Text

Units 1 and 2, Vogtle Electric Generating Plant - Units 1 and 2, Voluntary License Amendment Request to Use Beacon Power Distribution Monitoring System
	ML21160A257
Person / Time
Site:	Vogtle, Farley
Issue date:	06/09/2021
From:	Gayheart C; Southern Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NL-21-0042
	Download: ML21160A257 (117)
	v • d • e

3535 Colonnade Parkway Birmingham, AL 35243 205 992 5316 tel 205 992 7601 fax cagayhea@southernco.com Cheryl A. Gayheart Regulatory Affairs Director June 9, 2021 Docket Nos.: 50-348 50-424 NL-21-0042 50-364 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant - Units 1 and 2 Vogtle Electric Generating Plant - Units 1 and 2 Voluntary License Amendment Request to Use BEACON Power Distribution Monitoring System Ladies and Gentlemen:

Pursuant to 10 CFR 50.90, Southern Nuclear Operating Company (SNC) hereby requests a license amendment to Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2 renewed operating licenses NPF-2 and NPF-8, respectively, and Vogtle Electric Generating Plant (VEGP) Units 1 and 2 renewed facility operating licenses NPF-68 and NPF-81, respectively.

Specifically, SNC proposes to revise Technical Specification (TS) 3.1.7, "Rod Position Indication," TS 3.2.1, "Heat Flux Hot Channel Factor (FQ(Z))," and TS 3.3.1, "Reactor Trip System (RTS) Instrumentation" to allow the use of an alternate means of determining power distribution information.

The proposed TS changes will allow the use of a dedicated on-line core power distribution monitoring system (PDMS) to perform surveillance of core thermal limits. The PDMS to be used at FNP and VEGP is the NRC-approved Westinghouse proprietary core analysis system called Best Estimate Analyzer for Core Operations - Nuclear (BEACONTM).

The Enclosure to this letter provides the evaluation of the proposed TS changes for the PDMS. The Enclosure contains nine Attachments:

1. Evaluation for Excluding Power Distribution Monitoring System Requirements from the Technical Specifications

2. Farley Nuclear Plant 1&2 Marked-up TS Pages

3. Farley Nuclear Plant 1&2 Revised TS Pages

4. Vogtle Electrical Generating Plant 1&2 Marked-up TS Pages

5. Vogtle Electrical Generating Plant 1&2 Revised TS Pages

6. Farley Nuclear Plant 1&2 Marked-up TS Bases Pages (Information only)

7. Vogtle Electrical Generating Plant 1&2 Marked-up TS Bases Pages (Information only)

8. Farley Nuclear Plant 1&2 Marked-up TRM Pages (Information only)

9. Vogtle Electrical Generating Plant 1&2 Marked-up TRM Pages (Information only)

SNC has determined that the information for the proposed amendment does not involve a significant hazards consideration, authorize a significant change in the types or total amounts of effluent release, or result in any significant increase in individual or cumulative occupational

U. S. Nuclear Regulatory Commission NL-21-0042 Page 2 radiation exposure. Therefore, the proposed amendment meets the categorical exclusion requirements of 10 CFR 51.22(c)(9) and pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

In accordance with 10 CFR 50.91, SNC is notifying the State of Alabama and State of Georgia of this license amendment request by transmitting a copy of this letter and enclosures to the designated State Official. This letter contains no NRC commitments.

Approval of the proposed amendment is requested within one year from the date of this Submittal with implementation within 90 days following issuance of the amendment.

If you have any questions, please contact Jamie Coleman at 205.992.6611.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 9th day of June 2021.

Respectfully submitted, C. A. Gayheart Director, Regulatory Affairs Southern Nuclear Operating Company CAG/RMJ

Enclosure:

Evaluation of Proposed Change cc:

Regional Administrator, Region ll NRR Project Manager - Farley 1 & 2 Senior Resident Inspector - Farley 1 & 2 NRR Project Manager - Vogtle 1 & 2 Senior Resident Inspector - Vogtle 1 & 2 Director, State of Alabama Office of Radiation Protection State of Georgia Environmental Protection Division RType: CFA04.054, CVC7000

Joseph M. Farley Nuclear Plant - Units 1 and 2 Vogtle Electric Generating Plant - Units 1 and 2 Voluntary License Amendment Request to Use BEACON Power Distribution Monitoring System Enclosure Evaluation of Proposed Change

Enclosure NL-21-0042 Evaluation of Proposed Change E-1 ENCLOSURE Evaluation of the Proposed Change

Subject:

Voluntary License Amendment Request to Use BEACON Power Distribution Monitoring System

1.

SUMMARY

DESCRIPTION

2. DETAILED DESCRIPTION 2.1 System Design and Operation 2.2 Current Technical Specifications Requirements 2.3 Reason for the Proposed Change 2.4 Description of the Proposed Change

3. TECHNICAL EVALUATION 3.1 WCAP-12472-P-A 3.2 Applicability of Addenda to WCAP-12472-P-A

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration Analysis 4.4 Conclusions

5. ENVIRONMENTAL CONSIDERATION

6. REFERENCES ATTACHMENTS:

1. Evaluation for Excluding BEACON Power Distribution Monitoring System (PDMS)

Requirements from the Technical Specifications

2. Farley Nuclear Plant 1&2 Marked-up TS Pages

3. Farley Nuclear Plant 1&2 Revised TS Pages

4. Vogtle Electrical Generating Plant 1&2 Marked-up TS Pages

5. Vogtle Electrical Generating Plant 1&2 Revised TS Pages

6. Farley Nuclear Plant 1&2 Marked-up TS Bases Pages (Information only)

7. Vogtle Electrical Generating Plant 1&2 Marked-up TS Bases Pages (Information only)

8. Farley Nuclear Plant 1&2 Marked-up TRM Pages (Information only)

9. Vogtle Electrical Generating Plant 1&2 Marked-up TRM Pages (Information only)

Enclosure NL-21-0042 Evaluation of Proposed Change E-2

1.

SUMMARY

DESCRIPTION Pursuant to 10 CFR 50.90, Southern Nuclear Operating Company (SNC) hereby requests a license amendment to Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2 renewed operating licenses NPF-2 and NPF-8, respectively, and Vogtle Electric Generating Plant (VEGP) Units 1 and 2 renewed facility operating licenses NPF-68 and NPF-81, respectively.

Specifically, SNC proposes to revise Technical Specification (TS) 3.1.7, "Rod Position Indication," TS 3.2.1, "Heat Flux Hot Channel Factor (FQ(Z))," and TS 3.3.1, "Reactor Trip System (RTS) Instrumentation" to allow the use of an alternate means of determining power distribution information.

Enclosure NL-21-0042 Evaluation of Proposed Change E-3

2. DETAILED DESCRIPTION 2.1 System Design and Operation The Movable Incore Detector System (MIDS) consists of a set on miniature fission chamber detectors used to measure localized core power density. The MIDS is designed to insert these detectors into selected radial core locations. As the detectors traverse the length of the fuel assemblies, localized core power density measurements are taken at selected axial locations.

The core power density information in then processed to determine core power distribution information. This information can then be used to satisfy Technical Specification requirements for verification of control rod position, determining core power peaking factors, determining radial power tilts, and calibration of excore detector axial power input to reactor protection. The MIDS does not perform any reactor control or reactor protection functions.

The core power distribution monitoring system (PDMS) to be used at FNP and VEGP utilizes the NRC-approved Westinghouse proprietary core analysis system called the Best Estimate Analyzer for Core Operations - Nuclear (BEACON), together with continuous information from plant instrumentation. lncore detector measurements are used to periodically calibrate the BEACON PDMS. The BEACON PDMS serves as a three-dimensional (3-D) core monitor, operational analysis tool, and operational support package.

Westinghouse submitted topical report WCAP-12472-P, "BEACON Core Monitoring and Operations Support System," to the NRC on May 21, 1990. The NRC issued a Safety Evaluation Report (SER) approving the topical report on February 16, 1994. In its SER, the NRC concluded that BEACON is acceptable for performing core monitoring and operations support. The SER is contained in WCAP-12472-P-A (Reference 1).

BEACON has three monitoring levels that interface with plant instrumentation: BEACON-OLM (On-Line Monitor), BEACON-TSM (Technical Specification Monitor), and BEACON-DMM (Direct Margin Monitor).

The BEACON-OLM system level was developed to provide licensees with the same level of functionality and application that was being used before the licensing of BEACON. This system level provides the base functionality of the BEACON system which includes continuous core monitoring, core predictive capability and operational history analysis. This system level is used for information and analysis purposes and does not require operational action based on results from the core monitor displays. This level of the BEACON system is purely an information and analysis tool that plant operational personnel can use at their option. The use of the BEACON-OLM level can be integrated into the plant procedures. If this is done, then the flux map analysis and estimated critical condition (ECC) functions from BEACON can be used to replace other off-line codes and procedures used for these calculations.

The BEACON-TSM system level was developed to provide licensees with the functionality needed to integrate BEACON into the plant TS for monitoring of current TS thermal limits such as peak linear power density (TS 3.2.1, Heat Flux Hot Channel Factor (FQ(Z))) and peak enthalpy rise (TS 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor (FNH)). BEACON-TSM includes all the base functionality in the BEACON-OLM level. Added to this are the procedures, system operational status information and on-line calculations needed to provide the core monitoring capability for TS compliance. The licensing of BEACON for core monitoring allows the BEACON on-line monitoring functions to potentially eliminate most flux maps for normal and

Enclosure NL-21-0042 Evaluation of Proposed Change E-4 off-normal TS thermal limit verification. Once integrated into the plant TS and procedures, the BEACON-TSM system has the potential to provide the following benefits:

Essentially continuous monitoring of the core power distribution.

Increased interval for flux maps (using movable incore detectors) from 31 EFPD to 180 EFPD (flux maps are only required for BEACON calibration, when thermal power is less than 25% Rated Thermal Power (RTP), or when PDMS is non-functional).

Verification of the position of a rod with an inoperable rod position indicator (RPI) using the BEACON core monitor function.

Reduced movable incore detector instrumentation requirements to 50% after initial calibration for a fuel cycle. The BEACON system uses surface spline fitting to compensate for sparse instrumentation and automatically adjusts the applied thermal limit uncertainties allowing for operation with reduced instrumentation.

The BEACON-DMM level was developed to provide licensees with the full functionality and benefits of the BEACON license granted by the NRC. BEACON-DMM includes all the functionality of BEACON-TSM and provides for direct monitoring and use of Departure from Nucleate Boiling Ratio (DNBR) as a thermal limit in the plant TS. SNC does not propose to license the BEACON-DMM application of the PDMS for FNP and VEGP.

SNC proposes to license the BEACON-TSM application of the PDMS as an alternate means for obtaining power distribution information and performing surveillances when thermal power is greater than or equal to 25% RTP. The Technical Requirements Manual (TRM) will implement the associated PDMS functionality requirements. At thermal power levels less than 25% RTP, or when the PDMS is non-functional or as an alternative for obtaining power distribution information, the movable incore detector system will be used.

2.2 Current Technical Specifications Requirements The following Technical Specifications for FNP and VEGP Units specify the use of the Movable Incore Detector System (MIDS) for core power distribution determination (flux maps):

TS 3.1.7 Rod Position Indication TS 3.2.1 Heat Flux Hot Channel Factor (FQ(Z))

TS 3.3.1 Reactor Trip System (RTS) Instrumentation 2.3 Reason for the Proposed Change The BEACON-TSM system level was developed to provide licensees with the functionality needed to integrate BEACON into the plant TS for monitoring of current TS thermal limits such as peak linear power density and peak enthalpy rise. The licensing of BEACON for core monitoring allows the BEACON online monitoring functions to potentially eliminate most flux maps for normal and off-normal TS thermal limit verification. Once integrated into the plant TS and procedures, the BEACON-TSM system has the potential to provide the benefits previously described in Section 2.1 of this enclosure.

2.4 Description of the Proposed Change The following proposed Technical Specification changes are applicable to FNP and VEGP unless specifically noted otherwise.

Enclosure NL-21-0042 Evaluation of Proposed Change E-5 TS 3.1.7, Rod Position Indication In Required Actions A.1, A.2.1, C.1.1 (FNP), and C.1 (VEGP), movable incore detectors will be replaced with core power distribution information.

TS 3.2.1, Heat Flux Hot Channel Factor (FQ(Z)) (FQ Methodology)

In the NOTE above FNP SR 3.2.1.1, a power distribution map will be replaced with core power distribution information.

In the NOTE for SR 3.2.1.2, measurements indicate will be deleted.

In NOTE b in SR 3.2.1.2, flux maps will be replaced with surveillances.

TS 3.3.1, Reactor Trip System (RTS) Instrumentation In SR 3.3.1.3, the incore detector measurements will be replaced with core power distribution information.

In SR 3.3.1.9 (FNP) and SR 3.3.1.6 (VEGP), incore detector measurements will be replaced with core power distribution information.

Marked-up and Revised TS pages for FNP are included in Attachments 2 and 3 respectively.

Marked-up and Revised TS pages for VEGP are included in Attachments 4 and 5 respectively.

The TS Bases will be revised for consistency with the proposed TS changes. The Bases for TS 3.1.4, Rod Group Alignment Limits, TS 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor, and TS 3.2.4, Quadrant Power Tilt Ratio (QPTR), are also marked up to revise references to movable incore detectors and flux maps. FNP and VEGP marked-up TS Bases pages are included in Attachments 6 and 7 respectively for information only and will be updated in accordance with FNP and VEGP TS 5.5.14, Technical Specifications (TS) Bases Control Program.

SNC also proposes changes to the FNP and VEGP Technical Requirements Manual (TRM) to specify functionality requirements for the BEACON Power Distribution Monitoring System (PDMS). Technical Requirements (TR) and Technical Requirement Surveillances (TRS) are treated as plant procedures. Changes will be made in accordance with 10 CFR 50.59.

Attachments 8 and 9 provide (for information only) the proposed new Technical Requirement (TR) 13.3.10 (FNP) and 13.3.9 (VEGP), Power Distribution Monitoring System as well as an additional clarification note added to FNP and VEGP TR 13.3.1.

Enclosure NL-21-0042 Evaluation of Proposed Change E-6

3. TECHNICAL EVALUATION 3.1 WCAP-12472-P-A The PDMS to be used at FNP and VEGP utilizes the NRC-approved Westinghouse proprietary core analysis system called the Best Estimate Analyzer for Core Operations - Nuclear (BEACON), together with continuous information from plant instrumentation. lncore detector measurements are used to periodically calibrate the BEACON PDMS. The BEACON PDMS serves as a three-dimensional (3-D) core monitor, operational analysis tool, and operational support package.

Westinghouse submitted topical report WCAP-12472-P, "BEACON Core Monitoring and Operations Support System," to the NRC on May 21, 1990. The NRC issued a Safety Evaluation Report (SER) approving the topical report on February 16, 1994. In its SER, the NRC concluded that BEACON is acceptable for performing core monitoring and operations support for Westinghouse reactors. The SER is contained in WCAP-12472-P-A (Reference 1).

Westinghouse developed a reactor core power distribution monitoring system, which is described in WCAP-12472-P-A. This system uses instrumentation currently existing in Westinghouse reactors, but processes the information differently than is current practice, using a newly added on-line reactor neutronics calculation system.

Westinghouse reactors have two neutron flux measuring systems for power operation, incore and excore. These, however, do not by themselves provide direct, continuous determination of power distribution or direct relationship of the power distribution to fuel safety limits, i.e., peak power density or departure from nucleate boiling (DNB).

In addition to the neutron monitoring system, the Westinghouse reactors also measure (1) coolant temperature via resistance temperature detectors (RTDs) at the core inlet and outlet to measure power level as part of the protection system and (2) radial/azimuthal temperature distribution with thermocouples at the core outlet, which produces a measure of the radial distribution of the axially integrated power.

BEACON uses these elements in the core monitoring and support analysis system. The operation is based largely on, and tied together by, the use of a three-dimensional neutronics analysis code. SPNOVA, which was approved by the NRC in November 1990 (ML20077H136),

was the code originally used in BEACON. The primary role of the three-dimensional neutronics analysis code in BEACON is to generate detailed power distribution information. The code is calibrated periodically using the incore neutron flux measurement system to provide details of the power distribution and calibrated frequently (essentially continuously) using the core exit thermocouples for radial updating and the excore neutron detectors for axial updating. The incore information is also used to calibrate the thermocouples and the excore detectors.

WCAP-12472-P-A describes the system, the methodologies involved, the calibration processes, the uncertainties to be associated with the determined power distributions, xenon transient and criticality analysis, the calculation of more direct limiting conditions for operation (LCOs) than are currently used and Technical Specification modifications that would be necessary for BEACON operating and non-operating conditions.

WCAP-12472-P-A describes the minimum monitoring instrumentation requirements for BEACON to be considered functional. In particular, it discusses minimum requirements for the

Enclosure NL-21-0042 Evaluation of Proposed Change E-7 number of and configuration of movable incore detectors and core exit thermocouples for the calibration of BEACON. The thermocouple requirements are used to determine the frequency of calibration of BEACON. The accuracy of the power distribution information with decreased incore or thermocouple detector operability has been analyzed by Westinghouse, and penalties are applied to the calculated peaking factors.

In the NRC Safety Evaluation Report (SER) for WCAP-12472-P, the NRC staff evaluated the BEACON methodology, the uncertainty analysis, and the operation of the overall system and concluded that the BEACON PDMS is acceptable for performing core monitoring and operations support functions for Westinghouse pressurized water reactors (PWR) but subject to certain conditions as specified in the Technical Evaluation Report (TER). The SER and TER are contained in Reference 1. The conditions are listed below. After each condition listed, a description of how SNC will meet the condition for FNP and VEGP is provided.

1.

In the cycle-specific application of BEACON, the power peaking uncertainties UH and UQ must provide 95% probability upper tolerance limits at the 95% confidence level.

Cycle-specific BEACON calibrations performed before startup and at beginning-of-cycle conditions will ensure that power peaking uncertainties provide 95% probability upper tolerance limits at the 95% confidence level. These calibrations will be performed using the Westinghouse methodology. Until these calibrations are complete, more conservative default uncertainties will be applied. SNC has a database tracking item to ensure these calibration requirements will be documented in procedures and retained as records.

2.

In order to ensure that the assumptions made in the BEACON uncertainty analysis remain valid, the generic uncertainty components may require reevaluation when BEACON is applied to plant or core designs that differ sufficiently to have a significant impact on the WCAP-12472-P database.

FNP and VEGP utilize a Westinghouse 3-loop and 4-loop nuclear steam supply system (NSSS), respectively, with Westinghouse movable incore detector instrumentation, and fuel of Westinghouse manufacture. Therefore, FNP and VEGP do not differ significantly from the plants that form the WCAP database, and no additional review of the WCAP-12472-P-A applicability to FNP and VEGP is necessary. It should be noted that the VEGP Unit 2 core currently has a limited number of non-Westinghouse lead test assemblies (LTA) as permitted by Technical Specification 4.2.1. In accordance with TS 4.2.1, only a limited number of LTAs are used, and these are placed in nonlimiting core locations. The impact of the LTAs is assessed on a cycle-specific basis as part of the normal reload design process using NRC-approved methods as required by Technical Specification 4.2.1.

During the review of the Westinghouse topical report WCAP-12472-P, the NRC requested additional information on how BEACON treats core loadings with fuel designs from multiple fuel vendors, and the impact to the BEACON uncertainty analysis. Westinghouse responded that for all BEACON applications, the previous operating cycle is examined to establish reference uncertainties. This examination accounts for loading of fuel supplied by multiple vendors by comparing a BEACON model to actual operating data over the cycle. At the beginning of cycle, thermocouple data are verified, and calibration/uncertainty components are updated as necessary. In addition, the initial flux mapping at the start of the cycle ensures model calibration factors that reflect the actual

Enclosure NL-21-0042 Evaluation of Proposed Change E-8 fuel in the reactor before the BEACON system is declared FUNCTIONAL. Westinghouse also responded that the source of the BEACON model is the same set of nuclear and T&H models (PHOENIX-P, PARAGON, ANC, DNB correlations, etc.) that are used to perform the reload design calculations. Thus, there exists the same pedigree as the codes used to license the reload cycle, which can therefore be accurately captured by BEACON and remain within the uncertainty analysis for core designs when comparing to previous cycles and the reload cycle.

3. The BEACON Technical Specifications should be revised to include the changes described in Section 3 (of the BNL TER) concerning Specifications 3.1.3.1 and 3.1.3.2 and the Core Operating Limits Report.

WCAP-12472-P-A (Reference 1) describes an application of BEACON where the core operating limits are changed. As noted previously, SNC is proposing only to use BEACON as a core TS monitor for conformance to existing limits for FNP and VEGP. At the time of TER issuance, these Specifications referred to continued operation with one (trippable) inoperable rod, which is the equivalent of Specification 3.1.4 of the current Technical Specifications. The TS changes of concern per this question or condition are not applicable to the more limited changes being proposed by SNC for the intended use of BEACON and Specification 3.1.4 is not being changed. Therefore, this condition does not apply to the amendment requested for FNP and VEGP.

Subsequent to the approval of WCAP-12472-P in 1994, the NRC approved four addenda to the WCAP. Each addendum will be discussed below with respect to applicability to the proposed implementation of the BEACON PDMS at FNP and VEGP.

3.2 Applicability of Addenda to WCAP-12472-P-A Addendum 1 Addendum 1 of WCAP-12472-P-A was approved by the NRC on September 30, 1999 (ADAMS Accession Number ML003678190).

Addendum 1 describes additional features incorporated into the BEACON monitoring system:

1. Use of fixed incore self-powered neutron rhodium detectors, and

2. Use of three-dimensional advanced nodal code (ANC) neutronic model code.

Neither FNP nor VEGP use fixed incore detectors so this feature is not applicable. However, ANC is used for FNP and VEGP for cycle-specific reload design analyses. This ensures consistency between the reload design and BEACON models.

Addendum 2 Addendum 2 of WCAP-12472-P-A was approved by the NRC on February 1, 2002 (ADAMS Accession Number ML021270086).

Addendum 2 extends the previously licensed BEACON power distribution monitoring methodology to plants containing platinum and vanadium fixed incore self-powered detectors.

Neither FNP nor VEGP use fixed incore detectors so this Addendum is not applicable.

Enclosure NL-21-0042 Evaluation of Proposed Change E-9 Addendum 3 Addendum 3 of WCAP-12472-P-A was approved by the NRC on September 26, 2005 (ADAMS Accession Number ML052620347).

The objective of this addendum to the approved topical report (TR) is to provide the information and data necessary to approve an upgraded core monitoring system that merges three existing products, Best Estimate Analyzer for Core Operation - Nuclear (BEACON) core monitoring system, Core Operating Limit Supervisory System (COLSS), and the thermal hydraulic analysis computer code CETOP-D, into one, and an uncertainty analysis methodology that will be applied to this new product, BEACON-COLSS.

This Addendum is applicable to Combustion Engineering designed plants. FNP and VEGP are Westinghouse-designed plants so this Addendum is not applicable.

Addendum 4 Addendum 4 of WCAP-12472-P-A was approved by the NRC on August 9, 2012 (ADAMS Accession Number ML12158A263).

The purpose of the Addendum 4 to WCAP-12472-P-A is to:

1. Provide the information needed to review and approve the updated thermocouple uncertainty analysis process that will be applied in the BEACON on-line core monitoring system,

2. Affirm the continued use of the NRC approved Westinghouse design model methodology, currently PHOENIX-P/ANC, PARAGON/ANC, and NEXUS/ANC, in the BEACON system, and

3. Affirm that uncertainties applied to power distribution monitoring using fixed in-core detectors are valid using higher order polynomial fits of the detector variability and fraction of inoperable detectors.

The updated thermocouple uncertainty evaluation method presented in the submitted TR is based on the licensed methodology in the BEACON topical report but uses the current plant/cycle data in the evaluation process to generate cycle-specific uncertainty constants.

There are no new methods being developed for the BEACON system; this update is a change in the application of the approved method. Westinghouse stated in the submittal, that this thermocouple uncertainty methodology is only applied to plants with movable in-core detectors.

These plants use thermocouples to determine the measured power distribution as described in WCAP-12472-P-A, BEACON: Core Monitoring and Operations Support System and the request for additional information (RAI) responses for Addendum 4.

The use of the NRC approved Westinghouse design model methodology PARAGON/ANC, and NEXUS/ANC in the BEACON system is consistent with the methods used in performing FNP and VEGP cycle-specific reload design analyses.

Neither FNP nor VEGP use fixed incore detectors so the third purpose of this Addendum is not applicable.

Enclosure NL-21-0042 Evaluation of Proposed Change E-10

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50, Appendix A General Design Criterion 13 states:

Criterion 13Instrumentation and control. Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

Implementation of the PDMS at FNP and VEGP does not eliminate, replace, or modify existing plant instrumentation. The PDMS software runs on a workstation connected to the integrated plant computer system. The PDMS combines inputs from currently installed plant instrumentation and design data generated for each fuel cycle. Together, this provides a means to continuously monitor the power distribution limits including limiting peaking factors and quadrant power tilt ratio.

With regard to the FUNCTIONALITY and control requirements of the PDMS and its associated instrumentation, SNC has determined that no TS changes are needed for this purpose because the PDMS does not meet the selection criteria set forth in 10 CFR 50.36(c)(2)(ii) for inclusion in the TS. The evaluation for this determination is provided in Attachment 1 of this enclosure.

Further, precedent has been set by similar facilities with respect to use of the PDMS in the application as described herein. See Section 4.2 of this enclosure for details.

In lieu of TS requirements, requirements for the PDMS and associated instrumentation will be placed in the FNP and VEGP Technical Requirements Manual (TRM). Technical Requirements (TR) and Technical Requirement Surveillances (TRS) are treated as plant procedures. Changes will be made in accordance with 10 CFR 50.59. Attachments 8 and 9 provides (for information only) the proposed new Technical Requirement (TR) 13.3.10 (FNP) and 13.3.9 (VEGP), Power Distribution Monitoring System as well as an additional clarification note added to FNP and VEGP TR 13.3.1.

4.2 Precedent BEACON-TSM has been approved by the NRC for use at the following stations:

V. C. Summer, Amendment 142, April 9, 1999 (ADAMS Accession Number ML012260068)

Salem Units 1 and 2, Amendments 237 (Unit 1) and 218 (Unit 2), November 6, 2000, (ADAMS Accession Number ML003761792/ML003767901)

Diablo Canyon Units 1 and 2, Amendments 164 (Unit 1) and 166 Unit 2, March 31, 2004, (ADAMS Accession Number ML040920245)

STP Units 1 and 2, Amendments 175 (Unit 1) and 163 (Unit 2), March 31, 2006 (ADAMS Accession Number ML060760501/ML060950451)

Callaway, Amendment 182, March 21, 2007, (ADAMS Accession Number ML070460584/ML070680350)

Enclosure NL-21-0042 Evaluation of Proposed Change E-11 Commanche Peak Units 1 and 2, Amendments 144 (Unit 1) and 144 (Unit 2), April 2, 2008, (ADAMS Accession Number ML080510083/ML080500627)

Watts Bar Unit 1, Amendment 82, October 27, 2009, (ADAMS Accession Number ML092710381)

Wolf Creek, Amendment 188, July 23, 2010, (ADAMS Accession Number ML100820517)

Prairie Island Units 1 and 2, Amendments 201 (Unit 1) and 188 (Unit 2), May 4, 2011, (ADAMS Accession Number ML103430498)

The most recent amendments were for Prairie Island Units 1 and 2. In this application, a PDMS LCO was not added to the Technical Specifications, but technical requirements regarding FUNCTIONALITY were placed in the Technical Requirements Manual which is a document under licensee control.

Enclosure NL-21-0042 Evaluation of Proposed Change E-12 4.3 No Significant Hazards Consideration Analysis Pursuant to the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), Southern Nuclear Operating Company (SNC) hereby requests an amendment to the Vogtle Electric Generating Plant (VEGP) Units 1 and 2 renewed facility operating licenses NPF-68 and NPF-81, respectively, and Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2 renewed operating licenses NPF-2 and NPF-8, respectively.

The proposed change would revise Technical Specification (TS) 3.1.7, "Rod Position Indication," TS 3.2.1, "Heat Flux Hot Channel Factor (FQ(Z))," and TS 3.3.1, "Reactor Trip System (RTS) Instrumentation," to incorporate use of the Best Estimate Analyzer for Core Operations - Nuclear (BEACON') Power Distribution Monitoring System (PDMS) described in WCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System." The purpose of this system is to perform core power distribution surveillances.

SNC 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:

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

Response: No The PDMS performs continuous core power distribution monitoring with data input from existing plant instrumentation. The system passively supports Technical Specification (TS) surveillances which ensure that core power distribution is within the same limits that are currently prescribed.

Further, the proposed TS Actions are comparable to existing operator actions such that no new plant configurations are prompted by the proposed change. The system's physical interface with plant equipment is limited to an electronic link from the BEACON workstation to the plant process computer. The system is passive in that it provides no control or alarm functions and does not promote any new plant configuration which would affect the initiation, probability, or consequences of a previously evaluated accident.

Continuous on-line core monitoring through the use of PDMS provides significantly more information about the power distributions present in the core than is currently available. This system performance may result in an earlier determination of an adverse core condition and more time for operator action, thus reducing the probability of an accident occurrence and reduced consequences should a previously evaluated accident occur.

By virtue of its inherently passive surveillance function and limited interface with plant systems, structures, or components, the proposed changes will not result in any additional challenges to plant equipment that could increase the probability or occurrence of any previously evaluated accident. Further, the proposed changes will ensure conformance to the same core power distribution limits that form the basis for initial conditions of previously evaluated accidents.

Thereby, the proposed changes will not affect the consequences of any previously evaluated accident.

Therefore, the proposed change does not involve a significant increase in the probability or consequence of an accident previously evaluated.

Enclosure NL-21-0042 Evaluation of Proposed Change E-13 Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The system's physical interface with plant equipment is limited to an electronic link from the BEACON workstation to the plant process computer. The system is passive in that it provides no control or alarm functions, and the proposed changes (including operator actions prescribed by the proposed TS) do not promote any new plant configuration which would create the possibility for an accident of a new or different type.

The NRC previously evaluated the effects of using the PDMS to monitor core power distribution parameters and determined that all design standards and applicable safety criteria limits are met. The Technical Specifications will continue to require operation within the required core operating limits, and appropriate actions will continue to be taken when or if limits are exceeded.

Thus, the reactor core will continue to be operated within its reference bounds of design such that an accident of a new or different type is not credible.

The proposed change, therefore, does not create the possibility of a new or different kind of accident from any previously evaluated.

Do the proposed changes involve a significant reduction in a margin of safety?

Response: No No margin of safety is adversely affected by the implementation of the PDMS. The margins of safety provided by current TS requirements and limits remain unchanged, as the TS will continue to require operation within the core limits that are based on NRC-approved reload design methodologies. The proposed change does not result in changes to the core operating limits. Appropriate measures exist to control the values of these cycle-specific limits, and appropriate actions will continue to be specified and taken when or if limits are exceeded. Such actions remain unchanged.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

Enclosure NL-21-0042 Evaluation of Proposed Change E-14 4.4 Conclusions In conclusion, based on the considerations discussed herein, (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 Commissions 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.

Enclosure NL-21-0042 Evaluation of Proposed Change E-15

5. ENVIRONMENTAL CONSIDERATION 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 in 10 CFR Part 20, and would change an inspection or surveillance requirement. However, the proposed change 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 off site, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Enclosure NL-21-0042 Evaluation of Proposed Change E-16

6. REFERENCES

1. WCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System,",

August 1994 (ML19304C541) [NOTE: The approved version of the WCAP contains the NRC Safety Evaluation Report (SER), the Technical Evaluation Report (TER), and Requests for Additional Information (RAI) and RAI responses.]

Joseph M. Farley Nuclear Plant - Units 1 and 2 Vogtle Electric Generating Plant - Units 1 and 2 Voluntary License Amendment Request to Use BEACON Power Distribution Monitoring System Evaluation for Excluding Power Distribution Monitoring System Requirements from the Technical Specifications to NL-21-0042 Evaluation for Excluding Power Distribution Monitoring System Requirements from the Technical Specifications A1-1 ATTACHMENT 1 Evaluation for Excluding Power Distribution Monitoring System Requirements from the Technical Specifications The purpose for this attachment is to demonstrate that Limiting Conditions for Operation (LCOs) for the Best Estimate Analyzer for Core Operations - Nuclear (BEACON) Power Distribution Monitoring System (PDMS) and associated instrumentation are not required to be included in the FNP and VEGP Technical Specifications (TS). The justification for this statement is explained in the evaluation provided below. The evaluation demonstrates that the structures, systems, or components (i.e., instrumentation) that constitute the PDMS are not required to be contained in the TS in accordance with the requirements contained in 10 CFR 50.36(c)(2)(ii).

10 CFR 50.36(c)(2)(ii) requires that a TS LCO must be established for each item meeting one or more of the following criteria:

(A) Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

The PDMS instrumentation is not associated with monitoring of any aspect of the reactor coolant pressure boundary. Therefore, the PDMS cannot be used to detect or indicate any degradation of the reactor coolant pressure boundary.

(B) Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The limits for the power distribution parameters FQ(Z) and FNH are operating restrictions which ensure that the accident analyses and assumptions for all applicable, analyzed Design Basis Accidents (DBAs) remain valid. These limits are included in the TS and are not changed through implementation of the PDMS. The PDMS supports the capability to monitor core power distribution for verifying conformance to such limits, but it does not control core power distribution. In addition, the PDMS cannot by itself cause or affect any condition assumed in the accident/transient analyses.

The PDMS provides the capability to monitor power distribution parameters at more frequent intervals than is currently required by the TS. These parameters can be determined independent of the FUNCTIONALITY of PDMS. Therefore, the PDMS does not constitute a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(C) Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The PDMS performs only a monitoring function and does not affect any of the key safety parameter limits or levels of margin considered in the DBA evaluations. The PDMS performs no active/control functions, nor does the PDMS have an actuation capability.

to NL-21-0042 Evaluation for Excluding Power Distribution Monitoring System Requirements from the Technical Specifications A1-2 Therefore, the PDMS is not part of any primary success path for mitigation of a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(D) Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

The PDMS and its associated instrumentation provide the capability to monitor power distribution parameters at more frequent intervals than is currently required by the TS, but the PDMS has no active safety functions and its use has no impact on the results or consequences of any DBA or transient analysis. Further, the PDMS is only an alternative means for determining core power distribution information and performing related surveillances, because the current means of performing such activities (by use of the movable incore detectors) is still available. PDMS unavailability, therefore, is not significant relative to plant risk. Based on these considerations and facts, the PDMS is not a feature that is significant to public health and safety.

The evaluation completed above indicates that the BEACON PDMS does not meet any of the criteria for inclusion in the TS. The PDMS requirements and controls to be incorporated into the Technical Requirements Manual (TRM) are consistent with the recommendations in WCAP-12472-P-A and will suffice to provide the necessary FUNCTIONALITY and test requirements for the PDMS apart from the TS. Attachments 8 and 9 provides (for information only) the proposed new Technical Requirement (TR) 13.3.10 (FNP) and 13.3.9 (VEGP), Power Distribution Monitoring System as well as an additional clarification note added to TR 13.3.1.

Joseph M. Farley Nuclear Plant - Units 1 and 2 Vogtle Electric Generating Plant - Units 1 and 2 Voluntary License Amendment Request to Use BEACON Power Distribution Monitoring System Farley Nuclear Plant 1&2 Marked-up TS Pages

Rod Position Indication 3.1.7 Farley Units 1 and 2 3.1.7-1 Amendment No. 214 (Unit 1)

Amendment No. 211 (Unit 2) 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Rod Position Indication LCO 3.1.7 The Digital Rod Position Indication (DRPI) System and the Demand Position Indication System shall be OPERABLE.

APPLICABILITY:

MODES 1 and 2.

ACTIONS

NOTE-----------------------------------------------------------

Separate Condition entry is allowed for each inoperable DRPI and each demand position indicator.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One DRPI per group inoperable in one or more groups.

A.1 Verify the position of the rod with inoperable DRPI indirectly by using movable incore detectors.

OR A.2.1 Verify the position of the rod with inoperable DRPI indirectly by using movable incore detectors.

AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 hours AND Once per 31 EFPD thereafter AND 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after discovery of each unintended rod movement AND (continued) core power distribution information

Rod Position Indication 3.1.7 Farley Units 1 and 2 3.1.7-3 Amendment No. 214 (Unit 1)

Amendment No. 211 (Unit 2)

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C.

One or more DRPI inoperable in one or more groups and associated rod has been moved 24 steps in one direction since the last determination of the rods position.

C.1.1 Initiate action to verify the position of the rods with inoperable DRPIs indirectly by using movable incore detectors.

AND C.1.2 Complete rod position verification started in Required Action C.1.1.

OR C.2 Reduce THERMAL POWER to 50% RTP.

Immediately 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 hours D.

One or more demand position indicators per bank inoperable in one or more banks.

D.1.1 Verify by administrative means all DRPIs for the affected banks are OPERABLE.

AND D.1.2 Verify the most withdrawn rod and the least withdrawn rod of the affected banks are 12 steps apart.

OR D.2 Reduce THERMAL POWER to 50% RTP.

Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 hours E.

Required Action and associated Completion Time not met.

E.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months core power distribution information

FQ(Z) 3.2.1 Farley Units 1 and 2 3.2.1-3 Amendment No. 185 (Unit 1)

Amendment No. 180 (Unit 2)

SURVEILLANCE REQUIREMENTS

NOTE-------------------------------------------------------------

During power escalation at the beginning of each cycle, THERMAL POWER may be increased until an equilibrium power level has been achieved, at which a power distribution map is obtained SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify FQ(Z) is within steady state limit.

Once after each refueling prior to THERMAL POWER exceeding 75% RTP AND Once after achieving equilibrium conditions after exceeding, by t 20% RTP, the THERMAL POWER at which FQ(Z) was last verified AND In accordance with the Surveillance Frequency Control Program core information

FQ(Z) 3.2.1 Farley Units 1 and 2 3.2.1-4 Amendment No. 146 (Unit 1)

Amendment No. 137 (Unit 2)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.2

NOTE-------------------------------

If measurements indicate 1/4

º

¬

K(Z)

(Z)

F over maximum Q

has increased since the previous evaluation of FQ(Z):

a.

Increase FQ(Z) by the appropriate penalty factor specified in the COLR and reverify that this value is within the transient limits; or

b.

Repeat SR 3.2.1.2 once per 7 EFPD until either "a." above is met or two successive flux maps indicate 1/4

º

¬

K(Z)

(Z)

F over maximum Q

has not increased.

Verify FQ(Z) is within the transient limit.

Once after each refueling prior to THERMAL POWER exceeding 75%

RTP AND (continued) surveillances

RTS Instrumentation 3.3.1 Farley Units 1 and 2 3.3.1-9 Amendment No. 231 (Unit 1)

Amendment No. 228 (Unit 2)

SURVEILLANCE REQUIREMENTS

NOTE------------------------------------------------------------

Refer to Table 3.3.1-1 to determine which SRs apply for each RTS Function.

SURVEILLANCE FREQUENCY SR 3.3.1.1

NOTE--------------------------------

Not required to be performed for source range instrumentation until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after THERMAL POWER is < P-6.

Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.2

NOTE------------------------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is 15% RTP.

Compare results of calorimetric heat balance calculation to power range channel output. Adjust power range channel output if calorimetric heat balance calculation results exceed power range channel output by more than +2% RTP.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.3

NOTES------------------------------

1.

Not required to be performed until 7 days after THERMAL POWER is t 50% RTP.

2.

Performance of SR 3.3.1.9 satisfies this SR.

Compare results of the incore detector measurements to Nuclear Instrumentation System (NIS) AFD. Adjust NIS channel if absolute difference is 3%.

In accordance with the Surveillance Frequency Control Program FRUHSRZHUGLVWULEXWLRQLQIRUPDWLRQ

RTS Instrumentation 3.3.1 Farley Units 1 and 2 3.3.1-12 Amendment No. 185 (Unit 1)

Amendment No. 180 (Unit 2)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.9

NOTES------------------------------

1.

Neutron detectors are excluded from the calibration.

2.

Not required to be performed until 7 days after THERMAL POWER is t 50% RTP.

Calibrate excore channels to agree with incore detector measurements.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.10

NOTES------------------------------

1.

Neutron detectors are excluded from CHANNEL CALIBRATION.

2.

This Surveillance shall include verification that the time constants are adjusted to the prescribed values.

Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.11 Perform COT.

In accordance with the Surveillance Frequency Control Program AND

NOTE---------

Only required when not performed in accordance with the Surveillance Frequency Control Program.

(continued) core power distribution information

Joseph M. Farley Nuclear Plant - Units 1 and 2 Vogtle Electric Generating Plant - Units 1 and 2 Voluntary License Amendment Request to Use BEACON Power Distribution Monitoring System Farley Nuclear Plant 1&2 Revised TS Pages

Rod Position Indication 3.1.7 Farley Units 1 and 2 3.1.7-1 Amendment No. (Unit 1)

Amendment No. (Unit 2) 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Rod Position Indication LCO 3.1.7 The Digital Rod Position Indication (DRPI) System and the Demand Position Indication System shall be OPERABLE.

APPLICABILITY:

MODES 1 and 2.

ACTIONS

NOTE-----------------------------------------------------------

Separate Condition entry is allowed for each inoperable DRPI and each demand position indicator.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One DRPI per group inoperable in one or more groups.

A.1 Verify the position of the rod with inoperable DRPI indirectly by using core power distribution information.

OR A.2.1 Verify the position of the rod with inoperable DRPI indirectly by using core power distribution information.

AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 hours AND Once per 31 EFPD thereafter AND 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after discovery of each unintended rod movement AND (continued)

Rod Position Indication 3.1.7 Farley Units 1 and 2 3.1.7-3 Amendment No. (Unit 1)

Amendment No. (Unit 2)

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C.

One or more DRPI inoperable in one or more groups and associated rod has been moved 24 steps in one direction since the last determination of the rods position.

C.1.1 Initiate action to verify the position of the rods with inoperable DRPIs indirectly by using core power distribution information.

AND C.1.2 Complete rod position verification started in Required Action C.1.1.

OR C.2 Reduce THERMAL POWER to 50% RTP.

Immediately 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 hours D.

One or more demand position indicators per bank inoperable in one or more banks.

D.1.1 Verify by administrative means all DRPIs for the affected banks are OPERABLE.

AND D.1.2 Verify the most withdrawn rod and the least withdrawn rod of the affected banks are 12 steps apart.

OR D.2 Reduce THERMAL POWER to 50% RTP.

Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 hours E.

Required Action and associated Completion Time not met.

E.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

FQ(Z) 3.2.1 Farley Units 1 and 2 3.2.1-3 Amendment No. (Unit 1)

Amendment No. (Unit 2)

SURVEILLANCE REQUIREMENTS

NOTE-------------------------------------------------------------

During power escalation at the beginning of each cycle, THERMAL POWER may be increased until an equilibrium power level has been achieved, at which core power distribution information is obtained.

SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify FQ(Z) is within steady state limit.

Once after each refueling prior to THERMAL POWER exceeding 75% RTP AND Once after achieving equilibrium conditions after exceeding, by t 20% RTP, the THERMAL POWER at which FQ(Z) was last verified AND In accordance with the Surveillance Frequency Control Program

FQ(Z) 3.2.1 Farley Units 1 and 2 3.2.1-4 Amendment No. (Unit 1)

Amendment No. (Unit 2)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.2

NOTE-------------------------------

If 1/4

º

¬

K(Z)

(Z)

F over maximum Q

has increased since the previous evaluation of FQ(Z):

a.

Increase FQ(Z) by the appropriate penalty factor specified in the COLR and reverify that this value is within the transient limits; or b.

Repeat SR 3.2.1.2 once per 7 EFPD until either "a." above is met or two successive surveillances indicate 1/4

º

¬

K(Z)

(Z)

F over maximum Q

has not increased.

Verify FQ(Z) is within the transient limit.

Once after each refueling prior to THERMAL POWER exceeding 75%

RTP AND (continued)

RTS Instrumentation 3.3.1 Farley Units 1 and 2 3.3.1-9 Amendment No. (Unit 1)

Amendment No. (Unit 2)

SURVEILLANCE REQUIREMENTS

NOTE------------------------------------------------------------

Refer to Table 3.3.1-1 to determine which SRs apply for each RTS Function.

SURVEILLANCE FREQUENCY SR 3.3.1.1

NOTE--------------------------------

Not required to be performed for source range instrumentation until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after THERMAL POWER is < P-6.

Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.2

NOTE------------------------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is 15% RTP.

Compare results of calorimetric heat balance calculation to power range channel output. Adjust power range channel output if calorimetric heat balance calculation results exceed power range channel output by more than +2% RTP.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.3

NOTES------------------------------

Not required to be performed until 7 days after THERMAL POWER is t 50% RTP.

Performance of SR 3.3.1.9 satisfies this SR.

Compare results of core power distribution LQIRUPDWLRQ to Nuclear Instrumentation System (NIS) AFD. Adjust NIS channel if absolute difference is 3%.

In accordance with the Surveillance Frequency Control Program

RTS Instrumentation 3.3.1 Farley Units 1 and 2 3.3.1-12 Amendment No. (Unit 1)

Amendment No. (Unit 2)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.9

NOTES------------------------------

1.

Neutron detectors are excluded from the calibration.

2.

Not required to be performed until 7 days after THERMAL POWER is t 50% RTP.

Calibrate excore channels to agree with core power distribution information.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.10

NOTES------------------------------

1.

Neutron detectors are excluded from CHANNEL CALIBRATION.

2.

This Surveillance shall include verification that the time constants are adjusted to the prescribed values.

Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.11 Perform COT.

In accordance with the Surveillance Frequency Control Program AND

NOTE---------

Only required when not performed in accordance with the Surveillance Frequency Control Program.

(continued)

Joseph M. Farley Nuclear Plant - Units 1 and 2 Vogtle Electric Generating Plant - Units 1 and 2 Voluntary License Amendment Request to Use BEACON Power Distribution Monitoring System Vogtle Electrical Generating Plant 1&2 Marked-up TS Pages

Rod Position Indication 3.1.7 Vogtle Units 1 and 2 3.1.7-1 Amendment No. 193 (Unit 1)

Amendment No. 170 (Unit 2) 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Rod Position Indication LCO 3.1.7 The Digital Rod Position Indication (DRPI) System and the Demand Position Indication System shall be OPERABLE.

APPLICABILITY:

MODES 1 and 2.

ACTIONS

NOTE-----------------------------------------------------------

Separate Condition entry is allowed for each inoperable DRPI and each inoperable demand position indicator.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One DRPI per group inoperable in one or more groups.

A.1 Verify the position of the rod with inoperable DRPI indirectly by using movable incore detectors.

OR A.2.1 Verify the position of the rod with inoperable DRPI indirectly by using movable incore detectors.

AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 hours AND Once per 31 EFPD thereafter AND 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after discovery of each unintended rod movement AND (continued) core power distribution information

Rod Position Indication 3.1.7 Vogtle Units 1 and 2 3.1.7-3 Amendment No. 193 (Unit 1)

Amendment No. 176 (Unit 2)

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

One or more DRPI inoperable in one or more groups and associated rod has been moved 24 steps in one direction since the last determination of the rods position.

C.1 Verify the position of the rods with inoperable DRPIs by using movable incore detectors.

OR C.2 Reduce THERMAL POWER to d 50% RTP.

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 hours D.

One or more demand position indicators per bank inoperable in one or more banks.

D.1.1 Verify by administrative means all DRPIs for the affected banks are OPERABLE.

AND D.1.2 Verify the most withdrawn rod and the least withdrawn rod of the affected banks are 12 steps apart.

OR D.2 Reduce THERMAL POWER to 50% RTP.

Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 hours E.

Required Action and associated Completion Time not met.

E.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months core power distribution information

FQ(Z) 3.2.1 Vogtle Units 1 and 2 3.2.1-4 Amendment No. 96 (Unit 1)

Amendment No. 74 (Unit 2)

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.2.1.2

NOTE-------------------------------

If measurements indicate maximum over Z

>>1/4

º

<<¬

K(Z)

(Z)

FQ has increased since the previous evaluation of FQ(Z):

a.

Increase FQ(Z) by an appropriate penalty factor specified in the COLR and verify this value is within the transient limits; or

b.

Repeat SR 3.2.1.2 once per 7 EFPD until either

a. above is met or two successive flux maps indicate maximum over Z

>>1/4

º

<<¬

K(Z)

(Z)

FQ has not increased.

Verify FQ(Z) is within transient limit.

Once after each refueling after achieving equilibrium conditions at any power level exceeding 50% RTP AND (continued) surveillances

RTS Instrumentation 3.3.1 Vogtle Units 1 and 2 3.3.1-9 Amendment No. 194 (Unit 1)

Amendment No. 177 (Unit 2)

SURVEILLANCE REQUIREMENTS

NOTE-----------------------------------------------------------

Refer to Table 3.3.1-1 to determine which SRs apply for each RTS Function.

SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.2

NOTES-----------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER is t 15% RTP.

Compare results of calorimetric heat balance calculation to power range channel output. Adjust power range channel output if calorimetric heat balance calculation results exceed power range channel output by more than +2% RTP.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.3

NOTES------------------------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is t 50% RTP.

Compare results of the incore detector measurements to Nuclear Instrumentation System (NIS) AFD. Adjust NIS channel if absolute difference is t 3%.

In accordance with the Surveillance Frequency Control Program (continued) core power distribution LQIRUPDWLRQ

RTS Instrumentation 3.3.1 Vogtle Units 1 and 2 3.3.1-10 Amendment No. 158 (Unit 1)

Amendment No. 140 (Unit 2)

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.4

NOTE-------------------------------

This Surveillance must be performed on the reactor trip bypass breaker prior to placing the bypass breaker in service.

Perform TADOT.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.5 Perform ACTUATION LOGIC TEST.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.6

NOTES----------------------------

1.

Not required to be performed until 7 days after THERMAL POWER is t 75% RTP.

2.

Neutron detectors are excluded from CHANNEL CALIBRATION.

Calibrate excore channels to agree with incore detector measurements.

In accordance with the Surveillance Frequency Control Program (continued) core power distribution information

Joseph M. Farley Nuclear Plant - Units 1 and 2 Vogtle Electric Generating Plant - Units 1 and 2 Voluntary License Amendment Request to Use BEACON Power Distribution Monitoring System Vogtle Electrical Generating Plant 1&2 Revised TS Pages

Rod Position Indication 3.1.7 Vogtle Units 1 and 2 3.1.7-1 Amendment No. (Unit 1)

Amendment No. (Unit 2) 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Rod Position Indication LCO 3.1.7 The Digital Rod Position Indication (DRPI) System and the Demand Position Indication System shall be OPERABLE.

APPLICABILITY:

MODES 1 and 2.

ACTIONS

NOTE-----------------------------------------------------------

Separate Condition entry is allowed for each inoperable DRPI and each inoperable demand position indicator.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One DRPI per group inoperable in one or more groups.

A.1 Verify the position of the rod with inoperable DRPI indirectly by using core power distribution information.

OR A.2.1 Verify the position of the rod with inoperable DRPI indirectly by using core power distribution information.

AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 hours AND Once per 31 EFPD thereafter AND 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after discovery of each unintended rod movement AND (continued)

Rod Position Indication 3.1.7 Vogtle Units 1 and 2 3.1.7-3 Amendment No. (Unit 1)

Amendment No. (Unit 2)

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

One or more DRPI inoperable in one or more groups and associated rod has been moved 24 steps in one direction since the last determination of the rods position.

C.1 Verify the position of the rods with inoperable DRPIs by using core power distribution information.

OR C.2 Reduce THERMAL POWER to d 50% RTP.

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 hours D.

One or more demand position indicators per bank inoperable in one or more banks.

D.1.1 Verify by administrative means all DRPIs for the affected banks are OPERABLE.

AND D.1.2 Verify the most withdrawn rod and the least withdrawn rod of the affected banks are 12 steps apart.

OR D.2 Reduce THERMAL POWER to 50% RTP.

Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 hours E.

Required Action and associated Completion Time not met.

E.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

FQ(Z) 3.2.1 Vogtle Units 1 and 2 3.2.1-4 Amendment No. (Unit 1)

Amendment No. (Unit 2)

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.2.1.2

NOTE-------------------------------

If maximum over Z

>>1/4

º

<<¬

K(Z)

(Z)

FQ has increased since the previous evaluation of FQ(Z):

a.

Increase FQ(Z) by an appropriate penalty factor specified in the COLR and verify this value is within the transient limits; or b.

Repeat SR 3.2.1.2 once per 7 EFPD until either

a. above is met or two successive surveillances indicate maximum over Z

>>1/4

º

<<¬

K(Z)

(Z)

FQ has not increased.

Verify FQ(Z) is within transient limit.

Once after each refueling after achieving equilibrium conditions at any power level exceeding 50% RTP AND (continued)

RTS Instrumentation 3.3.1 Vogtle Units 1 and 2 3.3.1-9 Amendment No. (Unit 1)

Amendment No. (Unit 2)

SURVEILLANCE REQUIREMENTS

NOTE-----------------------------------------------------------

Refer to Table 3.3.1-1 to determine which SRs apply for each RTS Function.

SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.2

NOTES-----------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER is t 15% RTP.

Compare results of calorimetric heat balance calculation to power range channel output. Adjust power range channel output if calorimetric heat balance calculation results exceed power range channel output by more than +2% RTP.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.3

NOTES------------------------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is t 50% RTP.

Compare results of the core power distribution LQIRUPDWLRQ to Nuclear Instrumentation System (NIS) AFD. Adjust NIS channel if absolute difference is t 3%.

In accordance with the Surveillance Frequency Control Program (continued)

RTS Instrumentation 3.3.1 Vogtle Units 1 and 2 3.3.1-10 Amendment No. (Unit 1)

Amendment No. (Unit 2)

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.4

NOTE-------------------------------

This Surveillance must be performed on the reactor trip bypass breaker prior to placing the bypass breaker in service.

Perform TADOT.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.5 Perform ACTUATION LOGIC TEST.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.6

NOTES----------------------------

1.

Not required to be performed until 7 days after THERMAL POWER is t 75% RTP.

2.

Neutron detectors are excluded from CHANNEL CALIBRATION.

Calibrate excore channels to agree with core power distribution information.

In accordance with the Surveillance Frequency Control Program (continued)

Joseph M. Farley Nuclear Plant - Units 1 and 2 Vogtle Electric Generating Plant - Units 1 and 2 Voluntary License Amendment Request to Use BEACON Power Distribution Monitoring System Farley Nuclear Plant 1&2 Marked-up TS Bases Pages (Information only)

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6. WCAP-12472-P-A, BEACON Core Monitoring and Operations Support System, August 1994.

INSERT1FORFNPB3.2.2

BeforemakingcomparisonstotheFNHlimit,themeasuredvalueofFNHmustbemultipliedbya

measurementuncertaintyfactor.

IfthecorepowerdistributioninformationisobtainedwiththePowerDistributionMonitoringSystem

(PDMS),themeasuredvalueofFNHmustbemultipliedbythefactor[1.0+(UH/100)]whereUH

accountsforPDMSuncertaintydeterminedasdescribedinŽŽRef.6.

Ifthepowerdistributioninformationisobtainedfromafluxmapusingthemovableincoredetector

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Joseph M. Farley Nuclear Plant - Units 1 and 2 Vogtle Electric Generating Plant - Units 1 and 2 Voluntary License Amendment Request to Use BEACON Power Distribution Monitoring System Vogtle Electrical Generating Plant 1&2 Marked-up TS Bases Pages (Information only)

Rod Group Alignment Limits B 3.1.4 (continued)

Vogtle Units 1 and 2 B 3.1.4-4 Rev. 1-10/01

)

(FN H

N H

F N

H F

BASES APPLICABLE sufficient reactivity worth is held in the control rods to meet the SDM SAFETY ANALYSES requirement, with the maximum worth rod stuck fully withdrawn.

(continued)

Two types of analysis are performed in regard to static rod misalignment. With control banks at their insertion limits, one type of analysis considers the case when any one rod is completely inserted into the core. The second type of analysis considers the case of a completely withdrawn single rod from a bank inserted to its insertion limit. Satisfying limits on departure from nucleate boiling ratio in both of these cases bounds the situation when a rod is misaligned from its group by 12 steps.

Another type of misalignment occurs if one RCCA fails to insert upon a reactor trip and remains stuck fully withdrawn. This condition is assumed in the evaluation to determine that the required SDM is met with the maximum worth RCCA also fully withdrawn (Ref. 3).

The Required Actions in this LCO ensure that either deviations from the alignment limits will be corrected or that THERMAL POWER will be adjusted so that excessive local linear heat rates (LHRs) will not occur, and that the requirements on SDM and ejected rod worth are preserved.

Continued operation of the reactor with a misaligned control rod is allowed if the heat flux hot channel factor (FQ(Z)) and the nuclear enthalpy hot channel factor are verified to be within their limits in the COLR and the safety analysis is verified to remain valid.

When a control rod is misaligned, the assumptions that are used to determine the rod insertion limits, AFD limits, and quadrant power tilt limits are not preserved. Therefore, the limits may not preserve the design peaking factors, and FQ(Z) and must be verified directly by incore mapping. Bases Section 3.2 (Power Distribution Limits) contains more complete discussions of the relation of FQ(Z) and to the operating limits.

Shutdown and control rod OPERABILITY and alignment are directly related to power distributions and SDM, which are initial conditions assumed in safety analyses. Therefore they satisfy Criterion 2 of 10 CFR 50.36 (c)(2)(ii).

using core power distribution information

Rod Group Alignment Limits B 3.1.4 (continued)

Vogtle Units 1 and 2 B 3.1.4-5 Rev. 1-8/03 BASES (continued)

LCO The limits on shutdown or control rod alignments ensure that the assumptions in the safety analysis will remain valid. The requirements on OPERABILITY ensure that upon reactor trip, the assumed reactivity will be available and will be inserted. The OPERABILITY requirements (i.e., trippability) are separate from the alignment requirements which ensure that the RCCAs and banks maintain the correct power distribution and rod alignment.

The rod OPERABILITY (i.e., trippability) requirement is satisfied provided that the rod will fully insert in the required rod drop time assumed in the safety analyses. Rod control malfunctions that result in the inability to move a rod (e.g., rod lift coil failures), but that do not impact trippability, do not result in rod inoperability. However, where rod(s) are not moving, the rod(s) must be considered untrippable unless there is verification that a rod control system failure is preventing rod motion. If the rod control system is demanding motion properly and no motion occurs, the rod is considered untrippable (i.e.,

inoperable).

The requirement to maintain the rod alignment to within plus or minus 12 steps of their group step counter demand position is conservative. The safety analysis assumes a total misalignment from fully withdrawn to fully inserted. When required, movable incore detectors may be used to determine rod position and verify the rod alignment requirement of this LCO is met.

Failure to meet the requirements of this LCO may produce unacceptable power peaking factors and LHRs, or unacceptable SDMs, all of which may constitute initial conditions inconsistent with the safety analysis.

APPLICABILITY The requirements on RCCA OPERABILITY and alignment are applicable in MODES 1 and 2 because these are the only MODES in which a self-sustaining chain reaction (Keff t 1) occurs, and the OPERABILITY (i.e., trippability) and alignment of rods have the potential to affect the safety of the plant. In MODES 3, 4, 5, and 6, the alignment limits do not apply because the control rods are fully inserted and the reactor is shut down, with no self-sustaining chain reaction. In the shutdown MODES, the OPERABILITY of the shutdown and control rods has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the RCS. See LCO 3.1.1, "SHUTDOWN MARGIN (SDM)," for SDM in MODES 3, 4, and 5 and LCO 3.9.1, "Boron Concentration," for boron concentration requirements during refueling.

core power distribution information

Rod Group Alignment Limits B 3.1.4 (continued)

Vogtle Units 1 and 2 B 3.1.4-8 REVISION 51 BASES ACTIONS B.2, B.3, B.4, and B.5 (continued)

A Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient to ensure this requirement continues to be met.

Verifying that FQ(Z), as approximated by the steady state and transient FQ(Z), and N

H F

are within the required limits ensures that current operation at 75% RTP with a rod misaligned is not resulting in power distributions that may invalidate safety analysis assumptions at full power. The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months allows sufficient time to obtain flux maps of the core power distribution using the incore flux mapping system and to calculate FQ(Z) and N

H F

Once current conditions have been verified acceptable, time is available to perform evaluations of accident analysis to determine that core limits will not be exceeded during a Design Basis Event for the duration of operation under these conditions. A Completion Time of 5 days is sufficient time to obtain the required input data and to perform the analysis.

The following accident analyses require reevaluation for continued operation with a misaligned rod.

RCCA Insertion Characteristics RCCA Misalignment Decrease in Reactor Coolant Inventory x

Inadvertent Opening of a Pressurizer Safety or Relief Valve x

Break in Instrument Line or Other Lines From Reactor Coolant Pressure Boundary That Penetrates Containment x

Loss-of-Coolant-Accidents Increase in Heat Removal by the Secondary System (Steam System Piping Rupture) Spectrum of RCCA Ejection Accidents.

C.1 When Required Actions cannot be completed within their Completion Time, the unit must be brought to a MODE or Condition in which the LCO requirements are not applicable.

To achieve this status, the unit must be brought to at least information core power distribution information

Rod Position Indication B 3.1.7 (continued)

Vogtle Units 1 and 2 B 3.1.7-4 REVISION 51 BASES APPLICABILITY in which power is generated, and the OPERABILITY and (continued) alignment of rods have the potential to affect the safety of the plant.

In the shutdown MODES, the OPERABILITY of the shutdown and control banks has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the Reactor Coolant System.

ACTIONS The ACTIONS table is modified by a Note indicating that a separate Condition entry is allowed for each inoperable rod position indicator and each inoperable demand position indicator. This is acceptable because the Required Actions for each Condition provide appropriate compensatory actions for each inoperable position indicator.

A.1, A.2.1, and A.2.2 When one DRPI channel per group in one or more groups fails, the position of the rod may still be determined indirectly by use of the movable incore detectors. The Required Action may also be satisfied by ensuring at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months that FQ satisfies LCO 3.2.1, FdeltaH satisfies LCO 3.2.2, and SHUTDOWN MARGIN is within the limits provided in the COLR, provided the non-indicating rods have not been moved. Based on experience, normal power operation does not require excessive movement of banks. If a bank has been significantly moved, the Required Action of C.1 or C.2 below is required. Therefore, verification of RCCA position within the Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is adequate for allowing continued full power operation, since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small.

Required Action A.1 requires verification of the position of a rod with an inoperable DRPI once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months which may put excessive wear and tear on the movable incore detector system. Required Action A.2.1 provides an alternative. Required Action A.2.1 requires

verification of rod position using the movable incore detectors every

31 EFPD, which coincides with the normal use of the system to verify

core power distribution.

Required Action A.2.1 includes six distinct requirements for verification of the position of rods associated with an inoperable DRPI using the movable incore detectors:

using core power distribution information by using core power distribution information frequency core power distribution information

Rod Position Indication B 3.1.7 (continued)

Vogtle Units 1 and 2 B 3.1.7-6 REVISION 51 BASES ACTIONS B.1 and B.4 (continued)

When more than one DRPI per group in one or more groups fail, additional actions are necessary. Placing the Rod Control System in manual assures unplanned rod motion will not occur.

The immediate Completion Time for placing the Rod Control System in manual reflects the urgency with which unplanned rod motion must be prevented while in this Condition.

The inoperable DRPIs must be restored such that a maximum of one DRPI per group is inoperable, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time provides sufficient time to troubleshoot and restore the DRPI system to operation while avoiding the plant challenges associated with a shutdown without full rod position indication.

Based on operating experience, normal power operation does not require excessive rod movement. If one or more rods has been significantly moved, the Required Action of C.1 or C.2 below is required.

C.1 and C.2 With one DRPI inoperable in one or more groups and the affected groups have moved greater than 24 steps in one direction since the last determination of rod position, additional actions are needed to verify the position of rods within inoperable DRPI. Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , the position of the rods with inoperable position indication must be determined using the movable incore detectors to verify that these rods are still properly positioned, relative to their group positions.

Either the rod positions must be determined within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or THERMAL POWER must be reduced to d 50% RTP within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to avoid undesirable power distributions that could result from continued operation at > 50% RTP, if one or more rods are misaligned by more than 24 steps. The allowed Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months provides an acceptable period of time to verify the rod positions using the movable incore detectors.

core power distribution information

FQ(Z)

B 3.2.1 (continued)

Vogtle Units 1 and 2 B 3.2.1-1 Revision No. 0 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 Heat Flux Hot Channel Factor (FQ(Z)) (FQ Methodology)

BASES BACKGROUND The purpose of the limits on the values of FQ(Z) is to limit the local (i.e., pellet) peak power density. The value of FQ(Z) varies along the axial height (Z) of the core.

FQ(Z) is defined as the maximum local fuel rod linear power density divided by the average fuel rod linear power density, assuming nominal fuel pellet and fuel rod dimensions. Therefore, FQ(Z) is a measure of the peak fuel pellet power within the reactor core.

During power operation, the global power distribution is limited by LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)," which are directly and continuously measured process variables. These LCOs, along with LCO 3.1.6, "Control Bank Insertion Limits," maintain the core within power distribution limits on a continuous basis.

FQ(Z) varies with fuel loading patterns, control bank insertion, fuel burnup, and axial power distribution.

FQ(Z) is measured periodically using the incore detector system.

These measurements are generally taken with the core at or near steady state conditions.

Using the measured three dimensional power distributions, it is possible to derive a measured value for FQ(Z). However, because this value represents a steady state condition, it does not include the variations in the value of FQ(Z) that are present during nonequilibrium situations.

To account for these possible variations, the steady state value of FQ(Z) is adjusted by an elevation dependent factor that accounts for the calculated worst case transient conditions.

Core monitoring and control under non-steady state conditions are accomplished by operating the core within the limits of the appropriate LCOs, including the limits on AFD, QPTR, and control rod insertion.

FQ(Z)

B 3.2.1 (continued)

Vogtle Units 1 and 2 B 3.2.1-3 REVISION 15 BASES LCO (continued) where: F RTP Q

is the FQ(Z) limit at RTP provided in the COLR, K(Z) is the normalized FQ(Z) as a function of core height provided in the COLR, and RTP POWER THERMAL

=

P For this facility, the actual values of F RTP Q

and K(Z) are given in the COLR; however, F RTP Q

is normally a number on the order of 2.50, and K(Z) is a function that looks like the one provided in Figure B 3.2.1-1.

An FQ(Z) evaluation requires obtaining an incore flux map in MODE 1.

From the incore flux map results we obtain the measured value (F

M Q (Z)) of FQ(Z). Then, when using 44 detector thimbles FQ(Z) = F M

Q (Z) X 1.0815 where 1.0815 is a factor that accounts for fuel manufacturing tolerances (3%) and flux map measurement uncertainty (5%), or when using 29 and < 44 thimbles:

FQ(Z) = F M

Q (Z) x 1.03 x [1.05 + [2.0 {3-T/(14.5)}]/100],

where 1.03 accounts for fuel manufacturing tolerances with a more conservative flux map measurement uncertainty factor to account for the fewer detector thimbles available, and T is the number of thimbles being used. A bounding measurement uncertainty of 7.0 %, which is based on 29 thimbles, can be used for 29 and < 44 detector thimbles, if desired. FQ(Z) evaluations for comparison to the steady

0.5 P

for Z

K P

F Z

F RTP Q

Q d

0.5 P

for Z

K 0.5 F

Z F

RTP Q

Q d

d Insert 1 for B3.2.1 limits

INSERT1FORVEGPB3.2.1

AnFQ(Z)evaluationrequiresobtainingcorepowerdistributioninformationinMODE1.Fromthecore

powerdistributioninformation,themeasuredvalue(FMQ(Z))ofFQ(Z)isobtained.

IfthecorepowerdistributioninformationisobtainedwiththePowerDistributionMonitoringSystem

(PDMS),then:

FQ(Z)=FMQ(Z)X1.03X[1.0+(UQ/100)]

where1.03accountsforfuelmanufacturingtolerancesandUQaccountsforPDMSuncertaintyas

describedinŽŽRef.8.

Ifthecorepowerdistributioninformationisobtainedfromafluxmapusingthemovableincore

detectorsystemwith44thimbles,then:

FQ(Z)

B 3.2.1 (continued)

Vogtle Units 1 and 2 B 3.2.1-4 REVISION 55 BASES LCO state limits are applicable in all axial core regions, i.e., from 0 to 100%

(continued) inclusive.

Because flux maps are taken in steady state conditions, the variations in power distribution resulting from normal operational maneuvers are not present in the flux map data. These variations are, however, conservatively calculated by considering a wide range of unit maneuvers in normal operation. The maximum peaking factor increase over steady state values, calculated as a function of core elevation, Z, is called W(Z).

The W(Z) curve is provided in the COLR for discrete core elevations.

Provided the reload analysis determines that the limiting or peak FQ(Z) is not located within the upper and lower 8% of the core, FQ(Z) evaluations for comparison to the transient limits are not required for the following axial core regions, measured in percent of core height:

a.

Lower core region, from 0 to 8% inclusive; and b.

Upper core region, from 92 to 100% inclusive.

The top and bottom 8% of the core are typically excluded from the evaluation because of the low probability that these regions would be more limiting in the safety analyses and because of the difficulty of making a precise measurement in these regions. It may be necessary to exclude a smaller region or to exclude no region from the evaluation if the location of the limiting FQ(Z) is in the lower or upper 8% of the core based on the reload analysis.

To account for power distribution transients encountered during normal operation, the transient limits for FQ(Z) are established utilizing the cycle dependent function W(Z). To ensure that FQ(Z) will not become excessively high if a normal operational transient occurs, FQ(Z) shall be limited by the following relationships which define the transient limits:

0.5 P

for 0.5W(Z)

Z K

F Z

F 0.5 P

for PW(Z)

Z K

F Z

F RTP Q

Q RTP Q

Q d

d d

The FQ(Z) limits define limiting values for core power peaking that precludes peak cladding temperatures above 2200qF during either a large or small break LOCA.

core power distribution information is

FQ(Z)

B 3.2.1 (continued)

Vogtle Units 1 and 2 B 3.2.1-8 REVISION 17 BASES SURVEILLANCE SR 3.2.1.1 REQUIREMENTS (continued)

Verification that FQ(Z) is within its specified limits involves increasing FMQ(Z) to allow for manufacturing tolerance and measurement uncertainties in order to obtain FQ(Z). Specifically, FMQ(Z) is the measured value of FQ(Z) obtained from incore flux map results. When using 44 detector thimbles, FQ(Z) = FMQ (Z) X 1.0815 (Ref. 4), and when using 29 and < 44 thimbles, FQ(Z) =

FMQ (Z) x 1.03 x [1.05 + [2.0 {3-T/(14.5)}]/100], where T = the number of detector thimbles used (Ref. 6). A bounding measurement uncertainty of 7.0 %, which is based on 29 thimbles, can be used for 29 and < 44 detector thimbles, if desired. During the initial startup after a refueling outage up to and including performance of the first flux map at 100% RTP, t 44 detector thimbles, with t 2 detector thimbles per core quadrant as identified in TRM Figure 13.3.1-1 are required. This Note does not have to be met for Vogtle Unit 1, Cycle 17 based on the successful performance of the flux map at 30% RTP. FQ(Z) is then compared to its steady state and transient limits specified in the COLR.

Performing this Surveillance in MODE 1 after exceeding 50% RTP following refueling ensures that the FQ(Z) limit is met when RTP is achieved, because peaking factors generally decrease as power level is increased. In addition, at power levels above 50% RTP, equilibrium Xenon conditions approach those more closely at RTP. Therefore, performing the Surveillance at a power level above 50% RTP ensures a more accurate measurement of FQ(Z).

If THERMAL POWER has been increased by t 20% RTP since the last determination of FQ(Z), another evaluation of this factor is required after achieving equilibrium conditions at this higher power level (to ensure that FQ(Z) values are being reduced sufficiently with power increase to stay within the LCO limits).

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.2.1.2 This surveillance determines if FQ(Z) will remain within its limit during a normal operational transient. If FQ(Z) is determined to exceed the transient limit, Action B.1 requires that the AFD limit be reduced 1% for each 1% FQ(Z) exceeds the transient limit. This Insert 2 for B3.2.1

INSERT2FORVEGPB3.2.1

ThissurveillancedeterminesifFQ(Z)willremainwithinitslimitduringsteadystateoperation.

TRMTR13.3.1specifiesthemovableincoredetectorsystemFUNCTIONALrequirementsifafluxmapis

usedforperformingthissurveillance.

FQ(Z)

B 3.2.1 (continued)

Vogtle Units 1 and 2 B 3.2.1-9 REVISION 15 BASES SURVEILLANCE SR 3.2.1.2 (continued)

REQUIREMENTS will ensure that FQ(Z) will not exceed the transient limit during a normal operational transient within the reduced AFD limit.

Demonstrating that FQ(Z) is within the transient limit or reducing the AFD limit if the transient FQ(Z) limit was initially exceeded, only ensures that the transient FQ(Z) limit will not be exceeded at the time FQ(Z) was evaluated. This does not ensure that the limit will not be exceeded during the following surveillance interval. Both the steady state and transient FQ(Z) change as a function of core burnup.

If the two most recent FQ(Z) evaluations show an increase in the quantity maximum over z

>>1/4

º

<<¬

k(Z)

(Z)

FQ it is not guaranteed that FQ(Z) will remain within the transient limit during the following surveillance interval. SR 3.2.1.2 is modified by a Note to determine if there is sufficient margin to the transient FQ(Z) limit to ensure that the limit will not be exceeded during the following surveillance interval. This is accomplished by increasing FQ(Z) by a penalty specified in the COLR and comparing this value to the transient FQ(Z) limit. If there is insufficient margin, i.e., this value exceeds the limit, SR 3.2.1.2 must be repeated once per 7 EFPD until either FQ(Z) increased by the penalty factor is within the transient limit or, two successive (i.e., subsequent consecutive) flux maps indicate maximum over z

>>1/4

º

<<¬

k(Z)

(Z)

FQ has not increased.

Performing the Surveillance in MODE 1 after exceeding 50% RTP following refueling ensures that the FQ(Z) limits are met when RTP is achieved, because peaking factors are generally decreased as power level is increased. In addition, at power levels above 50% RTP, equilibrium Xenon conditions approach more closely those at RTP.

Therefore, performing the Surveillance at a power level above 50%

RTP ensures a more accurate measurement of FQ(Z).

FQ(Z) is verified at power levels t 20% RTP above the THERMAL POWER of its last verification, after achieving surveillances

FQ(Z)

B 3.2.1 Vogtle Units 1 and 2 B 3.2.1-10 REVISION 15 BASES SURVEILLANCE SR 3.2.1.2 (continued)

REQUIREMENTS equilibrium conditions to ensure that FQ(Z) is within its limit at higher power levels.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1.

10 CFR 50.46, 1974.

2.

FSAR Subsection 15.4.8.

3.

10 CFR 50, Appendix A, GDC 26.

4.

WCAP-7308-L-P-A, "Evaluation of Nuclear Hot Channel Factor Uncertainties," June 1988.

5.

WCAP-10216-P-A, Revision 1A, "Relaxation of Constant Axial Offset Control FQ Surveillance Technical Specification,"

February 1994.

6.

GP-18735, Evaluation of a Reduction in the Required Number of Movable Incore Detector Thimbles, January 31, 2011.

7.

GP-18767, Southern Nuclear Operating Company, Vogtle Electric Generating Plant Units 1 and 2, Cycle 17 Movable Incore Detector Thimble Evaluation, April 4, 2011.

8. WCAP-12472-P-A, BEACON Core Monitoring and Operations Support System, August 1994.

N H

F B 3.2.2 (continued)

Vogtle Units 1 and 2 B 3.2.2-1 Revision No. 0 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (

N H

F

)

BASES BACKGROUND The purpose of this LCO is to establish limits on the power density at any point in the core so that the fuel design criteria are not exceeded and the accident analysis assumptions remain valid. The design limits on local (pellet) and integrated fuel rod peak power density are expressed in terms of hot channel factors. Control of the core power distribution with respect to these factors ensures that local conditions in the fuel rods and coolant channels do not challenge fuel design limits at any location in the core during either normal operation or a postulated accident analyzed in the safety analyses.

is defined as the ratio of the integral of the linear power along the fuel rod with the highest integrated power to the average integrated fuel rod power. Therefore, is a measure of the maximum total power produced in a fuel rod.

is sensitive to fuel loading patterns, bank insertion, and fuel burnup. typically increases with control bank insertion and typically decreases with fuel burnup.

is not directly measurable but is inferred from a power distribution map obtained with the movable incore detector system. Specifically, the results of the three dimensional power distribution map are analyzed by a computer to determine. This factor is calculated at least every 31 EFPD. However, during power operation, the global power distribution is monitored by LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)," which address directly and continuously measured process variables.

The COLR provides peaking factor limits that ensure that the design criterion for the departure from nucleate boiling (DNB) is met for normal operation, operational transients, and any transient condition arising from events of moderate frequency. All DNB limited transient events are assumed to begin with an value that satisfies the LCO requirements.

N H

F N

H F

N H

F N

H F

N H

F N

H F

N H

F core information

N H

F B 3.2.2 (continued)

Vogtle Units 1 and 2 B 3.2.2-3 Rev. 1-10/01 BASES APPLICABLE transients that may be DNB limited are assumed to begin SAFETY ANALYSES with an initial as a function of power level defined by the (continued)

COLR limit equation The LOCA safety analysis indirectly models as an input parameter. The Nuclear Heat Flux Hot Channel Factor (FQ(Z)) and the axial peaking factors are inserted directly into the LOCA safety analyses that verify the acceptability of the resulting peak cladding temperature (Ref. 3).

The fuel is protected in part by Technical Specifications, which ensure that the initial conditions assumed in the safety and accident analyses remain valid. The following LCOs ensure this: LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)," LCO 3.1.7, "Control Bank Insertion Limits,"

LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor ( )," and LCO 3.2.1, "Heat Flux Hot Channel Factor (FQ(Z))."

and FQ(Z) are measured periodically using the movable incore detector system. Measurements are generally taken with the core at, or near, steady state conditions. Core monitoring and control under transient conditions (Condition 1 events) are accomplished by operating the core within the limits of the LCOs on AFD, QPTR, and Bank Insertion Limits.

satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).

LCO shall be maintained within the limits of the relationship provided in the COLR.

The limit identifies the coolant flow channel with the maximum enthalpy rise. This channel has the least heat removal capability and thus the highest probability for DNB.

The limiting value of, described by the equation contained in the COLR, is the design radial peaking factor used in the unit safety analyses.

N H

F N

H F

N H

F N

H F

N H

F N

H F

N H

F N

H F

core power distribution information

N H

F B 3.2.2 (continued)

Vogtle Units 1 and 2 B 3.2.2-5 REVISION 40 BASES ACTIONS A.1.1 (continued)

However, if power is reduced below 50% RTP, Required Action A.3 requires that another determination of must be done prior to exceeding 50% RTP, prior to exceeding 75% RTP, and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reaching or exceeding 95% RTP. In addition, Required Action A.2 is performed if power ascension is delayed past 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

A.1.2.1 and A.1.2.2 If the value of is not restored to within its specified limit either by adjusting a misaligned rod or by reducing THERMAL POWER, the alternative option is to reduce THERMAL POWER to < 50% RTP in accordance with Required Action A.1.2.1 and reduce the Power Range Neutron Flux High to d 55% RTP in accordance with Required Action A.1.2.2. Reducing RTP to < 50% RTP increases the DNB margin and does not likely cause the DNBR limit to be violated in steady state operation. The reduction in trip setpoints ensures that continuing operation remains at an acceptable low power level with adequate DNBR margin. The allowed Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for Required Action A.1.2.1 is consistent with those allowed for in Required Action A.1.1 and provides an acceptable time to reach the required power level from full power operation without allowing the plant to remain in an unacceptable condition for an extended period of time. The Completion Times of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for Required Actions A.1.1 and A.1.2.1 are not additive.

The allowed Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to reset the trip setpoints per Required Action A.1.2.2 recognizes that, once power is reduced, the safety analysis assumptions are satisfied and there is no urgent need to reduce the trip setpoints. This is a sensitive operation that may inadvertently trip the Reactor Protection System.

A.2 Once corrective action has been taken in accordance with Required Action A.1.1 or A.1.2.1, an incore flux map (SR 3.2.2.1) must be obtained and the measured value of N

H F

N H

F N

H F

performed

N H

F B 3.2.2 (continued)

Vogtle Units 1 and 2 B 3.2.2-6 Revision No. 0 BASES ACTIONS A.2 (continued) verified not to exceed the allowed limit. The unit is provided 20 additional hours to perform this task over and above the 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allowed by either Action A.1.1 or Action A.1.2.1. The Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is acceptable because of the low probability of having a DNB limiting event within this 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, and, in the event that power was reduced, the increase in DNB margin which is obtained at lower power levels. Additionally, operating experience has indicated that this Completion Time is sufficient to obtain the incore flux map, perform the required calculations, and evaluate.

A.3 Verification that is within its specified limits after an out of limit occurrence ensures that the cause that led to the exceeding its limit is corrected, and that subsequent operation proceeds within the LCO limit. This Action demonstrates that the limit is within the LCO limits prior to exceeding 50% RTP, again prior to exceeding 75% RTP, and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is t 95% RTP.

This Required Action is modified by a Note that clarifies that it is only applicable to the extent that THERMAL POWER has been reduced to comply with Required Actions A.1.1 or A.1.2.1. For example, if THERMAL POWER was reduced to less than 50%, SR 3.2.2.1 must be performed prior to THERMAL POWER exceeding 50%, 75%, and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reaching 95% RTP. If however, THERMAL POWER was only reduced to 70% RTP, then SR 3.2.2.1 must be performed prior to exceeding 75% RTP and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reaching 95% RTP. This course of action will provide assurance that has been restored to limits.

B.1 When Required Actions A.1.1 through A.3 cannot be completed within their required Completion Times, the plant must be placed in a mode in which the LCO requirements are not applicable. This is done by placing the plant in at least N

H F

N H

F N

H F

N H

F N

H F

core power distribution information

N H

F B 3.2.2 Vogtle Units 1 and 2 B 3.2.2-7 REVISION 17 BASES ACTIONS B.1 (continued)

MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience regarding the time required to reach MODE 2 from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.2.1 REQUIREMENTS The value of F N

H

' is determined by using the movable incore detector system to obtain a flux distribution map. A data reduction computer program then calculates the maximum value of F N

H

' from the measured flux distributions. Before making comparisons to the F

N H

' limit, the measured value of F N

H

' must be multiplied by a measurement uncertainty factor. When using 44 detector thimbles, the measured value of F N

H

' must be multiplied by 1.04. When using 29 and < 44 detector thimbles, the measured value of F N

H

' must be multiplied by 1.04 + [2.0 {3-T/(14.5)}]/100, where T = the number of detector thimbles used. A bounding measurement uncertainty of 6.0 %, which is based on 29 thimbles, can be used for 29 and < 44 detector thimbles, if desired. During the initial startup after a refueling outage up to and including performance of the first flux map at 100%

RTP, t 44 detector thimbles, with t 2 detector thimbles per core quadrant as identified in TRM Figure 13.3.1-1 are required. This Note does not have to be met for Vogtle Unit 1, Cycle 17 based on the successful performance of the flux map at 30% RTP.

After each refueling, F N

H

' must be determined in MODE 1 prior to exceeding 75% RTP. This requirement ensures that F N

H

' limits are met at the beginning of each fuel cycle.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1.

FSAR Subsection 15.4.8.

2.

10 CFR 50, Appendix A, GDC 26.

3.

10 CFR 50.46.

obtaining core power distribution information core power information Insert 1 for B3.2.2

4. WCAP-12472-P-A, BEACON Core Monitoring and Operations Support System, August 1994.

Insert1forVEGPB3.2.2

BeforemakingcomparisonstotheFNHlimit,themeasuredvalueofFNHmustbemultipliedbya

measurementuncertaintyfactor.

IfthecorepowerdistributioninformationisobtainedwiththePowerDistributionMonitoringSystem

(PDMS),themeasuredvalueofFNDHmustbemultipliedbythefactor[1.0+(UH/1000]whereUH

accountsforPDMSuncertaintydeterminedasdescribedinŽŽRef.4.

Ifthepowerdistributioninformationisobtainedfromafluxmapusingthemovableincoredetector

systemwith44thimbles,

QPTR B 3.2.4 (continued)

Vogtle Units 1 and 2 B 3.2.4-2 Rev. 1-10/01 BASES APPLICABLE established to preclude core power distributions that exceed SAFETY ANALYSES the safety analyses limits.

(continued)

The QPTR limits ensure that and FQ(Z) remain below their limiting values by preventing an undetected change in the radial power distribution.

In MODE 1, the and FQ(Z) limits must be maintained to preclude core power distributions from exceeding design limits assumed in the safety analyses.

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

LCO The QPTR limit of 1.02, at which corrective action is required, provides a margin of protection for both the DNB ratio and linear heat generation rate contributing to excessive power peaks resulting from X-Y plane power tilts. The value of 1.02 was selected because the purpose of the LCO is to limit, or require detection of, gross changes in core power distribution between monthly incore flux maps. In addition, it is the lowest value of quadrant power tilt that can be used for an alarm without spurious actuation.

APPLICABILITY The QPTR limit must be maintained in MODE 1 with THERMAL POWER > 50% RTP to prevent core power distributions from exceeding the design limits.

Applicability in MODE 1 d 50% RTP and in other MODES is not required because there is either insufficient stored energy in the fuel or insufficient energy being transferred to the reactor coolant to require the implementation of a QPTR limit on the distribution of core power. The QPTR limit in these conditions is, therefore, not important. Note that the FN'H and FQ(Z) LCOs still apply, but allow progressively higher peaking factors at 50% RTP or lower.

N H

F N

H F

GHWHUPLQDWLRQRIFRUHSRZHU

GLVWULEXWLRQLQIRUPDWLRQ

QPTR B 3.2.4 (continued)

Vogtle Units 1 and 2 B 3.2.4-3 REVISION 40 BASES (continued)

ACTIONS A.1 With the QPTR exceeding its limit, limiting THERMAL POWER to t 3% below RTP for each 1% by which the QPTR exceeds 1.00 is a conservative tradeoff of total core power with peak linear power. The Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months allows sufficient time to identify the cause and correct the tilt. Note that the power reduction itself may cause a change in the tilted condition.

A.2.1 and A.2.2 Because the QPTR alarm is already in its alarmed state, any additional changes in the QPTR are detected by requiring a check of the QPTR once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . If the QPTR continues to increase, THERMAL POWER has to be reduced accordingly within the following 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . A Note clarifies that the Completion Time of Required Action A.2.2 begins after Required Action A.2.1 is complete.

These Completion Times are sufficient because any additional change in QPTR would be relatively slow.

A.3 The peaking factors N

H F

and FQ(Z), as approximated by the steady state and transient FQ(Z), are of primary importance in ensuring that the power distribution remains consistent with the initial conditions used in the safety analyses. Performing SRs on N

H F

and FQ(Z) within the Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after H

achieving equilibrium conditions with THERMAL POWER limited by Required Action A.1 or A.2.2 ensures that these primary indicators of power distribution are within their respective limits.

The above Completion Time takes into consideration the rate at which peaking factors are likely to change, and the time required to stabilize the plant and perform a flux map. If these peaking factors are not within their limits, the Required Actions of these Surveillances provide an appropriate response for the abnormal condition. If the QPTR remains above its specified limit, the peaking factor surveillances are required each 7 days thereafter to evaluate FN and REWDLQFRUHpower distribution LQIRUPDWLRQ

QPTR B 3.2.4 (continued)

Vogtle Units 1 and 2 B 3.2.4-4 Revision No. 0 BASES ACTIONS A.3 (continued)

FQ(Z) with changes in power distribution. Relatively small changes are expected due to either burnup and xenon redistribution or correction of the cause for exceeding the QPTR limit.

A.4 When the QPTR exceeds its limit, it does not necessarily mean a safety concern exists. It does mean that there is an indication of a change in the radial power distribution that requires an investigation and evaluation that is accomplished by examining the power distribution using the incore detectors. Specifically, the core peaking must be evaluated because they are the factors that best characterize the core power distribution. This reevaluation is required to ensure that, for the duration of operation in accordance with Condition A of this LCO, before increasing THERMAL POWER to above the limit of Required Action A.1 and A.2.2, the reactor core conditions (peaking factors) are consistent with the assumptions in the safety analyses and will remain so after the return to RTP.

However, if prior to performing SR 3.2.1.1 and SR 3.2.2.1, QPTR is restored to within the limit, either due to prior completion of Required Actions or due to core performance characteristics that result in the QPTR out-of-limit condition correcting itself, Required Action A.3 and any other required actions would no longer apply because Condition A of LCO 3.2.4 would be exited in accordance with LCO 3.0.2 due to restoration of full compliance with LCO 3.2.4.

If it is determined that a sustained change in the radial power distribution has occurred, and Required Action A.3 has been completed with satisfactory results, an increase in THERMAL POWER above the limit of Required Action A.1 may be appropriate.

The necessary sequence of Required Actions, beginning with Required Action A.4, would be as follows prior to increasing THERMAL POWER above the limit of Required Action A.1 and A.2.2.

core

QPTR B 3.2.4 (continued)

Vogtle Units 1 and 2 B 3.2.4-7 REVISION 40 BASES SURVEILLANCE SR 3.2.4.1 (continued)

REQUIREMENTS This Surveillance verifies that the QPTR, as indicated by the Nuclear Instrumentation System (NIS) excore channels, is within its limits.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. Valid inputs to the detector current comparator from the upper and lower sections from 3 or 4 power range channels are required for the QPTR alarm to be OPERABLE.

For those causes of QPTR that occur quickly (e.g., a dropped rod),

there typically are other indications of abnormality that prompt a verification of core power tilt.

SR 3.2.4.2 This Surveillance is modified by a Note, which states that the surveillance is only required to be performed if input to QPTR from one or more Power Range Neutron Flux channels is inoperable with THERMAL POWER t75% RTP.

With an NIS power range channel inoperable, tilt monitoring for a portion of the reactor core becomes degraded. Large tilts are likely detected with the remaining channels, but the capability for detection of small power tilts in some quadrants is decreased. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

When one power range channel is inoperable, the incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the indicated QPTR. The incore detector monitoring is performed with a full incore flux map or two sets of four thimble locations with quarter core symmetry. The two sets of four symmetric thimbles is a set of eight unique detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, and N-8.

either the Power Distribution Monitoring System (PDMS) or movable

QPTR B 3.2.4 Vogtle Units 1 and 2 B 3.2.4-8 Revision No. 0 BASES SURVEILLANCE SR 3.2.4.2 (continued)

REQUIREMENTS The flux map can be used to generate power tilt. This can be compared to a reference tilt, from the most recent calibration flux map.

Therefore, the incore detectors can be used to confirm the accuracy of the QPTR as indicated by the excore detectors.

REFERENCES 1.

10 CFR 50.46.

2.

FSAR Subsection 15.4.8.

3.

10 CFR 50, Appendix A, GDC 26.

or the Power Distribution Monitoring System (PDMS)

RTS Instrumentation B 3.3.1 (continued)

Vogtle Units 1 and 2 B 3.3.1-56 REVISION 14 BASES SURVEILLANCE SR 3.3.1.2 (continued)

REQUIREMENTS contributor to the instrument uncertainty for a secondary side power calorimetric measurement is the feedwater flow measurement which is typically a ¨P measurement across a feedwater venturi. While the measurement uncertainty remains constant in ¨P as power decreases, when translated into flow, the uncertainty increases as a square term. Thus a 1% flow error at 100% RTP can approach a 10% error at 30% RTP even though the ¨P error has not changed. An evaluation of extended operation at part-power conditions would conclude that it is prudent to administratively adjust the setpoint of the Power Range Neutron Flux - High bistables to 90% RTP for a calorimetric power determined below 50% RTP, and to 75%

RTP for a calorimetric power determined below 20% RTP when:

1) the power range channel output is adjusted in the decreasing power direction due to a part-power calorimetric; or 2) for a post-refueling startup. While the part-power calorimetric uncertainty based on a feedwater flow measurement from the leading-edge flow meter (LEFM) is less than that based on the feedwater venturi, it is prudent to continue to apply the same adjustments to the setpoint.

Before the Power Range Neutron Flux - High bistables are reset to the nominal value in Table 3.3.1-1 of Specification 3.3.1, the power range channel adjustment must be confirmed based on a calorimetric performed at a power level 50% RTP.

The Note clarifies that this Surveillance is required only if reactor power is t 15% RTP and that 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is allowed for performing the first Surveillance after reaching 15% RTP. At lower power levels, calorimetric data are inaccurate.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.1.3 SR 3.3.1.3 compares the incore system to the NIS channel output.

core power distribution information

RTS Instrumentation B 3.3.1 (continued)

Vogtle Units 1 and 2 B 3.3.1-57 REVISION 52

)

Flux

+

Flux

( )

Power

(

)

Flux Flux

(

100

=

AO B

T B

T u

)/100)

(

Power

(

AO

=

AFD u

BASES SURVEILLANCE SR 3.3.1.3 (continued)

REQUIREMENTS If the absolute difference is t 3%, the NIS channel is still OPERABLE, but must be readjusted. If the NIS channel cannot be properly readjusted, the channel is declared inoperable. This surveillance is primarily performed to verify the (AFD) input to the overtemperature 'T function.

The Note clarifies that the Surveillance is required only if reactor power is t 50% RTP and that 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is allowed for performing the first Surveillance after reaching 50% RTP.

Axial offset is the difference between the power in the top half of the core and the bottom half of the core expressed as a fraction (percent) of the total power being produced by the core.

Mathematically, it is expressed as:

where FluxT = neutron flux at the top of the core, and FluxB = neutron flux at the bottom of the core The relationship between AFD and axial offset is:

AFD as displayed on the main control board and as determined by the plant computer use inputs from the power range NIS detectors which are located outside the reactor vessel. Axial offset is measured using incore detectors. For the performance of SR 3.3.1.3, WCAP-8648-A, EXCORE Detector Recalibration Using Quarter-Core Flux Maps, provides an acceptable method of measuring axial offset using incore detectors.

The surveillance assures that the AFD as displayed on the main control board and as determined by the plant computer is within 3% of the AFD as calculated from the axial offset equation.

Agreement is required so that the reactor is operated within the bounds of the safety analysis regarding axial power distribution.

either the Power Distribution Monitoring System (PDMS) or movable

RTS Instrumentation B 3.3.1 (continued)

Vogtle Units 1 and 2 B 3.3.1-59 REVISION 20 BASES SURVEILLANCE SR 3.3.1.6 REQUIREMENTS (continued)

SR 3.3.1.6 is a calibration of the excore channels to the incore channels. If the measurements do not agree, the excore channels are not declared inoperable but must be calibrated to agree with the incore detector measurements. If the excore channels cannot be adjusted, the channels are declared inoperable. This surveillance is primarily performed to verify the f(AFD) input to the overtemperature 'T function.

Two Notes modify SR 3.3.1.6. Note 1 states that this Surveillance is required only if reactor power is > 75% RTP and that 7 days is allowed for performing the first surveillance after reaching 75% RTP. Note 2 states that neutron detectors are excluded from the calibration.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.1.7 SR 3.3.1.7 is the performance of a COT.

A COT is performed on each required channel to ensure the entire channel will perform the intended Function. Setpoints must be conservative with respect to the Allowable Values specified in Table 3.3.1-1.

The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology. The setpoint shall be left set consistent with the assumptions of the current unit specific setpoint methodology.

The "as-found" and "as-left" values must also be recorded and reviewed for consistency with the assumptions of Reference 6.

This Surveillance Requirement is modified by two Notes that apply only to the Source Range instrument channels. Note 1 requires that the COT include verification that interlocks P-6 and P-10 are in the required state for the existing unit core power distribution information

RTS Instrumentation B 3.3.1 (continued)

Vogtle Units 1 and 2 B 3.3.1-63 REVISION 20 BASES SURVEILLANCE For channels determined to be OPERABLE but degraded, after REQUIREMENTS returning the channel to service the channels will be evaluated (continued) under the plant Corrective Action Program. Entry into the Corrective Action Program will ensure required review and documentation of the condition. The second Note requires that the as-left setting for the channel be returned to within the as-left tolerance of the NTSP. Where a setpoint more conservative than the NTSP is used in the plant surveillance procedures (field setting), the as-left and as-found tolerances, as applicable, will be applied to the surveillance procedure setpoint. This will ensure that sufficient margin to the Safety Limit and/or Analytical Limit is maintained. If the as-left channel setting cannot be returned to a setting within the as-left tolerance of the NTSP, then the channel shall be declared inoperable.

The second Note also requires that the methodologies for calculating the as-left and the as-found tolerances be in NMP-ES-033-006, Vogtle Setpoint Uncertainty Methodology and Scaling Instructions.

SR 3.3.1.11 SR 3.3.1.11 is the performance of a CHANNEL CALIBRATION, as described in SR 3.3.1.10. This SR is modified by a Note that states that neutron detectors are excluded from the CHANNEL CALIBRATION. The CHANNEL CALIBRATION for the power range neutron detectors includes a normalization of the detectors based on a power calorimetric and flux map performed above 75% RTP.

The CHANNEL CALIBRATION for the source range neutron detectors includes obtaining the detector preamp discriminator curves and evaluating those curves.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.1.11 is modified by two Notes as identified in Table 3.3.1-1. The first Note requires evaluation of channel performance for the condition where the as-found setting for the channel setpoint is outside its as-found tolerance but conservative with respect to the Allowable Value. Evaluation of channel performance will verify that the channel will continue to behave in accordance with safety analysis assumptions and the channel performance assumptions in the setpoint methodology.

The purpose of the assessment is to ensure confidence in the channel performance prior to returning the channel to service.

For channels determined to be OPERABLE but degraded, after core power distribution information

Joseph M. Farley Nuclear Plant - Units 1 and 2 Vogtle Electric Generating Plant - Units 1 and 2 Voluntary License Amendment Request to Use BEACON Power Distribution Monitoring System Attachment

Farley Nuclear Plant 1&2 Marked-up Technical Requirements Manual (Information only)

Movable Incore Detectors TR 13.3.1 Farley Units 1 and 2 13.3.1 - 1 Version 34.0 Technical Requirements 13.3 Instrumentation TR 13.3.1 Movable Incore Detectors TR 13.3.1 The movable incore detection system shall be FUNCTIONAL with:

1.1 38 detector thimbles, with 2 detector thimbles per quadrant as identified in Figure 13.3.1-1, OR 1.2 32 and < 38 detector thimbles, with 3 detector thimbles per quadrant as identified in Figures 13.3.1-1 and 13.3.1-2, OR 1.3 25 and < 32 detector thimbles, with 4 detector thimbles per quadrant as identified in Figures 13.3.1-1 and 13.3.1-2.

AND 2.

Sufficient movable detectors, drives, and readout equipment to map these thimbles.

NOTES------------------------------------------------------

1.

38 detector thimbles, with 2 detector thimbles per quadrant as identified in Figure 13.3.1-1 are required during initial startup after a refueling outage up to and including performance of the first flux map at 100% RTP. An exception is discussed in the Bases for TR 13.3.1 for recalibration of the excore neutron flux detection system and for monitoring QPTR. However, a flux map with 38 detector thimbles, with 2 detector thimbles per quadrant as identified in Figure 13.3.1-1 is still required prior to reaching 50% RTP to detect a core misload event.

2.

If a detector thimble is located on either the major axes of Figure 13.3.1-1 or minor axes of Figure 13.3.1-2, the detector thimble can be included in both quadrants that are divided by the axis for the purpose of determining the minimum number of detector thimbles per core quadrant.

APPLICABILITY:

When the movable incore detection system is used for:

a.

Recalibration of the excore neutron flux detection system, or b.

Monitoring the Quadrant Power Tilt Ratio (QPTR), or c.

Measurement of FN'H, FQ (Z) and Fxy.

3. Refer to TR 13.3.10 for movable incore detector requirements for the BEACON Power Distribution Monitoring System (PDMS).

Power Distribution Monitoring System (PDMS) 13.3.10 Farley Units 1 and 2 13.3.10-1 Revision ##

Technical Requirements 13.3 INSTRUMENTATION TR 13.3.10 Power Distribution Monitoring System (PDMS)

TR 13.3.10 The PDMS shall be Functional with the minimum inputs in Table 13.3.10-1.

APPLICABILITY:

,Q02'(573.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. TR not met.

A.1 Suspend the use of the PDMS.

Immediately TECHNICAL REQUIREMENT SURVEILLANCES SURVEILLANCE FREQUENCY TRS 13.3.10.1 Perform a CHANNEL CHECK.

31 days TRS 13.3.10.2 Perform calibration of the PDMS using the movable LQFRUHGHWHFWRUV\\VWHPZLWKDWOHDVWRIWKH

detector thimbles and at least 2 detector thimbles per quadrant, using the minimum thermocouple FRYHUDJHDQGZLWK7+(50$/32:(5!

RTP.

Once after each refueling prior to THERMAL POWER exceeding

573 Note: New TRM Section

Power Distribution Monitoring System (PDMS) 13.3.10 Farley Units 1 and 2 13.3.10-2 Revision ##

Technical Requirements SURVEILLANCE FREQUENCY TRS 13.3.10.3 Perform calibration of the PDMS using the movable incore GHWHFWRUV\\VWHPZLWKDWOHDVWRIWKH

detector thimbles and at least 2 detector thimbles per quadrant, using the minimum thermocouple FRYHUDJHDQGZLWK7+(50$/32:(5!

RTP.

31 EFPD with minimum thermocouple coverage OR 180 EFPD with optimum thermocouple coverage

Power Distribution Monitoring System (PDMS) 13.3.10 Farley Units 1 and 2 13.3.10-3 Revision ##

Technical Requirements BASES TRS 13.3.10.3 For PDMS calibration, the quantity and the coverage distribution of core exit thermocouples used as data input must meet certain criteria. With respect to thermocouple coverage, the available core exit thermocouple coverage can be optimum or minimum as described below. This criterion affects the TRS Frequency:

Optimum thermocouple coverage satisfies the minimum thermocouple Functionality requirement in Table 13.3.10-1 with the added requirement that the Functional pattern covers all internal fuel assemblies (no face along a baffle) within a chessboard knight move (an adjacent plus a diagonal square away).

Minimum thermocouple coverage satisfies thermocouple minimum Functionality requirements of Table 13.3.10-1 but does not meet the knight move pattern discussed above.

Power Distribution Monitoring System (PDMS) 13.3.10 Farley Units 1 and 2 13.3.10-4 Revision ##

Technical Requirements Table 13.3.10-1 (Page 1 of 1)

Power Distribution Monitoring System FUNCTION MINIMUM REQUIRED INPUTS

1. Control Bank Position 4 control banks (a)

2. RCS Cold Leg 2 RCS loops (b)

Temperature T-cold

3. Reactor Power 1(c)

Level

4. Power Range Excore 3(d)

Detector Signals

. Core Exit

ZLWKper Thermocouple quadrant Temperatures (a)

Determined from either valid demand position indication or the average of individual DRPI indications.

(b)

Either narrow range or wide range RTDs.

(c)

Either valid secondary calorimetric, average power range neutron flux power, or average 5&6ORRS7

(d)

An input is a channel which consists of corresponding upper and lower detector sections.

Joseph M. Farley Nuclear Plant - Units 1 and 2 Vogtle Electric Generating Plant - Units 1 and 2 Voluntary License Amendment Request to Use BEACON Power Distribution Monitoring System Attachment

Vogtle Electrical Generating Plant 1&2 Marked-up Technical Requirements Manual (Information only)

Movable Incore Detectors TR 13.3.1 Vogtle Units 1 and 2 13.3 - 1 REVISION 40 Technical Requirement 13.3 Instrumentation TR 13.3.1 Movable Incore Detectors TR 13.3.1 The movable incore detection system shall be FUNCTIONAL with:

1.

t 44 detector thimbles, with t 2 detector thimbles per core quadrant as identified in Figure 13.3.1-1, OR 2.

t 37 and < 44 detector thimbles, with t 3 detector thimbles per core quadrant as identified in Figures 13.3.1-1 and 13.3.1-2, OR 3.

t 29 and < 37 detector thimbles, with t 4 detector thimbles per core quadrant as identified in Figures 13.3.1-1 and 13.3.1-2.

AND Sufficient movable detectors, drives, and readout equipment to map these thimbles.

NOTES------------------------------------------------

1. t 44 detector thimbles, with t 2 detector thimbles per core quadrant as identified in Figure 13.3.1-1 are required during the initial startup after a refueling outage up to and including performance of the first flux map at 100% RTP. An exception is that for the performance of SR 3.3.1.3, WCAP-8648-A, EXCORE Detector Recalibration Using Quarter-Core Flux Maps, provides an acceptable method of measuring axial offset using incore detectors. However, a flux map with GHWHFWRUWKLPEOHVZLWKGHWHFWRU

thimbles per core quadrant as identified in Figure 13.3.1-1 is still required prior to reaching 50% RTP to detect a core misload event. This Note does not have to be met for Vogtle Unit 1, Cycle 17 based on the successful performances of the flux map at 30% RTP.

2. If a detector thimble is located on either the major axes of Figure 13.3.1-1 or minor axes of Figure 13.3.1-2, the detector thimble can be included in both core quadrants that are divided by the axis for the purpose of determining the minimum number of detector thimbles per core quadrant.

3. Refer to TR 13.3.9 for movable incore detector requirements for the BEACON Power Distribution Monitoring System (PDMS).

Power Distribution Monitoring System (PDMS) 13.3.9 Vogtle Units 1 and 2 13.3.9-1 Revision ##

Technical Requirements 13.3 INSTRUMENTATION TR 13.3.9 Power Distribution Monitoring System (PDMS)

TR 13.3.9 The PDMS shall be Functional with the minimum inputs in Table 13.3.9-1.

APPLICABILITY:

,Q02'(573.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. TR not met.

A.1 Suspend the use of the PDMS.

Immediately TECHNICAL REQUIREMENT SURVEILLANCES SURVEILLANCE FREQUENCY TRS 13.3.9.1 Perform a CHANNEL CHECK.

31 days TRS 13.3.9.2 Perform calibration of the PDMS using the movable LQFRUHGHWHFWRUV\\VWHPZLWKDWOHDVWRIWKH

detector thimbles and at least 2 detector thimbles per quadrant, using the minimum thermocouple FRYHUDJHDQGZLWK7+(50$/32:(5!

RTP.

Once after each refueling prior to THERMAL POWER exceeding

573 Note: New TRM Section

Power Distribution Monitoring System (PDMS) 13.3.9 Vogtle Units 1 and 2 13.3.9-2 Revision ##

Technical Requirements SURVEILLANCE FREQUENCY TRS 13.3.9.3 Perform calibration of the PDMS using the movable incore GHWHFWRUV\\VWHPZLWKDWOHDVWRIWKH

detector thimbles and at least 2 detector thimbles per quadrant, using the minimum thermocouple FRYHUDJHDQGZLWK7+(50$/32:(5!

RTP.

31 EFPD with minimum thermocouple coverage OR

()3'ZLWK optimum thermocouple coverage

Power Distribution Monitoring System (PDMS) 13.3.9 Vogtle Units 1 and 2 13.3.9-3 Revision ##

Technical Requirements BASES TRS 13.3.9.3 For PDMS calibration, the quantity and the coverage distribution of core exit thermocouples used as data input must meet certain criteria. With respect to thermocouple coverage, the available core exit thermocouple coverage can be optimum or minimum as described below. This criterion affects the TRS Frequency:

Optimum thermocouple coverage satisfies the minimum thermocouple Functionality requirement in Table 13.3.9-1 with the added requirement that the Functional pattern covers all internal fuel assemblies (no face along a baffle) within a chessboard knight move (an adjacent plus a diagonal square away).

Minimum thermocouple coverage satisfies thermocouple minimum Functionality requirements of Table 13.3.9-1 but does not meet the knight move pattern discussed above.

Power Distribution Monitoring System (PDMS) 13.3.9 Vogtle Units 1 and 2 13.3.9-4 Revision ##

Technical Requirements Table 13.3.9-1 (Page 1 of 1)

Power Distribution Monitoring System FUNCTION MINIMUM REQUIRED INPUTS

1. Control Bank Position 4 control banks (a)

2. RCS Cold Leg 2 RCS loops (b)

Temperature T-cold

3. Reactor Power 1(c)

Level

4. Power Range Excore 3(d)

Detector Signals

. Core Exit

ZLWKper Thermocouple quadrant Temperatures (a)

Determined from either valid demand position indication or the average of individual DRPI indications.

(b)

Either narrow range or wide range RTDs.

(c)

Either valid secondary calorimetric, average power range neutron flux power, or average 5&6ORRS7

(d)

An input is a channel which consists of corresponding upper and lower detector sections.

v t e Letter Sequence Request
EPID:L-2021-LLA-0109, Units 1 and 2, Vogtle Electric Generating Plant - Units 1 and 2, Voluntary License Amendment Request to Use Beacon Power Distribution Monitoring System (Approved, Closed)
Initiation Request Acceptance Administration Questions Answers Results Approval Other:
MONTHYEAR2021-06-09 [Table View]

NL-21-0042, Units 1 and 2, Vogtle Electric Generating Plant - Units 1 and 2, Voluntary License Amendment Request to Use Beacon Power Distribution Monitoring System

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