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| | document type = SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES, TEXT-SAFETY REPORT | | | document type = SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES, TEXT-SAFETY REPORT |
| | page count = 6 | | | page count = 6 |
| | | project = TAC:59440 |
| | | stage = Approval |
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| 4 taken are supplied for the new trip function (see Tables 1 and 2). Instru-mentation surveillance intervals are also designated for this function (see Table 3). These proposals are acceptable to us, as discussed above. | | 4 taken are supplied for the new trip function (see Tables 1 and 2). Instru-mentation surveillance intervals are also designated for this function (see Table 3). These proposals are acceptable to us, as discussed above. |
| In the original submittal, MP&L proposed to delete ECCS response times for low pressure systems. Subsequently, in a letter dated September 25, 1985, MD&L proposedtouse"f29 seconds"asresponsetimesforbothLPCIandLPCSinjection valves. This is acceptable to us because this response time is within the value used in analyses of loss of coolant accidents. | | In the original submittal, MP&L proposed to delete ECCS response times for low pressure systems. Subsequently, in a {{letter dated|date=September 25, 1985|text=letter dated September 25, 1985}}, MD&L proposedtouse"f29 seconds"asresponsetimesforbothLPCIandLPCSinjection valves. This is acceptable to us because this response time is within the value used in analyses of loss of coolant accidents. |
| One issue that has not been resolved is the interlock setpoint for the high pressure to low pressure interface valve. The setpoint 534 psig, is above the design pressure of the low pressure system, 500 psig. The ASME Code, dated 1977, which was the first Code edition to address such interlocks, required the interlocks to prevent the pressure from exceeding the design pressure of the low pressure side. Section 6.3.4 of GGNS SSER's (June 1982), | | One issue that has not been resolved is the interlock setpoint for the high pressure to low pressure interface valve. The setpoint 534 psig, is above the design pressure of the low pressure system, 500 psig. The ASME Code, dated 1977, which was the first Code edition to address such interlocks, required the interlocks to prevent the pressure from exceeding the design pressure of the low pressure side. Section 6.3.4 of GGNS SSER's (June 1982), |
| states that the valves in question should be " interlocked to prevent opening unless the reactor vessel pressure is lower than the design pressure of the ECCSs involved." From discussions with the licensee, we understand that main-taining the setpoint below the design pressure will result in increased peak cladding temperature for. the postulated loss of accident from 2151*F to 2157*F, which is still less than the limiting value of 2200*F required by 10 CFR 50.46. The staff concludes that exceeding the Code allowable design pressure for the interlocks is unacceptable because the Code position can be | | states that the valves in question should be " interlocked to prevent opening unless the reactor vessel pressure is lower than the design pressure of the ECCSs involved." From discussions with the licensee, we understand that main-taining the setpoint below the design pressure will result in increased peak cladding temperature for. the postulated loss of accident from 2151*F to 2157*F, which is still less than the limiting value of 2200*F required by 10 CFR 50.46. The staff concludes that exceeding the Code allowable design pressure for the interlocks is unacceptable because the Code position can be |
Letter Sequence Approval |
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Initiation
- Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request
- Acceptance...
- Supplement, Supplement, Supplement
Results
Other: AECM-82-492, Requests That License Condition 2.C.(44)(i)(a) Be Revised to Require Submittal Re Automatic Depressurization Sys Logic Mods by 821215.Previous Submittal Date of 821029 Erroneously Requested Due to Revised BWR Owners Group Schedule, AECM-82-603, Comments on BWROG-8260, NUREG-0737,Item II.K.3.18,Mod of Automatic Depressurization Sys Logic. Option 4,providing Bypass Timer Delay,Will Assure Adequate Core Cooling During Worst Case Conditions, AECM-85-0014, Forwards Addl Info,Requested in ,Re Concerns Involving Transfer of Control to Remote Shutdown Panel Resulting in Change in Operating Status of Load That May Be Detrimental to Safe Shutdown, AECM-85-0341, Forwards Addl Info Re ECCS Performance Analysis Which Supports Increase in Peak Cladding Temp of 51 F Compared to Current FSAR Analysis,Per Request.Fsar Changes Will Be Made If Design Change Implemented, AECM-85-0396, Advises That Tech Spec Changes Contained in Amends 4 & 6 to License NPF-29 Will Be Effective Before Startup After First Refueling Outage,Per 850918 & 1021 Requests, AECM-86-0006, Forwards Technical Justification for Proposal to Delete Peak Recording Accelerographs Currently Installed on Reactor Piping Locations,Per Section 16.3.1 of Sser 6 (NUREG-0831).Response by 860228 Requested.Fee Paid, AECM-86-0234, Forwards Clarification Re Proposed Mod to License Condition 2.C(20) Concerning Standby Svc Water Sys, AECM-86-0329, Submits List of Affected Tech Spec Pages,Effective Dates & Associated Amend Numbers,In Response to NRC 860827 & 1006 Requests, AECM-86-0332, Advises of Effective Dates for Amends 6 & 8 to License NPF-29 for Listed Tech Specs,In Response to 851021 & 860827 Requests, ML20133P171, ML20154R546, ML20155J934, ML20209D136, ML20209D152
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MONTHYEARAECM-82-492, Requests That License Condition 2.C.(44)(i)(a) Be Revised to Require Submittal Re Automatic Depressurization Sys Logic Mods by 821215.Previous Submittal Date of 821029 Erroneously Requested Due to Revised BWR Owners Group Schedule1982-10-21021 October 1982 Requests That License Condition 2.C.(44)(i)(a) Be Revised to Require Submittal Re Automatic Depressurization Sys Logic Mods by 821215.Previous Submittal Date of 821029 Erroneously Requested Due to Revised BWR Owners Group Schedule Project stage: Other AECM-82-603, Comments on BWROG-8260, NUREG-0737,Item II.K.3.18,Mod of Automatic Depressurization Sys Logic. Option 4,providing Bypass Timer Delay,Will Assure Adequate Core Cooling During Worst Case Conditions1982-12-14014 December 1982 Comments on BWROG-8260, NUREG-0737,Item II.K.3.18,Mod of Automatic Depressurization Sys Logic. Option 4,providing Bypass Timer Delay,Will Assure Adequate Core Cooling During Worst Case Conditions Project stage: Other AECM-85-0014, Forwards Addl Info,Requested in ,Re Concerns Involving Transfer of Control to Remote Shutdown Panel Resulting in Change in Operating Status of Load That May Be Detrimental to Safe Shutdown1985-01-17017 January 1985 Forwards Addl Info,Requested in ,Re Concerns Involving Transfer of Control to Remote Shutdown Panel Resulting in Change in Operating Status of Load That May Be Detrimental to Safe Shutdown Project stage: Other ML20151L0531985-07-12012 July 1985 Application for Amend to License NPF-29,revising Tech Specs to Reflect Organizational Changes & Addition of Two Primary Containment Penetration Conductor Overcurrent Protection Circuit Breakers Project stage: Request ML20133P1711985-08-12012 August 1985 Proposed Tech Specs,Reflecting Design Changes & Editorial Clarification Re Leak Rate Testing of RHR to Head Spray Isolation Valve & ECCS Actuation Instrumentation Project stage: Other AECM-85-0228, Application for Amend to License NPF-29,changing Tech Specs to Reflect Design Changes & Editorial Clarifications.Fee Paid1985-08-12012 August 1985 Application for Amend to License NPF-29,changing Tech Specs to Reflect Design Changes & Editorial Clarifications.Fee Paid Project stage: Request AECM-85-0248, Withdraws 850712 Proposed Change to License NPF-29,Item 7,to Reduce Burden of Requiring Rev to OL W/Each Tech Spec Amend. Change Inconsistent W/Past NRC Practices & Procedures1985-08-12012 August 1985 Withdraws 850712 Proposed Change to License NPF-29,Item 7,to Reduce Burden of Requiring Rev to OL W/Each Tech Spec Amend. Change Inconsistent W/Past NRC Practices & Procedures Project stage: Request ML20133A7611985-09-25025 September 1985 Proposed Changes to Tech Specs 6.3.1 & 6.4.1,providing Justification for Exemption from Requirements for Senior Reactor Operator Qualifications Project stage: Request AECM-85-0304, Application for Amend to License NPF-29,supplementing 850812,23 & 30 Submittals,To Provide Justification for Exemption from Tech Spec 6.3.1 & 6.4.1 Requirements for Senior Reactor Operator Qualifications1985-09-25025 September 1985 Application for Amend to License NPF-29,supplementing 850812,23 & 30 Submittals,To Provide Justification for Exemption from Tech Spec 6.3.1 & 6.4.1 Requirements for Senior Reactor Operator Qualifications Project stage: Supplement AECM-85-0314, Provides Addl Info to Support Proposed Change to Tech Specs 4.1.3.3.b.1.b.GE Info Ltr Entitled Hydraulic Control Unit Accumulator Pressure Switches, to Support Proposed Change,Encl1985-09-26026 September 1985 Provides Addl Info to Support Proposed Change to Tech Specs 4.1.3.3.b.1.b.GE Info Ltr Entitled Hydraulic Control Unit Accumulator Pressure Switches, to Support Proposed Change,Encl Project stage: Request AECM-85-0305, Forwards Resumes of Two Individuals Appointed as Supervisors to Organization Radiation Control Function,Per Request for Addl Info Re 850712 Proposed Change to Tech Spec Figure 6.2.2.-11985-10-0202 October 1985 Forwards Resumes of Two Individuals Appointed as Supervisors to Organization Radiation Control Function,Per Request for Addl Info Re 850712 Proposed Change to Tech Spec Figure 6.2.2.-1 Project stage: Request AECM-85-0306, Responds to Request for Addl Info Re 850812 Application for Amend to License NPF-29.Prompt Resolution of Questions & Approval of Previously Submitted Tech Specs Needed for Outage Scheduled to Start on 8510121985-10-0505 October 1985 Responds to Request for Addl Info Re 850812 Application for Amend to License NPF-29.Prompt Resolution of Questions & Approval of Previously Submitted Tech Specs Needed for Outage Scheduled to Start on 851012 Project stage: Request AECM-85-0341, Forwards Addl Info Re ECCS Performance Analysis Which Supports Increase in Peak Cladding Temp of 51 F Compared to Current FSAR Analysis,Per Request.Fsar Changes Will Be Made If Design Change Implemented1985-10-22022 October 1985 Forwards Addl Info Re ECCS Performance Analysis Which Supports Increase in Peak Cladding Temp of 51 F Compared to Current FSAR Analysis,Per Request.Fsar Changes Will Be Made If Design Change Implemented Project stage: Other ML20137Q8901985-11-22022 November 1985 Forwards Safety Evaluation Re 850812 Request for Amend to License NPF-29,adding Pressure Permissive Interlocks to Control Circuits for Low Pressure ECCS Injection Valves,Item 13 of 850812 Request Should Be Revised Project stage: Approval ML20137Q8971985-11-22022 November 1985 Safety Evaluation Re Low Pressure ECCS Sys Pressure Permissive Interlocks.Exceeding ASME Code Allowable Design Pressure for Interlocks Unacceptable Project stage: Approval AECM-85-0396, Advises That Tech Spec Changes Contained in Amends 4 & 6 to License NPF-29 Will Be Effective Before Startup After First Refueling Outage,Per 850918 & 1021 Requests1985-12-0606 December 1985 Advises That Tech Spec Changes Contained in Amends 4 & 6 to License NPF-29 Will Be Effective Before Startup After First Refueling Outage,Per 850918 & 1021 Requests Project stage: Other AECM-86-0006, Forwards Technical Justification for Proposal to Delete Peak Recording Accelerographs Currently Installed on Reactor Piping Locations,Per Section 16.3.1 of Sser 6 (NUREG-0831).Response by 860228 Requested.Fee Paid1986-01-10010 January 1986 Forwards Technical Justification for Proposal to Delete Peak Recording Accelerographs Currently Installed on Reactor Piping Locations,Per Section 16.3.1 of Sser 6 (NUREG-0831).Response by 860228 Requested.Fee Paid Project stage: Other ML20154R5461986-03-21021 March 1986 Proposed Tech Specs Revising License Condition 2.C.(28) Re Requirement for Corporate Mgt Officials on Nuclear Operations Staff & Modifying Tables 3.3.7.2-1 & 4.3.7.2-1 to Reflect Design Change Project stage: Other AECM-86-0082, Supplemental Application for Amend to License NPF-29, Modifying 851114 Proposed Change Re License Condition 2.C.(28) & Revising Tech Specs Concerning Design Changes to Be Implemented During First Refueling Outage.Fee Paid1986-03-21021 March 1986 Supplemental Application for Amend to License NPF-29, Modifying 851114 Proposed Change Re License Condition 2.C.(28) & Revising Tech Specs Concerning Design Changes to Be Implemented During First Refueling Outage.Fee Paid Project stage: Supplement ML20155D8101986-04-0808 April 1986 Summary of 860401 Meeting W/Utils Re Resolution of Issue of Interlocks for ECCS Injection Valves Which Also Function as Pressure Isolation Valves.Viewgraphs & List of Attendees Encl Project stage: Meeting ML20155J9341986-05-19019 May 1986 Proposed Tech Specs Adding Undervoltage Protection Device & Emergency Override of Test Mode for Div 3 HPCS Diesel Generator & Increasing Test Loads for Div 1 & 2 Batteries Project stage: Other AECM-86-0150, Application for Amend to License NPF-29,adding Another Undervoltage Protection Device & Emergency Override of Test Mode for Div 3 Diesel Generator & Increasing Test Loads for Div 1 & 2 Batteries.Fee Paid1986-05-19019 May 1986 Application for Amend to License NPF-29,adding Another Undervoltage Protection Device & Emergency Override of Test Mode for Div 3 Diesel Generator & Increasing Test Loads for Div 1 & 2 Batteries.Fee Paid Project stage: Request ML20205S2151986-05-27027 May 1986 Safety Evaluation Supporting Util 860110 Proposal to Remove Four Peak Recording Accelerographs Mounted on Reactor Piping Project stage: Approval ML20205S2021986-05-27027 May 1986 Forwards Safety Evaluation Supporting Util 860110 Proposal to Remove Four Peak Recording Accelerographs Installed on Reactor Piping Due to Inoperability Caused by Operating Transients & Adequacy of Other Seismic Instrumentation Project stage: Approval AECM-86-0111, Forwards Supplemental Info Re Design Change to Add Interlocks to ECCS Injection Valves,Per NRC 860403 Telcon. Approval of 850812 Proposed Amend to Tech Specs Also Requested1986-05-30030 May 1986 Forwards Supplemental Info Re Design Change to Add Interlocks to ECCS Injection Valves,Per NRC 860403 Telcon. Approval of 850812 Proposed Amend to Tech Specs Also Requested Project stage: Supplement AECM-86-0163, Provides Addl Info to 860321 Proposed Change PCOL-86/04 to License NPF-29 Re Mods to Automatic Depressurization Sys Per BWR Owners Group Evaluation of NUREG-0737,Item II.K.3.18. Info Clarifies Changes Made to Sys Initiation Circuitry1986-05-30030 May 1986 Provides Addl Info to 860321 Proposed Change PCOL-86/04 to License NPF-29 Re Mods to Automatic Depressurization Sys Per BWR Owners Group Evaluation of NUREG-0737,Item II.K.3.18. Info Clarifies Changes Made to Sys Initiation Circuitry Project stage: Request AECM-86-0221, Application for Amend to License NPF-29,changing Tech Spec Seismic Monitoring Instrumentation Tables to Delete four- Peak Recording Accelerographs on Reactor Piping & Make Measurement Values Consistent.Fee Paid1986-07-15015 July 1986 Application for Amend to License NPF-29,changing Tech Spec Seismic Monitoring Instrumentation Tables to Delete four- Peak Recording Accelerographs on Reactor Piping & Make Measurement Values Consistent.Fee Paid Project stage: Request ML20207E7261986-07-15015 July 1986 Proposed Changes to Tech Spec Seismic Monitoring Instrumentation Tables 3.3.7.2-1 & 4.3.7.2-1 to Delete four- Peak Recording Accelerographs on Reactor Piping & Make Measurement Values Consistent Project stage: Request AECM-86-0234, Forwards Clarification Re Proposed Mod to License Condition 2.C(20) Concerning Standby Svc Water Sys1986-07-25025 July 1986 Forwards Clarification Re Proposed Mod to License Condition 2.C(20) Concerning Standby Svc Water Sys Project stage: Other AECM-86-0232, Forwards Tabulation & Analyses for Standby Svc Water Cooling Towers Based on Max 30-day Average Wet Bulb & Refueling Outage 18,per NRC Request for Worst Case Scenario of Util 860519 Analyses1986-07-29029 July 1986 Forwards Tabulation & Analyses for Standby Svc Water Cooling Towers Based on Max 30-day Average Wet Bulb & Refueling Outage 18,per NRC Request for Worst Case Scenario of Util 860519 Analyses Project stage: Request ML20209D1521986-08-28028 August 1986 Notice of Consideration of Issuance of Amend to License NPF-29 & Proposed NSHC Determination & Opportunity for Hearing Re Addition of Specs to Table 3.3.3-1, ECCS Actuation Instrumentation Project stage: Other ML20209D1361986-09-0202 September 1986 Requests Notice of Consideration of Issuance of Amend to License NPF-29 & Proposed NSHC Determination & Opportunity for Hearing Re Addition of Specs to Table 3.3.3-1, ECCS Actuation Instrumentation Be Expeditiously Published in Fr Project stage: Other AECM-86-0329, Submits List of Affected Tech Spec Pages,Effective Dates & Associated Amend Numbers,In Response to NRC 860827 & 1006 Requests1986-10-14014 October 1986 Submits List of Affected Tech Spec Pages,Effective Dates & Associated Amend Numbers,In Response to NRC 860827 & 1006 Requests Project stage: Other AECM-86-0332, Advises of Effective Dates for Amends 6 & 8 to License NPF-29 for Listed Tech Specs,In Response to 851021 & 860827 Requests1986-10-17017 October 1986 Advises of Effective Dates for Amends 6 & 8 to License NPF-29 for Listed Tech Specs,In Response to 851021 & 860827 Requests Project stage: Other 1986-01-10
[Table View] |
Safety Evaluation Re Low Pressure ECCS Sys Pressure Permissive Interlocks.Exceeding ASME Code Allowable Design Pressure for Interlocks UnacceptableML20137Q897 |
Person / Time |
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Site: |
Grand Gulf ![Entergy icon.png](/w/images/7/79/Entergy_icon.png) |
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Issue date: |
11/22/1985 |
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NRC |
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Shared Package |
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ML20137Q894 |
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References |
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TAC-59440, NUDOCS 8512050379 |
Download: ML20137Q897 (6) |
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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20212F5641999-09-23023 September 1999 SER Concluding That All of ampacity-related Concerns Have Been Resolved & Licensee Provided Adequate Technical Basis to Assure That All of Thermo-Lag Fire Barrier Encl Cables Operating within Acceptable Ampacity Limits ML20196K4981999-07-0101 July 1999 Safety Evaluation Authorizing PRR-E12-01,PRR-E21-01, PRR-P75-01,PRR-P81-01,VRR-B21-01,VRR-B21-02,VRR-E38-01 & VRR-E51-01.Concludes That Alternatives Proposed by EOI Acceptable ML20195C4841998-11-0606 November 1998 SER Accepting QA Program Change to Consolidate Four Existing QA Programs for Arkansas Nuclear One,Grand Gulf Nuclear Station,River Bend Station & Waterford 3 Steam Electric Station Into Single QA Program ML20217Q6701998-05-0606 May 1998 SER Approving Proposed Postponement of Beginning Augmented Exam Requirements of 10CFR50.55a(g)(6)(ii)(A)(2) at Grand Gulf for Circumferential Shell Welds for Two Operating Cycles ML20217P8281998-04-0707 April 1998 Safety Evaluation Accepting Relief Authorization for Alternative to Requirements of ASME Section Xi,Subarticle IWA-5250 Bolting Exam for Plants,Per 10CFR50.55a(a)(3)(i) ML20217P0381998-04-0606 April 1998 Safety Evaluation Supporting Amend 135 to License NPF-29 ML20216J4211998-03-18018 March 1998 SER Approving Alternative to Insp of Reactor Pressure Vessel Circumferential Welds for Grand Gulf Nuclear Station ML20199H3711997-11-19019 November 1997 SER Accepting Approving Request Relief from Requirements of Section XI, Rule for Inservice Insp of NPP Components, of ASME for Current or New 10-year Inservice Insp Interval IAW 50.55(a)(3)(i) of 10CFR50 ML20199F3431997-11-18018 November 1997 SER Accepting Rev 15 of Operational Quality Assurance Manual for Grand Gulf Nuclear Station,Unit 1 ML20216E9921997-09-0404 September 1997 Safety Evaluation Accepting 970623 Request for Relief Re Authorization for Use of ASME Code Case N-416-1 & N-532,ISI Program for Listed Plants ML20134C5671997-01-30030 January 1997 SER Denying Request for Exemption from 10CFR50.55a Entergy Operating,Inc Et Al,Grand Gulf Nuclear Station,Unit 1 ML20149M4221996-12-12012 December 1996 Safety Evaluation Supporting Update Insvc Insp Programs to 1992 & Portions of 1993 ASME Boiler & Pressure Vessel Code, Sect XI for Licenses DPR-51,NPF-6,NPF-38,NPF-29 & NPF-47. Technical Ltr Rept Encl ML20138B5681996-12-11011 December 1996 Safety Evaluation Re Emergency Plan Change 28.001-95 to Entergy Operations,Inc,Grand Gulf Nuclear Station ML20134K5321996-11-18018 November 1996 Safety Evaluation Supporting Request for Relief Re 10CFR50.55a Inservice Testing of Main Steam Safety/Relief Valves for License NPF-29 ML20129D0441996-10-22022 October 1996 Safety Evaluation Supporting Request for Relief I-00014 ML20129B8751996-10-18018 October 1996 Safety Evaluation Accepting First 10-year Interval ISI Program Plan Addl Request for Relief ML20128H2491996-10-0707 October 1996 Safety Evaluation Authorizing Licensee Proposed Alternative to Use ASME Code Case N-508-1 for Rotation of Serviced Snubbers & Pressure Relief Valves for Sole Purpose of Testing in Lieu of ASME Requirements ML20107F5611996-04-17017 April 1996 Safety Evaluation Providing Guidance on Submitting plant- Specific Info W/Respect to IST Program Alternatives Request ML20128N1381993-02-17017 February 1993 Safety Evaluation Accepting Proposal to Raise Design Ground Water Level from 109 Ft Above Msl to 114.5 Ft Above Msl ML20058B4411990-10-25025 October 1990 Safety Evaluation Granting 900709 Request for Relief from Requirements for Leakage Testing of RCS in ASME Code,Section XI ML20062A1941990-10-16016 October 1990 Safety Evaluation Granting Relief Request from Inservice Insp Requirements ML20059N4971990-10-0101 October 1990 Safety Evaluation Accepting Util Response to Generic Ltr 88-01,w/listed Exceptions ML20058N3941990-08-0606 August 1990 Generic SER Re Mark III Containment Hydrogen Control, Concluding That HGN-112-NP, Generic Hydrogen Control Info for BWR-6 Mark III Containments Provides Acceptable Basis for Technical Resolution of Degraded Core Hydrogen Issue ML20055G5271990-07-18018 July 1990 Safety Evaluation Re Facility Procedures Generation Package. Util Needs to Revise Package to Address Programmatic Improvements Identified ML20055F9081990-07-16016 July 1990 Safety Evaluation Accepting Criticality Analysis for Cycle 5 Fuel in Spent Fuel Storage Racks ML20055G0341990-07-16016 July 1990 Safety Evaluation Re Boraflex Gaps in Spent Fuel Racks. Storage Racks Can Safely Accomodate Max Reactivity of Unit 1 Cycle 1 Through Cycle 4 Fuel.Storage Rack Surveillance Program Acceptable for long-term Storage of Spent Fuel ML20246D8021989-08-21021 August 1989 Safety Evaluation Granting 880524 Requests for Relief from Certain Requirements of ASME Code,Section XI ML20246B6311989-08-17017 August 1989 Safety Evaluation Supporting Amend 25 to License NPF-62 ML20247G7041989-07-21021 July 1989 SER Accepting Licensee Commitment to Install ex-core Neutron Flux Monitoring Instrumentation Meeting Requirements of Reg Guide 1.97,Rev 2,prior to Restart Following Fourth Refueling Outage ML20245A8251989-04-17017 April 1989 SER Re Proposed Changes to Administrative Controls Section of Tech Specs Concerning Min Shift Crew Composition,Unit Staff Qualifications,Training,Plant Safety Review Committee & Safety Review Committee Composition.Changes Unacceptable ML20244D5291989-04-14014 April 1989 Safety Evaluation Supporting Util Actions in Response to Generic Ltr 83-28,Position 4.5.2, Testing of Reactor Trip Sys. Justification for Not Testing Reactor Mode Switch or Backup Scram Valves Provided ML20248J9001989-04-0505 April 1989 Safety Evaluation Accepting Licensee Statements Confirming That Vendor Interface Program Exists W/Nsss Vendor for Components Required for Performance of Reactor Trip Function ML20151P0291988-08-0505 August 1988 Safety Evaluation Supporting Generic Ltr 83-28,Position 4.5.1 Re on-line Testing of Reactor Trip Sys ML20234D9021987-12-30030 December 1987 Safety Evaluation Supporting 870612,0814,1026 & 1119 Relief Requests from First 10-yr Interval Inservice Insp Program Requirements of Section XI of ASME Code ML20149E7091987-12-30030 December 1987 Safety Evaluation Supporting Amend 41 to License NPF-29 ML20234C7671987-12-23023 December 1987 Safety Evaluation Granting 870612,0814,1026 & 1119 Requests for Relief from Certain Section XI ASME Code Requirements Re First 10-yr Interval Inservice Insp Program ML20236B1021987-10-15015 October 1987 Safety Evaluation of 870630 Submittal Supplemented on 870821 & 0911 Re Deferral of Certain Insps of Tdi Div II Emergency Diesel Generators.Deferral of Certain Design Review & Quality Revalidation Baseline Insps Approved ML20235V1701987-10-0707 October 1987 Safety Evaluation Re 861022 Rev 2 to Process Control Program (PCP) for Processing & Packaging of Wet Radwastes.Pcp Acceptable ML20234A9471987-09-0404 September 1987 Safety Evaluation Re Util 870702 Submittal Re Containment Isolation for Various Instrument Lines.Util Proposed Alternate Basis to Meet GDC 55 to 10CFR50 App a Acceptable ML20237H1711987-08-21021 August 1987 Safety Evaluation Supporting Util 851014,870403 & 870622 Submittals Providing Info Re ATWS-related Design Features, Alternate Rod Injection,Standby Liquid Control Sys & Recirculation Pump Trip ML20236M3841987-07-31031 July 1987 Safety Evaluation on Generic Ltr 83-28,Items 3.1.1,3.1.2, 3.2.1 & 3.2.2 Re post-maint Testing.Licensee Programmatic Controls & Procedures for post-maint Testing of Components in Reactor Trip & safety-related Sys Acceptable ML20215J1351987-05-0404 May 1987 Safety Evaluation Re Spill of Sulphuric Acid at Plant on 860322.Licensee Monitoring & Recovery Sys Adequate to Contain & Eliminate Existing Acid Plume.No Addl Measures Recommended ML20210C3391987-04-30030 April 1987 Safety Evaluation Accepting Licensee 870129 Proposed Solution to Concern Re Control of Activities within Exclusion Areas ML20204J1451987-03-23023 March 1987 SER Supporting Licensee Responses to Humphrey Concerns on Safety of Mark III Containment Design ML20207R1051987-03-0505 March 1987 Safety Evaluation Supporting Util 861125 Rept on Conformance to Reg Guide 1.133,Rev 1, Loose-Part Detection Program for ...Sys of Light Water Cooled Reactors. License Condition 2.C(14) Adequately Addressed ML20212M3861987-03-0404 March 1987 Safety Evaluation Accepting Util 861223 Response to IE Bulletin 79-26, Boron Loss from BWR Control Blades, Per License Condition 2.C.(12) ML20212G3421987-01-12012 January 1987 Safety Evaluation Accepting Util 850228 & 860214 Submittals on Conformance to Reg Guide 1.97,Rev 2 ML20207N2761987-01-12012 January 1987 SER Supporting License Condition 2.C.(17) Which Replaces Check Valve Disc in Train B Feedwater Line Into Which Cold Water from Condensate Storage Tank Injected by RCIC ML20211P3901986-12-12012 December 1986 SER Supporting Util 860825 Request for Rev to Relief Request I-00007 Seeking Relief from Surface Exam for Welds within Flued Heads & Guard Pipes ML20214W8161986-12-0808 December 1986 Safety Evaluation Granting Util 861015 Request for Release from Commitment to Compare Performance of on-line Instrumentation Vs Grab Sampling Techniques 1999-09-23
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217F9921999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Grand Gulf Nuclear Station,Unit 1.With ML20212F5641999-09-23023 September 1999 SER Concluding That All of ampacity-related Concerns Have Been Resolved & Licensee Provided Adequate Technical Basis to Assure That All of Thermo-Lag Fire Barrier Encl Cables Operating within Acceptable Ampacity Limits ML20211Q3171999-09-0909 September 1999 Safety Evaluation Accepting BWROG Rept, Prediction of Onset of Fission Gas Release from Fuel in Generic BWR, Dtd July 1996 ML20216E4881999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Grand Gulf Nuclear Station.With ML20211A6921999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Grand Gulf Nuclear Station,Unit 1.With ML20209J1961999-07-12012 July 1999 Special Rept 99-001:on 990528,smoke Detectors Failed to Alarm During Performance of Routine Channel Functional Testing.Applicable TRM Interim Compensatory Measure of Establishing Roving Hourly Fire Patrol Was Implemented ML20196K4981999-07-0101 July 1999 Safety Evaluation Authorizing PRR-E12-01,PRR-E21-01, PRR-P75-01,PRR-P81-01,VRR-B21-01,VRR-B21-02,VRR-E38-01 & VRR-E51-01.Concludes That Alternatives Proposed by EOI Acceptable ML20209G0691999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Grand Gulf Nuclear Station,Unit 1.With ML20196A1161999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Grand Gulf Nuclear Station.With ML20206Q4831999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Grand Gulf Nuclear Station Unit 1.With ML20206J1201999-04-30030 April 1999 Redacted ME-98-001-00, Pressure Locking & Thermal Binding Test Program on Two Gate Valves with Limitorque Actuators ML18016A9011999-04-12012 April 1999 Part 21 Rept Re Defect in Component of DSRV-16-4,Enterprise DG Sys.Caused by Potential Problem with Connecting Rod Assemblies Built Since 1986,that Have Been Converted to Use Prestressed Fasteners.Affected Rods Should Be Inspected ML20205P8771999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Grand Gulf Nuclear Station,Unit 1.With ML20207M9231999-03-12012 March 1999 Amended Part 21 Rept Re Cooper-Bessemer Ksv EDG Power Piston Failure.Total of 198 or More Pistons Have Been Measured at Seven Different Sites.All Potentially Defective Pistons Have Been Removed from Svc Based on Encl Results ML20207K5141999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Grand Gulf Nuclear Station,Unit 1.With ML20206T7991999-01-31031 January 1999 Iodine Revolatizitation in Grand Gulf Loca ML20207A8301998-12-31031 December 1998 1998 Annual Operating Rept for Ggns,Unit 1 ML20206R9501998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Grand Gulf Nuclear Station,Unit 1.With ML20206D7721998-12-31031 December 1998 South Mississippi Electric Power Association 1998 Annual Rept ML20198E2481998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Grand Gulf Nuclear Station,Unit 1.With ML20195F4121998-11-13013 November 1998 Rev 16 to GGNS-TOP-1A, Operational QA Manual (Oqam) ML20195C4841998-11-0606 November 1998 SER Accepting QA Program Change to Consolidate Four Existing QA Programs for Arkansas Nuclear One,Grand Gulf Nuclear Station,River Bend Station & Waterford 3 Steam Electric Station Into Single QA Program ML20195C2791998-11-0505 November 1998 BWR Feedwater Nozzle Inservice Insp Summary Rept for GGNS, NUREG-0619-00006 ML20195F4801998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Grand Gulf Nuclear Station,Unit 1.With ML20155C1351998-10-26026 October 1998 Rev B to Entergy QA Program Manual ML20154K2391998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Grand Gulf Nuclear Station Unit 1.With ML20155F1961998-09-0101 September 1998 Engineering Rept for Evaluation of BWR CR Drive Mounting Flange Cap Screw ML20153B2161998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Grand Gulf Nuclear Station,Unit 1.With ML20237B6661998-07-31031 July 1998 Monthly Operating Rept for Jul 1998 for Grand Gulf Nuclear Station,Unit 1 ML20236R0231998-06-30030 June 1998 Monthly Operating Rept for June 1998 for Grand Gulf Nuclear Station,Unit 1 ML20155J0811998-05-31031 May 1998 10CFR50.59 SE for Period Jan 1997 - May 1998 ML20249B1251998-05-31031 May 1998 Monthly Operating Rept for May 1998 for Grand Gulf Nuclear Station,Unit 1 ML20248B6261998-05-11011 May 1998 Rev 6 to Grand Gulf Nuclear Station COLR Safety-Related ML20217Q6701998-05-0606 May 1998 SER Approving Proposed Postponement of Beginning Augmented Exam Requirements of 10CFR50.55a(g)(6)(ii)(A)(2) at Grand Gulf for Circumferential Shell Welds for Two Operating Cycles ML20206J1271998-04-30030 April 1998 Pressure Locking Thrust Evaluation Methodology for Flexible Wedge Gate Valves ML20247F3591998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for Grand Gulf Nuclear Plant,Unit 1 ML20217M8951998-04-30030 April 1998 QA Program Manual ML20217P8281998-04-0707 April 1998 Safety Evaluation Accepting Relief Authorization for Alternative to Requirements of ASME Section Xi,Subarticle IWA-5250 Bolting Exam for Plants,Per 10CFR50.55a(a)(3)(i) ML20217P0381998-04-0606 April 1998 Safety Evaluation Supporting Amend 135 to License NPF-29 ML20217A0291998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for Grand Gulf Nuclear Sation,Unit 1 ML20216J4211998-03-18018 March 1998 SER Approving Alternative to Insp of Reactor Pressure Vessel Circumferential Welds for Grand Gulf Nuclear Station ML20216J2021998-02-28028 February 1998 Monthly Operating Rept for Feb 1998 for Grand Gulf Nuclear Station,Unit 1 ML20203A2891998-01-31031 January 1998 Monthly Operating Rept for Jan 1998 for Grand Gulf Nuclear Station ML20247B4111997-12-31031 December 1997 1997 Annual Financial Rept for South Mississippi Electric Power Association ML20203H9741997-12-31031 December 1997 1997 Annual Operating Rept, for Ggns,Unit 1 ML20198P1121997-12-31031 December 1997 Monthly Operating Rept for Dec 1997 for Grand Gulf Nuclear Station,Unit 1 ML20203B5581997-12-0404 December 1997 Special Rept 97-003:on 971111,valid Failure of Div 2 EDG Occurred,Due to Jacket Water Leak.Failure Reported,Per Plant Technical Requirements Manual Section 7.7.2.2 ML20203K4031997-11-30030 November 1997 Monthly Operating Rept for Nov 1997 for Grand Gulf Nuclear Station,Unit 1 ML20199H3711997-11-19019 November 1997 SER Accepting Approving Request Relief from Requirements of Section XI, Rule for Inservice Insp of NPP Components, of ASME for Current or New 10-year Inservice Insp Interval IAW 50.55(a)(3)(i) of 10CFR50 ML20199F3431997-11-18018 November 1997 SER Accepting Rev 15 of Operational Quality Assurance Manual for Grand Gulf Nuclear Station,Unit 1 1999-09-09
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Enclosure SAFETY EVALUATION GRAND GULF UNIT 1 LOW PRESSURE ECCS SYSTEM PRESSURE PERMISSIVE INTERLOCKS Facility Operating License NPF-29, Condition 2.C.(18) requires the licensee to implement isolation protection against overpressurization of the low pressure emergency core cooling systems (RHR/LPCI and LPCS) through the implementation of reactor vessel pressure permissive interlocks. The licensee has proposed Technical Specification (T.S.) changes to support the required design change.
-The Instrumentation and Control Systems Branch (ICSB) focused its review on the revisions to T.S. Tables 3.3.3-1 and 4.3.3.1-1. The Reactor Systems Branch with the assistance of the Mechanical Engineering Branch focused its attention on systems and piping design aspects of the proposed changes.
Instrumentation and Control Systems Aspects The licensee proposes to include the new high/ low pressure permissive inter-locks for the low pressure ECCS injection valves in T.S. Table 3.3.3-1 (ECCSActuationInstrumentation). The Technical Specifications will require a minimum of three channels operable for all operating conditions with Action 31 (declare ADS trip system or ECCS inoperable) applicable for Operating Conditions 1, 2, and 3 and Action 35 (trip inoperable channel after one hour or declare associated systems inoperable) applicable for Operating Conditions 4 and 5. Upon request, the licensee provided information (Letter dated October 5,1985) to clarify what constitutes a channel for the pressure l
8512050379 851122 PDR ADOCK 05000416 P PDR
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.' 2 permissive interlocks. The licensee states that the minimum requirement of 3 operating channels per trip function is applicable to the "one-out-of-two twice" logic utilized in the design change and is adequate to assure operability of the required low pressure injection function considering the diversity of injection systems and logic channels available (four per trip function).
The licensee also proposes to add the surveillance requirements for the new interlock channels to T.S. Table 4.3.3.1-1 (ECCS Actuation Instrumentation Surveillance Requirements). The surveillance frequencies will be once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for channel check, once per month for channel functional test, and once per refueling cycle for channel calibration for all operating conditions.
Based on the above information, the staff finds the channel operability require-ments proposed for revised T.S. Table 3.3.3-1 to be acceptable. We also' find the channel surveillance frequencies proposed in T.S. Table 4.3.3.1-1 to be acceptable.
After implementation of the proposed design, the high/ low pressure permissive interlocks for the low pressure ECCS injection valves will be active for both automatic and manual control room operation. The licensee has proposed not to include such an interlock for the remote shutdown panel (RSP) control circuits for the ECCS injection valves. Based on the current information provided by the licensee, the staff finds the requested interlock omission to be unacceptable.
It is the staff's position that pressure permissive interlocks should be I installed as part of the RSP control circuits consistent with the corresponding control room control circuits. This position is based on the Standard Review Plan (SRP Section 7.4) interpretation of GDC 19 which requires that the remote shutdown station equipment be designed to the same standards as the corre-sponding equipment in the main control room. Also, Section 6.3.4 of the Grand Gulf, Unit 1 SER Supplement No. 2 requires that interlocks be present at all times for both manual and automatic actuation unless the reactor vessel pressure is lower than the design pressure of the ECCS involved. Thus, without special control over the operation of these valves, inadvertent operation must be assumed from the RSP which could result in overpressurization of the low pressure ECCS.
The following is an acceptable alternative (method of special control) to the implementation of pressure permissive interlocks for the RSP control circuits associated with the low pressure ECCS injection valves:
(1) the provision of spring-return-to-normal control switches on the RSPs which would alleviate the concern related to the change-of-state of equipment upon transfer of control to the l RSP and (2) (A) Continue to implement special Technical Specification j 4.4.3.2.2.b for LPCS and LPCI outboard check valves to ensure integrity since the licensee takes credit for these valves as part of the omission justification, or
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(B) Implementation of a keylocked control switch separate from the RSP controls (preferably at MCC) which would block operation of the valve via the RSP control. This should include administrative controls to ensure that RSP control
, of the valve is actually blocked when the valve is closed.
Also, valve position indication should not be negated in the control room or at the RSP by the implementation of this special control scheme.
Systems and Piping Design The licensee has proposed Technical Specification (TS) changes regarding the low pressure ECCS injection system, in response to an NRC position regarding the prevention of an intersystem LOCA which was expressed in Supplement No. 2 to the Grand Gulf Nuclear Station (GGNS) Safety Evaluation Report (NUREG-0831) and which became License Condition 2.C.(18).. These TS changes would implement design changes adding pressure interlocks to the injection valves on the low pressure ECCS systems, require respective trip setpoints, and require periodic l -
surveillance of associated ECCS actuation instrumentation.
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Two of the proposed changes are administrative in nature. For instance, in connection with the installment of pressure interlocks, an additional trip function has been added to the ECCS low pressure systems. The minimum number l of operable channels, applicable. operational conditions, and action to be
4 taken are supplied for the new trip function (see Tables 1 and 2). Instru-mentation surveillance intervals are also designated for this function (see Table 3). These proposals are acceptable to us, as discussed above.
In the original submittal, MP&L proposed to delete ECCS response times for low pressure systems. Subsequently, in a letter dated September 25, 1985, MD&L proposedtouse"f29 seconds"asresponsetimesforbothLPCIandLPCSinjection valves. This is acceptable to us because this response time is within the value used in analyses of loss of coolant accidents.
One issue that has not been resolved is the interlock setpoint for the high pressure to low pressure interface valve. The setpoint 534 psig, is above the design pressure of the low pressure system, 500 psig. The ASME Code, dated 1977, which was the first Code edition to address such interlocks, required the interlocks to prevent the pressure from exceeding the design pressure of the low pressure side. Section 6.3.4 of GGNS SSER's (June 1982),
states that the valves in question should be " interlocked to prevent opening unless the reactor vessel pressure is lower than the design pressure of the ECCSs involved." From discussions with the licensee, we understand that main-taining the setpoint below the design pressure will result in increased peak cladding temperature for. the postulated loss of accident from 2151*F to 2157*F, which is still less than the limiting value of 2200*F required by 10 CFR 50.46. The staff concludes that exceeding the Code allowable design pressure for the interlocks is unacceptable because the Code position can be
I met without reducing the acceptable margin of safety for peak cladding temperature. If meeting the Code requirement results in a hardship, the hardship should be described and justification should be provided for our consideration. Include in the justification the results of an analysis to show that pressure in the ECCS piping equal to the upper analytical limit of the interlock setpoint will satisfy the ASME Code stress allowable values for the piping, valves and other components of the affected portions of the ECCS.
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