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Category:OPERATING LICENSES-APPLIATION TO AMEND-RENEW EXISTING
MONTHYEARML17265A6921999-06-28028 June 1999 Application for Amend to License DPR-18 to Revise ITS Associated with RCS Leakage Detection Instrumentation.Lar Proposed as Result of Commitment That Rg&E Submit as Part of Staff Review of Application of leak-before-break Status ML17265A5691999-03-0101 March 1999 Application for Amend to License DPR-18,proposing Change in ITS to Revise Battery Cell Parameters Limit for Specific Gravity (SR 3.8.6.3 & SR 3.8.6.6) ML17265A4651998-11-24024 November 1998 Application for Amend to License DPR-18,revising Improved Tech Spec Description of Fuel Cladding Matl (TS 4.2.1) & Updating References Provided in TS 5.6.5 for COLR ML17265A2431998-04-27027 April 1998 Application for Amend to License DPR-18,revising Requirements Associated W/Sfp to Reflect Planned Mod to Storage Racks & Temporarily Addressing Boraflex Degradation within Pool ML17264B0431997-09-29029 September 1997 Application for Amend to License DPR-18,revising Administrative Controls W/Respect to RCS Pressure & Temp Limits Rept ML17264B0381997-09-29029 September 1997 Application for Amend to License DPR-18,revising TS Re Allowable Value & Trip Setpoint for High Steam Flow Input Into LCO Table 3.3.2-1,Function 4d (Main Steam Isolation) to Address Issues Identified in Revised Setpoint Analysis ML17264B0021997-08-19019 August 1997 Application for Amend to License DPR-18,allowing Testing of Certain ECCS motor-operated Valves in Mode 4 Which Currently Requires Entry in LCO 3.0.3 ML17264A9981997-08-19019 August 1997 Application for Amend to License DPR-18,correcting Specified Accumulator Borated Water Volume Values in SR 3.5.1.2 to Match Associated Accumulator Percent Level Values ML17264A8661997-04-24024 April 1997 Application for Amend to License DPR-18,revising RCP PT & Administrative Control Requirements ML17264A8471997-03-31031 March 1997 Application for Amend to License DPR-18,revising Spent Fuel Pool Storage Requirements ML20134D9911996-10-29029 October 1996 Application for Amend to License DPR-18,requesting Change to Mode of Applicability for Auxiliary Feedwater Pump Actuation on Main Feedwater Pump ML17264A7121996-10-29029 October 1996 Application for Amend to License DPR-18 Requesting Change to Required Actions for Inoperable Channels Associated with Auxiliary Feedwater Pump ML17264A5991996-09-13013 September 1996 Application for Amend to License DPR-18,revising RCS Pressure & Temp Limits Rept ML17264A4701996-05-0808 May 1996 Application for Amend to License DPR-18,correcting Typos ML17264A4041996-03-15015 March 1996 Application for Amend to License DPR-18,revising TS to Incorporate Methodology for Determining RCS P/T Limits & LTOP Limits in Administrative Controls Section of RCS P/T Limits Rept ML17264A3441996-02-0909 February 1996 Application for Amend to License DPR-18,revising TS Table 3.3.2-1,Function 5b, Feedwater Isolation,Sg Water Level - High to Raise Allowable Value & Trip Setpoint ML17264A3381996-02-0909 February 1996 Application for Amend to License DPR-18,revising TS to Incorporate Ref to Methodology for Determining LTOP Limits Into Administrative Controls Section for RCS PT Limits Rept ML17264A3331996-02-0909 February 1996 Application for Amend to License DPR-18,revising Requirements for Containment Closure During Mode 6 to Allow Use of Installed Overhead Door Assembly to Isolate Equipment Hatch Opening ML17264A3221996-01-26026 January 1996 Application for Amend to License DPR-18,incorporating Changes Necessary to Reflect Revised Nuclear Fuel Loading ML17264A2521995-11-27027 November 1995 Application for Amend to License DPR-18,revising TS to Implement 10CFR50,App J,Option B ML17264A2551995-11-20020 November 1995 Application for Amend to License DPR-18,submitting Changes to TS Instrumentation Requirements & Conversion to Improved TS ML17264A1501995-08-31031 August 1995 Application for Amend to License DPR-18,proposing Ts,By Implementing WCAP-10271,its Assoc Suppls & Other Re Changes W/Respect to RTS & ESFAS ML17263B0671995-05-26026 May 1995 Application for Amend to License DPR-18,revising TSs in Entirety to Convert to Improved TSs ML17263A9771995-03-13013 March 1995 Application for Amend to License DPR-18,to Revise TS 4.4.2.4.a to Replace Specific Leakage Testing Frequencies for Containment Isolation Valves ML17309A5501994-07-15015 July 1994 Application for Amend to License DPR-18,proposing TS 3.5, Providing NRC W/Opportunity to Communicate at Early Stage Any Concerns W/Respect to Differences from NUREG-1431 ML17263A7141994-06-24024 June 1994 Application for Amend to License DPR-18,re Upgrading of Administrative Controls Section 6 ML17263A7071994-06-15015 June 1994 Notarized Application for Amend to License DPR-18 Re Reactor Coolant Activity ML17263A6551994-05-23023 May 1994 Application for Amend to License DPR-18,increasing Allowable Reactor Coolant Activity Levels to Improved TS (NUREG-1431) ML17263A6411994-05-13013 May 1994 Application for Amend to License DPR-18,revising TS Administrative Controls Section 6.0 Consistent W/Improved TS ML17263A3181993-07-13013 July 1993 Application for Amend to License Revisions for License DPR-18 Re Removing Containment Isolation Valve Table 3.6-1 from TS ML17263A2081993-04-0505 April 1993 Application for Amend to License DPR-18,revising TS to Remove Containment Isolation Valve Table 3.6-1 & Maintain Required Listing in Updated FSAR ML17262B1171992-12-17017 December 1992 Application for Amend to License DPR-19,revising TS Sections 3.2 & 3.3 to Eliminate Use of High Concentration Boric Acid as safety-related Source of SI Pumps & Increasing Allowable Outage Times ML17262B0991992-11-30030 November 1992 Application for Amend to License DPR-18,revising TS to Remove Table of Containment Isolation Valves & Provide Ref in Bases to UFSAR ML17262B0571992-10-0808 October 1992 Corrected Application for Amend to License DPR-18,adding Addl Requirements to Auxiliary Electrical Sys TS & Action Statements for Instrument Bus Sys,Consisting of Notarized Page Two ML17262B0431992-10-0808 October 1992 Application for Amend to License DPR-18,adding Addl Requirements to Auxiliary Electrical Sys TS & Action Statements for Instrument Bus Sys,Per Guidance Provided in GL 91-11 Re Resolution of GI 48 ML17262B0201992-09-15015 September 1992 Application for Amend to License DPR-18,revising TS Sections 3.1.1.4,3.1.1.6,4.3.4 & Adding site-specific Basis Sections to Address GL 90-06 Re Resolution of Generic Issues 70 & 94 ML17262A9121992-06-22022 June 1992 Application for Amend to License DPR-18,revising TS 4.3.1 Per Guidance Provided in Generic Ltr 91-01, Removal of Schedule for Withdrawal of Reactor Vessel Matl Specimens from Ts. ML17262A8311992-04-23023 April 1992 Application for Amend to License DPR-18,changing TSs to Redefine Snubber Visual Insp Schedule,Per Generic Ltr 90-09 Guidance ML17262A8211992-04-21021 April 1992 Application for Amend to License DPR-18,instituting Std License Condition for Fire Protection Program & Removing Requirements for Fire Protection Sys from Tss,Per Guidance in Generic Ltrs 86-10 & 88-12 ML17262A7861992-03-23023 March 1992 Application for Amend to License DPR-18,revising Tech Spec Section 6.5.1, Plant Operations Review Committee Function. ML17262A7801992-03-20020 March 1992 Application for Amend to License DPR-18,revising TS 5.1 Figure 5.1-1 to Describe Rather than Depict Site Area Boundary to Evaluate Radiological Releases to Unrestricted Area ML17262A7901992-03-20020 March 1992 Application for Amend to License DPR-18,revising Tech Specs 6.9.1.2 & 6.9.2.5 ML17262A6321991-10-25025 October 1991 Application for Amend to License DPR-18 to Delete TS 5.1 & Figure 5.1-1 ML17262A5331991-06-20020 June 1991 Application for Amend to License DPR-18,to Incorporate Changes to Specifications & Action Statements for Offsite & Onsite Power Sources Available for Operation of Plant Auxiliaries ML17262A4121991-03-0808 March 1991 Application for Amend to License DPR-18,removing Table 3.6-1 Re Containment Isolation Valves from Tech Specs ML17262A3831991-02-15015 February 1991 Application for Amend to License DPR-18,revising Tech Specs to Modify Method of Locking Open Motor Operated Valve 856, Refueling Water Storage Tank Delivery Valve When Reactor Coolant Sys at or Above 350 F ML17262A3751991-02-15015 February 1991 Application for Amend to License DPR-18,revising Heatup/Cooldown Curves & Associated Low Temp Overpressure Protection Setpoint to Implement Reg Guide 1.99,Rev 2 Methodology ML17262A3721991-02-15015 February 1991 Application for Amend to License DPR-18,providing Clarification of Which QA Activities Subj to Audit Considerations & Changing Audit Frequency Described in Tech Spec 6.5.2.8d ML17262A3871991-02-15015 February 1991 Application for Amend to License DPR-18,revising Tech Spec Tables 3.5-5 & 4.1-5 Based on Mods to Radiation Monitoring Sys ML17262A2951991-01-14014 January 1991 Application for Amend to License DPR-18,changing Tech Specs to Limit Containment Internal Pressure to 1 Psig Vs 6 Psig Per Initial Condition Assumptions in Containment Integrity Analysis 1999-06-28
[Table view] Category:TEXT-LICENSE APPLICATIONS & PERMITS
MONTHYEARML17265A6921999-06-28028 June 1999 Application for Amend to License DPR-18 to Revise ITS Associated with RCS Leakage Detection Instrumentation.Lar Proposed as Result of Commitment That Rg&E Submit as Part of Staff Review of Application of leak-before-break Status ML17265A5691999-03-0101 March 1999 Application for Amend to License DPR-18,proposing Change in ITS to Revise Battery Cell Parameters Limit for Specific Gravity (SR 3.8.6.3 & SR 3.8.6.6) ML17265A4651998-11-24024 November 1998 Application for Amend to License DPR-18,revising Improved Tech Spec Description of Fuel Cladding Matl (TS 4.2.1) & Updating References Provided in TS 5.6.5 for COLR ML17265A2431998-04-27027 April 1998 Application for Amend to License DPR-18,revising Requirements Associated W/Sfp to Reflect Planned Mod to Storage Racks & Temporarily Addressing Boraflex Degradation within Pool ML17264B0431997-09-29029 September 1997 Application for Amend to License DPR-18,revising Administrative Controls W/Respect to RCS Pressure & Temp Limits Rept ML17264B0381997-09-29029 September 1997 Application for Amend to License DPR-18,revising TS Re Allowable Value & Trip Setpoint for High Steam Flow Input Into LCO Table 3.3.2-1,Function 4d (Main Steam Isolation) to Address Issues Identified in Revised Setpoint Analysis ML17264B0021997-08-19019 August 1997 Application for Amend to License DPR-18,allowing Testing of Certain ECCS motor-operated Valves in Mode 4 Which Currently Requires Entry in LCO 3.0.3 ML17264A9981997-08-19019 August 1997 Application for Amend to License DPR-18,correcting Specified Accumulator Borated Water Volume Values in SR 3.5.1.2 to Match Associated Accumulator Percent Level Values ML17264A8661997-04-24024 April 1997 Application for Amend to License DPR-18,revising RCP PT & Administrative Control Requirements ML17264A8471997-03-31031 March 1997 Application for Amend to License DPR-18,revising Spent Fuel Pool Storage Requirements ML20134D9911996-10-29029 October 1996 Application for Amend to License DPR-18,requesting Change to Mode of Applicability for Auxiliary Feedwater Pump Actuation on Main Feedwater Pump ML17264A7121996-10-29029 October 1996 Application for Amend to License DPR-18 Requesting Change to Required Actions for Inoperable Channels Associated with Auxiliary Feedwater Pump ML17264A5991996-09-13013 September 1996 Application for Amend to License DPR-18,revising RCS Pressure & Temp Limits Rept ML17264A4701996-05-0808 May 1996 Application for Amend to License DPR-18,correcting Typos ML17264A4041996-03-15015 March 1996 Application for Amend to License DPR-18,revising TS to Incorporate Methodology for Determining RCS P/T Limits & LTOP Limits in Administrative Controls Section of RCS P/T Limits Rept ML17264A3441996-02-0909 February 1996 Application for Amend to License DPR-18,revising TS Table 3.3.2-1,Function 5b, Feedwater Isolation,Sg Water Level - High to Raise Allowable Value & Trip Setpoint ML17264A3381996-02-0909 February 1996 Application for Amend to License DPR-18,revising TS to Incorporate Ref to Methodology for Determining LTOP Limits Into Administrative Controls Section for RCS PT Limits Rept ML17264A3331996-02-0909 February 1996 Application for Amend to License DPR-18,revising Requirements for Containment Closure During Mode 6 to Allow Use of Installed Overhead Door Assembly to Isolate Equipment Hatch Opening ML17264A3221996-01-26026 January 1996 Application for Amend to License DPR-18,incorporating Changes Necessary to Reflect Revised Nuclear Fuel Loading ML17264A2521995-11-27027 November 1995 Application for Amend to License DPR-18,revising TS to Implement 10CFR50,App J,Option B ML17264A2551995-11-20020 November 1995 Application for Amend to License DPR-18,submitting Changes to TS Instrumentation Requirements & Conversion to Improved TS ML17264A1501995-08-31031 August 1995 Application for Amend to License DPR-18,proposing Ts,By Implementing WCAP-10271,its Assoc Suppls & Other Re Changes W/Respect to RTS & ESFAS ML17263B0671995-05-26026 May 1995 Application for Amend to License DPR-18,revising TSs in Entirety to Convert to Improved TSs ML17263A9771995-03-13013 March 1995 Application for Amend to License DPR-18,to Revise TS 4.4.2.4.a to Replace Specific Leakage Testing Frequencies for Containment Isolation Valves ML17309A5501994-07-15015 July 1994 Application for Amend to License DPR-18,proposing TS 3.5, Providing NRC W/Opportunity to Communicate at Early Stage Any Concerns W/Respect to Differences from NUREG-1431 ML17263A7141994-06-24024 June 1994 Application for Amend to License DPR-18,re Upgrading of Administrative Controls Section 6 ML17263A7071994-06-15015 June 1994 Notarized Application for Amend to License DPR-18 Re Reactor Coolant Activity ML17263A6551994-05-23023 May 1994 Application for Amend to License DPR-18,increasing Allowable Reactor Coolant Activity Levels to Improved TS (NUREG-1431) ML17263A6411994-05-13013 May 1994 Application for Amend to License DPR-18,revising TS Administrative Controls Section 6.0 Consistent W/Improved TS ML17263A3181993-07-13013 July 1993 Application for Amend to License Revisions for License DPR-18 Re Removing Containment Isolation Valve Table 3.6-1 from TS ML17263A2081993-04-0505 April 1993 Application for Amend to License DPR-18,revising TS to Remove Containment Isolation Valve Table 3.6-1 & Maintain Required Listing in Updated FSAR ML17262B1171992-12-17017 December 1992 Application for Amend to License DPR-19,revising TS Sections 3.2 & 3.3 to Eliminate Use of High Concentration Boric Acid as safety-related Source of SI Pumps & Increasing Allowable Outage Times ML17262B0991992-11-30030 November 1992 Application for Amend to License DPR-18,revising TS to Remove Table of Containment Isolation Valves & Provide Ref in Bases to UFSAR ML17262B0571992-10-0808 October 1992 Corrected Application for Amend to License DPR-18,adding Addl Requirements to Auxiliary Electrical Sys TS & Action Statements for Instrument Bus Sys,Consisting of Notarized Page Two ML17262B0431992-10-0808 October 1992 Application for Amend to License DPR-18,adding Addl Requirements to Auxiliary Electrical Sys TS & Action Statements for Instrument Bus Sys,Per Guidance Provided in GL 91-11 Re Resolution of GI 48 ML17262B0201992-09-15015 September 1992 Application for Amend to License DPR-18,revising TS Sections 3.1.1.4,3.1.1.6,4.3.4 & Adding site-specific Basis Sections to Address GL 90-06 Re Resolution of Generic Issues 70 & 94 ML17262A9121992-06-22022 June 1992 Application for Amend to License DPR-18,revising TS 4.3.1 Per Guidance Provided in Generic Ltr 91-01, Removal of Schedule for Withdrawal of Reactor Vessel Matl Specimens from Ts. ML17262A8311992-04-23023 April 1992 Application for Amend to License DPR-18,changing TSs to Redefine Snubber Visual Insp Schedule,Per Generic Ltr 90-09 Guidance ML17262A8211992-04-21021 April 1992 Application for Amend to License DPR-18,instituting Std License Condition for Fire Protection Program & Removing Requirements for Fire Protection Sys from Tss,Per Guidance in Generic Ltrs 86-10 & 88-12 ML17262A7861992-03-23023 March 1992 Application for Amend to License DPR-18,revising Tech Spec Section 6.5.1, Plant Operations Review Committee Function. ML17262A7801992-03-20020 March 1992 Application for Amend to License DPR-18,revising TS 5.1 Figure 5.1-1 to Describe Rather than Depict Site Area Boundary to Evaluate Radiological Releases to Unrestricted Area ML17262A7901992-03-20020 March 1992 Application for Amend to License DPR-18,revising Tech Specs 6.9.1.2 & 6.9.2.5 ML17262A6321991-10-25025 October 1991 Application for Amend to License DPR-18 to Delete TS 5.1 & Figure 5.1-1 ML17262A5331991-06-20020 June 1991 Application for Amend to License DPR-18,to Incorporate Changes to Specifications & Action Statements for Offsite & Onsite Power Sources Available for Operation of Plant Auxiliaries ML17262A4121991-03-0808 March 1991 Application for Amend to License DPR-18,removing Table 3.6-1 Re Containment Isolation Valves from Tech Specs ML17262A3831991-02-15015 February 1991 Application for Amend to License DPR-18,revising Tech Specs to Modify Method of Locking Open Motor Operated Valve 856, Refueling Water Storage Tank Delivery Valve When Reactor Coolant Sys at or Above 350 F ML17262A3751991-02-15015 February 1991 Application for Amend to License DPR-18,revising Heatup/Cooldown Curves & Associated Low Temp Overpressure Protection Setpoint to Implement Reg Guide 1.99,Rev 2 Methodology ML17262A3721991-02-15015 February 1991 Application for Amend to License DPR-18,providing Clarification of Which QA Activities Subj to Audit Considerations & Changing Audit Frequency Described in Tech Spec 6.5.2.8d ML17262A3871991-02-15015 February 1991 Application for Amend to License DPR-18,revising Tech Spec Tables 3.5-5 & 4.1-5 Based on Mods to Radiation Monitoring Sys ML17262A2951991-01-14014 January 1991 Application for Amend to License DPR-18,changing Tech Specs to Limit Containment Internal Pressure to 1 Psig Vs 6 Psig Per Initial Condition Assumptions in Containment Integrity Analysis 1999-06-28
[Table view] |
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of))Rochester Gas and Electric Corporation
)(R.E.Ginna Nuclear Power Plant)))Docket No.50-244 APPLICATION FOR AMENDMENT TO OPERATING LICENSE Pursuant to Section 50.90 of the regulations of the U.S.Nuclear Regulatory Commission (the"Commission"), Rochester Gas and Electric Corporation
("RG&E"), holder of Facility Operating License No.DPR-18, hereby requests that the Technical Specifications set forth in Appendix A to that license be amended.This request for change in Technical Specifications is to increase allowable reactor coolant activity levels to the Improved Technical Specification values (NUREG-1431).
A description of the amendment request, necessary background information, justification of the requested change, safety'valuation and no significant hazards and environmental considerations are provided in Attachment A.A marked up copy of the current.Ginna Station Technical Specifications which shows the requested change is set forth in Attachment B.The proposed revised Technical Specifications are provided in Attachment C.These changes are consistent with Westinghouse Improved Technical Specifications (NUREG 1431)3.4.16.a,b and figure 3.4.16-1.94053iOih7 940523 PDR ADOCK 05000244.P'., PDR
WHEREFORE, Applicant respectfully requests that Appendix A to Facility Operating License No.DPR-18 be amended in the form attached hereto as Attachment C.Rochester Gas and Electric Corporation By Robert C.Mecredy Vice President Ginna Nuclear Production Subscribed and sworn to before me on this 23rd day of May, 1994.
ATTACHMENT A R.E.GINNA POWER PLANT LICENSE AMENDMENT REQUEST TECHNICAL SPECIFICATION 3.1.4, MAXIMUM COOLANT ACTIVITY This attachment provides a description of the amendment request and necessary justification for the proposed changes.The attachment is divided into seven sections as follows.Section A identifies all changes to the current Ginna Station Technical Specifications while Section B provides the background and history associated with the changes being requested.
Section C provides detailed justification for the proposed changes including a comparison to Improved Technical Specifications as applicable.
A safety evaluation, significant hazards consideration evaluation, and environmental consideration of the requested changes are provided in Sections D, E, and F, respectively.
Section G lists all references used in this attachment.
A.Description of Amendment Request This License Amendment Request (LAR)proposes to revise Ginna Station Technical Specifications 3.1.4.1.a, 3.1.4.1.b, figure 3.1.4-1 and associated Bases as follows: 1.Technical Specification 3.1.4.1.a i.The requirement is changed to"The total specific activity of the reactor coolant shall not exceed 100/E pCi/gm,..." ii.The bases are revised to change the referenced analysis (Reference 3)to"UFSAR Section 15.'6.3." 2.Technical Specification 3.1.4.1.b i.The requirement is revised to"The I-131 dose equivalent of the iodine activity in the reactor coolant shall not exceed 1.0 pCi/gm." ii.The bases are revised to change the referenced analysis (Reference 3)to"UFSAR Section 15.6.3." 3.Technical Specification Figure 3.1.4-1 i.The allowable operation region is modified consistent with Improved Technical Specifications (see Attachments B and C for revised figure).ii.The bases are revised to change the referenced analysis (Reference 3)to"UFSAR Section 15.6.3." B~Background History Prior to the January 25, 1982, steam generator tube rupture event at Ginna Station, reactor coolant activity limits were based on the original (1969)steam generator tube rupture analysis for the Ginna Station.The Commission's review of the 1982 tube rupture incident resulte in the requirement for a re sed steam generator tube rupture analysis.The staff required that this be completed within six months of the plant restart (NUREG-0916, Section 9.0), and imposed reduced allowable activity levels in the interim (Amendment No.51 to Provisional Operating License No.DPR-18, May 22, 1982).A bounding analysis using these reduced allowable activity levels was performed in order to satisfy the six month requirement, while a more detailed analysis supporting the standard technical specification values would follow.'he methodology for this new analysis (WCAP-10698-P-A) was submitted and approved by the Commission for'use on'estinghouse PWRs provided five plant specific inputs were verified to be consistent with the assumptions in the methodology (Reference a).RG&E has completed this verification, and therefore intends to update its analysis of record for the steam generator tube rupture to reflect use of this new methodology (UFSAR Section 15.6.3).This new analysis supports the activity limits proposed in this Amendment.
2.Hardware Modifications This LAR involves no hardware changes to Ginna Station.Justification This proposed Amendment imposes reactor coolant activity limits consistent with NUREG-1431,"Westinghouse Standard Technical Specifications." The applicability of these limits for Ginna Station are established by a plant specific steam generator tube rupture and radiological consequences analysis, WCAP-11668, which is consistent with the approved methodology of WCAP-10698-P-A for analysis of steam generator tube rupture transients.
All contingencies for usage of WCAP-10698-P-A methodology (Reference a)have been satisfied for Ginna Station as described in section D below.Safety Evaluation Potential environmental consequences of a steam generator tube rupture event at the R.E.Ginna nuclear power plant have been evaluated to verify that the Improved Technical Specification limit on primary coolant activity is adequate for Ginna.'This analysis, WCAP-11668 (attached) is consistent with the methodology described in WCAP-10698-P-A.
The Commission requires that five contingencies be met in order to use this methodology, specifically:
1~Demonstration that critical operator action times used in the analysis are realistic and consistent with those observed during simulator exercises.
2~3~A site specific Steam Generator Tube Rupture radiological offsite consequence analysis.A structural analysis of the main steam lines demonstrating adequacy under water-filled con itions.4~A list of systems, components, and instrumentation credited for accident mitigation and the specified safety grade for each.5.A comparison of the plant to the"bounding plant" used in WCAP-10698.
Compliance with those contingencies for Ginna Station has been satisfied and is described below.1~Demonstration that critical operator action times used in the analysis are realistic and consistent with those observed during simulator exercises.
During the week of August 19 through 23, 1991, simulator exercises were performed at the Ginna Station simulator to verify the assumptions used for both analyses cases presented in WCAP-11668.
The results are tabulated below.CASE 1, INTACT SG PORV FAILS CLOSED OPERATOR ACTION 1.Recognize and Isolate Ruptured SG 2.Recognize and locally open intact SG PORV open 3.Terminate SI 4.Terminate break flow WCAP 11668 TIME (SEC)600 1804 2798 3428 SIMULATOR TIME (SEC)423 1460*1916 2541*The simulator exercise imposed a 15 min.delay from when the operator identified the failed PORV to when the PORV was locally opened to account for operator actions outside the control room which could not be verified on the simulator.
This delay is consistent with the assumptions in WCAP-11668.
Simulation of these actions in the actual plant have demonstrated that these times are conservative.
.<5.%e)J CASE 2 RUPTURED SG PORV FAILS OPEN OPERATOR ACTION 1.Ruptured SG Isolated 2.Recognize and Locally Isolate Failed PORV 3.Terminate SI 4.Terminate Break Flow WCAP-11668 TIME (SEC)652 1558 3066 3438 SIMULATOR TIME (SEC)214 1116*2073 2424*The simulator exercise imposed a 15 min.delay from when the operator identified the failed PORV to when the PORV was locally isolated to account for operator actions outside the control room which could not be verified on the simulator.
This delay is consistent with the assumptions in WCAP-11668.
Simulation of these actions in the actual plant have demonstrated that these times are conservative.
These simulator exercises demonstrate that the critical operator action times assumed in WCAP-11668 are realistic and conservative and therefore this contingency is satisfied.
Provide a site specific Steam Generator Tube Rupture radiological offsite consequences analysis.WCAP-11668, provided with this LAR provides a Ginna site specific Steam Generator Tube Rupture radiation offsite consequences analysis, and therefore, this contingency is satisfied.
Provide a structural analysis of the main steam lines demonstrating adequacy under water-filled conditions.
Prior to restart of Ginna Station following the January 25, 1982, tube rupture incident, a main steam line structural analysis under water-filled conditions was performed and provided to the Commission.
The acceptability of this analysis is documented in the re@tart SER(NUREG-0916(section 6.0.Therefore, this contingency is met.A list of systems, components, and instrumentation credited for accident mitigation and the specified safety grade for each.In response to NUREG-0737, Supplement 1 Item 6.2, RG&E has provided post accident instrumentation qualification information.
A comprehensive table listing the credited equipment, its qualification, and all other attributes listed in Regulatory Guide 1.97, revision 3, was provided to the NRC by letter R.Mecredy to A.Johnson"Emergency c n Response Capability", dated October 4, 1992.An SER for~u this submittal was provided to RGGE by letter A.Johnson to R.Mecredy,"Emergency Response Capability," dated February 24, 1993.Therefore, this contingency has been satisfied.
A comparison of the plant to the"bounding plant" used in WCAP-10698.
/Plant parameters for the reference plant used in WCAP-10698-P-A are provided in Table 4.3-3 of the WCAP.WCAP-11668, the Ginna specific analysis, utilizes Ginna specific parameters.
All Ginna specific parameters fall within the bounds of the parameters listed in WCAP-10698-P-A as detailed below: PLANT PARAMETER RCS Pressure, sia Pressurizer Water Volume, ft~SG Secondary Mass, ibm Reactor Trip Delay, sec Turbine Trip Delay, sec Pressurizer Pressure for SI, sia Pressurizer Pressure for Reactor Trip, psia SG Relieve Pressure, psia SIS Pump Delay, sec AFW Delay, sec AFW Flow Rate, gpm AFW Tem erature, 4F f Decay Heat WCAP-10698 BASE CASE 2250 750 107759 2.0 0.3 1864 1960 1100 12 60 1839 40 100't ANS, WCAP-10698 CONSERVATIVE 2220868 118535 0.0 0.0 1889 1985 1050 0.0 0.0 1839 120 120%ANS WCAP-11668 GINNA 2220 800 103256 2.0 0.3 1750 1902 1060 0.0 0.0 800 120 120%ANS It should be noted that the methodology of WCAP-10698-P-A provides a benchmark against the 1982 Ginna tube rupture incident, and, therefore, its applicability to Ginna is explicit.Therefore, this contingency is satisfied.
Based on the above, the methodology described in WCAP-10698-P-A can be applied to Ginna.WCAP-11668 (enclosed) provides the results of this application, and demonstrates the acceptability of Improved Technical Specification coolant activity limits for Ginna.Therefore, the proposed amendment does not involve an unreviewed safety question and will not adversely affect or endanger the health and safety of the general public.
E.Significant zards Consideration Evalua ion The proposed changes~to the Ginna Station Technical Specifications do not involve a significant hazards consideration as discussed below: Operation of Ginna Station in accordance with the proposed changes does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed changes do not affect any accident initiators and therefore the probability of any accident is not increased.
Consequences of the changes are analyzed and shown acceptable in the enclosed analysis, WCAP-11668,Section III.2~Operation of Ginna Station in accordance with the proposed changes does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed changes involve no physical modifications to the plant;therefore, no new accident can be postulated.
3~Operation of Ginna Station in accordance with the proposed changes does not involve a significant reduction in a margin of safety, as no margin of safety is reduced by the proposed changes, as shown in WCAP-11668.
Based upon the above information, it has been determined that the proposed changes to the Ginna Station Technical Specifications do not involve a significant increase in the probability or consequences of an accident previously evaluated, does not create the possibility of a new or different kind of accident previously evaluated, and does not involve a significant reduction in a margin of safety.Therefore, it is concluded that the proposed changes meet the requirements of 10 CFR 50.92(c)and do not involve a significant hazards consideration.
F.Environmental Consideration RGGE has evaluated the proposed changes and determined that: 1.The changes do not involve a significant hazards consideration as documented in Section E above;2~The changes do not involve a significant change in the types or significant increase in the amounts of any effluents that may be released offsite as demonstrated in the enclosed analysis, WCAP 11668.3.The changes do not involve a significant increase in individual or cumulative occupational radiation exposure since the change does not affect allowable limits.Accordingly, the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR
.0 0<<4l 51.22(c)(9).Therefore, pursuant to 10 CFR 51.22(h), an environmental assessment of the proposed changes is not required.References (a): NRC Letter, C.Rossi to A.Ladieu (WOG),"Acceptance for Referencing of Licensing Topical Report WCAP-10698...", March 30, 1987.(b): NUREG-0916,"Safety Evaluation Report Related to the Restart of R.E.Ginna Nuclear Power Plant", May 1982.4 (c): RG&E Letter, R.Mecredy to A.Johnson (NRC),"Emergency Response Capability...", October 14, 1992.(d): NRC Letter, A.Johnson to R.Mecredy (RGGE), Emergency Response Capability
-Conformance to Regulatory Guide 1.97, revision 3", February 24, 1993.