RS-08-091, Request for a License Amendment to Revise Local Power Range Monitor Calibration Frequency

Request for a License Amendment to Revise Local Power Range Monitor Calibration Frequency

ML082110187
Person / Time
Site:	LaSalle
Issue date:	07/25/2008
From:	Simpson P Exelon Corp, Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML082110186	List: RS-08-091, Request for a License Amendment to Revise Local Power Range Monitor Calibration Frequency
References
RS-08-091 ANP-2739NP, Rev 1
Download: ML082110187 (39)

Text

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Nuclear Exelon Generation www.exeloncorp .com 4300 Winfield Road Warrenville, IL 60555 10 CFR 50 .90 RS-08-091 July 25, 2008 U . S . Nuclear Regulatory Commission Attn : Document Control Desk Washington, DC 20555-0001 LaSalle County Station, Units 1 and 2 Facility Operating License Nos . NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374

Subject:

Request for a License Amendment to Revise Local Power Range Monitor Calibration Frequency In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests the following amendment to Appendix A, Technical Specifications (TS), of Facility Operating License Nos.

NPF-11 and NPF-18 for LaSalle County Station (LSCS) Units 1 and 2. Specifically, the proposed changes will revise TS 3.3.1 .1, "Reactor Protection System (RPS) Instrumentation,"

Surveillance Requirement (SR) 3.3.1 .1 .8 and TS 3.3.1 .3, "Oscillation Power Range Monitor (OPRM) Instrumentation," SR 3.3.1 .3.2 to increase the frequency interval between Local Power Range Monitor (LPRM) calibrations from 1000 effective full power hours (EFPH) to 2000 EFPH .

Extending the LPRM calibration surveillance interval will increase the LPRM signal uncertainty value used in the LSCS Safety Limit Minimum Critical Power Ratio (SLMCPR) analysis .

However, this increase in the LPRM signal uncertainty value is justified based on a plant specific evaluation that confirms that the change in the core distribution uncertainty caused by the extended surveillance interval is bounded by the value used for the currently approved LSCS SLMCPR power distribution uncertainties .

The attached amendment request is subdivided as shown below.

Attachment 1 provides an evaluation of the proposed change.

Attachment 2 includes the marked-up TS pages with the proposed changes indicated .

Attachment 3 includes the marked-up TS Bases pages with the proposed changes indicated. The TS Bases pages are provided for information only, and do not require NRC approval.

July 25, 2008 U . S. Nuclear Regulatory Commission Page 2 Attachment 4 provides the LPRM calibration uncertainty analysis for the proposed 2000 EFPH calibration interval .

Attachment 5 provides the results of an additional analysis performed to confirm that the analytic and statistical treatments used are also bounding for the TS SR 3.0.2 25 percent grace (i .e., to 2500 EFPH).

EGC requests approval of the proposed change by July 27, 2009, with the amendment being implemented within 60 days of issuance . The requested implementation period will allow sufficient time for effective planning and scheduling of affected activities associated with LPRM calibrations .

The proposed changes have been reviewed by the LSCS Plant Operations Review Committee and approved by the Nuclear Safety Review Board in accordance with the requirements of the EGC Quality Assurance Program.

In accordance with 10 CFR 50 .91, "Notice for public comment ; State consultation," EGC is notifying the State of Illinois of this application for changes to the TS by transmitting a copy of this letter and its attachments to the designated State Official.

Attachments 4 and 5 contain information AREVA considers to be proprietary. AREVA requests that the proprietary information in these Attachments be withheld from public disclosure in accordance with the requirements of paragraph (a)(4) of 10 CFR 2 .390, "Public inspections, exemptions, requests for withholding ." The original signed affidavits supporting this request are provided in Attachment 6 to this letter. Attachments 7 and 8 to this letter provide non-proprietary versions of Attachments 4 and 5, respectively .

There are no regulatory commitments contained in this letter . Should you have any questions concerning this letter, please contact Mr. Timothy A. Byam at (630) 657-2804 .

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 25th day of July 2008.

Patrick R. Simpson Manager - Licensing Exelon Generation Company, LLC

July 25, 2008 U. S. Nuclear Regulatory Commission Page 3 : Evaluation of Proposed Change : Markup of Proposed Technical Specifications Page Changes : Markup of Proposed Technical Specifications Bases Page Changes : AREVA LPRM Calibration Uncertainty Analyses for 2000 EFPH Calibration Interval (Proprietary) : LaSalle LPRM Calibration Extension - Additional Analysis Result (Proprietary) : AREVA Affidavits : AREVA LPRM Calibration Uncertainty Analyses for 2000 EFPH Calibration Interval (Non-Proprietary) : LaSalle LPRM Calibration Extension - Additional Analysis Result (Non-Proprietary)

ATTACHMENT 1 Evaluation of Proposed Change

Subject:

Request for a License Amendment to Revise Local Power Range Monitor Calibration Frequency 1 .0

SUMMARY

DESCRIPTION 2 .0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration 4.4 Conclusions

6.0 REFERENCES

Page 1 of 11

ATTACHMENT 1 Evaluation of Proposed Change 1 .0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests the following amendment to Appendix A, Technical Specifications (TS), of Facility Operating License Nos.

NPF-11 and NPF-18 for LaSalle County Station (LSCS) Units 1 and 2. Specifically, the proposed changes will revise TS 3.3.1 .1, "Reactor Protection System (RPS) Instrumentation,"

Surveillance Requirement (SR) 3.3.1 .1 .8 and TS 3 .3.1 .3, "Oscillation Power Range Monitor (OPRM) Instrumentation," SR 3.3.1 .3.2 to increase the frequency interval between Local Power Range Monitor (LPRM) calibrations from 1000 effective full power hours (EFPH) to 2000 EFPH.

2.0 DETAILED DESCRIPTION The purpose of this proposed change is to revise the TS SRs for periodic calibration of the LPRMs. The current requirement is stipulated by SR 3.3.1 .1 .8 and SR 3.3.1 .3.2. These surveillance requirements currently specify that the LPRMs be calibrated at a frequency of every 1000 EFPH . The proposed change will revise the frequency of surveillance to every 2000 EFPH and will read as follows :

SURVEILLANCE FREQUENCY SR 3 .3 .1 .1 .8 Calibrate the local power range monitors . 2000 effective full power hours SURVEILLANCE FREQUENCY SR 3 .3 .1 .3 .2 Calibrate the local power range monitors . 2000 effective full power hours In addition, in support of this proposed TS change, the associated TS Bases Sections SR 3.3.1 .1 .8 and SR 3.3.1 .3.2 will be revised to reflect the change in the LPRM calibration frequency from 1000 EFPH to 2000 EFPH . The TS Bases changes are provided in Attachment 3 for information only and do not require NRC approval.

Extending the LPRM calibration surveillance interval will increase the LPRM signal uncertainty value used in the LSCS Safety Limit Minimum Critical Power Ratio (SLMCPR) analysis .

However, this increase in the detector uncertainty value is justified based on a plant specific evaluation that confirms that the change in the core distribution uncertainty caused by the extended surveillance interval is bounded by the value used for the currently approved LSCS SLMCPR power distribution uncertainties .

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ATTACHMENT 1 Evaluation of Proposed Change

3.0 TECHNICAL EVALUATION

The LPRM system consists of fission chamber detectors, signal conditioning equipment, and trip functions . The LPRMs also provide output to the Average Power Range Monitors (APRMs), the Rod Block Monitor (RBM), the Oscillation Power Range Monitor (OPRM) system, the core monitoring system, and the LSCS plant process computer.

The LPRM system includes 43 radially distributed LPRM detector strings having detectors located at different axial heights in the core ; each detector string contains four fission chambers (i .e ., 172 detectors) . The chambers are vertically spaced in the LPRM detector assemblies in a way that gives adequate axial coverage given by the horizontal arrangement of the LPRM detector assemblies . Each fission chamber produces a current that is coupled with the LPRM signal-conditioning equipment to provide the desired scale indications .

Each LPRM detector assembly also contains a calibration tube through which the Traversing Incore Probe (TIP) movable gamma detectors are periodically traversed to provide a continuous axial gamma flux profile at each LPRM string location. From these gamma flux profiles, thermal neutron flux profiles are calculated. Appropriate Gain Adjustment Factors (GAFs) are determined for each LPRM detector based on this information. These GAF values are then applied to LPRM signals during the calibration of the 172 fixed LPRM fission detectors. These calibrations compensate for small changes in detector sensitivity resulting from the depletion of the fissile material lining the individual LPRM fission chambers. LPRM calibrations are performed while the reactor is operating at power due to the limited sensitivity of the LPRM detectors. The LPRM calibration is performed by executing the LSCS on-demand computer program (i.e ., OD-1) that is used to collect axial neutron flux data and then LPRM output signal is adjusted if required, to match the TIP signal .

Numerous tests have been performed on the fission chamber assemblies including tests of linearity, lifetime, gamma sensitivity, and cable effects . These tests and operating experience provide confidence in the ability of the LPRM subsystem to monitor neutron flux to the design accuracy throughout the design lifetime .

The current signals from the LPRM detectors are transmitted to the LPRM amplifiers in the control room . The LPRM amplifier signals are indicated on the reactor control panel . When a control rod is selected for movement, the output signals from the amplifiers associated with the nearest LPRM detectors are individually displayed on the reactor control panel meters. The control room operator can obtain readings from all the LPRM amplifiers by selecting the control rods in order.

The trip circuits for the LPRM provide trip signals to activate lights, instrument inoperative signals, and annunciators . These trip circuits are set to trip when power is not available for the LPRM amplifiers .

The LPRM subsystem is designed to provide a sufficient number of LPRM signals to satisfy the safety design basis of the APRM subsystem. To fulfill its power generation design basis, the LPRM supplies :

a. Signals to the APRM that are proportional to the local neutron flux at various locations within the reactor core, Page 3 of 11

ATTACHMENT 1 Evaluation of Proposed Change

b. Signals to the RBM to indicate changes in local relative neutron flux during the movement of control rods,

c. Signals to alarm high or low local neutron flux, and

d. Signals proportional to the local neutron flux to drive indicating meters and auxiliary devices to be used for operator evaluation of power distribution, local heat flux, minimum critical heat flux, and fuel burnup rate .

The ABB Combustion Engineering Option III OPRM system was installed at LSCS from 1999 to 2000 and was fully implemented at LSCS on November 6, 2006. The ABB system utilizes the OPRM detect-and-suppress function to implement the Generic Letter 94-02, "Long-Term Solutions and Upgrade of Interim Operating Recommendations for Thermal Hydraulic Instabilities in Boiling Water Reactors," long-term solution designated as Option 111 . The system monitors LPRM signals for indications of neutron flux oscillations . The OPRM also monitors indicated power and indicated recirculation flow to automatically enable the OPRM trip when in a predefined region of the power-to-flow map . The OPRM initiates a trip whenever it detects an instability condition when in the predefined region of the power-to-flow map.

The OPRM Instrumentation System consists of four OPRM instrumentation trip channels . Each trip channel consists of two OPRM instrumentation modules, either of which can initiate the trip signal for that channel . Each OPRM instrumentation module receives input from 21 or 22 LPRMs . Each OPRM instrumentation module also receives input from the other OPRM instrumentation module in the trip channel, as well as from RPS APRM power and flow signals to automatically enable the trip function of the OPRM instrumentation module.

LSCS uses the POWERPLEX-III core monitoring system . The POWERPLEX-III Core Monitoring Software System (CMSS) is a group of computer programs that together with inputs from the nuclear instrumentation provide continuous online core monitoring . The POWERPLEX-III CMSS calculates heat balances and margins to thermal limits that are monitored to verify compliance with TS .

The APRM, RBM, and OPRM systems are the only nuclear instrumentation systems that utilize the LPRM readings. Above 25% rated thermal power (RTP), the APRM readings are maintained within +/- 2% of calculated thermal power by calibration against heat balance calculations . The verification that the absolute difference between the APRM channels and the calculated power (i.e., the heat balance calculation) is s 2% RTP is required by SR 3.3.1 .1 .2 on a weekly basis .

The purpose of the RBM system is to limit control rod withdrawal if localized neutron flux exceeds a predefined setpoint as identified in the Core Operating Limits Report (COLR) during non-peripheral control rod manipulations. It is assumed to function to block further control rod withdrawal to preclude a SLMCPR violation . The OPRM system is used to detect and suppress power oscillations that could be induced from operating the reactor within regions of instability.

Since the LPRM fission chamber responses are linear over the proposed surveillance interval, each of the APRM, RBM and OPRM system responses will not be impacted by the calibration interval extension .

As described above, LPRM gain settings are determined from the local flux profiles measured by the TIP system. This establishes the relative local flux profile for appropriate representative Page 4 of 11

ATTACHMENT 1 Evaluation of Proposed Change input to the APRM system. The current TS SR 3.3.1 .1 .8 and SR 3.3.1 .3.2 frequency interval of 1000 EFPH between required LPRM calibrations is based on operating experience with previous core monitoring systems at LSCS and older design LPRM detectors. LSCS currently uses an improved POWERPLEX-III core monitoring system and newer design LPRM chambers that exhibit more consistent sensitivity behavior than the older LPRM detectors .

Extending the LPRM calibration surveillance interval to the proposed 2000 EFPH will increase the LPRM signal uncertainty value used in the LSCS SLMCPR analysis . However, this increase in the LPRM signal uncertainty value is justified based on a plant specific evaluation that confirms that the change in the core distribution uncertainty caused by the extended surveillance interval is maintained within the uncertainties currently used in the LSCS SLMCPR analysis . Increasing the frequency interval between required LPRM calibrations is acceptable due to improvements in the fuel analytical bases and therefore, the revised surveillance interval continues to ensure that the LPRM detector signal is adequately calibrated .

The uncertainty in the measured power distribution is consistent with the uncertainties used to calculate the cycle specific SLMCPR . The current LSCS SLMCPR was implemented by Amendment No. 137 for Unit 1 (Reference 1) and Amendment No . 126 for Unit 2 (Reference 2).

The license amendments were obtained to support the fuel transition to ATRIUM-9B fuel and Siemens (currently AREVA) methodology and to conservatively bound a 5% power uprate . The SLMCPR is confirmed to be bounding for each reload design .

The current LSCS reload core designs (i.e., Unit 1 Cycle 13 and Unit 2 Cycle 12) are licensed using the NRC approved AREVA design methodologies as identified in TS 5 .6.5, "Core Operating Limits Report (COLR) ." As required by TS 5.6.5, subsequent planned LSCS Unit 1 and Unit 2 reloads will also use NRC approved design methodologies. The uncertainty used in the current LSCS SLMCPR analysis for two-loop operation and for single loop operation is based on NRC approved methods and supports an LPRM detector calibration interval of 2500 EFPH. These uncertainties are consistent with those identified in TS 5.6.5 which supports operation with up to two TIP machines out of service, up to 50% of the LPRMs out of service, and an LPRM calibration interval of 2500 EFPH.

The LPRM signal uncertainty normally used for a 1000 EFPH calibration interval is defined in Reference 3 . An LSCS plant specific LPRM uncertainty analysis for a 2000 EFPH LPRM calibration frequency has been performed and the results are documented in Attachment 4 to this letter . The analysis confirmed that the LPRM response uncertainty used in the current SLMCPR analysis remains bounding for LSCS if the LPRM calibration was extended from 1000 EFPH to 2000 EFPH . LSCS plant exposure data was collected from 1996 through 2006 . The data points used in the analysis were LPRM calibration current readings, which are proportional to the neutron flux level . Predicted calibration currents were compared with measured calibration currents using an effective exposure decay factor for the LPRM detector . The analysis determined that the increase in LPRM response uncertainty (i.e., standard deviation) resulting from the extended calibration interval was not significant. The maximum standard deviation increase is provided for Unit 1 and Unit 2 in Table A from Attachment 4. Both of these values are less than the value which is used for the increase in LPRM signal uncertainty from 1000 EFPH to 2500 EFPH calibration interval in the current SLMCPR analysis .

When the LPRM signal uncertainty increase from Table A of Attachment 4 is added to the LPRM response uncertainty for a calibration interval of 1000 EFPH, the final uncertainty value remains bounded by the LPRM response uncertainty used in the LSCS SLMCPR analysis . The Page 5 of 11

ATTACHMENT 1 Evaluation of Proposed Change result is very conservative since it is based on an arithmetic increase in the LPRM response uncertainty, because the impact of the calibration interval is independent of other components of the uncertainty (e.g., cable, amplifier) . A much smaller total LPRM response uncertainty would result if the individual components were statistically combined . Thus, the evaluation concluded that the actual LSCS LPRM response performance is as expected and that an increase in the calibration interval from 1000 EFPH to 2000 EFPH is bounded by the uncertainties currently applied to the LSCS licensing basis SLMCPR analysis .

An additional analysis was performed to evaluate the LSCS plant specific increase in LPRM response uncertainty when accounting for the TS SR 3 .0 .2 allowed 25% extension of the calibration interval (i.e., 2500 EFPH). The results of this analysis are provided in Attachment 5 .

To ensure conservative results from this analysis the exposure intervals were divided differently than those used in the analysis documented in Attachment 4. Therefore, the results differ slightly from those values included in Table A of Attachment 4 . This does not invalidate the results discussed above. The difference in standard deviations between the 1000 megawatt-days/metric ton (MWd/MTU) and upper bound calibration interval of 2500 MWd/MTU for Unit 1 and Unit 2 are provided in the Table provided in Attachment 5. Both of these values are less than the value which is used for the increase in LPRM signal uncertainty from 1000 EFPH to 2500 EFPH calibration interval in the current SLMCPR analysis .

The exposures used to generate the additional standard deviation results as provided in , are in MWd/MTU . The difference between exposure units of EFPH and MWd/MTU is small . The conversion value for LaSalle is approximately 1 .00 EFPH = 1 .07 MWd/MTU with small cycle-to-cycle deviation based on the amount of uranium in the core. The calibration intervals use data that is t 500 MWd/MTU around the specified intervals of 1000, 2000, and 2500 MWd/MTU. As such, the standard deviation for the 2000 MWd/MTU calibration interval contains all data between 1500 MWd/MTU and 2500 MWd/MTU and the standard deviation for the 2500 MWd/MTU calibration interval contains all data between 2000 MWd/MTU and 3000 MWd/MTU . The standard deviation calculated for the 2500 MWd/MTU calibration interval bounds all cases.

When the LPRM signal uncertainty increase identified in Attachment 5 is added to the LPRM response uncertainty for a calibration interval of 1000 EFPH, the final uncertainty value remains bounded by the LPRM response uncertainty used in the LSCS SLMCPR analysis .

The LPRM signal uncertainties associated with the extended LPRM calibration frequency were previously incorporated into the SLMCPR analysis for LSCS Units 1 and 2. Therefore, the calibration interval extension will not affect any safety analysis methods, core thermal limits documented in the COLR, or the current safety analysis results as documented in the LSCS Updated Final Safety Analysis Report (UFSAR).

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50.36(c)(3), "Surveillance requirements," states that SRs are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation (LCO) will be met.

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ATTACHMENT 1 Evaluation of Proposed Change The proposed change is a result of increasing the surveillance interval of the LPRM calibration frequency from 1000 EFPH to 2000 EFPH. Increasing the frequency interval between required LPRM calibrations is acceptable due to improvements in the fuel analytical bases. Therefore, the revised surveillance interval continues to ensure that the LPRM detector signal is adequately calibrated . Extending the LPRM calibration surveillance interval will increase the LPRM signal uncertainty value used in the LSCS SLMCPR analysis, however, this increase in the LPRM signal uncertainty value is acceptable since the increase is bounded by the values used by the AREVA analysis . This calibration continues to provide assurance that the LPRM accuracy remains within the total nodal power uncertainty assumed in the thermal analysis basis; and, therefore, the LCO will continue to be met.

4.2 Precedent The NRC has previously approved similar amendments for James A. Fitzpatrick Nuclear Power Plant (Reference 4), Vermont Yankee Nuclear Power Station (Reference 5), Grand Gulf Nuclear Station (Reference 6), River Bend Station (Reference 7), and Peach Bottom Atomic Power Station (Reference 8) . The Grand Gulf submittal included an AREVA plant specific statistical evaluation which confirmed that the uncertainty associated with the extended LPRM calibration interval is valid similar to that provided for LSCS.

4 .3 No Significant Hazards Consideration In accordance with 10 CFR 50 .90, "Application for amendment of license or construction permit," Exelon Generation Company, LLC (EGC) requests the following amendment to Appendix A, Technical Specifications (TS), of Facility Operating License Nos . NPF-11 and NPF-18 for LaSalle County Station (LSCS) Units 1 and 2. Specifically, the proposed changes will revise TS 3 .3.1 .1, "Reactor Protection System (RPS) Instrumentation," Surveillance Requirement (SR) 3.3.1 .1 .8 and TS 3.3 .1 .3, "Oscillation Power Range Monitor (OPRM)

Instrumentation," SR 3.3.1 .3.2 to increase the frequency interval between Local Power Range Monitor (LPRM) calibrations from 1000 effective full power hours (EFPH) to 2000 EFPH .

Extending the LPRM calibration surveillance interval will increase the LPRM signal uncertainty value used in the LSCS Safety Limit Minimum Critical Power Ratio (SLMCPR) analysis .

However, this increase in the LPRM signal uncertainty value is justified based on a plant specific evaluation that confirms that the change in the core distribution uncertainty caused by the extended surveillance interval is maintained within the uncertainties currently used in the LSCS SLMCPR analysis .

According to 10 CFR 50.92, "Issuance of amendment," paragraph (c), 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 .

Page 7 of 11

ATTACHMENT 1 Evaluation of Proposed Change EGC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three criteria set forth in 10 CFR 50.92 as described below.

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response : No The proposed change is a result of increasing the surveillance interval of the LPRM calibration frequency from 1000 EFPH to 2000 EFPH . Increasing the frequency interval between required LPRM calibrations is acceptable due to improvements in the fuel analytical bases and therefore, the revised surveillance interval continues to ensure that the LPRM detector signal is adequately calibrated . Extending the LPRM calibration surveillance interval will increase the LPRM signal uncertainty value used in the LSCS SLMCPR analysis, however, this increase in the LPRM signal uncertainty value is acceptable since the increase is bounded by the values used by the AREVA analysis .

This change will not alter the operation of process variables, structures, systems, or components as described in the LSCS Updated Final Safety Analysis Report (UFSAR) .

The proposed change does not alter the initiation conditions or operational parameters for the system and there is no new equipment introduced by the extension of the LPRM calibration frequency interval . The performance of the Average Power Range Monitor (APRM), Rod Block Monitor (RBM) and Oscillation Power Range Monitor (OPRM) systems are not significantly affected by the proposed surveillance interval increase .

The proposed LPRM calibration interval extension will have no significant effect on the Reactor Protection System (RPS) instrumentation accuracy during power maneuvers or transients and will, therefore, not significantly affect the performance of the RPS. As such, the probability of occurrence for a previously evaluated accident is not increased.

The radiological consequences of an accident can be affected by the thermal limits existing at the time of the postulated accident, however, increasing the surveillance interval frequency will not increase the calculated thermal limits since all uncertainties associated with the increased interval are currently implemented and are currently used to calculate the existing Safety Limits . Plant specific evaluation of LPRM sensitivity to exposure has determined that the extended calibration frequency increases the LPRM signal uncertainty value used in the LSCS SLMCPR analysis, however, the increase is bounded by the values currently used in the safety analysis . Therefore, the thermal limit calculation is not significantly affected by LPRM calibration frequency, and thus the radiological consequences of any accident previously evaluated are not increased .

Based on the above information, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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ATTACHMENT 1 Evaluation of Proposed Change

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The performance of the APRM, RBM and OPRM systems are not significantly affected by the proposed LPRM surveillance interval increase. The proposed change does not affect the control parameters governing unit operation or the response of plant equipment to transient conditions. For the proposed LPRM extended calibration interval frequency all uncertainties remain less than the uncertainties assumed in the existing thermal limit calculations . The proposed change does not change or introduce any new equipment, modes of system operation or failure mechanisms .

Based on the above information, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No The proposed change has no impact on equipment design or fundamental operation, and there are no changes being made to safety limits or safety system allowable values that would adversely affect plant safety as a result of the proposed LPRM surveillance interval increase . The performance of the APRM, RBM and OPRM systems are not significantly affected by the proposed change. The margin of safety can be affected by the thermal limits existing at the time of the postulated accident; however, uncertainties associated with LPRM chamber exposure have no significant effect on the calculated thermal limits . Plant specific evaluation of LPRM sensitivity to exposure has determined that the extended calibration frequency increases the LPRM signal uncertainty value used in the LSCS SLMCPR analysis, however, the increase is bounded by the values currently used in the safety analysis . The thermal limit calculation is not significantly affected since LPRM sensitivity with exposure is well defined . LPRM accuracy remains within the total nodal power uncertainty assumed in the thermal analysis basis, therefore maintaining thermal limits and the safety margin. The proposed change does not affect safety analysis assumptions or initial conditions and the margin of safety in the original safety analyses are therefore maintained .

Based on this information, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above evaluation, EGC concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c) .

4.4 Conclusions In conclusion, based on the considerations above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the Page 9 of 11

ATTACHMENT 1 Evaluation of Proposed Change issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public .

5.0 ENVIRONMENTAL CONSIDERATION

EGC has evaluated this proposed operating license amendment consistent with the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51 .21, "Criteria for and identification of licensing and regulatory actions requiring environmental assessments." EGC has determined that these proposed changes meet the criteria for a categorical exclusion set forth in paragraph (c)(9) of 10 CFR 51 .22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review," and as such, has determined that no irreversible consequences exist in accordance with paragraph (b) of 10 CFR 50 .92, "Issuance of amendment." This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10 CFR 50, "Domestic Licensing of Production and Utilization Facilities," which changes a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, "Standards for Protection Against Radiation," or which changes an inspection or a surveillance requirement, and the amendment meets the following specific criteria:

The amendment involves no significant hazards consideration.

As demonstrated in Section 4.3, "No Significant Hazards Consideration," the proposed changes do not involve any significant hazards consideration.

(ii) There is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite.

The proposed change to increase the surveillance interval of the LPRM calibration frequency from 1000 effective full power hours to 2000 effective full power hours does not result in an increase in power level, does not increase the production nor alter the flow path or method of disposal of radioactive waste or byproducts ; thus, there will be no change in the amounts of radiological effluents released offsite.

Based on the above evaluation, the proposed change will not result in a significant change in the types or significant increase in the amounts of any effluent released offsite .

(iii) There is no significant increase in individual or cumulative occupational radiation exposure.

The proposed change to increase the surveillance interval of the LPRM calibration frequency from 1000 effective full power hours to 2000 effective full power hours will not result in any changes to the previously analyzed configuration of the facility. There will be no change in the level of controls or methodology used for the processing of radioactive effluents or handling of solid radioactive waste, nor will the proposal result in any change in the normal radiation levels in the plant; therefore, there will be no increase in individual or cumulative occupational radiation exposure resulting from this change.

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ATTACHMENT 1 Evaluation of Proposed Change

6.0 REFERENCES

1 . Letter from U. S. NRC to O. D. Kingsley, (Commonwealth Edison Company),

"Issuance of Amendment," dated November 9, 1999

2. Letter from U. S . NRC to O. D. Kingsley, (Commonwealth Edison Company),

"LaSalle - Issuance of Amendment (TAC No. MA8306)," dated May 17, 2000

3. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors : Evaluation and Validation of CASMO-4/MICROBURN-B2"

4. Letter from U .S. NRC to M. Kansler, (Entergy Nuclear Operations, Inc), "James A.

Fitzpatrick Nuclear Power Plant - Amendment Re : Regarding Local Power Range Monitor Calibration Frequency," dated May 1, 2003 5 . Letter from U . S. NRC to S. L. Newton, (Vermont Yankee Nuclear Power Corporation), "Vermont Yankee Nuclear Power Station - Issuance of Amendment Re : Local Power Range Monitor Calibration Frequency," dated July 18, 2000 6 . Letter from U. S. NRC to W. R. Brian (Entergy Operations, Inc.), "Grand Gulf Nuclear Station, Unit 1 - Issuance of Amendment Re: Changes to Technical Specifications Surveillance Requirement 3.3.1 .1 .7, the Local Power Range Monitor Calibration Frequency (TAC No. MD3469)," dated October 24, 2007

7. Letter from U. S . NRC to R. K. Eddington, (River Bend Station), "River Bend Station, Unit 1 - Issuance of Amendment Re: Changes to Local Power Range Monitor Calibration Frequency," dated June 11, 1999

8. Letter from U . S. NRC to Mr. C. G . Pardee (Exelon Generation Company, LLC),

"Peach Bottom Atomic Power Station, Units 2 and 3 - Issuance of Amendments to Extend Local Power Range Monitor Calibration Interval (TAC Nos. MD3717 and MD3718)," dated February 29, 2008 Page 11 of 11

ATTACHMENT 2 LaSalle County Station Facility Operating License Nos. NPF-11 and NPF-18 Markup of Proposed Technical Specifications Page Changes REVISED TS PAGE 3.3.1 .1-4 3.3.1 .3-3

RPS instrumentation 3 .3 .1 .1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3 .3 .1 .1 .5 Perform CHANNEL FUNCTIONAL TEST . 7 days SR 3 .3 .1 .1 .6 Verify the source range monitor (SRM) and Prior to fully intermediate range monitor (IRM) channels withdrawing overlap . SRMs SR 3 .3 .1 .1 .7 ------------------ NOTE __----__-----__-___

Only required to be met during entry into MODE 2 from MODE l .

Verify the IRM and APRM channels overlap . 7 days a0o p SR 3 .3 .1 .1 .8 Calibrate the local power range monitors . ffective ruii power hours SR 3 .3 .1 .1 .9 Perform CHANNEL FUNCTIONAL TEST . 92 days SR 3 .3 . 1 .1 .10 Perform CHANNEL CALIBRATION . 92 days (continued)

LaSalle 1 and 2 3 .3 .1 .1-4 Amendment No . ~ 'z -

OPRM Instrumentation 3 .3 .1 .3 SURVEILLANCE REQUIREMENTS

---

_-_------------------__NOTE-_-----------------------------------

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 OPRM maintains trip capability .

SURVEILLANCE FREQUENCY SR 3 .3 .1 .3 .1 Perform CHANNEL FUNCTIONAL TEST . 184 days

~ooD° SR 3 .3 .1 .3 .2 Calibrate the local power range monitors . effective full power hours SR 3 .3 .1 .3 .3 ------------------- NOTE -_----------_-------

Neutron detectors are excluded .

Perform CHANNEL CALIBRATION . The setpoints 24 months for the trip function shall be as specified in the COLR .

SR 3 .3 .1 .3 .4 Perform LOGIC SYSTEM FUNCTIONAL TEST . 24 months SR 3 .3 .1 .3 .5 Verify OPRM is not bypassed when THERMAL 24 months POWER is >_ 28 .6% RTP and recirculation drive flow is < 60% of rated recirculation drive flow .

SR 3 .3 .1 .3 .6 ------------------- NOTE -_------------------

Neutron detectors are excluded .

Verify the RPS RESPONSE TIME is within 24 months on a limits . STAGGERED TEST BASIS LaSalle 1 and 2 3 .3 .1 .3-3 Amendment No . -r,- ,464

ATTACHMENT 3 LaSalle County Station Facility Operating License Nos. NPF-11 and NPF-18 Markup of Proposed Technical Specifications Bases Page Changes REVISED TS BASES PAGES B 3.3.1 .1-30 to B 3.3.1 .1-31 B 3.3.1 .3-7

RPS Instrumentation B 3 .3 .1 .1 BASES SURVEILLANCE SR 3 .3 .1 .1 .6 and SR 3 .3 .1 .1 .7 (continued)

REQUIREMENTS channel(s) declared inoperable . Only those appropriate channel(s) that are required in the current MODE or condition should be declared inoperable .

A Frequency of 7 days is reasonable based on engineering judgment and the reliability of the IRMs and APRMs .

SR 3 .3 .1 .1 .8 LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP)

System . This establishes the relative local flux profile for appropriate representative input to the APRM System .

The4448-effective full power hours (EFPH) Frequency is based on operating experience with LPRM sensitivity changes .

SR 3 .3 .1 .1 .8 also ensures the operability of the OPRM system (specification 3 .3 .1 .3) .

SR--a- 31-1 . 1 .9 and-S.R-a . 3 I .L12 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function . A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay . This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay . This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at lease once per refueling interval with applicable extensions . Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology .

The 92 day Frequency of SR 3 .3 .1 .1 .9 is based on the reliability analysis of Reference 10 .

The 24 month Frequency of SR 3 .3 .1 .1 .12 is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned (continued)

LaSalle 1 and 2 B 3 .3 .1 .1-30 Revision 23

RPS Instrumentation B 3 .3 . 1 .1 3ASES SURVEILLANCE SR 3 .3 .1 .1 .9 and SR 3 .3 .1 .1 .12 (continued)

REQUIREMENTS transient if the Surveillance were performed with the reactor at power . Operating experience has shown that these components usually pass the Surveillance when performed at the 24 month Frequency .

SR 3 .3 .1 .1 .10 . SR 3 .3 .1 .1 .11 . and SR 3 .3 .1 . 11`3 A CHANNEL CALIBRATION is a complete check of the instrument loop, including associated trip unit, and the sensor . This test verifies the channel responds to the measured parameter within the necessary range and accuracy . CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology .

Note 1 of SR 3 .3 .1 .1 .11 and SR 3 .3 .1 .1 .13 states that neutron detectors are excluded from CHANNEL CALIBRATION because of the difficulty of simulating a meaningful signal .

Changes in neutron detector sensitivity are compensated for by performing the 7 day calorimetric calibration (SR 3 .3 .1 .1 .2) and the 888-EFPH LPRM calibration against the TIPS (SR 3 .3 .1 .1 .8) . A second Note to SR 3 .3 .1 .1 .11 and SR 3 .3 .1 .1 .13 is provided that requires the APRM and IRM SRs to be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of entering MODE 2 from MODE 1 . Testing of the MODE 2 APRM and IRM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads, or movable links . This Note allows entry into MODE 2 from MODE 1 if the associated Frequency is not met per SR 3 .0 .2 .

Twenty-four hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR . The Frequencies of SR 3 .3 .1 .1 .10 and SR 3 .3 .1 .1 .11 are based upon the assumption of a 92 day and 184 day calibration interval, respectively, in the determination of the magnitude of equipment drift in the setpoint analysis . The Frequency of SR 3 .3 .1 .1 .13 is based on the assumption of a 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis .

(continued)

LaSalle 1 and 2 B 3 .3 .1 .1-31 Revision 13

OPRM Instrumentation B 3 .3 .1 .3 BASES SURVEILLANCE SR 3 .3 .1 .3 .1 (continued)

REQUIREMENTS other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions .

A Frequency of 184 days provides an acceptable level of system average unavailability over the Frequency interval and is based on the reliability analysis (Ref . 6) .

SR 3 .3 .1 .3 .2 LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP)

System . This establishes the relative local flux profile for appropriate representative input to the OPRM System .

The988-effective full power hours (EFPH) Frequency is based on operating experience with LPRM sensitivity changes .

SR 3 .3 .1 .3 .3 The CHANNEL CALIBRATION is a complete check of the instrument loop . This test verifies the channel responds to the measured parameter within the necessary range and accuracy . CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the plant specific setpoint methodology .

Calibration of the channel provides a check of the internal reference voltage and the internal processor clock frequency . It also compares the desired trip setpoint with those in the processor memory . Since the OPRM is a digital system, the internal reference voltage and processor clock frequency are, in turn, used to automatically calibrate the internal analog to digital converters . The nominal setpoints for the period based detection algorithm are specified in the COLR . As noted, neutron detectors are excluded from CHANNEL CALIBRATION because of difficulty of simulating a meaningful signal . Changes in neutron detector sensitivity are compensated for by performing the 7'~ effective full power hour (EFPH) calibration against the TIPS (SR 3 .3 .1 .3 .2) . SR 3 .3 .1 .3 .2 thus also ensures the operability of the OPRM instrumentation .

(continued)

LaSalle 1 and 2 B 3 .3 .1 .3-7 Revision 23

ATTACHMENT 6 LaSalle County Station Facility Operating License Nos. NPF-11 and NPF-18 AREVA Affidavits

AFFIDAVIT COMMONWEALTH OF VIRGINIA )

ss.

CITY OF LYNCHBURG )

1. My name is Gayle F. Elliott. I am Manager, Product Licensing, for AREVA NP Inc. and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary. I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria .

3. I am familiar with the AREVA NP information contained in the attachment to the letter from AREVA NP Inc. (D.B. McBumey) to Exelon Corporation (Frank Trikur) entitled "Transmittal of LPRM Calibration Uncertainty Analysis for 2000 EFPH Calibration Interval,"

FAB07-2078, dated February 9, 2007, and the letter from AREVA NP Inc. (D.B . McBumey) to Exelon Corporation (Carlos de la Hoz) entitled "Transmittal of LaSalle LPRM Calibration Extension - Additional Analysis Result," FAB08-2327, dated May 30, 2008, and referred to herein as "Documents ." Information contained in these Documents has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.

4. These Documents contain information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in these Documents as proprietary and confidential .

5. These Documents have been made available to the U .S. Nuclear Regulatory Commission in confidence with the request that the information contained in these Documents be withheld from public disclosure . The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary:

(a) The information reveals details of AREVA NP's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.

(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.

The information in these Documents is considered proprietary for the reasons set forth in paragraphs 6(b) and 6(c) above.

7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in these Documents have been made

available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information .

8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

SUBSCRIBED before me this day of 2008.

Sherry L. McFaden NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 10/31/10 Reg . # 7079129 SHERRY L . MCFAbEN Notary Public Commonwealth of Vlrglnla 7079129 MY Commission Expires Oct 31, 2010

AFFIDAVIT STATE OF VIRGINIA )

ss .

CITY OF LYNCHBURG )

1. My name is George L. Pannell. I am Manager, Product Licensing, for AREVA NP Inc. and as such 1 am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary. I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria .

3. I am familiar with the AREVA NP information contained in ANP-2739P Revision 1, entitled, "LaSalle LPRM Calibration Extension -Additional Analysis Result,"

dated July 2008 and referred to herein as the "Document."

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential .

5. This Document has been made available to the U S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure . The request for withholding of proprietary information is made in accordance with 10 CFR 2 .390 . The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information" .

6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary :

(a) The information reveals details of AREVA NP's research and development plans and programs or their results .

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.

(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.

The information in this Document is considered proprietary for the reasons set forth in paragraphs 6(b), 6(c) and 6(d) above .

7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief .

SUBSCRIBED before me this Z 3jmj-~A.ec--

Sherry L. McFaden NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES : 10/31110 Reg. # 7079129

- a, 11 mulumm DNMWL NO" Pubtle Commonweollh at Vlrginio 7079129 My Commiialon Expire: Oct 31, 2010

ATTACHMENT 7 LaSalle County Station Facility Operating License Nos. NPF-11 and NPF-18 AREVA LPRM Calibration Uncertainty Analyses for 2000 EFPH Calibration Interval (Non-Proprietary)

AREVA NP Inc , an AREVA and Shwwns cant TO: C . M. Powers 20553A-11 (3/21/20M)

FROM: K . Wei -7/Z/tt DISTRIBUTION: R. G. Grummer

," . A. N. Dado SUBJ.: LPRM Calibration Uncertainty Analysis for 2000 FILE. KW :07:002 EFPH calibration interval Non Proprietary REP . 1 : E-5078-010-1,LPRM Calibration Uncertainty DATE: July 1, 2008 Analysis for 2000 EFPH Calibration Interval for Lasalle, February,2007

1. Introduction The LaSalle specific LPRM uncertainty analysis for extended calibration interval of 2000 EFPH (including an additional 25% for a total 2500 EFPH) was performed by AREVA based on LPRM data from LaSalle supplied by Exelon . The purpose of the analysis was to identify the increase in LPRM signal uncertainties with the LPRM calibration interval being extended from 1000 EFPH to 2000 EFPH. This is being done to ensure that the extended LPRM calibration interval uncertainties are still bounded by the value being used for the currently approved SLMCPR power distribution uncertainties.

2. Summary This analysis scope included LaSalle LPRM data from 15.96 through 2405 that was collected by LaSalle for both units. Since LaSalle data was used exclusively, this analysis is applicable only to LaSalle Nuclear Power Station .

Table A provides the summary of the relative standard deviations for actual LPRM calibration currents compared to predicted values . The valises increased by [ J for Unit 1 and [ I for Unit 2 which are both less than the value of [ J which is used for the increase in LPRM

AREVA NP Inc .

C_ M. Powers KWD7 002 July 1, 2008 Page 2 signal uncertainty from 1000 EFPH to 2500 EFPH calibration interval in the current SLMCPR analysis .

The [ I value for the LPRM signal uncertainty that is used in the LaSalle SLMCPR analysis is applicable to operation with LPRM calibration intervals up to 2500 EFPH.

Table A Summary of the Relative Standard Deviation

• 1000 EFPH corresponds to approximately 1070 MWDIMT for LaSalle Approved :

R . G.'Grummer ; Manager BWR Methods & Codes

ATTACHMENT 8 LaSalle County Station Facility Operating License Nos. NPF-11 and NPF-18 LaSalle LPRM Calibration Extension - Additional Analysis Result (Non-Proprietary)

ANP-2739NP Revision 1 LaSalle LPRM Calibration Extension -Additional Analysis Result July 2008 AR EVA Page 1 of 7

AREVA NP Inc .

ANP-2739NP Revision 1 LaSalle LPRM Calibration Extension - Additional Analysis Result Copyright © 2008 AREVA NP Inc.

All Rights Reserved

LaSalle LPRM Calibration ANP-2739NP Extension - Additional Analysis Result Revision 1 Nature of Changes Item Page Description and Justification

1. 5 Added reference .

AREVA NP Inc ., an AREVA and Siemens company Page 3

LaSalle LPRM Calibration ANP-2739NP Extension_- Additional Analysis Result Revision 1 Contents 1 .0 Introduction . . . . .. . . . . . . . . .. .. .. . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .. .. . . . . . . . . ...... . . . . . . . . . . . . . . . . . . . . . . . . . ... . . . . . . . . . . . .5 2.0 Reference . .. . . . . . . . . . . .. .. . . . . . . . . . . . . . . . . . . . .. . .. . . . . . . . . . . . . . .. .. .. . . . . . . . . . ... .. . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . .. . . .5 Appendix A LaSalle LPRM Calibration Extension - Additional Analysis Result . . . . . . . . . . .. .. .. .. . . . . . 6 A.1 Summary .. . . . . . . . . . . .. .. . . . . . . . . . . . . . . . . . .. ... . . . . . . . . . . . . . . . . . ... . . . . . . . . . .. .... . . . . . . . . . . . .. . . ... .. . . . . . . . . . . . . . . . . . . . .. . . .. . . . . .6 A.2 Result . .. .. . . . . . . . . . . .. .. . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . .. .. .. . . . . . . . . . . .... . . . . . . . . . . . . . . . ... .. . . . . . . . . . . . . . . . . . . . . . .. .... . . . . . . . 6 This document contains a total of 7 pages AREVA NP Inc ., an AREVA and Siemens company Page 4

LaSalle LPRM Calibration ANP-2739NP Extension - Additional Analysis Result Revision 1 1 .0 Introduction Appendix A provides additional information for LaSalle LPRM calibration uncertainty analysis requested by Exelon .

2.0 Reference

1. E-5079-010-1, "LPRM Calibration Uncertainty Analysis for 2000 EFPH Calibration Interval for LASALLE," February 2007.

AREVA NP Inc., an AREVA and Siemens company Page 5

LaSalle LPRM Calibration ANP-2739NP Extension -Additional Analys is Result Revision 1 Appendix A LaSalle LPRM Calibration Extension - Additional Analysis Result A.1 Summary Additional Analysis was performed by AREVA for LaSalle LPRM calibration extension. LPRM uncertainty analysis results for calibration interval of 1000, 2000, and 2500 MWD/T were provided in this document . The analysis showed the LPRM instrumentation was performing as expected, and the uncertainty increase caused by the extended calibration interval from 1000 to 2500 MWD/T is bounded by the value used for current reload analysis and the currently approved SLMCPR power distribution uncertainties.

A.2 Result The additional analysis results are provided below. Both units use fixed detector sensitivity value which is for C-lattice core and GN-NA-300 LPRM type . The LPRM uncertainty calculations show that the increased uncertainty for the extended calibration interval is within the uncertainty used in the safety limit analysis .

These results may differ from the results in the original calculation notebook (reference 1) since the exposure intervals have been divided differently in order to address the plant request for additional information . The exposure intervals for this analysis were defined as the collection of data t 500 MWd/MTU around the specified interval. For example, the standard deviation for 2000 MWd/MTU is determined by evaluating all the data between calibration intervals of 1500 MWd/MTU and 2500 MWd/MTU .

AREVA NP Inc., an AREVA and Siemens company Page 6

LaSalle LPRIVI Calibration ANP-2739NP Extension -Additional Analysis Result _ Revision 1 The difference in standard deviations between the 1000 MWd/MTU and the upper bound calibration interval of 2500 MWd/MTU are [ ] for units 1 and 2. Since the value of 3.4% has been used for the current calibration interval, the appropriate value for the extended calibration interval would be [ ] for units 1 and 2 . The safety limit analysis used an LPRM uncertainty value of [ ], which is conservative .

AREVA NP Inc ., an AREVA and Siemens company Page 7