed ML18107A4751999-08-0505 August 1999 Forwards Fitness for Duty Performance Data Rept for Six Month Period Ending 990630 ML20210M7571999-08-0404 August 1999 Forwards Response to Requesting Addl Info Re Status of Decommissioning Funding for Lgs,Pbaps & Sngs. Attachment Provides Restatement of Questions Followed by Response ML18107A4431999-07-0606 July 1999 Informs That Nj Dept of Environ Protection Bureau of Nuclear Engineering Has Reviewed Pse&G License Change Request S98-17 Re Permissible Enrichment Values for New Fuel Storage & Has No Comments ML18107A4181999-06-30030 June 1999 Submits Response to NRC Request for Info Re Y2K Readiness at Npps,Per GL 98-01,suppl 1.Disclosure Encl ML18107A4131999-06-25025 June 1999 Provides Further Clarification of Licensing & Design Basis for 125 Vdc Battery Margins for Sgs & HCGS for Meeting Station SBO & Loca/Loop Loading Requirements,Per Util 990426 Ltr & Discussion with NRC ML18107A4061999-06-21021 June 1999 Provides Supplemental Info to Proposed License Amend, Modifying TS 3/4 7.7, Auxiliary Bldg Exhaust Air Filtration Sys. Info Re Acceptance Criterion Discussed ML18107A3691999-06-11011 June 1999 Forwards Corrected Monthly Operating Rept for Apr 1999 for Salem Generating Station,Unit 1.Original Submittal Contained Typo for year-to-date Value for Numbers of Hours Generator Was on Line (Service Hours) ML18107A3641999-06-0404 June 1999 Requests Enforcement Discretion for TS 3/4.6.2.3 Re Containment Cooling Sys for Salem Generating Station,Unit 1 ML18107A3561999-06-0303 June 1999 Informs That Nj Dept of Environ Protection Bureau of Nuclear Engineering Has Reviewed Pse&G License Change Request S99-05 & Has No Comments ML18107A3611999-05-27027 May 1999 Forwards Responses to NRC 990301 & 990323 RAIs for Salem & Hope Creek Generating Stations Relating to GL 96-05 ML18107A3301999-05-24024 May 1999 Forwards Suppl Info Re GL 95-07, Pressure Locking & Thermal Binding of Safety Related Power Operated Gate Valves. Encl Contains Methodology Used in Determination of Pressure Locking Susceptibility of PORVs Block Valves ML18107A3291999-05-20020 May 1999 Forwards Redacted Response to NRC 990322 RAI Re Notification of Licensed Operator That Tested Positive for Alcohol. Attachment 2 Withheld,Per 10CFR2.790(a)(6) ML18107A3031999-05-18018 May 1999 Provides Summary of Changes to NRC Commitments That Have Been Made But Not Reported by Other Means,Iaw with NEI Process for Managing NRC Commitments ML18107A2891999-05-13013 May 1999 Forwards Rev 36 to Pse&G Nuclear Business Unit Emergency Plan. Rev 36 Incorporates Changes to Section 1-3,6 & 7 & 9-17.Attached Copy Includes All Sections of EP for Completeness ML18107A2951999-05-12012 May 1999 Submits SG Tube Plugging Rept,Per Plant TS 4.4.6.5.a.Total of 47 Tubes Were Plugged During SG Tube Insps,Which Were Completed During Plant Tenth RFO ML18107A2861999-05-11011 May 1999 Forwards Rev 0 to NFS-0174, COLR for Salem Unit 2 Cycle 11. COLR Rept Was Received by Util as Part of Reload SE ML18107A2481999-04-29029 April 1999 Forwards 1998 Annual Radiological Environ Operating Rept for Salem & Hope Creek Generating Stations. Rept Summarizes Results of Radiological Environ Surveillance Program for 1998 ML18107A2511999-04-27027 April 1999 Submits 30-day Fuel Clad Temp Rept for Salem Generating Station,Units 1 & 2.Rept Describes Changes to Calculated Peak Clad Temp (PCT) for Plant Large & Small LOCA & Small Break LOCA Analyses ML18107A2371999-04-26026 April 1999 Forwards Corrected Response to NRC RAI Re Licensee Request for Change to TS Permissible Enrichment Values for New Fuel Storage.Incorrect Attachment Was Provided with Util 990412 Ltr to Nrc.Encl Supersedes 990412 Submittal ML18107A2631999-04-26026 April 1999 Provides Clarification on Licensing & Design Basis for 125 Vdc Battery Margins for Sgs & HCGS for Meeting SBO & Loca/ LOOP Loading Requirements ML18107A2411999-04-22022 April 1999 Forwards Draft Revised Pages 4.1 & 4.2 of Nuclear Business Unit Emergency Plan for Hope Creek & Salem Generating Stations.Changes Are Noted in Italics ML18107A1841999-04-14014 April 1999 Forwards PSEG Annual Rept for 1998, & PECO Annual Rept for 1998. Stockholders Annual Rept of Each Owner & Cash Flow Statements Showing 1998 Actual & 1999 Projected Cash Flow with Explanation Encl ML18107A1981999-04-12012 April 1999 Responds to 990312 RAI Re Request for Change to TSs Permissible Enrichment Values for New Fuel Storage,Which Was Submitted on 990202 ML18107A1691999-04-12012 April 1999 Forwards Proprietary & non-proprietary Epips,Including Rev 17 to EPIP 807,rev 1 to NC.EP-EP.ZZ-0801(Q) & Rev 2 to NC.EP-EP.ZZ-0806(Q) & Revised EPIPs Table of Contents. Proprietary Info Withheld ML20205K4541999-04-0808 April 1999 Forwards Revised Info Re 990330 NRC Nuclear Power Reactor Licensee Financial Qualifications & Decommissioning Funding Assurance Status Rept ML18106B1491999-04-0505 April 1999 Forwards Drafts of Proposed Changes to Pages 4.1 & 4.2 of Emergency Plan,Which Are Contained on Page 4.2 & Noted in Italics & Underlined ML20205F8981999-03-31031 March 1999 Provides Info Re Status of Decommissioning Funding for LGS, Units 1 & 2,PBAPS,Units 1,2 & 3 & Sgs,Units 1 & 2,per Requirements of 10CFR50.75(f)(1) ML18106B1431999-03-31031 March 1999 Forwards Pse&G Rept on Financial Min Assurance for Period Ending 981231 for Hope Creek,Salem,Units 1 & 2 & Pbaps,Units 2 & 3,IAW 10CFR50.75 ML18107A2201999-03-30030 March 1999 Forwards Final Exercise Rept for 980303,full-participation Plume Exposure Pathway Exercise & 980505-07, full-participation Ingestion Pathway Exercise of Offsite Radiological Emergency Response for Salem & Hope Creek 1999-09-08
[Table view] Category:UTILITY TO NRC
MONTHYEARML18095A4881990-09-17017 September 1990 Requests Regional Waiver of Compliance from Tech Spec 3.6.2.3, Containment Cooling Sys. Waiver Requested in Order to Allow Replacement of Containment Fan Cooler Unit Motor #22 W/O Requiring Plant Shutdown ML18095A4901990-09-13013 September 1990 Provides Supplemental Info Applicable to Clarification of 10CFR50,App R Exemption Request Re Fire Suppression Sys for Panel 335,per NRC Request ML20059E6821990-09-0404 September 1990 Forwards Info Re Temporary Mod to Security Plan Concerning Protected Area.Info Withheld ML18095A4641990-08-31031 August 1990 Forwards Revised Response to NRC Bulletin 88-004 Re Potential pump-to-pump Interaction.Util Pursuing Permanent Solution to Issue & Will Implement Appropriate Permanent Field Change by End of Unit 1 10th Refueling Outage ML18095A4621990-08-31031 August 1990 Provides Revised Response to Generic Ltr 89-13, Svc Water Problems Affecting Safety-Related Equipment. Only HXs Exhibiting Unsatisfactory Test Results Will Be Inspected & Possibly Cleaned ML18095A4431990-08-30030 August 1990 Forwards Salem Generating Station Semiannual Radioactive Effluent Release Rept,Jan-June 1990 & Rev 6 to Odcm. ML18095A4531990-08-30030 August 1990 Forwards RERR-28, Salem Generating Station Semiannual Radioactive Effluent Release Rept for Jan-June 1990 & Revised Odcm.W/O Revised ODCM ML18095A4391990-08-29029 August 1990 Forwards Semiannual Rept Re fitness-for-duty Performance Data for 6-month Period Ending 900630,per 10CFR26.71(d).Rept Includes Testing Results,Random Testing Program Results & Confirmed Positive Tests for Specific Substances ML18095A4421990-08-28028 August 1990 Clarifies 900710 Request for Amends to Licenses DPR-70 & DPR-75,changing Sections I & M.Under Proposed Change,Section I Should Be Changed to Read Section 2.J for License DPR-70 & Section M Changed to Read Section 2.N for License DPR-75 ML20059B6611990-08-22022 August 1990 Confirms That 10 Anchor/Darling Model S350W Swing Check Valves Installed at Plant,Per NRC Bulletin 89-002.All 18 Valves Inspected & Retaining Block Studs Replaced W/Upgraded Matl.No Crack Noted on Any Studs Which Were Replaced ML20059C2861990-08-21021 August 1990 Provides Correction to 900810 Response to Request for Addl Info Re Util Request for Restatement of OL Expiration Dates ML18095A4151990-08-10010 August 1990 Forwards Response to Request for Addl Info Re Reinstatement of OL Expiration Dates Based on Original Issuance of Ols. Advises That Correct Expiration Date for OL Proposed to Be 200418 ML18095A4091990-08-0909 August 1990 Forwards Responses to NRC Comments Re Plant Simulator Certification for 10CFR55.45(b)(2),per 891228 Ltr ML18095A4061990-08-0808 August 1990 Forwards Corrected marked-up Pages for Tech Spec Table 3.3-11 Re Subcooling Margin Monitor & Reactor Vessel Level Instrumentation Sys,Per 900223 Ltr.Administrative Changes Made as Indicated ML18095A3861990-07-30030 July 1990 Forwards Listing of Station Blackout Major Audit Items Resolution Scope,Per Station Blackout Schedule Commitment ML18095A3661990-07-26026 July 1990 Forwards Decommissioning Repts for Hope Creek,Peach Bottom & Salem Nuclear Generating Stations ML18095A3761990-07-26026 July 1990 Forwards Decommissioning Repts & Certification of Financial Assurance for Plants ML18095A3721990-07-24024 July 1990 Forwards Rept & Certification of Financial Assurance for Decommissioning for Plants,Per 10CFR50.75 ML18095A3751990-07-18018 July 1990 Provides Status of Commitments Made to NRC by Util in 900109 Ltr Re NUREG-0737,Item II.D.1,per 900628 Telcon ML18095A3741990-07-18018 July 1990 Provides Supplemental Info Re Facility sub-cooling Margin Monitor ML18095A3611990-07-18018 July 1990 Responds to NRC Bulletin 90-001, Loss of Fill-Oil in Transmitters Mfg by Rosemount. ML18095A3621990-07-18018 July 1990 Forwards Corrected Tech Spec Page 3/4 3-5 for License Change Request 89-12 Submitted on 891227 & 900521 ML18095A3591990-07-13013 July 1990 Corrects Typo in 900702 Response to Generic Ltr 90-04 Re Schedule for Completion of Remaining Open Items ML18095A3471990-07-11011 July 1990 Responds to NRC 900611 Ltr Re Violations Noted in Insp Repts 50-272/90-14 & 50-311/90-14.Corrective Actions:Directive from Radiation Protection Mgt to All Radiation Protection Personnel Issued Re Control of Compliance Agreement Sheets ML18095A3451990-07-10010 July 1990 Forwards Addl Info Re License Change Request 89-03 Re Reactor Trip Sys Instrumentation ML18095A3461990-07-10010 July 1990 Responds to NRC 900608 Ltr Re Violations Noted in Insp Repts 50-272/90-12 & 50-311/90-12.Corrective Actions:Assessment of ECCS & Component Performance Undertaken & ECCS Flow Testing Procedure Upgraded to Address Human Factors ML18095A3491990-07-10010 July 1990 Forwards Jn Steinmetz of Westinghouse 900614 Ltr Re Reassessment of Util Response to Bulletin 88-002 ML18095A3481990-07-10010 July 1990 Submits Supplemental Rept Identifying Root Cause of Missed Commitment & Corrective Actions to Assure Future Compliance Re Implementation of Mods to Facility PASS ML18095A3441990-07-0909 July 1990 Provides Written Notification Re Change in Calculated Peak Clad Temp,Per 900606 Verbal Notification ML18095A3281990-07-0202 July 1990 Responds to NRC 900530 Ltr Re Violations Noted in Insp Repts 50-272/90-09 & 50-311/90-09.Corrective Actions:Util Intends to Use Nuclear Shift Supervisor as Procedure Reader & EOP, Rev 2 Under Development ML18095A3301990-07-0202 July 1990 Responds to Generic Ltr 90-04 Re Status of Licensee Implementation of Generic Safety Issues.Table Describing Status of Generic Safety Issue Implementation Encl ML18095A3391990-06-29029 June 1990 Forwards Correction to 890913 License Change Request 88-09, Consisting of Tech Spec Page 3/4 4-13 ML18095A3221990-06-28028 June 1990 Provides Supplemental Info Re 900223 Proposed Revs to Tech Specs for Reactor Vessel Level Instrumentation Sys.Tables 3.3-11a & 3.3-11b Should Be Combined Into Single Table ML18095A3231990-06-28028 June 1990 Responds to NRC 900518 Ltr Re Violations Noted in Insp Repts 50-272/90-10,50-311/90-10 & 50-354/90-07.Two Noncited Violations Disputed.Util fitness-for-duty Program Exceeds Part 26 Requirements for Positive Blood Alcohol Limits ML18095A3241990-06-28028 June 1990 Forwards Retyped Pages to 871224 License Change Request 87-15 & Modified,Per 900620 Ltr ML18095A3211990-06-26026 June 1990 Requests 30-day Extension Until 900730 to Provide Completion Schedule to Resolve Audit Findings Re Station Blackout ML18095A3161990-06-25025 June 1990 Forwards Supplemental Info Re Response to Generic Ltr 88-14. All Committed Actions Complete as of 900613 ML18095A3141990-06-25025 June 1990 Provides Schedule Change for Implementation of Control Room Mods.Schedule Modified to Address Overhead Annunciator Human Engineering Discrepancies During Phase III ML18095A3201990-06-25025 June 1990 Responds to NRC 900524 Ltr Re Violations Noted in Insp Repts 50-272/90-11 & 50-311/90-11.Corrective Actions:All Overdue Operations & Maint Procedure Files Reviewed for Outstanding Rev Requests & Procedure Upgrade Program Initiated ML18095A3001990-06-20020 June 1990 Provides Addl Info Re Application for Amend to Licenses DPR-70 & DPR-75 Concerning Turbine Valve Surveillance Interval,Per 900320 Request.Util Adding Direction to Personnel If Unnacceptable Flaws Found ML20043H6221990-06-20020 June 1990 Provides Supplemental Info Re NRC Bulletin 88-008 for Fifth Refueling Outage.Detailed Test Rept Being Prepared to Document Results of Each Individual Insp Re Insulation, Hangers & High Energy Break Areas ML18095A2991990-06-20020 June 1990 Forwards Westinghouse Affidavit Supporting 900412 Request for Withholding Proprietary Info from Public Disclosure Per 10CFR2.790 ML18095A2721990-06-0808 June 1990 Responds to NRC 900329 Ltr Re Weaknesses Noted in Insp Repts 50-272/90-80 & 50-311/90-80.Corrective Actions:Fire Doors Placed on Blanket Preventive Maint Work Order & Damaged Fire Doors Will Be Repaired Immediately ML18095A2711990-06-0606 June 1990 Submits Info in Support of 900522 Verbal Request for Relief from Requirements of ASME Section XI ML18095A2611990-06-0101 June 1990 Forwards Corrected Operating Data Rept, Page for Apr 1990 Monthly Operating Rept ML18095A2521990-06-0101 June 1990 Forwards Application in Support of Request for Renewal of NJPDES Permit NJ0005622,per Requirements of Subsection 3.2 of Plant Environ Protection Plan,Nonradiological ML18095A2591990-06-0101 June 1990 Forwards Corrected Unit Shutdown & Power Reductions, Page for Apr 1990 Monthly Operating Rept ML18095A2411990-05-30030 May 1990 Submits Special Rept 90-4 Addressing Steam Generator Tube Plugged During Fifth Refueling Outage.Plugging Completed on 900516.Cause of Tube Degradation Attributed to Normal Wear Due to Erosion/Corrosion Factors ML18095A2431990-05-30030 May 1990 Informs of Util Plans Re Facility Cycle 6 Reload Core, Expected to Achieve Burnup of 16600 Mwd/Mtu.All Postulated Events within Allowable Limits Based on Review of Basis of Cycle 6 Reload Analysis & Westinghouse SER ML18095A2531990-05-29029 May 1990 Provides Addl Info Re End of Life Moderator Temp Coefficient.Feedback Used in Steam Line Break Has No Relationship to Full Power Moderator Density Coefficient 1990-09-04
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Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge. New Jersey 08038 Nuclear Department January 30, 1984 Director of Nuclear Reactor Regulation
- u. s. Nuclear Regulatory Commission 7920 Nor(olk Avenue Bethesda, MD 20014 Attention: Mr. Steven Varga, Chief Operating Reactors Branch 1 Division of Licensing
Dear Mr. Varga:
SAFETY PARA.METER DISPLAY SYSTEM SAFETY ANALYSIS AND IMPLEMENTATION PLAN REQUIREMENTS FOR EMERGENCY RESPONSE CAPABILITY SALEM GENERATING STATION NO. 1 AND 2 UNITS DOCKET NOS. 50-272 AND 50-311 PSE&G hereby submits its Safety Analysis and implementation plan for the Safety Parameter Displ~y System in accordance with the requirements of Generic Letter 82-33, Requirements for Emergency Response Capability.
Should you have any questions, please do not _hesitate tq contact us.
Sincerely, Manager - Nuclear Licensing and Regulation
~/1/
RSP: jab cc: Mr. Donald c. Fischer Licensing Project Manager Mr. James Linville Senior Resident Inspector 8402070374 84020i ~~
PDR ADOCK 05000272 .
F * *.* PDR . .
95 2
SAFETY ANALYSIS FOR SPDS PARAMETERS Functjonal Description The Safety Parameter Display System will serve as an aid to the control room personnel during abnormal and emergency conditions in determining the safety status of the plant. It will also function as an operator aid during normal operation by monitoring other parameters or graphic displays that are determined to be important to the operator for maintaining safe operation of the plant. The displays will serve to concentrate a set of plant parameters to aid in assessing plant safety status without surveying the entire control room. The primary display will provide an overview of plant conditions and the secondary displays will provide more detailed information on specific plant systems and equipment.
System Description
General The Safety Parameter Display System will be a redundant computer system with CRTs located in the TSC, EOF and Units 1 and 2 Control Room. This system is independent of the Plant Computer. The major components are as follows:
- three lE multiplexer cabinets per unit
- two NON-lE multiplexer cabinets per unit
- two SEL 32/8705. Central Processing Units
- two color CRT/keyboards per unit control room
- one line printer per unit
- four color CRT/keyboards for TSC
- two color CRT/keyboards for EOF
- one video copier for EOF The data concentrators and the two Central Processing Units will be shared by both Units. The CRT/keyboard assemblies and video copiers in the TSC and EOF will not be dedicated to any one unit. Attachment 1 gives a general layout of the above mentioned components and other peripheral equipment.
Data Acquisition Subsystem Each multiplexer in the subsystem functions as an independent unit utilizing a 16 bit microprocessor. Complete isolation of field inputs is maintained by use of fiber optic communication links to the rest of the system. Signal conditioning and buffers necessary to isolate the P-250 process computer is included.
DR2 1/4
Computer Subsystem The computer subsystem utilizes two SEL 32/8705 processors in a fully redundant configuration. Each CPU acquires and processes the data from all multiplexers and maintains its own data base. One CPU is designated as the primary unit and handles all display subsystem interfacing. This allows the other CPU to be utilized for development work while maintaining a hot standby condition for smooth fail-over. A full duplex RS-232 "watchdog" communication channel is provided so that the CPUs can monitor each other. All communication with equipment outside the computer environs is via fiber optic links or standard RS-232 modems.
Display Subsystem The display subsystem comprises high resolution color graphics CRTs, color video hard copy units and printers for data output. The IDT #2200 color graphics CRTs are used and full graphics editing capabilities are provided for building and modifying color displays.
Isolation of Class lE Signals At the output of the multiplexer cabinets, the corrununication link to the computer will be by fiber optic cables which will perform an isolation function. All class lE signals will be isolated prior to entering the multiplexer cabinets. These isolators will be qualified based on their function.
Availability The Host processor/display system will be designed to achieve an availability of 99.0% under the following conditions:
- All of the ERF on-line functions are executing without degradation and the following minimum complement of hardware is operational.
- 1. One of the two CPUs with all of its main memory and its prograrruner's I/O device, and with sufficient hardware in the CPU interfaces to communicate with all of the field multiplexers communication circuits at the specified scan rates.
- 2. One of the two auxiliary memories.
- 3. One printer in either unit control room.
- 4. One of the two unit CRTs in the control room, one of the two unit CRTs in the TSC and one of the two CRTs in the EOF excluding the moderns and phone lines .
- Each multiplexer will be designed to achieve the availability under the following conditions:
- 1. The multiplexer is considered available unless:
DR2 2/4
- a. Any function is lost for all points of a single type, or
- b. More than one input card of the same type fails, or
- c. One input card of each type fails.
Human Factors The Safety Parameter Display System display will be designed to incorporate accepted Human Factor Principles. The following Human Factors Principles references will be used:
- "Human Engineering Principles for Control Room Design Review", Section 3.7, published by the Nuclear Utility Task Action Committee.
Parameter Selection PSE&G has selected a total of sixty-one parameters to be displayed on the SPDS using the parameters listed in Regulatory Guide 1.97 as a guideline. These parameters are listed in Attachment 2.
The basis of this safety analysis is the Critical Safety Function Status Trees. The Critical Safety Functions were identified and Status Trees developed by PSE&G based on the Westinghouse Emergency Response Guidelines, Revision 1. The Status Trees and the procedures associated with them are contained within the Emergency Operating Procedure Set, which was also developed based on the Westinghouse Owners Group Emergency Response Guidelines. For any transient or accident condition, the Emergency Operating Procedures will direct the operator to monitor the Status Trees. Operator training also addresses the use of the Status Trees during transient or accident conditions. The following is a list of the six Critical Safety Functions for Salem Generating Station:
- 1. Shutdown Margin
- 2. Core Cooling
- 3. Heat Sink
- 4. Thermal Shock
- 5. Containment Environment
- 6. Coolant Inventory is "The Critical Safety Function Status Trees Bas is Document", and Attachment 4 is "The Emergency Operating Procedure EOP-CFST-1 and Status Trees". These documents are in draft form. They will be made final when the Emergency Operating Procedures are implemented.
DR2 3/4
The "Critical Safety Function Status Trees Basis Document" basically lists the Critical Safety Functions and describes the use and organization of the Status Trees. It also explains how the Status Trees are used in evaluating the Critical Safety Functions. The "Emergency Operating Procedure EOP-CFST-1 and Status Trees" document shows graphically the Status Tree for each Critical Safety Function and explains the significance of the colors used.
Of the total parameters that were selected for the Safety Parameter Display System, fifteen are utilized in satisfying the Critical Safety Functions. The parameters are as follows:
- 1. Neutron Flux
- 2. RCS Cold Leg Water Temperature
- 3. RCS Pressure
- 4. Core Exit Temperature
- 5. Reactor Vessel Level
- 6. Degrees of Subcooling
- 7. Containment Sump Water Level
- 8. Containment Pressure
- 9. Containment Area Radiation
- 10. Reactor Coolant Pump Status
- 11. Pressurizer Level
- 12. Steam Generator Level
- 13. Steam Generator Pressure
- 14. Auxiliary Feedwater Flow
- 15. RCS Loop Average Temperature.
The other forty-six parameters will be included in the SPDS data base because they have been determined to be important in aiding the operator in determining the status of the plant. Most of these parameters will be used in developing graphic displays which will be used as an operator aid.
DR2 4/4
Attachment 1
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- - -- - - - - - - -. - - - - - - -r-- - -- - --- - -- _,.
f 61U CONTROL ROOM UNIT~ I CONTROL ROOM UNIT o 2 SPARE I RMS COLOR COLOR COLOR COLOll CRT/ CRT/ CRT/ CATI - *so KEY* KEY* KEY* KEY*
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PSE&G Emergency Response Fdciliries..
11 /22/83.
ATTACHMENT 2 SALEM GENERATING STATION UNITS 1 AND 2 SAFETY PARAMETER DISPLAY SYSTEM PARAMETERS
- 1. Neutron Flux - Source, Power, and Intermediate Ranges, Start-up Rate.
- 2. Rod Control Positions
- 3. RCS Soluble Boron Concentration
- 4. RCS Cold Leg Water Temperature
- 5. RCS Hot Leg Water Temperature
- 6. RCS Pressure
- 7. Core Exit Temperature
- 8. Coolant Level in Reactor
- 9. Degrees of Subcooling (calculated)
- 10. Containment Sump Water Level
- 11. Containment Pressure (Wide and Narrow Range)
- 12. Containment Isolation Valve Position (excluding check valves)
- 13. Containment Area Radiation
- 14. Noble Gas Effluent Radioactivity from Condenser Air Removal System.
- 15. Containment Hydrogen Concentration
- 16. Containment Effluent Radioactivity (Plant Vent)
- 17. Radiation Exposure Rate (Fuel Storage Room, Charging Pump Room, Fuel Handling Building, and Mechanical Penetration Area)
- 18. Radiation Exposure Rate (Electrical Penetration Area)
- 19. RHR System Flow
- 20. RHR Heat Exchanger Outlet Temperature DFl.l 1/03
ATTACHMENT 2 (Continued)
- 21. Accumulator Tank Level and Pressure
- 22. Accumulator Isolation Valve Position
- 23. Boric Acid Charging Flow
- 24. Flow in HPI System (Charging Pumps Discharge)
- 25. Flow in LPI System (Safety Inspection Pumps Discharge)
- 26. Refueling Water Storage Tank Level
- 27. Reactor Coolant Pump Status 28~ Primary System Safety Relief Valve Position
- 29. Pressurizer Level
- 30. Pressurizer Heater Status
- 31. Pressurizer Relief Tank Level
- 32. Pressurizer Relief Tank Temperature
- 33. Pressurizer Relief Tank Pressure
- 34. Steam Generator Level
- 35. Steam Generator Pressure
- 36. Main Steam Flow
- 37. Main Feedwater Flow
- 38. Auxiliary Feedwater Flow
- 39. Auxiliary Feedwater Storage Tank Level
- 40. Containment Spray Flow Additive Rate
- 41. Heat Removal by the Containment Fan Heat Removal System
- 42. Containment Atmosphere Temperature
- 43. Letdown Flow
- 44. Volume Control Tank Level
- 45. Component Cooling Water Temperature to ESF System DFl.l 2/03
l ATTACHMENT 2 (Continued)
- 46. Component Cooling Water Flow to ESF System
- 47. High Level Radioactive Liquid Tank Level
- 48. Radioactive Gas Hold Up Tank Pressure
- 49. Control Room Emergency Ventilation Damper Position
- 50. Auxiliary Building Emergency Damper Position
- 51. Fuel Handling Building Emergency Damper Position
- 52. Status of Stanby Power and Other Emergency Energy Sources Important to safety.
- 53. Control Air
- 54. Main Steam Radiation
- 55. Wind Direction
- 56. Wind Speed
- 57. Estimation of Atmospheric Stability
- 58. Steam Generator Blowdown Radiation
- 59. Condenser Availability (Condenser Vacuum and Circulator Amperes)
- 60. RCS heat up/cool down rate (Average Loop Temperature)
- 61. Main Steam Isolation Valve Position DFl.l 3/03
~ Attachment 3 CRITICAL SAFETY FUNCTION STATUS TREES (CFST)
BASIS DOCUMENT
1.0 INTRODUCTION
The Critical Safety Function Status Trees ares used to monitor specific plant conditions while the Emergency Operating Procedures are in use. The conditions that are monitored relate directly to the barriers to release of fission products to the environment. These barriers are the fuel matrix and cladding, RCS pressure boundary and Containment.
Protection and Control Systems, augmented by trained operator response to annunciator alarms and backed by Technical Specifications, serve to ensure that small departures from preferred operating conditions are rectified before any challenge to the Critical Safety Functions develops.
Failures in system components and the Protection System can create conditions which threaten the integrity of one or more barriers.
The Status Trees determine when these challenges are present and designate Functional Restoration Procedures to use to correct the condition.
2.0 ORGANIZATION The six Critical Safety Functions evaluated by the Status Trees are necessary to maintain the integrity of the three barriers to fission product release.
The first barrier is the fuel matrix and clad. Three conditions are necessary to maintain fuel integrity during accident conditions:
- 1. Maintenance of subcriticality to prevent power generation and excessive fuel temperatures.
- 2. Maintenance of adequate Reactor Coolant inventory to allow Core Cooling.
- 3. Maintenance of Core Cooling to remove core decay heat.
The second barrier is the RCS pressure boundary. Three conditions necessary to maintain RCS integrity are:
- 1. Maintenance of the secondary Heat Sink to provide heat removal from the RCS.
- 2. Prevention of Thermal Shock to the Reactor Vessel which could lead to vessel brittle £racture.
Salem Unit 1 Draft A Rev.
9 Attachment 3 CFST Basis
- 3. Control of Reactor Coolant inventory to prevent filling the pressurizer and loss of RCS pressure control.
The third barrier is the Containment. The Containment Environment (pressure) is controlled to prevent overpressurization of the Containment structure.
The six Status Trees relate to the above conditions as shown in the table below.
Critical Safety Function Status Tree Functional Restoration Subcriticality 3.1 Shutdown FRSM Margin Core Cooling 3.2 Core FRCC Cooling Secondary Heat Sink 3.3 Heat Sink FRHS Thermal Shock 3.4 Thermal FRTS Shock Containment 3.5 Containment FRCE Environment Reactor Coolant Inventory 3.6 Coolant FRCI Inventory Also shown is the Functional Restoration block used by each Status Tree to restore threatened Critical Safety Functions.
3.0 CFST USE 3.1 Status Tree Scanning The Status Trees are used by an SRO licensed individual in the Control Room to monitor Critical Safety Functions while the Desk Operator and Control Operator respond to a unit trip or Safety Injection with the Emergency Operating Procedures.
Status Tree scanning begins when EOP-TRIP-1, "Reactor Trip or Safety Injection" is departed. EOP-TRIP-1 also directs Status Tree use if the SI cannot be terminated but the problem has not been diagnosed. The Status Trees are evaluated in order while the fault specific EOP is conducted. The Status Trees are scanned continuously until all Critical Safety Functions are satisfied. The Status Trees are then scanned periodically until the event is terminated.
Salem Unit 2 Draft A Rev.
Attachment 3 CFST Basis 3.2 Functional Restoration Priorities Priority of a Status Tree designated Functional _
Restoration is determined by the color of the condition and the order of the Status Trees. Red is the highest priority condition, followed by orange and yellow.
Green is used to signify that a Critical Safety Function is satisfied. The Status Trees are arranged in descending order of priority.
Color is considered first, then order. Thus a Red condition on Status Tree 3.1 would have priority over all other challenges to Critical Safety Functions.
Likewise an Orange condition on Status Tree 3.5 would have priority over a Yellow condition on any Status Tree.
3.3 Response to an Unsatisfied CSF When a CSF is evaluated as un~atisfied a Functional Restoration is identified. Performing the Function Restoration removes the challenge to the CSF.
A Red condition requires immediate suspension of the EOP in use. The current step is noted and the page marked
.for later reference. The Functional Restoration is initiated and continues until the challenge is removed.
The EOP in effect is then resumed unless an additional Red condition is present. Note that if a Red condition is identified while a Functional Restoration is in progress for a lower priority Red condition, the lower priority procedure is suspended and the higher priority Functional Restoration initiated.
When an Orange condition is encountered, note the associated Functional Restoration and continue tree evaluation. When the current pass through the Status Trees is complete, initiate the Orange related Functional Restorations in order of importance.
A Yellow condition is a slight challenge to a CSF and could lead to a serious challenge if not corrected.
Initiate Yellow condition Functional Restorations when practical.
4.0 REFERENCES
4.1 WOG Guideline F-0 "Critical Safety Function Status Trees" Rev HP-Basic.
END OF PROCEDURE FINAL PAGE Salem Unit 3 Draft A Rev.
Attachment' 4 EMERGENCY OPERATING PROCEDURE EOP-CFST-1 CRITICAL SAFETY FUNCTION STATUS TREES 1.0 ENTRY CONDITIONS 1.1 EOP-TRIP-1.
2.0 STATUS TREE USAGE 2.1 Initiate CRT tests 23 and 41 to facilitate monitoring CORE EXIT TC's. If PRODAC 250 not available, then direct Performance Department to perform Emergency Surveillance Procedure PD-14.3.010, "Extended Range Reading of Incore Thermocouples" and establish contact with operator monitoring CSFT.
2.2 START Status Tree evaluation after departing EOP-TRIP-1, "Reactor Trip or Safety Injection."
2.3 IF a Red is encountered, immediately go to the designated functional restoration procedure. The EOP in effect is resumed when the Function Restoration is completed unless otherwise directed.
2.4 IF an Orange is encountered, note the designated functional restoration procedure and continue status tree evaluation. When the current pass through the trees is complete, initiate the designated procedures in order of importance unless otherwise directed.
2.5 IF a Yellow is encountered, note the nature of the deficiency and continue status tree evaluation. When practical, initiate the designated procedures unless otherwise directed.
2.6 The Status Trees are arranged in descending order of importance. Consider the condition color and tnen the procedure order to determine the priority, among a group of Functional Restorations.
2.7 Red conditions require suspension of the procedure in effect. Orange and Yellow condition Functional Restorations take precedence over any conflicting procedure steps in the EOP in effect.
Salem Unit 1 1 DRAFT *c
~ Attachment 4 EOP-CFST-1 3.0 Critical Safety Function Status Trees 3.1 Shutdown Margin.
3.2 Core Cooling.
3.3 Heat Sink.
3.4 Thermal Shock.
3.5 Containment Environment.
3.6 Coolant Inventory.
END OF PROCEDURE FINAL PAGE Salem Unit 1 2 DRAFT C
e Attachment 4 CRITICAL SAFETY FUNCTION STATUS TREES
[
('.,() 'ID FRSM-1 C-0 TO FRSM-1 C-0 TO FRSM-2 POV\!$ RANGE N LESS THAN 5%
y INTERMEDIATE RANGE SUR N MORE NEGATIVE THAN -.2 DPM INTERMEDIA'IE y RANGE SUR N ZERO OR NEGATIVE CS?
SAT SOURCE RANGE N ENERGIZED y
GO TO FRSM-2 SOURCE RANGE N Si:JB. NEGATIVE OR ZERO y
CSF SAT
{L_3_:11~_co_RE_*_c_o_o_L_IN_G________~ltt GO TO FRCC-1 GO TO CORE EXIT N FRCC-1 Tes LESJ;? -
THAN 1200°F IS NARRCW N y_ RANGE GREATER THAN 40%
y CORE EXIT Tes LESS THAN 700° H
~ GO TO y " " " " FReC-2 AT LEAST N C-0 TO ONE RCP FRCC-2 RUNNIN3 y IS NARROW GE GREATER N
RCS
'IRAN 40 %
SUBCOOLING GREATER THAN y 10 F y
GO TO FRCC-3 00 TO FRCC-2 RVLIS
'WIDE RANGE N GREATER TP.AN 44% 4 RCP 30% 3 RCP 20% 2 RCP y -
13% 1 RCP
~C-OTO
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-CSF SAT
3 - HEAT SINK le GO T0 RHS-1
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~TER IN FI..av TO SGs GREATER
. THAN 2-2E04lbm I y NARRCW RANGE GO TO LEVEL GREATER IN FRHS-2 THAN io/. IN AT LEAST ONE PRESSURE LESS SG _I y THAN 1125_ PSI1 IN ALL SGs C-0 TO FRHS-4 i.\JARIDW RANGE I LEVEL LESS '.N T.:fAN 6 7% IN ALL SGs
.Y C-0 TO FRHS-3 PRESS LESS N THAN 1070 .
PSIG IN ALL SGs y
lL-3-~
___ __L__s_Ho_c_K________
TH_E_RMA ~l4t
[ 3 .CONTAINMENT ENVIORNMENTJ
PRESSURE LESS N
---i THAN 47 PSIG y
,.....__________ ...,..:-----------------------~~~
CONTAINMENT PRESSURE LESS N THAN 23.5 PSIG y
C-0 TO FRCE-2 CONI'AilMENT N SUMP LEVEL LESS THAN
.MAX* FLCX>D LEVEL y GO TO FRCE-3 CONTAINMENT N RADIATION LESS THAN R-44 ALARm y CSF SAT
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___N_T_o_R_Y__ ~__I~
GO TO.
FRCI-3 RVLIS N INDICATES UPPER HEAD
. FULL y C-0 TO FRCI-1 PRESSURIZER
.LEVEL LESS N THAN 92%*
C-0 TO y
FRCT-2 THAN 17%
y C-0 TO
- 5'RCI-3 RVLIS INDICATES N L-----i UPPER HEAD FULL y CSF SAT
SAFETY PARAMETER DISPLAY SYSTEM IMPLEMENTATION PLAN -
- 1. SCHEDULE
- a. DESIGN PHASE 9/84
- b. DEVELOPMENT PHASE 9/85
- c. INSTALLATION PHASE 12/85
- d. FIELD TE;STING, OPERATION AND ACCEPTANCE PHASE 5/86
- e. FULLY OPERATIONAL 12/86
- 2. VERIFICATION AND VALIDATION PLAN Verification and validation will be conducted by the computer system vendor. The program will be developed using NSAC-39 "Verification and Validation for Safety Parameter Display Systems" as guidance and will address the traceability of requirements of hardware and software and provide independent review. The V & V activities will be performed by a team which is completely independent of the development effort.
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