ML072970339

License Amendment Request, Revise Local Power Range Monitor Calibration Frequency

ML072970339
Person / Time
Site:	Limerick
Issue date:	10/19/2007
From:	Cowan P Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML072970339 (17)

Text

E Y e 1 on N LL c 1 e a r ;do Exelon Way Kerinett Square. PA 19348 vww exeloncorp corn 1 OCFR50.90 October 19,2007 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Limerick Generating Station, Units 7 and 2 Facility Operating License Nos. NPF-39 and NPF-85 Docket Nos. 50-352 and 50-353

Subject:

License Amendment Request Revise Local Power Range Monitor Calibration Frequency Pursuant to 10 CFR 50.90 Exelon Generation Company, LLC, (Exelon) hereby requests the following amendment to the Technical Specifications (TS), Appendix A, of Facility Operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station (LGS), Units 1 and 2. The proposed amendment would increase the interval between Local Power Range Monitor (LPRM) calibrations from 1000 effective full power hours (EFPH) to 2000 EFPH. Increasing the frequency interval between required LPRM calibrations is acceptable due to improvements in core monitoring processes and nuclear instrumentation.

In addition, this proposed change will reduce the time that certain Primary Containment Isolation Valves (PCIVs) are open and will reduce wear and tear on the Traversing tncore Probe (TIP) system potentially resulting in fewer repairs in a high radiation area. The NRC has previously approved similar amendment requests to the TS for James A. Fitzpatrick Nuclear Power Plant, Vermont Yankee Nuclear Power Station, and River Bend Station - Unit 1. The subject License Amendment Request proposes to adopt surveillance testing requirements similar to those discussed in the previously approved amendments.

Exelon requests approval of the proposed amendment by October 19,2008, with the amendment being implemented within 60 days of issuance.

The requested approval date and implementation period will allow sufficient time for effective planning and scheduling of affected activities associated with LPRM calibration.

There are no regulatory commitments contained in this letter.

These proposed changes have been reviewed by the Plant Operations Review Committee, and approved by the Nuclear Safety Review Board. 7 License Amendment Request Revise LPRM Calibration Frequency Docket Nos. 50-352 and 50-353 October 19,2007 Page 2 Pursuant to 10 CFR 50.91 (b)(l), a copy of this License Amendment Request is being provided to the designated official of the Commonwealth of Pennsylvania.

If any additional information is needed, please contact Mr. Richard Gropp at 61 0-765-5557.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the lgth day of October 2007. Res pectf u I I y, Pamela B. Gowan Director, Licensing and Regulatory Affairs Exelon Generation Company, LLC Attachments:

1 - Evaluation of Proposed Changes 2 - Markup of Proposed Technical Specifications Page Changes 3 - Markup of Proposed Technical Specifications Bases Page Changes cc: Regional Administrator - NRC Region I w/ attachment NRC Senior Resident Inspector - LGS NRC Project Manager, NRR - LGS Director, Bureau of Radiation Protection - Pennsylvania Department of Environmental Protection (6 IL IL ATTACHMENT 1 Evaluation of Proposed Changes LGS, Units 1 and 2 Facility Operating License Nos. NPF-39 and NPF-85 "Revise Local Power Range Monitor Calibration Frequency" I .O DESCRIPTION

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration 5.2 Ap p I i ca b le Regulatory Req u i reme n tslC rite ri a 6.0 ENVIRONMENTAL EVALUATION

7.0 REFERENCES

ATTACHMENT 1 Evaluation of Proposed Changes 1 .O DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit," Exelon Generation Company, LLC (Exelon) requests the following amendment to Appendix A, Technical Specifications (TS), of Facility Operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station (LGS), Units 1 and 2. Specifically, the proposed changes will revise Surveillance Requirement (SR) 4.3.1.1 (Table 4.3.1.1-1) to increase the frequency interval between Local Power Range Monitor (LPRM) calibrations from 1000 effective full power hours (EFPH) to 2000 EFPH. Increasing the frequency interval between required LPRM calibrations is acceptable due to improvements in core monitoring processes and nuclear instrumentation.

In addition, this proposed change will reduce the time that certain Primary Containment Isolation Valves (PCIVs) are open and will reduce wear and tear on the Traversing lncore Probe (TIP) system potentially resulting in fewer repairs in a high radiation area. The NRC has previously approved similar amendment requests to the TS for James A. Fitzpatrick Nuclear Power Plant (Reference 6), Vermont Yankee Nuclear Power Station (Reference 7), and River Bend Station - Unit 1 (Reference 8). The subject License Amendment Request proposes to adopt surveillance testing requirements similar to those discussed in the previously approved amendments.

We request approval of the proposed license amendment by October 19,2008. Once approved, the amendment will be implemented within 60 days of issuance.

2.0 PROPOSED CHANGE

The purpose of this proposed change is to revise the Technical Specifications (TS) Surveillance Requirement (SR) for periodic calibration of the LPRMs. The current requirement is stipulated by SR 4.3.1.1 (Table 4.3.1.1-1 ) contained in TS 3/4.3.1, "Reactor Protection System Instrumentation.

SR 4.3.1 .I (Table 4.3.1 .I-I) specifies that LPRMs be calibrated at a frequency of every 1000 EFPH. The proposed change will revise the frequency of the surveillance to every 2000 EFPH and will read as follows: (f) The LPRMs shall be calibrated at least once per 2000 effective full power hours (EFPH). Additionally, in support of this proposed TS change, the associated TS Bases Section 3/4.3.1 will be revised to reflect the change in the LPRM calibration frequency from 1000 EFPH to 2000 EFPH. The Bases change is being provided for information only.

3.0 Background The LPRM subsystem consists of fission chamber detectors, signal conditioning equipment, display and alarm equipment, associated power supplies, cabling, and trip functions. The LPRM system also provides neutron flux signal inputs to the Average Power Range Monitor (APRM) system, Oscillation Power Range Monitor (OPRM) system, Rod Block Monitor (RBM) system, and the 30 MONICORE core monitoring system. The APRM system provides indication of core average thermal power and input to the Reactor Protection System (RPS). Page 1 of 8 ATTACHMENT 1 Evaluation of Proposed Changes The OPRM system is capable of detecting thermal-hydraulic instability by monitoring the local neutron flux within the reactor core. The OPRM system also provides input to the RPS. The RBM system prevents the withdrawal of selected control rods when local power is above a preset limit.

LPRM inputs to the 3D MONICORE system are used to calculate core power distribution and ensure operation within established fuel thermal operating limits. The LPRM system is comprised of 43 LPRM detector strings radially distributed throughout the core. Each detector string contains four (4) fission chambers located at fixed axial elevations. Each fission chamber produces an output current that is processed by the LPRM signal- conditioning equipment to provide the desired scale indications. Adjacent to each LPRM string is a calibration tube through which Traversing In-Core Probe (TIP) movable gamma detectors are periodically traversed to provide a continuous axial gamma flux profile at each LPRM string location. This data is used in the calibration of the 172 fixed LPRM fission detectors.

The LPRM output signals are transmitted to the Power Range Neutron Monitoring System (PRNMS) operator display assemblies in the Main Control Room. LPRM readings are also directly displayed on the reactor control panel for the detectors adjacent to a selected control rod. The LPRM system is designed to provide a sufficient number of LPRM signals to satisfy the safety design basis of the APRM, OPRM, RBM, and 3D MONICORE systems. This safety design basis is to detect conditions in the core that threaten the overall integrity of the fuel barrier due to excessive power generation and provide signals to the RPS so that the release of radioactive material from the fuel barrier is limited. The LPRM system also incorporates features designed to diagnose and display various system trip and inoperative conditions.

As discussed above, gamma TIP data is used to perform periodic LPRM channel calibrations. These calibrations compensate for small changes in detector sensitivity resulting from the depletion of 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. Adjacent to each LPRM string is a calibration tube, through which TIP movable gamma detectors are traversed to provide a continuous gamma flux profile at each LPRM 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 LPRM calibration process. At rated thermal power (RTP), 1000 EFPH is approximately 42 days. The proposed change to the SR frequency will essentially double the effective time interval between successive LPRM calibrations from 1000 EFPH to 2000 EFPH which is approximately 84 days. The LGS, Units 1 and 2, Updated Final Safety Analysis Report (UFSAR)

Sections 7.6.1.4 and 7.7.1.6, both titled "Neutron Monitoring System - Instrumentation and Controls,"

provide additional discussion on LPRM, APRM, RBM, and TIP systems.

The accuracy of the LPRM system and its impact on overall power distribution uncertainty are documented in General Electric (GE) Licensing Topical Report NEDO-10958-P-A (Reference 1). Page 2 of 8

ATTACHMENT I Evaluation of Proposed Changes

4.0 TECHNICAL ANALYSIS

LPRM gain settings are determined based upon local neutron flux profiles derived from the TIP system. Appropriately gain adjusted LPRM readings establish the relative local neutron flux profile for input to the APRM, OPRM and RBM systems. The current SR frequency interval between LPRM calibrations is based upon original GE recommendations. SR 4.3.1.

I (Table 4.3.1.1-1) in TS 3/4.3.1 establishes an LPRM calibration frequency of 1000 EFPH average core exposure.

The proposed change would increase the interval between whole core LPRM calibrations to 2000 EFPH. The APRM, OPRM, RBM, 30 MONICORE systems are the only systems that use LPRM readings.

In accordance with TS requirements, APRM readings are maintained within 2% of core thermal power by weekly calibration to heat balance calculations. Since LPRM chamber responses are very linear and do not vary significantly with exposure over the period of a week, the LPRM calibration interval extension will have no significant effect on APRM accuracy during power maneuvers or transients. The purpose of the OPRM system is to monitor the LPRM signals to detect thermal-hydraulic instabilities in the reactor core and to generate an automatic suppression signal to terminate the instability if oscillation amplitude, growth, or period exceed predefined levels. The accuracy of the OPRM function is not dependent on the absolute value of individual LPRM readings when the reactor is at equilibrium. The proposed LPRM calibration interval extension will have no significant effect on OPRM accuracy during plant transients.

For the RBM, when a rod is selected, the RBM channel readings are automatically nulled to 100% of scale and the rod block trips are set to a percentage calculated during the reload licensing process. The system monitors for relative changes in LPRM response in the vicinity of a selected control rod, and is insensitive to the absolute value of individual LPRM readings.

Therefore, it is concluded that the performance of the APRM, OPRM, and RBM, Nuclear Instrumentation systems will not be significantly affected by the proposed LPRM surveillance interval increase.

With regard to the 3D MONICORE core monitoring system, the justification to increase the surveillance interval is based on maintaining the overall uncertainty in power distribution calculation within the limits contained in an NRC-approved Licensing Topical Report, NEDO- 1 0958-P-A, "General Electric BWR Thermal Analysis Basis (GETAB) Data, Correlation and Design Application, dated January 1977 (Reference 1). Global Nuclear Fuel (GNF) evaluations of LPRM calibration nodal and bundle power standard deviation data indicate that the calibration frequency has only a small effect on the overall nodal power distribution uncertainty associated with LPRM based operation between successive LPRM calibrations. This small additional uncertainty will not cause an increase in the total power distribution uncertainty to a value in excess of the 8.7% value allowed by the GE Thermal Analysis Basis (GETAB) safety limit analysis. The existing surveillance frequency (i.e., 1000 EFPH) was based on using the older GE P-I, Periodic Core Evaluation, software in the evaluation of core power distribution and fuel operating limits. This original software did not contain the sophisticated neutron diffusion and adaptive learning models used by the current 3D MONICORE system. Furthermore, the original GETAB analysis was based on core monitoring with first generation GE LPRM detectors. These older design LPRM chambers for core monitoring experienced certain inaccuracies Page 3 of 8 ATTACHMENT 1 Evaluation of Proposed Changes related to depletion and loss of fissile material and fill gas between calibrations.

These detectors introduced larger uncertainties into the GETAB analysis than the LPRM designs (i.e., NA 300 Series) currently in-service in the LGS, Units 1 and 2, reactors.

GE evaluation of data from several plants has confirmed that, given the improved performance of the current generation LPRM chambers and the improved analytical methods incorporated into the 30 MONICORE core monitoring software, the nodal power distribution uncertainty is not substantially dependent upon the exposure interval between LPRM calibrations.

The GE evaluation confirms that the LPRM calibration interval may be increased to 2000 EFPH without exceeding the total power distribution uncertainty limit of 8.7% cited in the original GETAB analysis. The technical bases for extending the interval between LPRM calibrations to 2000 EFPH have been previously reviewed and approved by the NRC staff (Reference 2). The licensing topical reports considered in Reference 2, provide detailed statistical evaluations of the uncertainties associated with LPRM-adaptive 3D MONICORE core monitoring calculations. Based on the data examined, it has been shown that the nodal power distribution uncertainty does not significantly change with LPRM exposure. These evaluations provide a basis for confidence that the GETAB equivalent power distribution uncertainty of 8.7% will not be exceeded as a result of extending the LPRM calibration frequency to 2000 EFPH. This is due to improved performance of current generation LPRM chambers, which consistently exhibit less sensitivity loss as a function of exposure throughout their useful nuclear life (up to 40,000 MWD/T).

Also, LGS uses the 30 MONICORE core monitoring software which is substantially more accurate, and less dependent on LPRM inputs, than the original GE P-I power distribution calculation due to incorporation of a more sophisticated methodology (i.e., nodal diffusion theory coupled with plant data and improved nuclear instrumentation).

The analysis is inherently conservative for LGS applications due to the assumed use of neutron TIPs, which are substantially less accurate than gamma TIPs. It should be noted that the analysis also supports LPRM calibration at 2000 EFPH intervals in the bounding situation of up to one-third of all TIP string data being unavailable. Other bounded conditions in the analysis include: failure of up to 25% of all LPRM detectors, significantly asymmetric control rod patterns and minor core loading pattern asymmetries, and significant control rod pattern adjustments in the middle of the LPRM calibration interval. Note that typical LGS practice is to operate with most or all LPRM detectors and TIP machines in service, symmetric rod patterns and core loading patterns, and relatively long intervals between significant control rod pattern adjustments.

GE has performed detailed statistical evaluations (Reference

9) of the uncertainty in LPRM- based monitoring cases at exposure intervals up to 2991 EFPH. Analysis of the data shows that nodal power uncertainty did not significantly deviate with exposure. These evaluations provide the basis that the Reference 1 equivalent safety limit of 8.7% would not be exceeded even with a 25% surveillance extension (2500 EFPH) as permitted by TS SR 4.0.2.

This is because of the use of improved NA 300 series LPRM chambers at LGS, which exhibit consistent LPRM sensitivity throughout their useful nuclear life (up to 40,000 MWD/T), and improved core monitoring systems.

LGS uses the 3D MONICORE system which utilizes nodal diffusion theory, coupled with plant data, and the improved nuclear instrumentation. The 30- MONICORE model is based on accepted BWR calculation methods used to monitor on-line core performance. Page 4 of 8 ATTACHMENT 1 Evaluation of Proposed Changes Conclusion The performance of the APRM, OPRM, RBM, and 30 MONICORE systems will not be significantly affected by the proposed LPRM calibration surveillance interval increase to 2000 EFPH. The significant decrease in overall power distribution uncertainty that results from the use of the improved 3DMONICORE reactor analysis software, and from the improved current generation LPRM detectors, will more than compensate for any small increase in uncertainty which may arise from the increased calibration interval. Evaluations previously reviewed and approved by the NRC, as documented in Reference 2, show that the total power distribution uncertainty for the increased calibration interval of 2000 EFPH will remain bounded by the requirement of 8.7% specified in the Reference 1 GETAB analysis. Improvements in the accuracy of core monitoring methods, and the quality of nuclear instrumentation, support extending the LPRM calibration frequency from 1000 EFPH to 2000 EFPH for LGS, Units 1 and 2. For analyzed cases up to 2991 EFPH, the total nodal uncertainty remains less than the original Reference I requirement of 8.7%.

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit," Exelon Generation Company, LLC (Exelon) requests the following amendment to Appendix A, Technical Specifications (TS), of Facility Operating License Nos. NPF-39 and NPF- 85 for Limerick Generating Station (LGS), Units 1 and 2. Specifically, the proposed change will revise Surveillance Requirement (SR) 4.3.1.1 (Table 4.3.1.1-1) in TS 3/4.3.1 to increase the frequency interval between Local Power Range Monitor (LPRM) calibrations from 1000 EFPH to 2000 EFPH. Increasing the frequency interval between required LPRM calibrations is acceptable due to improvements in core monitoring processes and nuclear instrumentation.

In addition, this proposed change will reduce the time that certain Primary Containment Isolation Valves (PCIVs) are open and will reduce wear and tear on the Traversing lncore Probe (TIP) system potentially resulting in fewer repairs in a high radiation area. According to I0 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: (I) 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 support of this determination, Exelon has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below: Page 5 of 8 ATTACHMENT 1 Evaluation of Proposed Changes Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response:

No The proposed amendment revises the surveillance interval for the Local Power Range Monitor (LPRM) calibrations from 1000 Effective Full Power Hours (EFPH) to 2000 EFPH. Increasing the frequency interval between required LPRM calibrations is acceptable due to improvements in core monitoring processes and nuclear instrumentation and therefore, the revised surveillance interval continues to ensure that the LPRM detector signal is adequately calibrated.

This proposed change will not alter the operation of process variables, structures, systems, or components as described in the LGS Updated Final Safety Analysis Report (UFSAR). The proposed change does not alter the initiation conditions or operational parameters for the LPRM system and there is no new equipment introduced by the extension of the LPRM calibration interval. The performance of the APRM, OPRM, RBM, and 3D MONICORE systems is not significantly affected by the proposed surveillance interval increase.

As such, the probability of occurrence of 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, LPRM chamber exposure has no significant effect on the calculated thermal limits since LPRM accuracy does not significantly deviate with exposure.

For the LPRM extended calibration interval, the total nodal power uncertainty remains less than the uncertainty assumed in the thermal analysis basis safety limit, maintaining the accuracy of the thermal limit calculation. 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 significantly increased.

Therefore, based on the above information, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 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, OPRM, RBM, and 30 MONICORE systems is 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. The proposed change does not change or introduce any new equipment, modes of system operation or failure mechanisms. Therefore, based on the above information, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Page 6 of 8 ATTACHMENT 1 Evaluation of Proposed Changes

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, OPRM, RBM, and 30 MONICORE systems is 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. 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; thereby maintaining thermal limits and the safety margin. The proposed change does not affect safety analysis assumptions or initial conditions and therefore, the margin of safety in the original safety analyses are maintained.

Therefore, based on the above information, the proposed change does not involve a significant reduction in a margin of safety. Based on the above evaluation, Exelon concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c).

5.2 Applicable Regulatory RequirementsKriteria 10 CFR 50.36(~)(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 will be met. The proposed change involves 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 core monitoring processes and nuclear instrumentation and therefore, the revised surveillance interval continues to ensure that the LPRM detector signal is adequately calibrated. This calibration provides assurance that the LPRM accuracy remains within the total nodal power uncertainty assumed in the thermal analysis basis and therefore, the limiting conditions for operation will be met. In conclusion, based on the considerations discussed 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 issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. 6.0 ENVIRONMENTAL EVALUATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined Page 7 of 8 ATTACHMENT 1 Evaluation of Proposed Changes in 10 CFR 20, "Standards for Protection Against Radiation," or would change an inspection or surveillance requirement. However, the proposed amendment does not involve: (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review," Paragraph (c)(9). Therefore, pursuant to 10 CFR 51.22, Paragraph (b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 1. 2. 3. 4. 5. 6. 7. 8. 9. REFERENCES General Electric Licensing Topical Report NEDO-10958-P-A, "General Electric BWR Thermal Analysis Basis (GETAB) Data, Correlation and Design Application," dated January 1977. Letter from F. Akstulewicz (NRR) to G.A Watford (GE), "Acceptance for Referencing of Licensing Topical Reports NEDC-32601 P, 'Methodology and Uncertainties for Safety Limit MCPR Evaluations';

NEDC-32694P, 'Power Distribution Uncertainties for Safety Limit MCPR Evaluation'; and 'Amendment 25 to NEDE-24011 -P-A on Cycle-Specific Safety Limit MCPR' (TAC Nos. M97490, M99069 and M97491), dated March 11,1999. General Electric Licensing Topical Report NEDC-32601 P-A, "Methodology and Uncertainties for Safety Limit MCPR Evaluations," dated August 1999. General Electric Licensing Topical Report NEDC-32694P-A, "Power Distribution Uncertainties for Safety Limit MCPR Evaluations," dated August 1999. General Electric Licensing Topical Report NEDE-32321, "3D MONICORE (RL3D) Performance Evaluation Accuracy," dated January 1994.

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

Fitzpatrick Nuclear Power Plant - Amendment No.

277 Re: Regarding local Power Range Monitor Calibration Frequency (TAC No. MB6945), dated May 1, 2003. Letter from U.S. NRC to S. L. Newton (Vermont Yankee Nuclear Power Corporation), "Vermont Yankee Nuclear Power Station - Issuance of Amendment No. 191 Re: Local Power Range Monitor Calibration Frequency (TAC No. MA9053), dated July 18, 2000. Letter from U.S. NRC to R. K. Edington (Entergy Operations, Inc.), River Bend Station, Unit 1, Amendment No. 107, Re: Changes to Local Power Range Monitor (LPRM) Calibration Frequency (TAC No. M98883), dated June 11, 1999.

General Electric Report entitled "Justification for Operating 2000 EFPH Between OD-I and LPRM Calibration (Rev 3) and Justification for Allowing LPRM GAF Range of .85 to 1 .I5 Following LPRM Calibration (Rev 3)," prepared by G. R. Parkos, dated October 7, 1993, revised June 16, 1994 Page 8 of 8 ATTACHMENT 2 LGS, Units 1 and 2 Facility Operating License Nos. NPF-39 and NPF-85 "Revise Local Power Range Monitor Calibration Frequency" Markup of Proposed Technical Specifications Page Changes REVISED TS PAGES Unit 1 Unit 2 3/4 3-8 3 14 3-8 TABLE 4.3.1 . I -1 (Conti nued) REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS FUNCTIONAL UNIT 9. Turbine Stop Valve - Closure 10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low 11. Reactor Mode Switch 12. Manual Scram Shutdown Posi ti on CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH CHECK (n ) TEST(n) CALIBRATION(a) (n) SURVEILLANCE REQUIRED N.A. 1 N.A. N.A. N.A. N .A. N.A. 1 1, 2, 3, 4, 5 1, 2, 3, 4, 5 Neutron detectors may be excluded from CHANNEL CALIBRATION.

The IRM and SRM channels shall be determined to overlap for at least 1/2 decades during each startup after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlap for at least 1/2 decades during each controlled shutdown, if not performed within the previous 7 days. Cali bration includes verification that the OPRM Upscale trip auto-enable (not-bypass) setpoint for APRM Simulated Thermal Power is 2 30% and for recirculation drive flow is < 60%. The more frequent calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER 225% of RATED THERMAL POWER. Adjust the APRM channel if the absolute difference is greater than 2% of RATED THERMAL POWER. CHANNEL FUNCTIONAL TEST shall include the flow input function, excluding the flow transmitter.

The LPRMs shall be calibrated at least once per 2000 4-WQ effective full power hours (EFPH). The 1 ess frequent cal i brati on i ncl udes the flow input function.

This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1. With any control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

If the RPS shorting links are required to be removed per Specification 3.9.2, they may be reinstalled for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance.

During this time, CORE ALTERATIONS shall be suspended, and no control rod shall be moved from its existing position.

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

2. With THERMAL POWER 2 25% of RATED THERMAL POWER. Frequencies are speci f i ed in the Survei 11 ance Frequency Control Program unless otherwi se noted i n the tab1 e. LIMERICK - UNIT 1 314 3-8 Amendment No. 29, 41, 43, 66, W,-l-u., *,*, Iff, 4-86, xxx TABLE 4.3.1.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS FUNCTIONAL UNIT 9. Turbine Stop Valve - Closure 10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low 11. Reactor Mode Switch 12. Manual Scram Shutdown Position CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH CHECK (n) TEST ( n) CALIBRATION(a) (n) SURVEILLANCE REQUIRED N.A. 1 N.A. N.A. N.A. N.A. N.A. 1 Neutron detectors may be excluded from CHANNEL CALIBRATION.

The IRM and SRM channels shall be determined to overlap for at least 1 /2 decades during each startup after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlap for at least 1/2 decades during each controlled shutdown, if not performed within the previous 7 days. Calibration includes verification that the OPRM Upscale trip auto-enable (not-bypass) setpoint for APRM Simulated Thermal Power is 2 30% and for recirculation drive flow is < 60%. The more frequent calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER 225% of RATED THERMAL POWER. Adjust the APRM channel if the absolute difference is greater than 2% of RATED THERMAL POWER. CHANNEL FUNCTIONAL TEST shall include the flow input function, excluding the flow transmitter.

The LPRMs shall be calibrated at least once per 2000 4-0.W effective full power hours (EFPH) . The less frequent Cali bration includes the flow input function.

This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1. With any control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

If the RPS shorting links are required to be removed per Specification 3.9.2, they may be reinstalled for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance.

During this time, CORE ALTERATIONS shall be suspended, and no control rod shall be moved from its existing position.

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

2. With THERMAL POWER 2 25% of RATED THERMAL POWER. Frequencies are speci f i ed i n the Survei 11 ance Frequency Control Program unl ess otherwi se noted i n the tab1 e . LIMERICK - UNIT 2 314 3-8 Amendment No. 7,~,48,75,?9,~,~,4-39,WlXXx ATTACHMENT 3 LGS, Units 1 and 2 Facility Operating License Nos. NPF-39 and NPF-85 "Revise Local Power Range Monitor Calibration Frequency" Markup of Proposed Technical Specifications Bases Page Changes REVISED TS BASES PAGES Unit 1 Unit 2 B 314 3-la B 314 3-la 314.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued)

Action b, so the voter Function 2.e must be declared inoperable if any of its functionality is inoperable.

inoperable due to any failure affecting only the APRM Interface hardware portion of the Two-Out-Of-Four Logic Module. The voter Function 2.e does not need to be declared Three of the four APRM channels and all four of the voter channels are required to be OPERABLE to ensure that no single failure will preclude a scram on a valid signal. provide adequate coverage of the entire core, consistent with the design bases for the APRM Functions 2.a, 2.b, and 2.c, at least 20 LPRM inputs, with at least three LPRM inputs from each of the four axial levels at which the LPRMs are located, must be operable for each APRM channel. In addition, no more than 9 LPRMs may be bypassed between APRM calibrations (weekly gain adjustments).

For the OPRM Upscale Function 2.f, LPRMs are assigned to "cells" of 3 or 4 detectors.

A minimum of 23 cells (Reference 9), each with a minimum of 2 OPERABLE LPRMs, must be OPERABLE for each APRM channel for the OPRM Upscale Function 2.f to be OPERABLE in that channel. 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 input to the APRM System. The 2000 EFPH frequency is based on operating experience with LPRM sensitivity changes. To References 4, 5 and 6 describe three algorithms for detecting thermal- hydraul i c i nstabi 1 i ty re1 ated neutron f 1 ux osci 11 ations : the period based detection algorithm, the amplitude based algorithm, and the growth rate algorithm.

All three are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the period based detection algorithm.

The remaining algorithms provide defense in depth and additional protection against unanticipated osci 11 ati ons . purposes is based only on the period based detection algorithm.

OPRM Upscal e Functi on OPERABILITY for Techni cal Speci f i cati on An OPRM Upscale trip is issued from an APRM channel when the period based detection algorithm in that channel detects oscillatory changes in the neutron flux, indicated by the combined signals of the LPRM detectors in any cell, with period conf i rmati ons and re1 ati ve cell amp1 i tude exceedi ng speci f i ed setpoi nts . One or more cells in a channel exceeding the trip conditions will result in a channel trip. An OPRM Upscale trip is also issued from the channel if either the growth rate or amplitude based algorithms detect growing oscillatory changes in the neutron flux for one or more cells in that channel. The OPRM Upscale Function is required to be OPERABLE when the plant is at 2 25% RATED THERMAL POWER. The 25% RATED THERMAL POWER level is selected to provide margin in the unlikely event that a reactor power increase transient occurring while the plant is operating below 30% RATED THERMAL POWER causes a power increase to or beyond the 30% RATED THERMAL POWER OPRM Upscale trip auto-enable point without operator action. Upscale trip automatic-enable function will be OPERABLE when required.

This OPERABILITY requirement assures that the OPRM Actions a, b and c define the Action(s) required when RPS channels are discovered to be inoperable.

for each inoperable RPS channel. clock(s) for Actions a, b or c start upon discovery of inoperability for that specific channel. Restoration of an inoperable RPS channel satisfies only the action statements for that particular channel. Action statement(s) for remaining inoperable channel(s) must be met according to their original entry time. For those Actions, separate entry condition is allowed Separate entry means that the allowable time Because of the diversity of sensors available to provide trip signals and the redundancy of the RPS design, an allowable out of service time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> has been shown to be acceptabl e (NEDC-30851 P-A and NEDC-3241 OP-A) to permi t restoration of any inoperable channel to OPERABLE status. However, this out of service time is only acceptable provided that the associated Function's (identified as a "Functional Unit" in Table 3.3.1-1) inoperable channel is in one trip system and the Function still mai ntai ns RPS tri p capabi 1 i ty . LIMERICK - UNIT 1 B 3/4 3-la Amendment No. 43, *49, 4-32, 4-44, W, XM:

3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued)

Action b, so the voter Function 2.e must be declared inoperable if any of its functionality is inoperable.

inoperable due to any failure affecting only the APRM Interface hardware portion of the Two- Out -Of - Four Logi c Modul e , The voter Function 2.e does not need to be declared Three of the four APRM channels and all four of the voter channels are required to be OPERABLE to ensure that no single failure will preclude a scram on a valid signal. provide adequate coverage of the entire core, consistent with the design bases for the APRM Functions 2.a, 2.b, and 2.c, at least 20 LPRM inputs, with at least three LPRM inputs from each of the four axial levels at which the LPRMs are located, must be operable for each APRM channel. In addition, no more than 9 LPRMs may be bypassed between APRM calibrations (weekly gain adjustments).

For the OPRM Upscale Function 2.f, LPRMs are assigned to "cells" of 3 or 4 detectors.

A minimum of 23 cells (Reference 9), each with a minimum of 2 OPERABLE LPRMs, must be OPERABLE for each APRM channel for the OPRM Upscale Function 2.f to be OPERABLE in that channel. 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 input to the APRM System. The 2000 EFPH frequency is based on operating experience with LPRM sensitivity changes. To References 4, 5 and 6 describe three algorithms for detecting thermal- hydraul i c i nstabi 1 i ty re1 ated neutron f 1 ux osci 11 ati ons : the period based detection algorithm, the amplitude based algorithm, and the growth rate algorithm.

takes credit only for the period based detection algorithm.

The remaining a1 gori thms provi de defense in depth and additional protection against unanti ci pated osci 1 1 at i ons purposes is based only on the period based detection algorithm.

All three are implemented in the OPRM Upscale Function, but the safety analysis OPRM Upscal e Funct i on OPERABILITY for Techni cal Speci f i cati on An OPRM Upscale trip is issued from an APRM channel when the period based detection algorithm in that channel detects oscillatory changes.in the neutron flux, indicated by the combined signals of the LPRM detectors in any cell, with period conf i rrnati ons and re1 ati ve cell amp1 i tude exceeding speci fi ed setpoi nts. One or more cells in a channel exceeding the trip conditions will result in a channel trip. An OPRM Upscale trip is also issued from the channel if either the growth rate or amplitude based algorithms detect growing oscillatory changes in the neutron flux for one or more cells in that channel, The OPRM Upscale Function is required to be OPERABLE when the plant is at 2 25% RATED THERMAL POWER. The 25% RATED THERMAL POWER level is selected to provide margin in the unlikely event that a reactor power increase transient occurring while the plant is operating below 30% RATED THERMAL POWER causes a power increase to or beyond the 30% RATED THERMAL POWER OPRM Upscale trip auto-enable point without operator action. Upscale trip automatic-enable function will be OPERABLE when required.

This OPERABILITY requirement assures that the OPRM Actions a, b and c define the Action(s) required when RPS channels are di scovered to be i noperabl e. For those Actions, separate entry condition i s a1 1 owed for each inoperable RPS channel. clock(s) for Actions a, b or c start upon discovery of inoperability for that specific channel. Restoration of an inoperable RPS channel satisfies only the action statements for that particular channel. Action statement(s) for remaining inoperable channel (s) must be met according to their original entry time. Separate entry means that the allowable time Because of the diversity of sensors available to provide trip signals and the redundancy of the RPS design, an allowable out of service time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> has been shown to be acceptable (NEDC-30851P-A and NEDC-32410P-A) to permit restoration of any inoperable channel to OPERABLE status. However, this out of service time is only acceptable provided that the associated Function's (identified as a "Functional Unit" in Table 3.3.1-1) inoperable channel is in one trip system and the Function still maintains RPS tri p capabi 1 i ty , LIMERICK - UNIT 2 B 3/4 3-la Amendment No. W,52,93,W,-WJ,XXX