JAFP-06-0015, Application for Technical Specification Amendment to Support Implementation of Arts/Meod

Application for Technical Specification Amendment to Support Implementation of Arts/Meod

ML060390370
Person / Time
Site:	FitzPatrick
Issue date:	01/26/2006
From:	Ted Sullivan Entergy Nuclear Northeast, Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
JAFP-06-0015, TAC MC3390
Download: ML060390370 (38)

Text

Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.

James A. Fitzpatrick NPP P.O. Box 110 EnteLLgy

____g Lycoming, NY 13093 Tel 315 349 6024 Fax 315 349 6480 T.A. Sullivan Site Vice President - JAF January 26, 2006 JAFP-06-001 5 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Entergy Nuclear Operations, Inc.

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 License No. DPR-59 Application for Technical Specification Amendment to Support Implementation of ARTS/MEOD

Dear Sir:

Pursuant to 10 CFR 50.90, Entergy Nuclear Operations, Inc. (ENO) hereby requests an amendment to the Technical Specifications (TS) for the James A. FitzPatrick Nuclear Power Plant (JAF).

The proposed amendment will modify TS requirements to support the implementation of Average Power Range Monitor, Rod Block Monitor, Technical Specifications/Maximum Extended Operating Domain Analysis (ARTS3/MEOD). Similar TS changes have been approved for other BWRs. The NRC is currently reviewing a similar change for the Hope Creek Generating Station, TAC No. MC3390.

The proposed changes implement a more direct power- and flow-dependent thermal limits application in lieu of the periodic APRM scram and rod block trip setdown process. The changes also support operation in an expanded operating domain. These changes will reduce reactivity adjustments by control rod manipulations during xenon transient conditions and normal operations. Maintaining rated power conditions with changes in core flow is a more uniform method of maintaining power as compared to control rod manipulations.

Attachment 1 provides a description and assessment of the proposed changes, as well as the no significant hazards consideration and environmental assessment.

Attachment 2 provides the existing TS pages marked-up to show the proposed changes.

Attachment 3 provides revised TS pages (re-typed).

Attachment 4 provides the existing TS Bases pages marked-up to show the proposed changes.

The Bases changes are provided for information only. The final TS Bases pages will be submitted with a future update in accordance with TS 5.5.11, "Technical Specifications (TS)

Bases Control Program."

contains the "J. A. FitzPatrick Nuclear Power Plant APRM/RBM/Technical Specifications/Maximum Extended Operating Domain (ARTS/MEOD)," NEDC-33087P, Revision 1, September 2005, including the safety analysis prepared by General Electric (GE) to support this proposed changes. GE considers Attachment 5 proprietary information. In accordance with 10CFR2.390(b)(1), an affidavit attesting to the proprietary nature of the information (report) is enclosed within Attachment 5. The original affidavit is provided at the end of Attachment 5. This analysis was reviewed and accepted in accordance with JAF's design review and verification process. contains a non-proprietary version (NEDO-33807) of Attachment 5.

In accordance with 10 CFR 50.91, a copy of this application, with the associated attachments, is being provided to the designated New York State official.

ENO requests approval of the proposed license amendment by January 31, 2007.

There are no commitments made by the Licensee in this letter.

Should you have any questions, please cont1act Mr. Jim Costedio at 315-349-6358.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on the 2- D' day of January, 2006.

Sincerely, Site Vice President TAS:tp Attachments: 1. Description and Assessment

2. Proposed Technical Specification Pages (Marked-Up)

3. Proposed Technical Specification Pages (Re-Typed)

4. Proposed Technical Specification Bases Pages (Marked-Up)

5. "J. A. FitzPatrick Nuclear Power Plant APRM/RBM/Technical Specifications/Maximum Extended Operating Domain (ARTS/MEOD),"

NEDC-33087P, Revision 1, September 2005 - Proprietary, and original affidavit

6. "J. A. FitzPatrick Nuclear Power Plant APRM/RBM/Technical Specifications/Maximum Extended Operating Domain (ARTS/MEOD),"

NEDC-33087, Revision 1, September 2005 - Non-proprietary cc: next page

cc:

Regional Administrator, Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415 Resident Inspector's Office U.S. Nuclear Regulatory Commission James A. FitzPatrick Nuclear Power Plant P.O. Box 136 Lycoming, NY 13093 Mr. John P. Boska, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop O-8-C2 Washington, DC 20555 Mr. Paul Eddy (w/o Attachment 5)

New York State Department of Public Service 3 Empire State Plaza Albany, NY 12223-1350 Mr. Peter R. Smith, President (w/o Attachment 5)

New York State Energy, Research, and Development Authority 17 Columbia Circle Albany, NY 12203-6399

Attachment 1 to JAFP-06-0015 Entergy Nuclear Operations, Inc. - FitzPatrick Docket No. 50-333 DESCRIPTION AND ASSESSMENT

Attachment 1 to JAFP-06-0015 Entergy Nuclear Operations, Inc. - FitzPatrick Docket No. 50-333 Description and Assessment I. DESCRIPTION The proposed Technical Specifications (TS) changes support the implementation of Average Power Range Monitor, Rod Block Monitor, Technical Specifications/Maximum Extended Operating Domain Analysis (ARTS/MEOD) at the James A. FitzPatrick Nuclear Power Plant (JAF). The BWR/4 Standard Technical Specifications (NUREG-1433) contain two alternate thermal limit formulations. Plants either impose power and flow dependent limits (ARTS), or use an Average Power Range Monitor (APRM) setdown procedure to provide off-rated protection. JAF has always used the latter method, but intends to implement ARTS/MEOD in conjunction with an expansion of the power-flow map commonly known as Maximum Extended Load Line Limit Analysis (MELLLA). MEOD is defined as the combination of the power/flow operating map expansion with MELLLA and increased core flow (ICF).

JAF thermal limits are currently adjusted by a Kf curve that applies to Minimum Critical Power Ratio (MCPR) only for offrated flow conditions, and indirectly by the APRM setdown or adjustment method when the Maximum Fraction of Limiting Power Density (MFLPD) is greater than the Fraction of Rated Power (FRP). In lieu of these methods, JAF proposes to adjust MCPR, Linear Heat Generation Rate (LHGR) and Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) at offrated conditions, (i.e.,

less than rated power and flow), using a set of power and flow corrections. These corrections are commonly known as ARTS thermal limit curves, and are used at many domestic Boiling Water Reactors (BWRs).

Several factors restrict the flexibility of a BWR during power ascension from the low-power / low-core flow condition to the high-power / high-core flow condition. Once rated power is achieved, numerous control rod manipulations are required to compensate for reactivity changes due to xenon effects. In addition, control rods are adjusted to compensate for gadolinium (poison) and fuel burn up during the cycle at steady state conditions and at end of cycle. To reduce the number of control rod manipulations and enhance flexibility, Entergy Nuclear Operations, Inc. (ENO) is proposing a MEOD region to the currently licensed allowable operating power/flow map.

Similar TS changes have been approved for other BWRs. The NRC is currently reviewing a similar change for the Hope Creek Generating Station, TAC No. MC3390.

PROPOSED CHANGES The proposed changes to the JAF TS support implementation of ARTS/MEOD. The specific TS changes are described below. The appropriate changes will also be made to the associated TS Bases in accordance with TS 5.5.11, TS Bases Control Program.

1. TS Section 1.1, Definitions Delete the definition of 'MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLPD)"

1

Attachment 1 to JAFP-06-0015 Entergy Nuclear Operations, Inc. - FitzPatrick Docket No. 50-333 Description and Assessment

2. TS Section 3.2, Power Distribution Limits Delete Limiting Condition for Operation (LCO) 3.2.4, "Average Power Range Monitor (APRM) Gain and Setpoint," in its entirety and revise the Table of Contents accordingly.

3. TS Section 3.3.1.1, Reactor Protection System (RPS) Instrumentation Revise Surveillance Requirement (SR) 3.3.1.1.2 by deleting the words 'plus any gain adjustment required by lICO 3.2.4, 'Average Power Range Monitor (APRM)

Gain and Setpoint,"'

Ill. BACKGROUND The JAF ARTS/MEOD program utilizes the results of anticipated operational occurrence (AOO) analyses to define initial condition operating thermal limits over the range of allowed power and flow combinations, to conservatively ensure that all licensing criteria are satisfied without setdown of the flow-referenced APRM scram and rod block trip setpoints. The objective of the APRMA improvements is to justify removal of the APRM trip setdown requirement. Two licensing areas that can be affected by the elimination of the APRM trip setdown requirement are: (1) fuel thermal-mechanical integrity and (2)

Loss of Coolant Accident (LOCA) analysis.

The following criteria ensure satisfaction of the applicable licensing requirements. They were applied to demonstrate the acceptability of elimination of the APRM trip setdown requirement:

1. The Safety Limit MCPR (SLMCPR) shall not be violated as a result of any AOO.

2. All fuel thermal-mechanical design bases shall remain within the licensing limits described in the General Electric (GE) generic fuel licensing report (GESTAR-Il).

3. Peak cladding temperature and maximum cladding oxidation fraction following a LOCA shall remain within the limits defined in 10 CFR 50.46.

The safety analyses used to evaluate the Operating Limit MCPR (OLMCPR) such that the SLMCPR is satisfied and that the fuel thermal-mechanical design bases are satisfied are documented in Attachment 5 ("J. A. FitzPatrick Nuclear Power Plant APRM/RBM/Technical Specifications/Maximum Extended Operating Domain (ARTS/MEOD)," NEDC-33087P, Revision 1, September 2005). These analyses also establish the fuel type specific power-dependent and flow-dependent MCPR, LHGR, and MAPLHGR curves for JAF. The effect on the LOCA response due to the ARTS program implementation is discussed in this attachment as well.

JAF is currently licensed to operate in the Extended Load Line Limit Analysis (ELLLA) region up to the 108% rod line, which results in a core flow window of 87% to 100% at rated thermal power (RTP). This operation is analyzed in NEDC-32016P-1, "Power 2

Attachment 1 to JAFP-06-0015 Entergy Nuclear Operations, Inc. - FitzPatrick Docket No. 50-333 Description and Assessment Uprate Safety Analysis for the James A. FitzPatrick Nuclear Power Plant," Revision 1, April 1993 (Reference I of this Attachment). The proposed TS changes support JAF operation in a region which is above the 108% rod line (100% power / 87% flow). The current operating envelope is modified to include the extended operating region bounded by a line which passes through the 100% power / 80% flow point (the MELLLA line). The MELLLA line is shown in Figure 1-1 of Attachment 5.

Expansion of the operating domain (MELLLA) and implementation of ARTS will reduce reactivity adjustments by control rod manipulations during xenon transient conditions and normal operations. A larger fraction of reactivity changes to maintain 100% RTP will be made by changing core flow which is a more uniform method of maintaining power while imposing lower fuel duty compared to control rod manipulations.

The technical analysis is referred to as MEOD, and is defined as the combination of the power/flow operating map expansion with MELLLA and ICF. The safety analyses and system evaluations performed to justify operation in the MEOD region consist of a non-fuel dependent portion and a fuel dependent portion that is fuel cycle dependent. In general, the limiting AOO MCPR calculation and the reactor vessel overpressure protection analysis are fuel dependent. The non-fuel dependent evaluations such as containment response are based on the current hardware design and plant geometry.

The analytical results also indicate that performance in the MEOD region is within allowable design limits for overpressure protection, LOCA, containment dynamic loads, flow-induced vibration and reactor internals structural integrity, and meets the Anticipated Transient Without Scram (ATWS) licensing criteria.

Physical changes to support ARTS/MEOD operations were completed and tested during Cycle 17 and included upgrading to digital APRM Flow Control Trip Reference (FCTR) cards and revising APRM and Rod Block Monitor (RBM) instrument channel setpoints.

These modifications were completed using the JAF nuclear change process and were engineered to meet applicable design and regulatory criteria.

This ARTS application does not include modifications to the RBM system as some ARTS programs have in the past. The purpose of the RBM is to limit the consequences of a Rod Withdrawal Error (RWE) AOO to protect the SLMCPR and thermal-mechanical design basis of the fuel. An RBM setpoint change was made based on the assumption that no credit is taken for RBM actuation to limit a RWE at JAF. Since no credit is taken for the RBM in mitigating the RWE, the resultant OLMCPR for the RWE increases.

However, the pressurization AQOs remain bounding and continue to establish the limiting OLMCPR. Therefore, SLMCPR and the thermal-mechanical design basis of the fuel continue to be met.

3

Attachment 1 to JAFP-06-0015 Entergy Nuclear Operations, Inc. - FitzPatrick Docket No. 50-333 Description and Assessment IV. TECHNICAL ANALYSIS The JAF TS require that the plant operate such that the core MFLPD, which is equivalent to the ratio of Maximum Total Peaking Factor (MTPF) to the Design Total Peaking Factor (DTPF), does not exceed the FRP. This requirement limits the maximum local power at lower core power and flows to a fraction of that allowed at rated power and flow. If the MTPF exceeds the DTPF, the flow-referenced APRM trips must be lowered (setdown) or APRMs adjusted to MTPF to limit the maximum power that the plant can achieve. The basis for this 'APRM trip setdown" requirement originated under the original BWR design Hench-Levy Minimum Critical Heat Flux Ratio (MCHFR) thermal limit criterion and provides conservative restrictions with respect to current fuel thermal limits.

JAF currently operates under the GE Thermal Analysis Basis critical power correlation, which replaced the minimum critical heat flux basis. The GE Thermal Analysis Basis emphasis on bundle critical power rather than local critical heat flux allows for a more direct determination of fuel thermal limits.

The proposed TS changes support the implementation of ARTS/MEOD. The safety analyses used to evaluate the OLMCPR such that the SLMCPR is satisfied and to ensure that the fuel thermal-mechanical design bases are satisfied are documented in Attachment 5. These analyses establish the fuel type specific power-dependent and flow-dependent MCPR, LHGR, and MAPLHGR curves for JAF. The effect on the LOCA response due to the ARTS program implementation is also discussed in Attachment 5.

Operation in an expanded operating domain is based on a MEOD analysis performed by GE using the methods described in Attachment 5. This analysis was reviewed and accepted in accordance with JAF's design review and verification process. Operation at high loadlines (in the MELLLA region) may result in high local peaking that, though acceptable by analyses presented in this application, would not be achievable by the current APRM setdown method (e.g., a plant startup could not proceed if a resultant setdown inserted an APRM rod block).

The current power-flow map is modified to include the extended operating region bounded by a line which passes through the 100% power / 80% core flow point (i.e., the MELLLA line, the rated power line and the ELLLA line). Plant operating efficiency is increased by the power-flow map expansion, which updates the thermal limits requirements.

The safety analyses and system evaluations performed to justify operation in the MEOD region consist of a non-fuel dependent portion and a fuel dependent portion that is cycle dependent. In general, the limiting ADO MCPR calculation and the reactor vessel overpressure protection analysis are fuel dependent. These analyses, as discussed in Attachment 5, are based on the assumption of a representative GE12 and GE14 core (Cycle 16 core design). Subsequent cycle-specific analyses are performed in conjunction with the reload licensing activities. The non-fuel dependent evaluations are 4

Attachment 1 to JAFP-06-0015 Entergy Nuclear Operations, Inc. - FitzPatrick Docket No. 50-333 Description and Assessment based on the current hardware design and plant geometry. The limiting AQOs were reviewed for the MEOD region based on a review of existing thermal analysis limits at plants similar to JAF and use of generic power-dependent and generic flow-dependent MCPR and MAPLHGR limits. For the fuel-dependent evaluations of reactor pressurization events, these reviews indicate that there is a small difference in the OLMCPR for operation in the MEOD region and the current licensed thermal power (CLTP) condition (100% of CLTP / 100% of RCF (rated core flow)). The actual operating limit is calculated on a cycle specific basis to bound the entire operating domain. The analysis results also indicate that performance in the MEOD region is within allowable design limits for overpressure protection, LOCA, containment dynamic loads, flow-induced vibration and reactor internals structural integrity, and ATWS licensing criteria.

The analyses which justify operation in the MEOD region under the stated conditions are discussed in Attachment 5 and its supporting references. These analyses include fuel performance event evaluations, mechanical evaluations of the reactor internals, structural vibration assessment, LOCA evaluations, and containment loads evaluations.

NRC-approved or industry-accepted computer codes and calculational techniques are used in the ARTS/MEOD analyses. A 10 CFR Part 21 notice concerning the effect of the Turbine Control System on the Transient Analysis was addressed in establishing the power-dependent MCPR, LHGR and MAPLHGR limits, as described in Section 3.3.1 of . The adequacy of the margin to the Standby Liquid Control System (SLCS) relief valve lifting as described in NRC Information Notice 2001-13, Inadequate Standby Liquid Control System Relief Valve Margin," was also addressed, as described in Section 11 of Attachment 5.

CONCLUSION The proposed TS changes implement a more direct thermal limit application process that avoids periodic APRM bypassing for adjustment under the current process. The changes support operation in an expanded operating domain. Operation in the expanded operating domain is based on analysis performed by GE using methods described in Attachment 5. This analysis was reviewed and accepted in accordance with JAF's design review and verification process. Plant operating efficiency is increased by the ARTS changes which updates thermal limits requirements. Physical changes to the plant were completed and tested to accommodate the expanded operating region and included upgrading to digital APRM FCTR cards and implementing revised APRM and RBM setpoints. These modifications were completed using the JAF nuclear change process and were engineered to meet applicable design and regulatory criteria.

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Attachment 1 to JAFP-06-0015 Entergy Nuclear Operations, Inc. - FitzPatrick Docket No. 50-333 Description and Assessment V. REGULATORY ANALYSIS NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION According to 10 CFR 50.92(c), "Issuance of amendment," a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

(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 from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

In accordance with 10 CFR 50.90, Application for amendment of license or construction permit," Entergy Nuclear Operations, Inc. (ENO) is proposing changes to the James A.

FitzPatrick Nuclear Power Plant (JAI-) Technical Specifications (TS). The proposed changes delete the definition for MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLPD), delete LCO 3.2.4, Average Power Range Monitor (APRM) Gain and Setpoint, in its entirety, and revise SR 3.3.1.1.2 by deleting words referring to LCO 3.2.4 which state 'plus any gain adjustment required by LCO 3.2.4, 'Average Power Range Monitor (APRM) Gain and Setpoint, ".

The proposed TS changes support an expanded operating domain resulting from the proposed implementation of the Average Power Range Monitor, Rod Block Monitor, Technical Specifications/Maximum Extended Operating Domain Analysis (ARTS/MEOD).

1. Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes revise thermal limit structure employed to comply with TS Section 3.2 LCOs. The proposed changes will replace the flow-biased APRM scram and rod block trip setdown requirements with power and flow dependent adjustments to the Minimum Critical Power Ratio (MCPR) and Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) or Linear Heat Generation Rate (LHGR) thermal limits. The adjustments to the thermal limits have been determined using NRC approved analytical methods as required by Technical Specifications 5.6.5.b and topical reports as specified in the Core Operating Limits Report (COLR).

The proposed changes will not affect any accident initiating mechanism.

Adjustments to thermal limits will be determined using NRC approved methodologies. The power and flow dependent adjustments will ensure that the MCPR safety limit will not be violated as a result of any anticipated operational occurrence (AOO), that the fuel thermal and mechanical design bases will be maintained, and that the consequences of the postulated loss of coolant accident 6

Attachment 1 to JAFP-06-0015 Entergy Nuclear Operations, Inc. - FitzPatrick Docket No. 50-333 Description and Assessment (LOCA) will remain within acceptable limits. There are no changes to radioactive source terms or release pathways. Operation within the expanded operating domain has been evaluated and the affect on plant accidents was found to be within acceptable parameters. The proposed changes do not result in any significant change in the availability of logic systems or safety-related systems themselves.

Required protective functions will be maintained. The proposed changes do not degrade plant design, operation, or the performance of any safety system assumed to function in the accident analysis.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes do not introduce any new accident initiators or failure mechanisms because the changes and the affects on existing structures, systems and components have been evaluated and found to not have any adverse affects.

The proposed changes eliminate the requirement for setdown of the flow-biased APRM scram and rod block trip setpoints or APRM adjustments under specified conditions and will substitute adjustments to the MCPR and MAPLHGR or LHGR thermal limits. Because the thermal limits will continue to be met, no transient event will escalate into a new or different type of accident due to the initial starting conditions permitted by the adjusted thermal limits.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident than those previously evaluated.

3. Involve a significant reduction in a margin of safety.

The proposed changes do not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. There is no affect on the conclusions of any safety analysis. Replacement of the APRM setpoint requirement with power and flow dependent adjustments to the MCPR and MAPLHGR or LHGR thermal limits will continue to ensure that margins to the fuel cladding Safety Limit are preserved during operation at other than rated conditions.

The fuel cladding safety limit will not be violated as a result of any anticipated operational occurrence. The flow and power dependent adjustments will be determined using NRC approved methodologies. The flow and power dependent adjustments will also ensure that all fuel thermal-mechanical design bases shall remain within the licensing limits. The proposed changes do not involve any increase in calculated off-site dose consequences. Operability of protective instrumentation and the associated systems is assured, and performance of equipment will not be significantly affected.

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Attachment 1 to JAFP-06-0015 Entergy Nuclear Operations, Inc. - FitzPatrick Docket No. 50-333 Description and Assessment Therefore, there is no significant reduction in the margin of safety as a result of the proposed changes.

Based on the above evaluation, ENO has concluded that the proposed changes do not involve a significant hazards consideration.

VI. ENVIRONMENTAL ASSESSMENT ENO has evaluated the proposed changes and determined the changes do not involve (1) a significant hazards consideration, (2) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (3) a significant increase in the individual or cumulative occupational radiation exposure. Accordingly, the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9), and an environmental assessment of the proposed changes is not required.

VII. REFERENCES

1. NEDC-32016P-1, 'Power Uprate Safety Analysis for the James A. FitzPatrick Nuclear Power Plant," Revision 1, April 1993 8

Attachment 2 to JAFP-06-0015 Entergy Nuclear Operations, Inc. - FitzPatrick Docket No. 50-333 PROPOSED TECHNICAL SPECIFICATION PAGES (MARKED-UP)

TS PAGES TOC i 1.1-4 3.2.4-1 3.2.4-2 3.3.1.1-3

TABLE OF CONTENTS 1.0 USE AND APPLICATION 1.1 Definitions . . . . ... . . . . . . . . . . . . . 1.1-1 1.2 Logical Connectors . ... . . . . . . . . . . . . . 1.2-1 1.3 Completion Times . . ... . . . . . . . . . . . . . 1.3-1 1.4 Frequency . . . . . ... . . . . . . . . . . . . . 1.4-1 2.0 SAFETY LIMITS (SLs) 2.1 SLs . . . . . . . .. 2.0-1 2.2 SL Violations . . . ... . . . . . . . . . . . . . 2.0-1 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY 3.0-1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY . . . . . 3.0-3 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM) . . . . . . . . . . . . . . 3.1.1-1 3.1.2 Reactivity Anomalies . . . . . . . . . . . . . . . 3.1.2-1 3.1.3 Control Rod OPERABILITY . . . . . . . . . . . . . 3.1.3-1 3.1.4 Control Rod Scram Times . . . . . . . . . . . . . 3.1.4-1 3.1.5 Control Rod Scram Accumulators . . . . . . . . . . 3.1.5-1 3.1.6 Rod Pattern Control . . . . . . . . . . . . . . . 3.1.6-1 3.1.7 Standby Liquid Control (SLC) System . . . . . . . 3.1.7-1 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves 3.1.8-1 3.2 POWER DISTRIBUTION LIMITS 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) . 3.2.1-1 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR) . . . . . . . . . 3.2.2-1 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) . . . . . . . 3.2.3-1 gg#< ,-'vera 24ower Mon tr R~e (!yoam epo . . .4-3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation . . . 3.3.1.1-1 3.3.1.2 Source Range Monitor (SRM) Instrumentation . . . . . . 3.3.1.2-1 3.3.2.1 Control Rod Block Instrumentation . . . . . . . . . . 3.3.2.1-1 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentat'ion .............. 3.3.2.2-1 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation . . . . 3.3.3.1-1 3.3.3.2 Remote Shutdown System . . . . . . . . . . . . . . . . 3.3.3.2-1 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RFPT) Instrumentation . . . . . . 3.3.4.1-1 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation 3.3.5.1-1 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation . . . . . . . . . . . . . . . . . 3.3.5.2-1 3.3.6.1 Primary Containment Isolation Instrumentation . . . . 3.3.6.1-1 3.3.6.2 Secondary Containment Isolation Instrumentation . . . 3.3.6.2-1 3.3.7.1 Control Room Emergency Ventilation Air Supply (CREVAS)

System Instrumentation . . . . . . . . .. . . . 3.3.7.1-1 3.3.7.2 Condenser Air Removal Pump Isolation Instrumentation 3.3.7.2-1 3.3.7.3 Emergency Service Water (ESW) System Instrumentation 3.3.7.3-1 (continued)

JAFNPP i Amendment .74-

Definitions 1.1 1.1 Definitions (continued)

LINEAR HEAT GENERATION The LHGR shall be the heat generation rate per RATE (LHGR) unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL A LOGIC: SYSTEM FUNCTIONAL TEST shall be a test TEST of all logic components required for OPERABILITY of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

The MFL.PD spen be the lar value of t fraction limiting pow density in t core.

The fr- ion of limiti power densit shall be the lR existing at given locati divided by th j specified LHGR imit for that undle type.

MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power RATIO (MCPR) ratio (CPR) that exists in the core for each type of fuel. The CPR is that power in the assembly i

that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature. and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

OPERABLE -OPERABILITY A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem. division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

(continued)

)

JAFNPP 1.1-4 Amendment 274-

APRM Gain and Setpoint 3.2.4 3.2 POWER DISTRIBUTION LIMITS

)

JAFNPP 3.2.4-1 Amendment 4274

APRM Gain and Setpoint 3.2.4 SURVEILLANCE RthIREMENTS 7" v/SUdI H LLANCE / S ENCY SR ..4.1 ---- ..------------NOTE ..............---

N required to be me if SR 3.2.4.2 is atisfied for LCO 3 .4.b or LCO 3.2.4.c requirements.

Verify MFLP is within limits. Once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af r k 25% RTP AND 2 /ours ereafter SR 3. .4.2--------- . NOTE-------------...

Not requiredo be met if SR 3.2.4.1 s satisfied or LCO 3.2.4.a requirem ts.

JAFNPP 3.2.4-2 Amendment -2i7+

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS &

..- ------------------ NOTES------------------------------------

1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.

2. When a channel is placed in an Iinoperable status solely for performance of required Surveillances, entry i into associated Conditions and Required Actions may be delayed for up Ito 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function maintains RPS trip capability.

SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.3.1.1.2 ----------------- NOTE-------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER 2 25X RTP.

Verify the absolute difference between 7 days the average power range monitor (APRM) channels and the calculated wr i

/rfiredb5LCO KS4.)>^verasge'Powr>' ,

an onitor M in a Se in,'

while operating al: 2 25% RTP.

SR 3.3.1.1.3 ----------------- NOTE-------------------

Not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering MODE 2.

Perform CHANNEL FUNCTIONAL TEST. 7 days SR 3.3.1.1.4 Perform a functional test of each RPS 7 days automatic scram contactor.

(continued)

JAFNPP 3.3.1.1-3 Amendment Attachment 3 to JAFP-06-0015 Entergy Nuclear Operations, Inc. - FitzPatrick Docket No. 50-333 PROPOSED TECHNICAL SPECIFICATION PAGES (RE-TYPED)

TS PAGES TOC i 1.1-4

3.3.1.1-3

TABLE OF CONTENTS 1.0 USE AND APPLICATION 1.1 Definitions ............. 1.1-1 1.2 Logical Connectors ............. 1.2-1 1.3 Completion Times ............ 1.3-1 1.4 Frequency ............ 1.4-1 2.0 SAFETY LIMITS (SLs) 2.1 SLs .................................................. 2.0-1 2.2 SL Violations .................................................. 2.0-1 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY .................. 3.0-1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY ................................ 3.0-3 3.1 REACTIVITY CONTROL SYSTEM";

3.1.1 SHUTDOWN MARGIN (SDMVI) .............. .......................... 3.1.1-1 3.1.2 Reactivity Anomalies ........................................ 3.1.2-1 3.1.3 Control Rod OPERABILITY ........................................ 3.1.3-1 3.1.4 Control Rod Scram Times ........................................ 3.1.4-1 3.1.5 Control Rod Scram Accumulators ....................... ................. 3.1.5-1 3.1.6 Rod Pattern Control ........................................ 3.1.6-1 3.1.7 Standby Liquid Control (SLC) System ........................................ 3.1.7-1 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves ................... 3.1.8-1 3.2 POWER DISTRIBUTION LIMITS 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) ...... 3.2.1-1 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR) ...................................... 3.2.2-1 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) ......................................... 3.2.3-1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation ............... ........... 3.3.1.1-1 3.3.1.2 Source Range Monitor (SRM) Instrumentation .................... ............ 3.3.1.2-1 3.3.2.1 Control Rod Block Instrumentation .............................................. 3.3.2.1-1 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation ............. 3.3.2.2-1 3.3.3.1 Post Accident Monitoring (IPAM) Instrumentation .................. .......... 3.3.3.1-1 3.3.3.2 Remote Shutdown System .......................................... 3.3.3.2-1 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPTX Instrumentation ................................... 3.3.4.1-1 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation ............... 3.3.5.1-1 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation........................................................................... 3.3.5.2-1 3.3.6.1 Primary Containment Isolation Instrumentation .............................. 3.3.6.1-1 3.3.6.2 Secondary Containment Isolation Instrumentation ......................... 3.3.6.2-1 3.3.7.1 Control Room Emergency Ventilation Air Supply (CREVAS)

System Instrumentation ................ 3.3.7.1-1 3.3.7.2 Condenser Air Removal Pump Isolation Instrumentation .......... ...... 3.3.7.2-1 3.3.7.3 Emergency Service Water (ESW) System Instrumentation .............. 3.3.7.3-1 (continued)

JAFNPP i Amendment

Definitions 1.1 1.1 Definitions (continued)

LINEAR HEAT GENERATION The LHGR shall be the heat generation rate per unit RATE (LHGR) length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all TEST logic components required for OPERABILITY of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power that I

RATIO (MCPR) exists in the core for each type of fuel. The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

OPERABLE-OPERABILITY A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

(continued)

JAFNPP 1.1-4 Amendment

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS

---

N------NOTES--------------------------------

1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function maintains RPS trip capability.

SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.3.1.1.2 --------------- NOTE---------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER 2 25% RTP.

Verify the absolute difference between the 7 days average power range monitor (APRM) channels and the calculated power is

• 2% RTP while I operating at 2 25% RTP.

SR 3.3.1.1.3 ---------------NOTE---------------

Not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering MODE 2.

Perform CHANNEL FUNCTIONAL TEST. 7 days SR 3.3.1.1.4 Perform a functional test of each RPS automatic 7 days scram contactor.

(continued)

JAFNPP '}.3.1.1-3 Amendment

Attachment 4 to JAFP-06-0015 Entergy Nuclear Operations, Inc. - FitzPatrick Docket No. 50-333 PROPOSED TECHNICAL SPECIFICATION BASES PAGES (MARKED-UP)

TS BASES PAGES TOC i 13 3.2.1-1 13 3.2.2-1 E3 3.2.2-2 F3 3.2.2-4 F3 3.2.4-1

[3 3.2.4-2

[3 3.2.4-3 B3 3.2.4-4

[3 3.2.4-5

[3 3.2.4-6 B 3.3.1.1-26

TABLE OF CONTENTS B 2.0 SAFETY LIMITS (SLs)

B 2.1 Reactor Core SLs . . . . . . . . . 6 . . . . . . . . B 2.1.1-1 B 2.2 Reactor Coolant System (RCS) Pressure SL . . . . . B 2.1.2-1 B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY . . B 3.0-1 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY . . . . . . B 3.0-12 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM) . . . . . . . . . . . . . . B 3.1.1-1 B 3.1.2 Reactivity Anomalies . . . . . . . . . . . . . . . B 3.1.2-1 B 3.1.3 Control Rod OPERABILITY . . . . . . . . . . . . . B 3.1.3-1 B 3.1.4 Control Rod Scram Times . . . . . . . . . . . . . B 3.1.4-1 B 3.1.5 Control Rod Scram Accumulators . . . . . . . . . . B 3.1.5-1 B 3.1.6 Rod Pattern Control .. . . . . . . . . . . . . B 3.1.6-1 B 3.1.7 Standby Liquid Control (SLC) System . . . .l.e. B 3.1.7-1 B 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves B 3.1.8-1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) B 3.2.1-1 B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR) . . . . . . . . B 3.2.2-1 B 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) . . . . . . . . . B 3.2.3-1

.4

( Aver ge or1 n tePower B 3.3 ][NSTRUMENTATION B 3.3.1.1 Reactor Protection System (RPS) Instrumentation . . I3 3.3.1.1- 1 B 3.3.1.2 Source Range Monitor (SRM) Instrumentation . . . . .I 3 3.3.1.2-1 B 3.3.2.1 Control Rod Block Instrumentation . . . . . . . . .I 3 3.3.2.1-1 B 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation . . . . . . . . . . . . . 3 3.3.2.2-1 B 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation . . . 3 3.3.3.1-1 B 3.3.3.2 Remote Shutdown System . . . . . . . . . . . . . . . 3 3.3.3.2-1 B 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation . . . . . 3 3.3.4.1-1 B 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation I3 3.3.5.1-1 B 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation .......... 1 3 3.3.5.2-1 B 3.3.6.1 Primary Containment Isolation Instrumentation . . . IB 3.3.6.1-1 B 3.3.6.2 Secondary Containment Isolation Instrumentation . . F3 3.3.6.2-1 B 3.3.7.1 Control Room Emergency Ventilation Air Supply (CREVAS)

System Instrumentation 3 3.3.7.1-1 B 3.3.7.2 Condenser Air Removal Pump Isolation Instrumentation i3 3.3.7.2-1 B 3.3.7.3 Emergency Service Water (ESW) System Instrumentation I3 3.3.7.3-1 B 3.3.8.1 Loss of Power (LOP) Instrumentation . . . . . . . . 3 3.3.8.1-1 B 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring . . . . . . . . . . . . . . . . . . E3 3.3.8.2-1 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1 Recirculation Loops Operating . . . . . B 3.4.1-1 B 3.4.2 Jet Pumps . . . . . . . . . . . . . . . B 3.4.2-1 (continued)

JAFNPP i Revision APLHGR B 3.2.1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)

BASES BACKGROUND The APLHGR is a measure of the average LHGR of all the fuel rods in a fuel assembly at any axial location. Limits on the APLHGR are specified to ensure that the peak cladding temperature (PCT) during the postulated design basis loss of coolant accident MLOCA) does not exceed the limits specified in 10 CFR 50.46.

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the fuel design limits are presented in Reference 1. The analytical methods and assumptions used in evaluating the Design Basis Accident (DBA) that determines the APLHGR limits are presented in References 1, 3. 4, 5, 6, and 7.

APLHGR limits are equivalent to the assumed APLHGR in the LOCA analysis or the LHGR design limit (assuming a 1.0 local peaking factor) whichever is less. APLHGR limits are developed as a function of exposure for each fuel bundle design (Refs. 5 and 6).

GE~Z 2 LOCA analyses are then performed to ensure that the above determined APLHGR limits are adequate to meet the PCT and maximum oxidation limits of 10 CFR 50.46. The analysis is performed using calculational models that are consistent with the requirements of 10 CFR 50, Appendix K. A complete discussion of the analysis code is provided in Reference 7. I The PCT following a postulated LOCA is a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is not strongly influenced by the rod to rod power distribution within an assembly. A conservative multiplier is applied to the LHGR I assumed in the LOCA analysis to account for the uncertainty associated with the measurement of the APLHGR.

(continued)

JAFNPP B 3.2.1-1 Revision is

INSERT 1 Flow dependent Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits, MAPFAC(F), were designed to assure adherence to all fuel thermal-mechanical design bases. The same transient events used to support the MCPR(F) operating limits were analyzed, and the resulting overpowers were statistically evaluated as a function of the initial and maximum core flow. From the bounding overpowers, the MAPFAC(F) limits were derived such that the peak transient LHGR would not exceed fuel mechanical limits. The flow-dependent MAPLHGR limits are generic, cycle-independent, and are specified in terms of multipliers, MAPFAC(F), to be applied to the rated MAPLHGR values.

Power-dependent MAPLHGR limits, expressed in terms of a MAPLHGR multiplier, MAPFAC(P), is substituted to assure adherence to the fuel thermal-mechanical design bases. Both incipient centerline melting of fuel and plastic strain of the cladding are considered in determining the power dependent MAPLHGR limit.

Generally, the limiting criterion is incipient centerline melting. The power-dependent MAPFAC(P) multipliers were generated using the same database as used to determine the MCPR multiplier (Kp). Appropriate MAPFAC(P) multipliers are selected based on plant-specific transient analyses with suitable margin to assure applicability to future JAF reloads. These limits are derived to assure that the peak transient MAPLHGR for any transient is not increased above the fuel design bases values.

MCPR B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)

BASES BACKGROUND MCPR is a ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power. The MCPR Safety Limit (SL) is set such that 99.9% of the fuel rods avoid boiling transition if the limit is not violated (refer to the Bases for SL 2.1.1.2). The operating limit MCPR is established to ensure that no fuel damage results during abnormal operational transients.

Although fuel damage does not necessarily occur if a fuel rod actually experienced boiling transition (Ref. 1). the critical power at which boiling transition is calculated to occur has been adopted as a fuel design criterion.

The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs. Based on these experimental data, correlations have been developed to predict critical bundle power (i.e.,

the bundle power level at the onset of transition boiling) for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling). Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur.

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the abnormal operational transients to establish the operatin limit MCPR are presented in References 2, 3, 4, 5, 6, 7, 8 To ensure that the MCPR SL is not exceeded auring any ransient event that occurs with moderate frequency, limiting transients have been analyzed to

% t JiI Cdetermine the largest reduction in critical power ratio (CPR). The types of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest change in CPR (ACPR). When the largest ACPR is added to the MCPR SL, the required operating limit MCPR is obtained.

The MCPR operating limits derived from the transient analysis are dependent on the operating core flow and core exposure to ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency (continued)

JAFNPP B 3.2.2-1 Revision MCPR B 3.2.2 BASES APPLICABLE (Refs. 6, 7, 8. and 9). A generator load reject without SAFETY ANALYSES bypass and a feedwater controller transient normally result (continued) in the worst case MCPR transients for a given fuel cycle.

During operations at low core flows the MCPR operating limit must be increased by a factor of Kf (specified in the COLR) which is derived from the recirculation flow runout transient and is a function of core flow. This will ensure the MCPR safety limit is not exceeded during a recirculation flow runout event.

The MCPR satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii)

(Ref. 10).

LCO The MCPR operating limits specified in the COLR are the result of the Design Basis Accident (DBA) and transient analysis. The operating limit MCPR is a function of exposure, control rod scram times and core flow. The MCPR values for each fuel assembly must remain above the operating limit MCPR.

APPLICABILITY The MCPR operating limits are primarily derived from the analyses of transients that are assumed to occur at high power levels. Below 25% RTP, the reactor is operating at a minimum recirculation pump speed and the moderator void ratio is small. Surveillance of thermal limits below 25% RTP is unnecessary due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs. Statistical analyses indicate that the nominal value of the initial MCPR expected at 25% RTP is

> 3.5. Studies of the variation of limiting transient behavior have been performed over the range of power and flow conditions. These studies encompass the range of actual values for key plant parameters important to typically limiting transients. The results of these studies demonstrate that a margin is expected between performance and the MCPR requirements, and that margins increase as power is reduced to 25% RTP. This trend is expected to continue to the 5' to 15% power range when entry into MODE 2 occurs. When in MODE 2, the intermediate range monitor provides rapid scram initiation for any significant power increase transient, which effectively eliminates any MCPR compliance concern. Therefore, at THERMAL POWER levels

< 25% RTP, the reactor is operating with substantial margin to the MCPR limits and this LCO is not required.

(continued)

JAFNPP B 3.2.2-2 Revision -6

INSERT 2 Flow-dependent MCPR limits, MCFPR(F), are necessary to assure that the Safety Limit MCPR (SLMCPR) is not violated during recirculation flow increase events. The design basis flow increase event is a slow-flow power increase event which is not terminated by scram, but which stabilizes at a new core power corresponding to the maximum possible core flow. Flow runout events were analyzed along a constant xenon flow control line assuming a quasi steady-state plant heat balance. The MCPR(F) limit is specified as a absolute value and is generic and cycle-independent.

The operating limit is dependent on the maximum core flow limiter setting in the Recirculation Flow Control System.

Above the power at which the scram is bypassed (Pbypass), bounding power-dependent trend functions have been developed. This trend function, Kp, is used as a multiplier to the rated MCPR operating limits to obtain the power-dependent MCPR limits, MCPR(P). Below the power at which the scram is automatically bypassed (Below Pbypass), the MCPR(P) limits are actual absolute Operating Limit MCPR (OLMCPR) values, rather than multipliers on the rated power OLMCPR.

MCPR B 3.2.2 BASES SURVEILLANCE SR 3.2.2.2 (continued)

REQUIREMENTS the value of T, which is a measure of the actual scram speed distribution compared with the assumed distribution.

The MCPR operating limit is then determined based on an interpolation between the applicable limits for Option A (scram times of LCO 3.1.4,"Control Rod Scram Times") and Option B (realistic scram times) analyses. The parameter X must be determined once within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after each set of scram time tests required by SR 3.1.4.1, SR 3.1.4.2, and SR 3.1.4.4 because the effective scram speed distribution may change during the cycle or after maintenance that could affect scram times. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is acceptable due to the relatively minor changes in r expected during the fuel cycle.

REFERENCES 1. NUREG-0562, Fuel Rod Failure as a Consequence of Departure From Nucleate Boiling or Dry Out, June 1979.

2. NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel, (Revision specified in the COLR).

3. UFSAR, Chapter 3.

4. UFSAR. Chapter 6.

5. UFSAR, Chapter 14.

6. NEDO-24281, FitzPatrick Nuclear Power Plant Single-Loop Operation, August 1980.

7. NEDO-24243, General Electric Boiling Water Reactor Load Line Limit Analysis For James A. FitzPatrick Nuclear Power Plant, February 1980.

8. NEDC-32016P-1, Power Uprate Safety Analysis For James A. FitzPatrick Nuclear Power Plant, April 1993, including Errata and Addenda Sheet No. 1, dated January 1994.

9. Supplemental Reload Licensing Report for James A. FitzPatrick (Revision specified in the COLR).

10. 10 CFR 50.36(c)(2)(ii).

JAFNPP B 3.2.2-4 Revision INSERT 3

11. NEDC-33087P, J. A. FitzPatrick Nuclear Power Plant APRM/RBM/Technical Specifications/ Maximum Extended Operating Domain (ARTS/MEOD),

Revision 1, September 2005.

APRM Gain and Setpoint B 3.2.4 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.4 Average Power Ra e Monitor (APRM) Gain and etpoint BASES BACKGROUND he OPERABILITY of the APR and their setpoints is a initial condition of all fety analyses that assume od insertion upon reactor ram. Applicable design c teria is discussed in UFSAR, Se tion 16.6 (Ref. 1). This CO is provided to require e APRM gain or APRM Neutr n Flux- High (Flow Biased) Func on Allowable Value (LCO 3 .1.1, "Reactor Protection System (RPS) Instrument ion,"

Function 2.b) t be adjusted when operati under conditions of excessive twer peaking to maintain a eptable margin to the fuel cladding integrity Safety Lin (SL) and the fuel cladding 1 plastic strain limit.

The con tion of excessive power aking is determined by the r io of the actual power pe ing to the limiting power pea ng at RTP. This ratio is/qual to the ratio of the co limiting MFLPD to the Fr ction of RTP (FRTP), wher F TP is the measured THERM POWER divided by the RTP.

xcessive power peaking eists when:

/ / MFLPD FRTP , 1,/

1 indicating that MF D is not decreasing propo ionately to the overall power reduction, or conversely, hat power peaking is incr asing. To maintain margin similar to those at RTP conditj ns, the excessive power paking is compensated y a gain adjustment on th APRMs or adjustment of the AP Neutron Flux- High (Flow ased) Function AllowablVValue. Either of these a ustments has effectyfely the same result as mai taing MFLPD less than or e al to FRTP and thus mainta s RTP margins for APLHGR/

MC , and LHGR.

e normally selected APRM utron Flux -High (Flow Bia d)

Function Allowable Value sitions the scram above th upper bound of the normal powe /flow operating region that as been considered in the esign of the fuel rods. T Allowable Value is f w biased with a slope that approximates the up er flow control line, such at an approximately cons ant margin is maintained b ween the flow biased trip 1ev and the upper operating b ndary for core flows in exces of about 45X of rated core flow. In the range of infp quent operations below 45%Xf rated core flow, the margi o scram is reduced because f the nonlinear core (continued)

JAFNPP B 3.2.4-1 Revision AS

APRM Gain and Setpoint B 3.2.4 BASES / / I E3ACKGROUND flow v rsus drive flow relation ip. The normally selectg (continued) APRM llowable Value is suppor ed by the analyses presejted in eferences 2 and 3 that c centrate on events ini tted fr m rated conditions. De gn experience has shown that nimum deviations occur thin expected margins t operating limits (APLHG , MCPR. and LHGR), at rattd conditions for normal wer distributions. How er, at other than rated con tions, control rod patte ns can be established that si ificantly reduce the mar in to thermal limits. Therefore, the APRM Neutron Flux igh (Flow Biased) Functions llowable Value may be r uced during operation when tIe combination of THERM POWER and MFLPD indicates an scessive power peaking dstribution. In addition. th IAPRH Neutron Flux- Hig (Flow Biased) Function provides pr6tection from reactor the mal hydraulic instability consistent with Boili Water Reactors Owners' Group L g*Term Solution. Option -D (Refs. 4, 5 and 6). /

APPLICABLE The ceptance criteria for he APRM gain or setpoint SAFETY ANALYSES ad ustments are that accep ble margins (to APLHGR, PR, d LHGR) be maintained the fuel cladding integr ty SL and the fuel cladding 1 plastic strain limit.

The safety analyses efs. 2 and 3) concentrat on the rated power condition for which the minimum expect margin to the operating limits PLHGR. MCPR, and LHGR) o curs.

LCO 3.2.1, "AVE GE PLANAR LINEAR HEAT GEERATION RATE (APLHGR)," and CO 3.2.2. "MINIMUM CRITI AL POWER RATIO (MCPR)," and CO 3.2.3, "Linear Heat G eration Rate (LHGR) " iifit the initial margins to/these operating limits at rated 'nditions so that specifi acceptable fuel design limits ace met during transients itiated from rated condit ns. At initial power leels less than rated levels/

the m rgin degradation of eithe the APLHGR. the MCPR. or the HGR during a *transient c be greater than at the r ed co dition event. This greater margin degradation durrnn the ansient is primarily off t by the larger initial m gin o limits at the lower t n rated power levels. How ver, power distributions can e hypothesized that would esult in reduced margins to th pre-transient operating liit. When combined with the in eased severity of certain ransients at other than rate conditions, the SLs could e approached.

At substantially Preduced power levels, highl peaked power distributions c Id be obtained that could educe thermal margins to the in-Imum levels required fo transient events.

To prevent or mitigate such situations, ither the APRM gain is adjusted upward by the ratio of th core limiting MFLPD (continued)

Yr JAFNPP B 3.2.4-2 Revision 6'

APRM Gain and Setpoint B 3.2.4 11 BASES / /

APPLICABLE to the F P, or the APRM Neutron F - High (Flow Biased)

SAFETY ANALYSES Functio Allowable Value is requited to be reduced by t (continued) ratio f FRTP to the core limit g MFLPD. Either of t ese adju ments effectively count s the increased severny of son/ events at other than ratd conditions by propo ionally ireasing the APRM gain or/proportionally lowering the

,ARMNeutron Flux- High (bow Biased) Function A owable

/Value, dependent on the ncreased peaking that ay be encountered.

The reactor thermal ydraulic stability a yses (Ref. 6) indicates that thi APRM Neutron Flux-Hi (Flow Biased)

Function will suppress power oscillatio s prior to exceeding the fuel safet limit (MCPR). This p tection is provided at a high sta istical confidence lev for core wide mode oscillation and at a nominal stat' tical confidence level for region mode oscillations. his protection is adequ e since cor wide oscillation is e dominant mode becaus the plant i designed with relativ ly tight fuel inlet on cing (Ref./ ).//

Th APRM gain and setpoin s satisfy Criteria 2 an 3 of 1 CFR 50.36(c)(2)(ii) ef. 7).

LCO Meeting any one of /e following conditions nsures acceptable operat'sg margins for events dew ibed above:

a. Limiting cess power peaking;

b. Reduci the APRM Neutron Flux - igh (Flow Biased)

Func on Allowable Value by m tiplying the APRM Neu on Flux--High (Flow Bia d) Function Allowable V ue by the ratio of FRTP nd the core limiting val MFLPD: or

c. Increasing APRM gains cause the APRM to read greater than 100% tifs MFLPD. This condition s to account for the red tion in margin to the fu cladding integrity SL and the fuel cladding plastic strain limit.

MFLPD is the ratio f the limiting LHGR to e LHGR limit for the specific Candle type. As power is educed, if the design power di tribution is maintained, FLPD is reduced in proportion to he reduction in power. wever, if power peaking incr ses above the design val e the MFLPD is not reduced in roportion to the reducti in power. Under (continued)

JAFNPP B 3.2.4-3 Revision --

APRM Gain and Setpoint B 3.2.4 BASES / /

LCO these c itions, the APRM gain s adjusted upward or tpi (continued ) APRM utron Flux -High (Flow ased) Function AllowabYe Valu is reduced accordingly. When the reactor is oy4rating wit peaking less than the sign value, it is not Necessary to/modify the APRM Neutron lux -High (Flow Biaseo'5 Function lowable Value. Adjust g APRM gain or modifyiJng the eutron Flux- High (Flo Biased) Function Allowable Value is equivalent to maintai ng MFLPD less than or qual to FRTP, as stated in the LCO For compliance wil LCO Item b (APRM Neu on Flux- High (Flow Biased) Function Allowable Value odification) or Item c (APRM gwn adjustment), only APMs required to be OPERABLE per 'CO 3.3.1.1, Function 2. are required to be modified or djusted. In addition each APRM may be allowed to have it gain or Allowable Valie adjusted or modified independ tly of other APRMs th are having their gain or

/

/ Allowab Value adjusted.

APPLICAIBILITY The FLPD limit, APRM gai adjustment, and APRM Neu on F -High (Flow Biased) unction Allowable Value odification is provid to ensure that the fuel ladding integrity SL and the 6el cladding 1% plastic s ain limit are not violated dur ng design basis transient. As discussed in the B es for LCO 3.2.1 and LCO .2.2, sufficient margi fto these limits exists b ow 25X RTP and, therefore, thesg requirements are only ne ssary when the reactor is op ,ating at 2 25%XRTP.

ACTIO A.1 If the PRM gain or Neutron Flux High (Flow Biased)

Func 'on Allowable Value is not within limits while the MFLRD has exceeded FRTP, the rgin to the fuel claddin i tegrity SL and the fuel cl dding 1X plastic strain mit piay be reduced. Therefore prompt action should be aken to 1/restore the MFLPD Ito withn its required limit or ke

/ acceptable APRM adjustm ts such that the plant operating

/ within the assumed mar in of the safety analyse The 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Completi n Time is normally suffi ent to restore

/ either the MFLPD ) within limits or the AP gain or Neutron Flux- Hi (Flow Biased) Function llowable Value to within limits d is acceptable based o/he low probability of a transien or Design Basis Acciden occurring simultaneou y with the LCO not met.

//(continued)

_ -4,,

JAFNPP B :3.2.4-4 Revision 4-

APRM Gain and Setpoint B 3.2.4 e

BASES / / /

ACTIONS (cont-inued)

If MFLPD, APRM gain, or Ne ron Flux- High (Flow Bia ed)

F nction Allowable Value annot be restored to wit n its equired limits within e associated Completion me, the plant must be brought a MODE or other specifid condition in which the LCO does not apply. To achieve t s status.

THERMAL POWER is re ced to < 25% RTP within hours. The allowed Completion/ime is reasonable, base on operating experience, to re} uce THERMAL POWER to < X RTP in an orderly manner #" without challenging pant systems.

SURVE LANCE SR 3.2.4.1 nd SR 3.2.4.2 REU EMENTS The MFLP is required to be calcu ated and compared to FRTP or APRM gain or Neutron Fl - High (Flow Biased)

Functi'n Allowable Value to endure that the reactor is oper ing within the assumpt ns of the safety analysis.

The e SRs are only required/to determine the MFLPD an a uming MFLPD is greater han FRTP, the appropriate gain or eutron Flux- High (Flow iased) Function Allowabl Value, and is not intended tore a CHANNEL FUNCTIONAL T for the APRM gain or Neutron ux -High (Flow Biased) F ction circuitry. SR 3.2. .1and SR 3.2.4.2 have bee modified by Notes which clarif that the respective SR d s not have to be met if the al trnate requirement demonst ated by the other SR is sat'sfied. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequ cy of SR 3.2.4.1 is chosen to coincide with the determina ion of other thermal limi /, specifically those for the APLHGR (LCO 3.2.1) and LHGR (LCO 3.2.3). Th 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on e recognition of the slow ess of changes in power distribu ion during normal operati . The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance after TFERMAL POWER 2 25% RTP is chieved is acceptable give the large inherent margin o operating limits at lo powef levels.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency of S 3.2.4.2 requires a more equent verification than if MFLP is less than or equal to RTP.

When MFLPD is greater th>4n FRTP. more rapid chang in power distribution are typic ly expected.

REFERENCES/ 1. UFSAR, Sectio 16.6.

2. UFSAR, Se ion 14.5.

3. NEDE-24 11-P-A, General Electric tandard Application for R ctor Fuel, (Revision sp ified in the COLR).

(continued)

JAFNPP B 3.2.4-5 Revision --

APRM Gain and Setpoint B 3.2.4 BASES /

REFERENCES 4. NEDO-31960-A / WR Owners' Group Long erm Stability (contin d) Solutions censing Methodology, J e 1991.

5. NEDO-3 I0-A, Supplement 1. BWR wners' Group Long-Term tability Solutions Lice ing Methodology, Mar 1992.

6. ENE-637-044-0295, Applic ion Of The "Regional Exclusion With Flow-Bia d APRM Neutron Flux Scr Stability Solution (0 ion I-D) To The James A.

FitzPatrick Nuclear ower Plant, February 199

7. 10 CFR 50.36(c)C (ii).

JAFNPP B 3.2.4- 6 Revision 4-

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.1 REQUIREMENTS (continued) Performance of the CHANNEL CHECK once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. For Functions 8 and 9, this SR is associated with the enabling circuit sensing first stage pressure.

Channel agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.

The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.

SR 3.3.1.1.2 To ensure that the APRMs are accurately indicating the true core average power, the APRMs are calibr ted to the oweccculated from a heat balance. CO 3.2.4, "verage Power,Range Mo cr I and S points," ows the APR tobe r ading grea r than ac ual THERMA OWER to c pensate or localiz power pe ing. When his djustme is made. e require nt for the PRMs to indica within 2%,TP of cal lated power is modified o req e the APRMt indica within 2X Pof calc ted MF 0. he Frequency of once per 7 days is based on minor changes in LPRM sensitivity, which could affect the APRM reading between performances of SR 3.3.1.1.8.

A restriction to satisfying this SR when < 25Z RTP is provided that requires the SR to be met only at 2 25% RTP because it is difficult to accurately maintain APRM indication of core THERMAL POWER consistent with a heat (continued)

JAFNPP B 3.3.1.1-26 Revision-0-