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Category:TECHNICAL SPECIFICATIONS
MONTHYEARML17250B3061999-10-20020 October 1999 Proposed Tech Specs,Consisting of 1999 Changes to TS Bases ML17265A6941999-06-28028 June 1999 Proposed Tech Specs,Revising ITS Associated with RCS Leakage Detection Instrumentation,As Result of Commitment Submitted as Part of Staff Review of Application of leak-before-break Status to Protions of RHR Piping ML17265A5911999-03-0101 March 1999 Proposed Tech Specs Change Revising Battery Cell Parameters Limit for Specific Gravity (SR 3.8.6.3 & SR 3.8.6.6) ML17265A4661998-11-24024 November 1998 Proposed Tech Specs 4.2.1,revising Description of Fuel Cladding Matl & Updating List of References Provided in TS 5.6.5 for COLR ML17265A3151998-06-0404 June 1998 Proposed Tech Specs Basis Re Main Steam Isolation Setpoint ML17265A2941998-05-21021 May 1998 Tech Specs Consisting of Submittal Changes for 1998 ML17265A2461998-04-27027 April 1998 Proposed Tech Specs Revising Requirements Associated W/Sfp to Reflect Planned Mod to Storage Racks & Temporarily Addressing Boraflex Degradation within Pool ML17264B0671997-10-0808 October 1997 Proposed Tech Specs,Correcting & Clarifying Info Re RCS Pressure & Temp Limits Rept Administrative Controls Requirements ML17264B0441997-09-29029 September 1997 Proposed Tech Specs Revising Adminstrative Controls W/ Respect to Reactor Coolant Sys Pressure & Temp Limits Rept ML17264B0391997-09-29029 September 1997 Proposed Tech Specs Revising 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 Rev Revised Setpoint Analysis Study ML17264A9991997-08-19019 August 1997 Proposed Tech Specs Correcting Specified Accumulator Borated Water Volume Values in SR 3.5.1.2 to Match Associated Accumulator Percent Level Values ML17264B0031997-08-19019 August 1997 Proposed Tech Specs Allowing Testing of Three ECCS motor- Operated Valves in Mode 4 Which Currently Requires Entry Into LCO 3.0.3 ML17264A9241997-06-0303 June 1997 Proposed Tech Specs Clarifying Issues Re Low Temperature Overpressure Protection ML17264A8671997-04-24024 April 1997 Proposed Tech Specs,Revising Rcs,Pt & Administrative Control Requirements ML17264A8491997-03-31031 March 1997 Proposed Tech Specs 3.7.12 Re Spent Fuel Pool Boron Concentration ML17264A7601996-12-16016 December 1996 Proposed Annual TS Bases ML17264A7131996-10-29029 October 1996 Proposed Tech Specs Re Actions for Inoperable Channels Associated with Auxiliary Feedwater Pump ML17264A6001996-09-13013 September 1996 Proposed Tech Specs Re RCS Pressure & Temp Limits Rept ML17264A5301996-06-0303 June 1996 Proposed Tech Specs Re Update to Correction of Typographical Error Request ML17264A5111996-05-29029 May 1996 Proposed Tech Specs Pages 3.3-18 & 3.3-19,reflecting Revs to LCO 3.3.1 ML17264A4711996-05-0808 May 1996 Proposed Tech Specs,Correcting Typos ML17264A4081996-03-20020 March 1996 Proposed Tech Specs 3.9.3,Containment Penetrations Re Use of Roll Up Door & Associated Encl Building for LCO 3.9.3 Requirements ML17264A4051996-03-15015 March 1996 Proposed Tech Specs Changing Section 5.6.6 Page 5.0-22 to Incorporate Methodology for Determining RCS P/T & LTOP Limits Into Administrative Controls Section for RCS P/T Limits Rept ML17264A3451996-02-0909 February 1996 Proposed Tech Specs,Revising Setpoints for SG Water Level High Feedwater Isolation Function ML17264A3401996-02-0909 February 1996 Proposed Tech Specs Re LTOP Limits Using Util Proposed Methodology ML17264A3341996-02-0909 February 1996 Proposed Tech Specs Containment Requirements During Mode 6 Cost Beneficial Licensing Action ML17264A3231996-01-26026 January 1996 Proposed Tech Specs Re Nuclear Fuel Cycle for NRC Approval ML17264A2811995-12-0808 December 1995 Proposed Tech Specs Re Changes to Current RCS Pressure & Temp Limits for Heatup,Cooldown,Criticality & Hydrostatic Testing ML17264A2531995-11-27027 November 1995 Proposed Tech Specs for Implementation of 10CFR50,App J, Option B ML17264A2571995-11-20020 November 1995 Proposed Tech Specs,Discussing Changes to TS Instrumentation Requirements & Conversion to Improved TS ML17264A2491995-11-20020 November 1995 Proposed Tech Specs Re Ventilation Filter Testing Program ML17264A1511995-08-31031 August 1995 Proposed Tech Specs Implementing WCAP-10271,its Assoc Suppls & Other Re Changes W/Respect to RTS & ESFAS ML17263A9781995-03-13013 March 1995 Proposed Tech Specs to Revise TS 4.4.2.4.a,replacing Specific Leakage Testing Frequencies for Containment Isolation Valves ML17263A7901994-09-27027 September 1994 Proposed TS Section 6.0, Administrative Controls. ML17263A7281994-07-15015 July 1994 Proposed Tech Specs Providing NRC W/Opportunity to Communicate at Early Stage Any Concerns W/Respect to Differences from NUREG-1431 ML17263A6581994-05-23023 May 1994 Proposed Tech Specs,Increasing Allowable Reactor Coolant Activity Levels ML17263A6431994-05-13013 May 1994 Proposed Tech Specs Administrative Section 6.0 ML17263A3701993-08-20020 August 1993 Corrected Proposed TS Page 5.1-1,changing Word Released to Leased ML17263A3571993-08-0606 August 1993 Proposed Tech Specs Re Alternative Requirements for Snubber Visual Insp Intervals ML17263A3191993-07-15015 July 1993 Proposed Tech Specs,Removing Containment Isolation Valve Table 3.6-1 from TS ML17263A2101993-04-0505 April 1993 Proposed TS Table 3.6-1, Containment Isolation Valves. ML17262B1191992-12-17017 December 1992 Proposed Tech Specs Sections 3.2 & 3.3 Re Acid Storage Tank Boron Concentration Reduction Study ML17309A5021992-11-30030 November 1992 Proposed Tech Specs Reflecting Removal of Table of Containment Isolation Valves ML17262B0441992-10-0808 October 1992 Proposed TS 3.7.1 Re Auxiliary Electrical Sys ML17262B0211992-09-15015 September 1992 Proposed Tech Specs Sections 3.1.1.4,3.1.1.6 & 4.3.4, Addressing GL 90-06, Resolution of Generic Issue 70, 'Porv & Block Valve Reliability' & Generic Issues 94, 'Addl LTOP for Lwrs.' ML17262A9141992-06-22022 June 1992 Proposed TS 4.3.1 Re Reactor Vessel Matl Surveillance Testing ML17262A8321992-04-23023 April 1992 Proposed Tech Specs Re Snubber Visual Insp Schedule ML17262A8221992-04-21021 April 1992 Proposed Tech Specs Re Fire Protection Program ML17262A7871992-03-23023 March 1992 Proposed Tech Spec Revising Section 6.5.1 Re Plant Operations Review Committee Function ML17262A7911992-03-20020 March 1992 Proposed Tech Specs Revising 6.9.1.2 & 6.9.2.5 1999-06-28
[Table view] Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARML17250B3061999-10-20020 October 1999 Proposed Tech Specs,Consisting of 1999 Changes to TS Bases ML17265A6941999-06-28028 June 1999 Proposed Tech Specs,Revising ITS Associated with RCS Leakage Detection Instrumentation,As Result of Commitment Submitted as Part of Staff Review of Application of leak-before-break Status to Protions of RHR Piping ML17265A6401999-05-12012 May 1999 Rev 11 to Technical Requirements Manual for Ginna Station. ML17265A6261999-04-18018 April 1999 Rev 10 to Technical Requirements Manual (Trm), for Ginna Station ML17311A0701999-04-14014 April 1999 Rev 10 to AP-PRZR.1, Abnormal Pressurizer Pressure. ML17309A6521999-04-14014 April 1999 Rev 14 to AP-RCS.1, Reactor Coolant Leak. ML17309A6531999-04-13013 April 1999 Rev 1 to FIG-2.0, Figure Sdm. with 990413 Ltr ML17265A6111999-03-26026 March 1999 Rev 9 to Technical Requirements Manual for Ginna Station. ML17265A5911999-03-0101 March 1999 Proposed Tech Specs Change Revising Battery Cell Parameters Limit for Specific Gravity (SR 3.8.6.3 & SR 3.8.6.6) ML17265A5571999-02-25025 February 1999 Rev 4 to Technical Requirements Manual (Trm). ML17265A5491999-02-12012 February 1999 to EOP FR-H.5, Response to SG Low Level. ML17265A5481999-02-12012 February 1999 to EOP ATT-22.0, Attachment Restoring Feed Flow. ML17265A5461999-02-12012 February 1999 0 to EOP AP-TURB.1, Turbine Trip Without Rx Trip Required. ML17309A6481999-01-25025 January 1999 Revised Ginna Station Emergency Operating Procedures. with 990125 Ltr ML17265A5151999-01-14014 January 1999 Revised Emergency Operating Procedures,Including Rev 14 to AP-SW.1,rev 3 to ATT-5.2,rev 5 to ATT-8.0,rev 1 to ATT-14.6, Rev 17 to E-1,rev 16 to ECA-1.1,rev 26 to ES-1.3,rev 4 to FR-Z.2 & Index ML17265A5081999-01-0808 January 1999 Rev 7 to Technical Requirements Manual, for Ginna Station ML17265A4961998-12-18018 December 1998 Rev 6 to Technical Requirements Manual. ML20198C0361998-12-14014 December 1998 Rev 11 to ECA-0.2, Loss of All AC Power Recovery with SI Required ML20198C0121998-12-14014 December 1998 Rev 10 to AP-RCS.2, Loss of Reactor Coolant Flow ML20198C0581998-12-14014 December 1998 Rev 5 to FR-Z.1, Response to High Containment Pressure ML20198C0411998-12-14014 December 1998 Rev 16 to FR-C.1, Response to Inadequate Core Cooling ML20198C0521998-12-14014 December 1998 Rev 13 to FR-S.1, Response to Reactor Restart/Atws ML17265A4661998-11-24024 November 1998 Proposed Tech Specs 4.2.1,revising Description of Fuel Cladding Matl & Updating List of References Provided in TS 5.6.5 for COLR ML20155G0301998-10-30030 October 1998 Rev 13 to EOP AP-CCW.1, Leakage Into Component Cooling Loop ML17265A4091998-08-24024 August 1998 Rev 13 to AP-SW.1, Svc Water Leak. W/980824 Ltr ML17265A3721998-07-16016 July 1998 Rev 14 to EOP AP-CW.1, Loss of Circ Water Pump. W/980716 Ltr ML17265A3151998-06-0404 June 1998 Proposed Tech Specs Basis Re Main Steam Isolation Setpoint ML17265A2941998-05-21021 May 1998 Tech Specs Consisting of Submittal Changes for 1998 ML17265A2861998-05-0606 May 1998 Rev 19 to EOP ECA-3.2, SGTR W/Loss of Reactor Coolant Saturated Recovery Desired & Updated ECA Index.W/980506 Ltr ML17265A2461998-04-27027 April 1998 Proposed Tech Specs Revising Requirements Associated W/Sfp to Reflect Planned Mod to Storage Racks & Temporarily Addressing Boraflex Degradation within Pool ML17265A1821998-02-20020 February 1998 Rev 1 to EWR 5111, MOV Qualification Program Plan, Calculation Assumption Verification Criteria. ML17265A1491997-12-31031 December 1997 Rev 1 to Inservice Testing Program. ML17264B0731997-10-14014 October 1997 Rev 4 to Technical Requirements Manual (TRM) for Ginna Station. ML17264B0671997-10-0808 October 1997 Proposed Tech Specs,Correcting & Clarifying Info Re RCS Pressure & Temp Limits Rept Administrative Controls Requirements ML17264B0441997-09-29029 September 1997 Proposed Tech Specs Revising Adminstrative Controls W/ Respect to Reactor Coolant Sys Pressure & Temp Limits Rept ML17264B0391997-09-29029 September 1997 Proposed Tech Specs Revising 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 Rev Revised Setpoint Analysis Study ML17265A1541997-09-16016 September 1997 Rev 1 to DA EE-92-089-21, Design Analysis Ginna Station Instrument Loop Performance Evaluation & Setpoint Verification. ML17264B0031997-08-19019 August 1997 Proposed Tech Specs Allowing Testing of Three ECCS motor- Operated Valves in Mode 4 Which Currently Requires Entry Into LCO 3.0.3 ML17264A9991997-08-19019 August 1997 Proposed Tech Specs Correcting Specified Accumulator Borated Water Volume Values in SR 3.5.1.2 to Match Associated Accumulator Percent Level Values ML17264A9791997-08-0505 August 1997 Rev 7 to AP-RCS.3, High Reactor Coolant Activity. ML17264A9781997-08-0505 August 1997 Rev 12 to AP-RCS.1, Reactor Coolant Leak. ML17264A9771997-08-0505 August 1997 Rev 11 to AP-RCP.1, RCP Seal Malfunction. ML17264A9751997-08-0505 August 1997 Rev 14 to AP-ELEC.1, Loss of 12A & 12B Busses. ML17264A9851997-08-0404 August 1997 Rev 2 to Summary Description of Compliance w/10CFR73 Amend Protection Against Malevolent Use of Vehicles at Nuclear Power Plants. ML17264A9391997-07-0707 July 1997 Rev 3 to Technical Requirements Manual for Ginna Station, Inserting New Tabs Accordingly Per Previous Transmittal for Rev ML17264A9341997-07-0101 July 1997 to Technical Requirements Manual (TRM) & inter-office Correspondence Dtd 970624 ML17264A9461997-06-20020 June 1997 to Inservice Insp (ISI) Program. ML17264A9241997-06-0303 June 1997 Proposed Tech Specs Clarifying Issues Re Low Temperature Overpressure Protection ML17311A0471997-05-22022 May 1997 Revised Eops,Including Procedures Index,Rev 5 to ATT-15.0, Rev 3 to ATT-15.2,rev 3 to AP-ELEC.3,rev 13 to ECA-0.1 & Rev 9 to ECA-0.2.W/970522 Ltr ML17264A8671997-04-24024 April 1997 Proposed Tech Specs,Revising Rcs,Pt & Administrative Control Requirements 1999-06-28
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Attachment B Revise the Technical Specification pages as follows:
Remove Insert 5.4-1 5.4-1 5.4-2 5.4-2 5.4-3 5.4-3 5.4-4 5.4-4
- 5. 4-5 5.4-5 960%300309 960126 PDR ADOCK 05000244 P PDR
Attachment B 5.4 Fuel Stora e S ecification 5.4. 1 The new and spent fuel pit structures are designed to withstand the anticipated earthquake loadings as Class I structures. The spent fuel pit has a stainless steel liner to ensure against loss of water.
5.4.2 The new and spent fuel storage racks are designed so that it is impossible to insert fuel assemblies in other than the prescribed locations. The spent fuel storage racks are divided into two regions as depicted on Figure 5.4-1.
The fuel is stored vertically in an array with sufficient center to center distance between assemblies to assure Keff g 0.95 for (1) unirradiated fuel assemblies delivered prior to January 1, 1984 (Region 1-15) containing no more than 39.0 gms U-235 per axial cm, and (2) unirradiated fuel assemblies delivered between January 1, 1984 and February 1, 1996 containing no more than 41.9 gms U-235 per axial cm, and (3) unirradiated fuel assemblies delivered after Feb. 1, 1996 containing no more than 49.8 grams U-235 per axial cm. All cases assume unborated water used in the pool.
5.4.3 Xn Region 2 of the spent fuel storage racks, fuel is stored in a close packed array utilizing fixed neutron poisons in each of the stored locations. For discharged fuel assemblies to be stored in Region 2, (1) 60 days must have elapsed since the core reached hot shutdown prior to discharge and (2) the combination of assembly average burnup and initial U-235 enrichment must be such that the point identified by these two parameters on Figure 5.4-2 is above the line applicable to that particular fuel assembly design, therefore assuring that Keff 5 0.95.
5.4-1
5.4.4 Canisters containing consolidated fuel rods may be stored in either Region 1 or 2 provided that:
average burnup and initial enrichment of the fuel I'he a~
assemblies from which the rods were removed satisfy the requirements of 5.4.2 and 5.4.3 above, and
- b. the average decay heat of the fuel assembly from which the rods were removed is less than 2150 BTU/hr 5.4.5 The requirements of 5.4.4a may be excepted for those I
consolidated fuel assemblies of Region RGAF2.
5.4.6 The spent fuel storage pit is filled with borated water at a concentration to match that used in the reactor cavity and refueling canal during refueling operations whenever there is fuel in the pit.
Basis The center to center spacing of Region 1 insures that Keff 5 0.95 for the enrichment limitations specified in 5.4.2'4, and for, a postulated missile impact the resulting dose at the EAB would be within the guidelines of 10CFR100~. Fuel assemblies with an enrichment of < 4.05 w/o can be stored in any available location. Fuel assemblies with an enrichment > 4.05 w/o can also be stored in Region I provided that integral burnable poisons are present in the assemblies such that k-infinity is ~ 1.458.
In Region 2, Keff ~ 0.95 is insured by the addition of fixed neutron poison (boraflex) in each of the Region 2 storage locations, and a minimum burnup requirement as a function of initial enrichment for each fuel assembly design. The 60 day cooling time requirement insures that for a postulated missile impact the resulting dose at the EAB would be within the guidelines of 10CFR100.
5.4-2
1 The two curves of Figure 5.4-2 divide the fuel assembly designs into two groups. The first group is all fuel except Exxon fuel delivered prior to January 1, 1984, which incorporates all Westinghouse HIPAR designs used at Ginna.4 The second curve is for all Exxon fuel, as well as the Westinghouse Optimized Fuel assembly design delivered to .Ginna beginning in February 1984.~
The assembly average burnup is calculated using INCORE generated power sharing data and the actual plant operating history. The calculated assembly average burnup should be reduce'd by 104 to account for uncertainties. An uncertainty .of 4% is associated with the measurement of power sharing. The additional 64 provides additional margin to bound the burnup uncertainty associated with the time between measurements and updates of core burnup.
The calculations of fuel assembly burnup for comparison to the curves of Figure 5.4-2 to determine the acceptability for storage in Region 2 shall be independently checked. The record of these calculations shall be kept for as long as .fuel assemblies remain in the pool.
The fuel storage canisters are designed so that, normally, they can contain the equivalent number of fuel rods from two fuel assemblies in a close packed array, and can be stored in either Region 1 or Region 2 rack locations. The close packed array will insure the K~ of the rack configuration containing any number of canisters will be less than that for stored fuel assemblies at the same burnup and initial enrichment. The exception of paragraph 5.4.5 is possible because the consolidated configuration is substantially less reactive than that of a fuel assembly. The maximum decay heat requirement will ensure that local and film boiling will not occur between the close packed fuel rods
- 5. 4-3
4
~ ~
if the pool temperature is maintained at or below 150'F. The decay heat of the assembly will be determined using ANS 5.1, ASB 9-2 or other acceptable substitute standards.
With the addition of the storage of consolidated fuel canisters, the theoretical storage capacity of the pool would be increased to 2032 fuel assemblies (2xl016). Moreover, due to limitation on the heat removal capability of the spent fuel pool cooling system, the storage capacity is limited to 1016 fuel assemblies.~
References 1 ~ Letter, J.E. Maier to H.R. Denton, January 18, 1984.
2 ~ Safety Evaluation from John Zwolinski to Roger Kober, November 14, 1984, "Increase of the Spent Fuel Pool Storage Capacity."
3 ~ Criticality Analysis of Region 2 of the Ginna MDR Spent Fuel Storage Rack, Pickard, Lowe and Garrick, Inc. March 8, 1984.
4 ~ Letter, T.R. Robbins, Pickard, Lowe and Garrick, Inc. to J.D.
Cook, RGGE March 15, 1984.
- 5. Letter, D.M. Crutchfield to J.E. Maier, November 5. 1981.
- 6. Safety Evaluation from Allen Johnson to Dr. Robert C. Mecredy, August 30. 1995, "Proposed Criticality Analysis of Ginna New and Spent Fuel Racks/Consolidated Rod Storage Canisters.
- 5. 4-4
40000 ACCEPTABLE 30000 R
CL K
20000 i5 E
UNACCEPTABLE 10000
~
OFA Fuel & Exxon Fuel
---. S 1O Fuel 1.8 2.6 3.0 3.4 3.8 4.2 5.0 Nominal U Ennc hment (m/o)
Figure 5.4-2 Fuel Assembly Burnup Limits in Region 2 5.4-5
Attachment C The purpose of this amendment is to allow storage of Westinghouse OFA fuel with a nominal enrichment of up to 5.0 w/o U-235 in the new and spent fuel storage racks.
In Reference (a) RGGE submitted an engineering calculation that demonstrated that the allowable enrichment of fuel to be stored in the new and spent fuel storage racks could be increased to a nominal 5.0 w/o U-235. There were no restrictions on placement in the new fuel storage racks. Placement in Region 1 of the spent fuel storage racks requires a minimum number of integral fuel burnable absorber (IFBA) rods. Placement in Region 2 of the spent fuel storage racks requires a minimum fuel burnup. This requirement is illustrated on Figure 5.4-2 of the proposed Amendment.
The NRC staff has reviewed the engineering calculations and issued an SER in Reference (b). The SER concluded that the criticality aspects of the proposed enrichment increase was acceptable and met the requirements of General Design Criteria 62 for the prevention of criticality in fuel storage and handling.
In accordance with 10 CFR 50.91, these changes to Technical Specifications have been evaluated to determine if the operation of the facility in accordance with the proposed amendment would:
- 1. involve a significant increase in the probability or consequences of an accident previously evaluated; or
- 2. create the possibility of a new or different kind of accident previously evaluated; or
- 3. involve a significant reduction in a margin of safety.
As presented in the SER, the referenced engineering calculations demonstrate that the acceptance criterion are satisfied. The proposed change does not increase the probability or consequences of a previously evaluated accident or create a new or different kind of accident. Further, there is no unacceptable reduction in the margin of safety for any Technical Specification.
Therefore, Rochester Gas and Electric submits that the issues associated with this Amendment are outside the criteria of 10 CFR 50.91, and a no significant hazards finding is warranted.
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