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Category:TECHNICAL SPECIFICATIONS
MONTHYEARML20217G4341999-10-14014 October 1999 Rev C to Proposed TS Change Re Conversion to Improved Standard TSs ML20217G5121999-10-14014 October 1999 Revised Page 285 to TS Re Allowed Containment Leakage Rate, Changing Rev 0 to Rev 1 ML20216J3871999-09-29029 September 1999 Proposed Tech Specs Pages,Extending LCO Allowable Out of Service Time for RHRSW Sys from 7 Days to 11 Days with Special Conditions to Allow for Installation of Mod to Division a RHRSW Strainer ML20196F6071999-06-22022 June 1999 Proposed Tech Specs Re pressure-temp Limits ML20195B8831999-06-0101 June 1999 Proposed Tech Specs,Converting to Improved Std TS ML20206U1421999-05-19019 May 1999 Proposed Tech Specs Revising AOTs for Single Inoperable EDG ML20205K1091999-04-0505 April 1999 Proposed Tech Specs,Removing Position Title of General Manager from Sections & Will State That If Site Executive Officer Is Unavailable,Responsibilities Will Be Delegated to Another Staff Member,In Writing ML20199H3611999-01-15015 January 1999 Proposed Tech Specs Table 4.1-2 Re Local Power Range Monitor (LPRM) Signal Calibr ML20198M8321998-12-30030 December 1998 Proposed Tech Specs Page 258f Re Configuration Risk Mgt Program ML20197G6181998-12-0303 December 1998 Proposed Tech Specs Reducing Size of Spent Fuel Rack Assembly N3 from 8x13 Cells to 8x12 Cells & Deleting Proposed Inclusion of Fuel Pool Water Level Inadvertent Drainage Into Amend ML20154M7181998-10-16016 October 1998 Proposed Corrected Tech Specs Section 3/4.6.C,relocating Portions of Reactor Coolant Sys - Coolant Chemistry ML20236X8041998-08-0303 August 1998 Proposed Tech Specs Section 1.1.A Re SLMCPR to Be Applicable During Cycle 14 ML20236M6231998-07-10010 July 1998 Proposed Tech Specs Pages Re Amend to Relocate TS 3/4.6.C, RCS - Coolant Chemistry, from TS to UFSAR & Applicable Plant Procedures Controlled by 10CFR50.59 Process ML20249B1631998-06-16016 June 1998 Proposed Tech Specs Re Relocation of Safety Review Committee Review & Audit Requirements ML20236L2931998-06-0707 June 1998 Proposed Tech Specs Section 3.5.b.1 Re Main Condenser Steam Jet Air Ejector & Table 3.10-1 Re Radiation Monitoring Sys That Initiates &/Or Isolates Sys ML20217K0391998-03-30030 March 1998 Proposed Tech Specs Changing Interval of Selected LSFT from Semiannually to Once Per 24 Months & Revising Definition for LSFT to Be Consistent w/NUREG-1433 ML20247F7981998-02-26026 February 1998 Proposed Tech Specs Re Allowed Containment Leakage Rate ML20202G8841998-02-0606 February 1998 Revised Proposed TS Pages,Revising Allowed Outage Times for 4kV Emergency Bus Trip Functions & Replace Generic Actions for Inoperable Instrument Channels w/function-specific Actions ML20202D4021998-02-0606 February 1998 Proposed Tech Specs Revising RPS Normal Supply Electrical Protection Assembly Undervoltage Trip Setpoint as Result of Reanalysis Based on Most Limiting Min Voltage Requirements of Applied Loads ML20202D5621998-02-0606 February 1998 Proposed Tech Specs Allowing RCS Pressure Tests to Be Performed While Remaining in Cold Shutdown Mode ML20203J6971997-12-12012 December 1997 Proposed Tech Specs,Revising Administrative Controls for Normal Working Hours of Plant Staff Who Perform Safety Related Functions ML20217J0021997-10-14014 October 1997 Proposed TS Pages Re Changes to Design Features Section, Including Revised Limits for Fuel Storage ML20217G3071997-10-0808 October 1997 Proposed Tech Specs Re Distribution of Inoperable Control Rods ML20211H0801997-09-26026 September 1997 Revised Proposed TS Changes to ASME Section XI, Surveillance Testing ML20236N7061997-09-0909 September 1997 Proposed Tech Specs,Describing Licensee'S Configuration Risk Mgt Program Which Supports Rev of Allowed out-of-svc Times for Single Inoperable EDGs to Accommodate on-line Maint of EDGs ML20140D8131997-04-14014 April 1997 Proposed Tech Specs Re SRC Audit Requirements & Mgt Title Change ML20135B1921996-11-26026 November 1996 Proposed Tech Specs,Requesting That Snubber Operability, Surveillance & Records Requirements in TS Be Relocated to Plant Controlled Documents ML20134M1751996-11-20020 November 1996 Proposed Tech Specs Reflecting Interposed Amend That Was Issued,Updating References to Repts on TS Pp & Changing Rev Bars on Previously Submitted Update Pp Which Were Erroneously Positioned on Pp ML20149L7191996-11-0808 November 1996 Proposed Tech Specs Re Min Critical Power Ratio Safety Limit ML20129G0091996-10-23023 October 1996 Proposed Tech Specs Re Page 134 Deleted Under Amend 236 & Remain Deleted ML20128Q7441996-10-11011 October 1996 Proposed Tech Specs Re Extension of Instrumentation & Miscellanous Surveillance Test to Accommodate 24 Month Cycles ML20115G0641996-07-12012 July 1996 Proposed Tech Specs Re Cycle 12 Min Critical Power Ratio Safety Limit ML20113B6511996-06-20020 June 1996 Proposed Tech Specs Re Option B to 10CFR50,App J for Primary Containment Leakage Rate Testing Program ML20112D1531996-05-30030 May 1996 Proposed Tech Specs,Revising Minimum Critical Power Ratio Safety Limit & Associated Basis ML20112D5711996-05-30030 May 1996 Proposed Tech Specs,Eliminating Selected Response Time Testing Requirements ML20112D2721996-05-30030 May 1996 Proposed Tech Specs Re ATWS Recirculation Pump Trip Instrumentation Requirements JPN-96-023, Proposed Tech Specs Re Deletion of Requirement for PORC to Review Fire Protection Program & Implementing Procedures. Addl Deletion of Insp & Audit Requirements of Specs 6.14.A & 6.14.B1996-05-16016 May 1996 Proposed Tech Specs Re Deletion of Requirement for PORC to Review Fire Protection Program & Implementing Procedures. Addl Deletion of Insp & Audit Requirements of Specs 6.14.A & 6.14.B ML20108D1171996-04-24024 April 1996 Proposed Tech Specs,Supporting Adoption of Primary Containment Lrt Requirements of Option B to 10CFR50,App J & Clarifying Numerical Value of Allowable Containment Leakage Rate as 1.5% Per Day ML20107M4511996-04-24024 April 1996 Proposed Tech Specs 3.11.B/4.11.B Re Crescent Area Ventilation ML20101H6741996-03-27027 March 1996 Proposed Tech Specs,Supporting Adoption of Primary Containment Lrt Requirements of Option B to 10CFR50,App J at Plant & Clarifies Numerical Value of Allowable Containment Lrt as 1.5% Per Day ML20101H3821996-03-22022 March 1996 Proposed TS Table 3.2-2 Re Core & Containment Cooling Sys Initiation & Control Instrumentation Operability Requirements ML20101F8411996-03-22022 March 1996 Proposed Tech Specs,Implementing BWROG Option I-D long-term Solution for Thermal Hydraulic Stability ML20097A2271996-02-0101 February 1996 Proposed Tech Specs,Allowing RCS Pressure Tests to Be Performed While Remaining in Cold Shutdown Mode ML20100C2991996-01-25025 January 1996 Proposed Tech Specs Re EDGs Surveillance Testing ML20097J6691996-01-25025 January 1996 Proposed Tech Specs Re Extension of Instrumentation & Miscellaneous Surveillance Test Intervals to Accommodate 24- Month Operating Cycles ML20095F7091995-12-14014 December 1995 Proposed Tech Specs,Incorporating IST Requirements of Section XI of ASME Boiler & Pressure Vessel Code ML20094R6981995-11-30030 November 1995 Proposed Tech Specs,Extending Surveillance Test Intervals for SLC Sys to Support 24 Month Operating Cycles ML20094B6641995-10-25025 October 1995 Proposed Tech Specs Extending Containment Sys Surveillance Test Intervals to Accommodate 24 Month Operating Cycles ML20092H5401995-09-15015 September 1995 Proposed Tech Specs Extending Surveillance Test Intervals for Auxiliary Electrical Sys to Support 24 Month Operating Cycles ML20086P6561995-07-21021 July 1995 Proposed Tech Specs Re Replacement of title-specific List of PORC Members W/More General Statement of Membership Requirements 1999-09-29
[Table view] Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARML20217G5121999-10-14014 October 1999 Revised Page 285 to TS Re Allowed Containment Leakage Rate, Changing Rev 0 to Rev 1 ML20217G4341999-10-14014 October 1999 Rev C to Proposed TS Change Re Conversion to Improved Standard TSs ML20217D9961999-10-13013 October 1999 Risk-Informed ISI Program Plan for Ja Fitzpatrick ML20216J3871999-09-29029 September 1999 Proposed Tech Specs Pages,Extending LCO Allowable Out of Service Time for RHRSW Sys from 7 Days to 11 Days with Special Conditions to Allow for Installation of Mod to Division a RHRSW Strainer ML20196F6071999-06-22022 June 1999 Proposed Tech Specs Re pressure-temp Limits ML20195B8831999-06-0101 June 1999 Proposed Tech Specs,Converting to Improved Std TS ML20206U1421999-05-19019 May 1999 Proposed Tech Specs Revising AOTs for Single Inoperable EDG ML20205K1091999-04-0505 April 1999 Proposed Tech Specs,Removing Position Title of General Manager from Sections & Will State That If Site Executive Officer Is Unavailable,Responsibilities Will Be Delegated to Another Staff Member,In Writing ML20204B6321999-03-21021 March 1999 Plant Referenced Simulation Facility Four Year Performance Testing Rept ML20199H3611999-01-15015 January 1999 Proposed Tech Specs Table 4.1-2 Re Local Power Range Monitor (LPRM) Signal Calibr ML20206P0541998-12-31031 December 1998 Rev 3.2 to EDAMS/RADDOSE-V ML20198M8321998-12-30030 December 1998 Proposed Tech Specs Page 258f Re Configuration Risk Mgt Program ML20197G6181998-12-0303 December 1998 Proposed Tech Specs Reducing Size of Spent Fuel Rack Assembly N3 from 8x13 Cells to 8x12 Cells & Deleting Proposed Inclusion of Fuel Pool Water Level Inadvertent Drainage Into Amend ML20154M7181998-10-16016 October 1998 Proposed Corrected Tech Specs Section 3/4.6.C,relocating Portions of Reactor Coolant Sys - Coolant Chemistry ML20155E7831998-09-15015 September 1998 Rev 2 to Ja FitzPatrick NPP IST Program for Pumps & Valves Third Interval Plan ML20236X8041998-08-0303 August 1998 Proposed Tech Specs Section 1.1.A Re SLMCPR to Be Applicable During Cycle 14 ML20236M6231998-07-10010 July 1998 Proposed Tech Specs Pages Re Amend to Relocate TS 3/4.6.C, RCS - Coolant Chemistry, from TS to UFSAR & Applicable Plant Procedures Controlled by 10CFR50.59 Process ML20249B1631998-06-16016 June 1998 Proposed Tech Specs Re Relocation of Safety Review Committee Review & Audit Requirements ML20236L2931998-06-0707 June 1998 Proposed Tech Specs Section 3.5.b.1 Re Main Condenser Steam Jet Air Ejector & Table 3.10-1 Re Radiation Monitoring Sys That Initiates &/Or Isolates Sys ML20217K0391998-03-30030 March 1998 Proposed Tech Specs Changing Interval of Selected LSFT from Semiannually to Once Per 24 Months & Revising Definition for LSFT to Be Consistent w/NUREG-1433 B110073, Rev 1 to GE-NE-B1100732-01, FitzPatrick Reactor Pressure Vessel Surveillance Matls Testing & Analysis Rept of 120 Degree Capsule at 13.4 Efpy1998-02-28028 February 1998 Rev 1 to GE-NE-B1100732-01, FitzPatrick Reactor Pressure Vessel Surveillance Matls Testing & Analysis Rept of 120 Degree Capsule at 13.4 Efpy ML20247F7981998-02-26026 February 1998 Proposed Tech Specs Re Allowed Containment Leakage Rate ML20202G8841998-02-0606 February 1998 Revised Proposed TS Pages,Revising Allowed Outage Times for 4kV Emergency Bus Trip Functions & Replace Generic Actions for Inoperable Instrument Channels w/function-specific Actions ML20202D4021998-02-0606 February 1998 Proposed Tech Specs Revising RPS Normal Supply Electrical Protection Assembly Undervoltage Trip Setpoint as Result of Reanalysis Based on Most Limiting Min Voltage Requirements of Applied Loads ML20202D5621998-02-0606 February 1998 Proposed Tech Specs Allowing RCS Pressure Tests to Be Performed While Remaining in Cold Shutdown Mode ML20199G5921998-01-0707 January 1998 Rev 0 to JAF-ISI-0003, Third ISI Interval,Ten-Yr ISI Plan ML20199G5661998-01-0606 January 1998 Rev 0 to JAF-ISI-0002, Third ISI Interval,Isi Program. W/28 Oversize Drawings ML20203J6971997-12-12012 December 1997 Proposed Tech Specs,Revising Administrative Controls for Normal Working Hours of Plant Staff Who Perform Safety Related Functions ML20217J0021997-10-14014 October 1997 Proposed TS Pages Re Changes to Design Features Section, Including Revised Limits for Fuel Storage ML20217G3071997-10-0808 October 1997 Proposed Tech Specs Re Distribution of Inoperable Control Rods ML20217K5841997-09-30030 September 1997 Rev 1 to Ja FitzPatrick Nuclear Power Plant IST Program for Pumps & Valves,Third Interval ML20211H0801997-09-26026 September 1997 Revised Proposed TS Changes to ASME Section XI, Surveillance Testing ML20236N7061997-09-0909 September 1997 Proposed Tech Specs,Describing Licensee'S Configuration Risk Mgt Program Which Supports Rev of Allowed out-of-svc Times for Single Inoperable EDGs to Accommodate on-line Maint of EDGs ML20149G2571997-07-14014 July 1997 JAFNPP ISI Program Relief Requests for 2nd Ten-Yr Interval Closeout ML20140D8131997-04-14014 April 1997 Proposed Tech Specs Re SRC Audit Requirements & Mgt Title Change ML20198G7211997-04-0303 April 1997 Hot Rolled XM-19 Stainless Steel Core Shroud Tie-Rod Matl - Crevice Corrosion Investigation ML20136H3771997-03-11011 March 1997 Rev 0 to Power Uprate Startup Test Rept for Cycle 13 ML20135B1921996-11-26026 November 1996 Proposed Tech Specs,Requesting That Snubber Operability, Surveillance & Records Requirements in TS Be Relocated to Plant Controlled Documents ML20134M1751996-11-20020 November 1996 Proposed Tech Specs Reflecting Interposed Amend That Was Issued,Updating References to Repts on TS Pp & Changing Rev Bars on Previously Submitted Update Pp Which Were Erroneously Positioned on Pp ML20149L7191996-11-0808 November 1996 Proposed Tech Specs Re Min Critical Power Ratio Safety Limit ML20129G0091996-10-23023 October 1996 Proposed Tech Specs Re Page 134 Deleted Under Amend 236 & Remain Deleted ML20128Q7441996-10-11011 October 1996 Proposed Tech Specs Re Extension of Instrumentation & Miscellanous Surveillance Test to Accommodate 24 Month Cycles ML20135D0311996-07-31031 July 1996 Rev 4 to Radiological Effluent Controls & Offsite Dose Calculation Manual ML20115G0641996-07-12012 July 1996 Proposed Tech Specs Re Cycle 12 Min Critical Power Ratio Safety Limit ML20113B6511996-06-20020 June 1996 Proposed Tech Specs Re Option B to 10CFR50,App J for Primary Containment Leakage Rate Testing Program ML20112D1531996-05-30030 May 1996 Proposed Tech Specs,Revising Minimum Critical Power Ratio Safety Limit & Associated Basis ML20112D2721996-05-30030 May 1996 Proposed Tech Specs Re ATWS Recirculation Pump Trip Instrumentation Requirements ML20112D5711996-05-30030 May 1996 Proposed Tech Specs,Eliminating Selected Response Time Testing Requirements JPN-96-023, Proposed Tech Specs Re Deletion of Requirement for PORC to Review Fire Protection Program & Implementing Procedures. Addl Deletion of Insp & Audit Requirements of Specs 6.14.A & 6.14.B1996-05-16016 May 1996 Proposed Tech Specs Re Deletion of Requirement for PORC to Review Fire Protection Program & Implementing Procedures. Addl Deletion of Insp & Audit Requirements of Specs 6.14.A & 6.14.B ML20108D1171996-04-24024 April 1996 Proposed Tech Specs,Supporting Adoption of Primary Containment Lrt Requirements of Option B to 10CFR50,App J & Clarifying Numerical Value of Allowable Containment Leakage Rate as 1.5% Per Day 1999-09-29
[Table view] |
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ATTACHMENT l
i o PROPOSED TECHNICAL SPECIFIC TION CHANGES REGARDING -
SAFti V UMIT MINIMUM CRITICAL POWER RATIO.
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- JPTS-89 036 :
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4 New York Power Authority
- JAMES A. FITZPATRICK NvCLEAR POWER PLANT Docket No. 50-333 l
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1.1 FUEL Cl. ADDING INTEGRITY 2.1 FUEL Ct. ADDING INTEGRl'iY '
Applicability: Applicability-The Safety Umits established to preserve the fuel cladding integrity The Umiting Safety System Settings apply to trip settings of the instruments and devices which are provided to prevent the fuel cladding apply to those variables which monitor the fuel thermal behavior.
Objective:
UN U U* N ** -
ective:
The objective of the Safety Umits is to establish limits below which the The objective of the Umiting Safety System Settings '.s to define the level integrity of the fuel cladding is preserved.
of the proce;,s variables at which automatic protective action is irwtiated to prevent the fuel cladding integrity Safety Umits from being exceeded.
Specifications:
Spacifications:
A. Trip Settings A. Reactor Pressure >785 psig and Core Flow > 10% of Rated The limiting safety system trip settings shall be as specEisd The existence of a minimum critical power ratio (MCPR) less below-than 1.07 shall constitute violation of the fuel cladding integrity l l 1. Neutron Flux Trip Settings safety limit, hereafter called the Safety Umit. An MCPR Umit of 1.08 shall apply during single-loop operation. IRM - The IRM flux scram setting shall be set at l
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<120/125 of full scale.
l Amendment No.14, &f,30,46,96,1//
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JAFNPP_
1.1 BASES 1.1 FUEL CLADDING INTEGRITY A.- Reactor Pmssum ' M @ W Ne h MM N Onset of transition boiling results in a decrease in heat transfer The fuel cladding integrity limit is set such that no calculated from the clad and, therefore, elevated clad temperature and the '
fuel damage would occur as a result of an abnormal possibility of clad failure. However, the existence of cntcal operational transient.' Because fuel damage is not directly .
observable, a step-back approach is used to establish a Safety power, or boiling transition,is not a directly observable Umit such that the minimum critical power ratio (MCPR) is no parameter in an Opmeting reactor. Therefore, the margin to less than 1.07. MCPR > 1.07 represents a conservative margin boiling transstion is calculated from plant operating parameters l such as core power, core flow, feedwater temperature, and -
relative to the conditions required to maintain fuel cladding core power distribution. The margin for each fuel assembly is integrity. The fuel cladding is one of the physical barriers which characterized by the critical power ratio (CPR) which is the separate radioactive materials from the environs. The integnty ratio of the bundle power which would produce onset of of this cladding barrier is related to its relative freedom from transition boiling dmded by the actual bundle power. The perforations or cracking. Although some corrosion or use minimum value of this ratio for any bundle in the core is the related cracking may occur during the life of the cladding, minimum critical power ratio (MCPR). It is assumed that the fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding plant operation is controlled to the reinnai protective setpoints -
via the instrumented variable, i.e., the operating domain. The perforations, however, can result from thermal stresses which current load line limit analysis contains the current operating occur from reactor operation significantly above design domain map. The Safety Umit (MCPR of 1.07) has sufficient conditions and the protection system safety settings. While fission product migration from cladding perforation is just as - conservatism to assure that in the event of an isrema! '
measurable as that from use related cracking, the thermally - operational transient initiated from the MCPR operating conditions in specification 3.1.8, more than 99.9% of the fuel caused cladding perforations signal a threshold, beyond which rods in the core are expected to avoid boiling transition. The '
still greater thermal stresses may cause gross rather than MCPR fuel cladding safety limit is increased by 0.01 for single-incremental cladding deterioration. Therefore, the fuel cladding loop operation as discussed in Reference 2.' The margin Safety Umit is defined with margin to the conditions which l.
between MCPR of 1.00 (onset of transition boiling) and the would produce onset of transition boiling, (MCPR of 1.00).
l These conditions represent a significant departure from the Safety Umit is derived from a detailed statistical analysis ]
considering all of the uncertainties in monitoring the core condition intended by design for planned operation.
operating state including the uncertainty in the boiling transition -
correlation as described in Reference 1. The uncertanties employed in deriving the Sabty Umit are Amendment No. 14,18,21,30,46,7f,98,Jd 12
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- 5. ',~
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. ATTACHMENT 11
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SAFETY EVALUATION FOR PROPOSED TECH 6GES REGARDING SAFETY UMIT MINIMUM CRITICAL POWER RATIO JPTS-89 036
'k New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333
. ."" ' --'."- Attachment II SAFETY EVALUATION ~
Page 1 of 3 -
U
- 12 DESCRIPTION OF THE PROPOSED CHANGES T
The proposed changes to the James A. FitzPatrick Technical Specifications revise-Specification 1.1.A and the associated Bases on pages 7 and 12.
Page 7, Specification 1.1.A Replace "1.04" with *1.07" and replace "1.05" with '1.08.*
Page 12, Bases 1.1 In two places, replace *1.04' with '1.07*
Replace "1.0" with "1.00.*
Page 12,- Bases 1.1.A g Replace "1.04* with "1.07."
Replace "1.0" with "1.00."
- 11. -- PURPOSE OF THE PROPOSED CHANGES The purpose of the proposed Technical Specification changes is to revise the safety limit Minimum Critical Power Ratio (MCPR) to the appropriate value to support the Reload
, 9/ Cycle 10 core. Reload 9 will consist nf General Electric GE-10 fuel assemblies
. designated GE8x8NB 3 in GE correspondence with the NRC and four GE 11 Lead Test Assemblies. The revised safety limit MCPR will be conservatively applied to all the fuel in the core. The MCPR value of 1.0 in the Bases is modified to 1.00 to show the same number of signficant figures as the safety limit MCPR.
These fuel assemblies have two unique features: an interactive channel design and an offset lower tie plate. The offset lower tio plate shifts the fuel bundle 40 mils towards the
' control blado, making the D-lattice core of the FitzPatrick plant more like the C-lattice cores of later BWR plants.
Although this reload core affects other cycle-specific parameter limits contained in the technical specifications, no other changes are required. The Authority is removing the remainder of the cycle-specific limits from the technical specifications in a separate amendment requcat. These other changes will be reflected in a Cycle 10 Core Operating Umits Report (COLR) in accordance with the guidance contained in Generic Letter 8816.
The COLR will be provided to the NRC prior to the startup of Cycle 10. Because the change requested in this application involves revision to a safety limit, it is inappropriate to remove it from the specifications, even though its value may change based upon the fuel design selected for the reload core (Reference 3).
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- w. .: Y Attachment ll
. SAFETY EVALUATION Page 2 of 3
- Ill. IMPACT OF THE PROPOSED CHANGES -
The unique features of the GE8x8NB 3 fuel design have been reviewed fully and approved by the NRC for use in BWRs. In the NRC acceptance of this fuel design (Reference 1), the NRC stated that the GE8x8NB C lattice safety limit MCPR is acceptable for the GE8x8NB-3 fuel design. -This safety limit of 1.07 is provided in Table 4 2, " Fuel Cladding Integrity Safety Umit MCPR" of General Electric's " Standard Application for Reactor Fuel" (Reference 2). This revised MCPR safety limit provides the same assurance as the previous safety limit value in preventing boiling transition. All aspects of this reload fuel design and its use at the FitzPatrick plant have been reviewed and approved by the NRC.
The application of this fuel design at the FitzPatrick plant allows greater operational
. flexibility and improved fuel performance, in addition, the enrichment and gadolinium loadings are designed to support the longer,24 month cycle length planned for Cycle 11.
- l. IV. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION -
Operation of the FitzPatrick plant in accordance with the proposed Amendment would not involve a significant hazards consideration as defined in 10 CFR 50.92, since It would not: ,
- 1. . Involve a s!gnificant increase in the probability or consequences of an accident previously evaluated.
L NRC approved methodologies and codes have been used to perform all analyses concerning General Electric Co. fuel to be loaded at this refueling (Reference 1). The fuel design has been reviewed and approved for use at the FitzPatrick plant under - the constraints and methodologies detailed in References 1 and 2. There are no unique aspects of this fuel or its application L which have not undergone prior NRC review and approval. The refueling of l the FitzPatrick reactor and Cycle 10 operation does not result in an increase in the probability or consequences of any accident previously evaluated.
- 2. create the possibility of a new or different kind of accident from any "
accident previously evaluated.
Refueling the FitzPatrick reactor Is a periodic evolution performed in g- accordance with appropriate procedures and controlled by the Technical Specifications. The GE-10 fuel bundles inserted as Reload 9 have been fully L reviewed by the NRC (Reference 1), and their use will not create the possibility l of a new or different type of accident. The nuclear characteristics of the E individual fuel bundles and the core loading pattem have been fully analyzed L by the General Electric Co. and do not create the possibility of a new or different type of accident.
!' 3. Involve a significant reduction in a margin of safety.
The analyses performed in support of this reload assure maintenance of all existing margins of safety. These analyses have resulted in core wide and
! bundle specific limits for General Electric Co. fuel which, when applied to the i
reloaded core, assure operation within the design criteria previously approved in References 1 and 2. The revised MCPR safety limit provides an equivalent margin of safety as the previous safety limit value in preventing boiling L
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p l.,,,*** Attachment 11 SAFETY EVALUATION Page 3 of 3 ,
i transition. Therefore, the proposed change does not result in the reduction of any margin of safety.
V. IMPLEMENTATION OF THE PROPOSED CHANGE Implementation of the proposed changes will not impact the ALARA or Fire Protection Programs at FitzPatrick, nor will the changes impact the environment.
=
VI. CONCLUSION The change, as proposed, does not constitute an unreviewed safety question as defined in 10 CFR 50.59. That is, it:
- a. will not change the probability nor the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the Safety Analysis Report;
- b. will not increase the possibility of an accident or malfunction of a type different from any previously evaluated in the Safety Analysis Report;
- c. will not reduce the margin of safety as defined in the basis for any technical specification; and
- d. involves no significant hazards consideration, as defined in 10 CFR 50.92.
Vll. REFERENCES
- 1. NRC letter, A. C. Thadani to J. S. Charnley (GE), " Acceptance for Referencing of Amendment 21 to General Electric Ucensing Topical Report NEDE 24011-P-A, ' General Electric Standard Application for Reactor Fuel,'" dated March
= 17,1989.
- 2. " General Electric Standard Application for Reactor Fuel," Revision 9, dated September,1988.
- 3. GE letter, J. S. Charnley to M. W. Hodges (NRC), " Acceptance implementation of Generic Letter 88-16," dated August 8,1989.
- 4. James A. FitzPatrick Nuclear Power Plant Updated Final Safety Analysis Report.
- 5. James A. FitzPatrick Nuclear Power Plant Safety Evaluation Report (SER),
dated November 20,1972, and Supplements.
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