View source for NRC 2022-0025, License Amendment Request 295, Beacon Power Distribution Monitoring System

{{Adams
| number = ML22270A084
| issue date = 09/26/2022
| title = License Amendment Request 295, Beacon Power Distribution Monitoring System
| author name = Strand D
| author affiliation = NextEra Energy Point Beach, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000266, 05000301
| license number = DPR-024, DPR-027
| contact person = 
| case reference number = NRC 2022-0025
| document type = Letter type:NRC, License-Application for Facility Operating License (Amend/Renewal) DKT 50
| page count = 55
| project = 
| stage = Request
}}

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{{#Wiki_filter:NEXTeraM ENERGY~
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September 26, 2022 NRC 2022-0025 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington , D. C. 20555-0001 RE:      Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 Renewed Facility Operating Licenses DPR-24 and DPR-27 License Amendment Request 295, BEACON Power Distribution Monitoring System NextEra Energy Point Beach, LLC (NextEra) hereby requests amendments to Renewed Facility Operating Licenses DPR-24 and DPR-27 for Point Beach Nuclear Plant (PBNP)
Units 1 and 2, respectively. Specifically, NextEra proposes to revise Technical Specification (TS) 3.2.4, "Quadrant Power Tilt Ratio (QPTR)" and TS 3.3.1 , "Reactor Protection System (RPS) 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 is an NRC approved Westinghouse proprietary core analysis system called Best Estimate Analyzer for Core Operations - Nuclear (BEACON').
The Enclosure to this letter provides the evaluation of the proposed TS changes for the PDMS. The Enclosure contains five Attachments:
: 1. Evaluation for Excluding Power Distribution Monitoring System Requirements from the Technical Specifications
: 2. Marked-up TS Pages
: 3. Revised TS Pages
: 4. Marked-up TS Bases Pages (Information only)
: 5. Marked-up TRM Pages (Information only)
Next Era 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 radiation exposure. Therefore, the proposed amendment meets the categorical exclusion requirements of 10 CFR 51 .22(c)(9) and pursuantto 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
NextEra En ergy Point Beach, LLC 6610 Nu clear Road, Two Rive rs, W I 54241

Point Beach Nuclear Plant, Units 1 and 2                                       NRC 2022-0025 Docket Nos. 50-266 and 50-301                                                       Page 2 of 2 The Point Beach Onsite Review Group (ORG) has reviewed the enclosed amendment request.
In accordance with 10 CFR 50.91 (b)(1), a copy of this license amendment request is being forwarded to the designee for the State of Wisconsin .
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 120 days following issuance of the amendment.
Should you have any questions regarding this submittal, please contact Mr. Kenneth Mack, Licensing Manager, at 561-904-3635.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on the    2-VJ    day of September 2022.
Sincerely, Dianne Strand General Manager Regulatory Affairs NextEra Energy cc:    USNRC Regional Administrator, Region Ill Project Manager, USNRC, Point Beach Nuclear Plant Resident Inspector, USNRC , Point Beach Nuclear Plant Public Service Comm ission of Wisconsin

==Enclosure:==
Evaluation of Proposed Change

Point Beach Nuclear Plant, Units 1 and 2                                                                                               NRG 2022-0025 Docket Nos. 50-266 and 50-301                                                                                                                     Enclosure Page 1 of 53 ENCLOSURE Evaluation of the Proposed Change License Amendment Request 295, BEACON Power Distribution Monitoring System
: 1. 

==SUMMARY==
DESCRIPTION ..... ..... .......... .. ... ..... ... ... ... .. . ..... ...... .. .. ... ........ . ..... .. .. 2
: 2. DETAILED DESCRIPTION .. ...... . ...... .. ... ..... ...................... .. ........ ............. ...... ... 2 2.1 System Design and Operation .. ... . .... ......... ....... . ..... .. ..... ....... ..... ... .... ........... . 2 2.2 Current Technical Specifications Requirements .. .. .... .. ...................................... 3 2.3 Reason for the Proposed Change .......................... . ......... .. ........ .. ............. .. ... 4 2.4 Description of the Proposed Change ..... .. .. .............. .. ........................ .. ........... 4
: 3. TECHNICAL EVALUATION .. ... . .. ... ...... ... ............................... ... .. ...... .... .. ... .. ... .. 5 3.1 WCAP-12472-P-A ... .. .................................... .......... .. ...... ...... .. .... .... ... ...... .. .. .. .. ......... .. 5 3.2 Applicability of Addenda to WCAP-12472-P-A .. ...... .... .. .. .. ............ ...... .. ... .... ........... .. .. 7 4 . REGULATORY EVALUATION ......... .. ............ .......... .. .. ............... ... ..... ........ ............ .......... 9 4.1 Applicable Regulatory Requirements/Criteria ... ... ..... ............ .. ....... .. ........ .. ................. 9 4.2 Precedent .... .. ...... .... ........... ..... .. ..... ..... .... ......... ....... ... ... ... ........... .... ..... .. ..... ........... ..... 9 4.3 Significant Hazards Consideration ..... .... ... ... ... ... ..... .. .... .. ............ .... ... .. ........... ..... .. .. .... 10 4.4 Conclusions ...... ... .. .. ............ .. .... ..... ..... .... .............. ..... ... .. .. .. .......... .. ............... ........... .. 12
: 5. ENVIRONMENTAL CONSIDERATION ..... ... ........... ...... .. ..... .. ...... .. ....... .......... ...... ..... .. .... 12
: 6. REFERENCES ... ......... .... .. .. ..... ......... .. ..... .. ...... .. .......................................... .. .... ... ... .. .. .... . 12 ATTACHMENTS :
: 1. Evaluation for Excluding BEACON Power Distribution Monitoring System (PDMS)
Requirements from the Technical Specifications
: 2. Marked-up TS Pages
: 3. Revised TS Pages
: 4. Marked-up TS Bases Pages (Information on ly)
: 5. Marked-up TRM Pages (Information only)

Point Beach Nuclear Plant, Units 1 and 2                                            NRC 2022-0025 Docket Nos. 50-266 and 50-301                                                             Enclosure Page 2 of 53
: 1. 

==SUMMARY==
DESCRIPTION Pursuant to 10 CFR 50.90, NextEra Energy Point Beach, LLC (NextEra) hereby requests a license amendment to Units 1 and 2 renewed operating licenses DPR-24 and DPR-27, respectively. Specifically, NextEra proposes to revise Technical Specification (TS) 3.2.4, "Quadrant Power Tilt Ratio (QPTR)" and TS 3.3.1, "Reactor Protection System (RPS)
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.
: 2. DETAILED DESCRIPTION 2.1 System Design and Operation The Movable lncore Detector System (MIDS) consists of a set of 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 is then processed to determine core power distribution information. This information can then be used to satisfy requirements for 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 reactor control or reactor protection functions.
The core power distribution monitoring system (PDMS) to be used at PBNP 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

Point Beach Nuclear Plant, Units 1 and 2                                            NRC 2022-0025 Docket Nos. 50-266 and 50-301                                                             Enclosure Page 3 of 53 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 (ECG) 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 (Fo(Z))) and peak enthalpy rise (TS 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor (FN~H)). 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 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).
* 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 NRG. 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 . NextEra does not propose to license the BEACON-DMM application of the PDMS .
NextEra 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 specify the use of the MIDS for core power distribution determination (flux maps):
TS 3.2.4, Quadrant Power Tilt Ratio (QPTR),
TS 3.3.1, Reactor Protection System (RPS) Instrumentation

Point Beach Nuclear Plant, Units 1 and 2                                           NRC 2022-0025 Docket Nos. 50-266 and 50-301                                                            Enclosure Page 4 of 53 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 will 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.
TS 3.2.4, Quadrant Power Tilt Ratio (QPTR)
In SR 3.2.4.2, "the movable incore detectors" will be replaced with "core power distribution information".
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 3.3.1.6, "incore detector measurements" will be replaced with "core power distribution information".
Marked-up and Revised TS pages are in Attachments 2 and 3 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 3.2.1, "Heat Flux Hot Channel Factor (FQ(Z))," TS 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor'', and TS 3.2.3, "Axial Flux Difference (AFD)", are also marked up to revise references to movable incore detectors and flux maps.
Marked-up TS Bases pages are included in Attachments 4 for information only and will be updated in accordance with TS 5.5.13, "Technical Specifications (TS) Bases Control Program".
NextEra also proposes changes to the TRM to specify functionality requirements for the BEACON (PDMS). TRM and Technical Requirement Manual Surveillance Requirements (TS Rs) are treated as plant procedures. Changes will be made in accordance with 10 CFR 50.59.

Point Beach Nuclear Plant, Units 1 and 2                                                NRC 2022-0025 Docket Nos. 50-266 and 50-301                                                                   Enclosure Page 5 of 53
: 3. TECHNICAL EVALUATION 3.1 WCAP-12472-P-A The PDMS 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 TS modifications that would be necessary for BEACON operating and non-operating conditions.

Point Beach Nuclear Plant, Units 1 and 2                                              NRC 2022-0025 Docket Nos. 50-266 and 50-301                                                               Enclosure Page 6 of 53 WCAP-12472-P-A describes the minimum monitoring instrumentation requirements for BEACON to be considered functional. In particular, it discusses minimum requirements for the 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 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 the condition will be met at PBNP is provided.
: 1. In the cycle-specific application of BEACON, the power peaking uncertainties Ut:iH 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. 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.
NextEra utilizes a Westinghouse 2-loop pressurized water reactor (PWR) nuclear steam supply system (NSSS) with movable incore instrumentation and other core power distribution monitoring instrumentation described by Section 1.0 of the SER for WCAP-12472-P-A. That SER states the general applicability of the WCAP to Westinghouse PWRs. Therefore, PBNP does not differ significantly from the plants that form the WCAP database.
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 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

Point Beach Nuclear Plant, Units 1 and 2                                                NRC 2022-0025 Docket Nos. 50-266 and 50-301                                                                Enclosure Page 7 of 53 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, NextEra is proposing only to use BEACON as a core TS monitor for conformance to existing TS limits. 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 TS. The TS changes of concern per this question or condition are not applicable to the more limited changes being proposed by NextEra 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.
3.2 Applicability of Addenda to WCAP-12472-P-A 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 PBNP.
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 add itional features incorporated into the BEACON monitoring system:
: 1. Use of fixed in core self-powered neutron rhodium detectors, and
: 2. Use of three-dimensional advanced nodal code (ANC) neutronic model code.
PBNP does not use fixed incore detectors so this feature is not applicable. However, ANC is used 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.
PBNP does not use fixed incore detectors so this Addendum is not applicable.

Point Beach Nuclear Plant, Units 1 and 2                                              NRC 2022-0025 Docket Nos. 50-266 and 50-301                                                               Enclosure Page 8 of 53 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. PBNP is a Westinghouse-designed plant so this Addendum is not applicable.
Addendum 4 Addendum 4 of WCAP-12472-P-A was approved by the NRG 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 NRG 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 PBNP cycle-specific reload design analyses.
PBNP does not use fixed incore detectors so the third purpose of this Addendum is not applicable.

Point Beach Nuclear Plant, Units 1 and 2                                             NRC 2022-0025 Docket Nos. 50-266 an d 50-301                                                             Enclosure Page 9 of 53
: 4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria The Point Beach General Design Criteria (GDCs) are similar in content to the Atomic Industrial Forum (AIF) version of the Proposed 1967 GDCs.
* Point Beach GDC 13 (Fission Process Monitors and Controls) states that means shall be provided for monitoring or otherwise measuring and maintaining control over the fission process throughout core life under all conditions that can reasonably be anticipated to cause variations in reactivity of the core.
Implementation of the PDMS at PBNP 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, NextEra 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 TRM. Technical Requirements Manual Limiting Condition for Operation (TLCOs) and TS Rs are treated as plant procedures. Changes will be made in accordance with 10 CFR 50.59. Attachments 4 and 5 provide (for information only) the proposed new TLCO 3.2.2, "Power Distribution Monitoring System" as well as an additional clarification note added to TLCO 3.2.1 under TRM 3.2.1 , "Movable In-core Instrumentation".
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 (Un it 1) and 163 (Unit 2), March 31 , 2006 (ADAMS Accession Number ML060760501 /ML060950451)
* Callaway, Amendment 182, March 21 , 2007, (ADAMS Accession Number

Point Beach Nuclear Plant, Units 1 and 2                                              NRC 2022-0025 Docket Nos. 50-266 and 50-301                                                              Enclosure Page 10 of 53 ML070460584/ML070680350)
* Comanche 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)
* Farley Units 1 and 2, Amendments 239 (Unit) 1 and 236 (Unit 2), and Vogtle Units 1 and 2 Amendments 212 (Unit) 1 and 195 (Unit 2), January 26, 2022, (ADAMS Accession Number ML21344A003)
In the more recent amendments, a PDMS LCO was not added to the TS, but technical requirements regarding FUNCTIONALITY were placed in the TRM which is a document under licensee control.
4.3 No Significant Hazards Consideration Analysis The proposed change would revise Technical Specification (TS) 3.2.4, "Quadrant Power Tilt Ratio (QPTR)" 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.
NextEra 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 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.

Point Beach Nuclear Plant, Units 1 and 2                                              NRC 2022-0025 Docket Nos. 50-266 and 50-301                                                                Enclosure Page 11 of 53 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 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.
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 TS 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.

Point Beach Nuclear Plant, Units 1 and 2                                                NRC 2022-0025 Docket Nos. 50-266 an d 50-301                                                               Enclosure Page 12 of 53 Such actions remain unchanged.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, NextEra 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.
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 Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
: 5. ENVIRONMENTAL CONSIDERATION 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.
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.]

Point Beach Nuclear Plant, Units 1 and 2                                                 NRC 2022-0025 Docket Nos. 50-266 and 50-301                                                                  Enclosure Page 13 of 53 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 PBNP 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.
(8) Criterion 2. A process varia.ble, 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 FN11H 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 OBA 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 OBA evaluations. The PDMS performs no active/control functions, nor does the PDMS have an actuation capability.
A1-1

Point Beach Nuclear Plant, Units 1 and 2                                               NRC 2022-0025 Docket Nos. 50-266 and 50-301                                                                Enclosure Page 14 of 53 Therefore, the PDMS is not part of any primary success path for mitigation of a OBA 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 OBA 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 w ill suffice to provide the necessary FUNCTIONALITY and test requirements for the PDMS apart from the TS. Attachment 5 provides (for information only) the proposed new TRM 3.2.2, "Power Distribution Monitoring System" as well as an additional clarification note added to TRM 3.2.1 .
A1-2

Point Beach Nuclear Plant, Units 1 and 2                    NRC 2022-0025 Docket Nos. 50-266 and 50-301                                   Enclosure Page 15 of 53 Attachment 2 Marked-up TS Pages (3 Pages Follow)

QPTR 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                        FREQUENCY SR 3.2.4 .1     ---- ---------------- ------NOTES-------------------------
: 1.       With input from one Power Range Neutron Flux channel inoperable and THERMAL POWER ~ 75% RTP, the remaining three power range channels can be used for calculating QPTR.
: 2.       SR 3.2.4.2 may be performed in lieu of this Surveillance.
Verify QPTR is within limit by calculation .                 In accordance with the Surveillance Frequency Control Program SR 3.2.4.2      ---------------------------NOTE--------------------------
N ot requ ired to be performed until 12 hours after input from one or more Power Range Neutron Flux channels are inoperable with THERMAL POWER > 75 % RTP.
Verify QPTR is within limit using tl=le F11ovab le            In accordance iRCOFC EleteetoFS .                                           with the core power distribution V                                                              Surveillance information                                                                            Frequency Control Program Point Beach                                    3.2.4-4               Unit 1 - Amendment No . 253 Unit 2 - Amendment No . 257

RPS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS
--------------------------------------------------NOTE-----------------------------------------------------
Refer to Table 3.3.1-1 to determine which SRs apply for each RPS 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---------------------- ---
: 1.      Adjust NIS channel if absolute difference is> 2% .
: 2.      Not required to be performed until 12 hours after THERMAL POWER is 2 15%
RTP.
In accordance Compare results of calorimetric heat balance                  with the calculation to Nuclear Instrumentation System                 Surveillance (NIS) channel output.                                         Frequency Control Program SR 3.3.1.3            ---------------------------NOTES-------------------------
: 1.      Adjust NIS channel if absolute difference is ~ 3%.
: 2.      Not required to be performed until 24 hours after THERMAL POWER is 2 50%
RTP.
In accordance Compare results of the ii,csre deteotor                       with the ffieosureFRen~s to N IS AFD.                                  Surveillance Frequency Icore power distribution information                                        Control Program (continued)
Point Beach                                          3.3.1-7             Unit 1 - Amendment No. 253 Unit 2 - Amendment No . 257

RPS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE                                        FREQUENCY SR 3.3.1.4       ----------------------------NOTE--------------------------
Th is Surveillance must be performed on the reactor trip bypass breaker prior to placing the bypass breaker in service.
In accordance Perform TADOT.                                              with the Surveillance Frequency Control Program SR 3.3.1.5       ---------------------------NOTES----------- ----------- ---
: 1. Not required to be performed for the Source Range Neutron Flux Trip Function until 8 hours after power is below P-6.
: 2. Not required to be performed for the RCP Breaker Position (Two Loops), Reactor Coolant Flow - Low (Two Loops) and Underfrequency Bus A01 and A02 Trip Functions and the P-6, P-7 , P-8, P-9 and P-10 Interlocks.
In accordance Perform ACTUATION LOGIC TEST.                               with the Surveillance Frequency Control Program SR 3.3 .1.6      ----------------------------NOTE--------------------------
N ot requ ired to be performed until 24 hours after THERMAL POWER is 2 50% RTP.
In accordance Calibrate excore channels to agree with inoore              with the detector rneasuremen~s .                                    Surveillance
!core power distribution information ~                                          Frequency Control ProQram (continued)
Point Beach                                   3.3.1-8              Unit 1 - Amendment No. 253 Unit 2 - Amendmen t No . 257

Point Beach Nuclear Plant, Units 1 and 2                NRG 2022-0025 Docket Nos. 50-266 an d 50-301                               Enclosure Page 19 of 53 Attachment 3 Revised TS Pages (3 Pages Follow)

QPTR 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                        FREQUENCY SR 3.2.4.1   --------------------------NOTES-------------------------
: 1.      With input from one Power Range Neutron Flux channel inoperable and THERMAL POWER ~ 75% RTP, the remaining three power range channels can be used for calculating QPTR.
: 2.      SR 3.2 .4 .2 may be performed in lieu of this Surveillance .
Verify QPTR is within limit by calculation.                  In accordance with the Surveillance Frequency Control Program SR 3.2.4.2  ---------------------------NOTE--------------------------
Not required to be performed until 12 hours after input from one or more Power Range Neutron Flux channels are inoperable with THERMAL POWER> 75% RTP.
Verify QPTR is within limit using core                       In accordance power distribution information.                              with the Surveillance Frequency Control Program Point Beach                              3.2.4-4                    Unit 1 - Amendment No .
Unit 2 - Amendment No .

RPS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS
--------------------------------------------------NOTE-----------------------------------------------------
Refer to Table 3.3.1-1 to determine which SRs apply for each RPS 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---------------- --- ---- --
: 1.      Adjust NIS channel if absolute difference is> 2%.
: 2.      Not required to be performed until 12 hours after THERMAL POWER is ~ 15%
RTP.
In accordance Compare results of calorimetric heat balance                      with the calculation to Nuclear Instrumentation System                     Surveillance (N IS) channel output.                                            Frequency Control Program SR 3.3.1.3            --------- ------------------NOTES-------------------- --- --
: 1.      Adjust NIS channel if absolute difference is ~ 3% .
: 2.      Not required to be performed until 24 hours after THERMAL POWER is ~ 50%
RTP.
In accordance Compare results of core power distribution                        with the information to NIS AFD.                                           Surveillance Frequency Control Program (continued)
Point Beach                                            3.3.1-7                     Unit 1 - Amendment No.
Unit 2 - Amendment No .

RPS Instrumentation 3.3.1 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 .
In accordance Perform TADOT.                                                    with the Surveillance Frequency Control Program SR 3.3.1.5  --- -------------- -- -------- NOTES-------------------------
: 1.       Not required to be performed for the Source Range Neutron Flux Trip Function until 8 hou rs after power is below P-6.
: 2.       Not required to be performed for the RCP Breaker Position (Two Loops), Reactor Coolant Flow - Low (Two Loops) and Underfrequency Bus A01 and A02 Trip Functions and the P-6 , P-7, P-8 , P-9 and P-10 Interlocks.
In accordance Perform ACTUATION LOGIC TEST .                                    with the Surveillance Frequency Control Program SR 3.3.1.6  -- ---- ---------------- --- ---NOTE------ --------------------
Not re quired to be performed until 24 hours after THERMAL POWER is ~ 50% RTP .
In accordance Calibrate excore channels to agree with core                      with the power distribution information.                                   Surveillance Frequency Control Proqram (continued)
Point Beach                                  3.3.1-8                     Unit 1 - Amendment No .
Unit 2 - Amendment No .

Point Beach Nuclear Plant, Units 1 and 2                         NRG 2022-0025 Docket Nos . ~0-266 and 50-301                                       Enclosure Page 23 of 53 Attachment 4 Marked-up TS Bases Pages (Information only)
(24 Pages Follow)

Rod Group Alignment Limits 83.1.4 BASES BACKGROUND   analyses are limited. Operability of the control rod position indicators is (continued)  required to determine control rod position and thereby ensure compliance with the control rod alignment and insertion limi Core power distribution Permitted control rod misalignments              cated by the RPI System information, as referenced                                       otion) are; a)+/- 18 steps of the bank below, can be provided by  demand positio            1cient peaking factor margin exists, the power the power distribution     leve
* ater than 85 percent of rated power, and bank D demand is monitoring system           ess than 215 steps withdrawn), b) +/- 24 steps of the bank demand (PDMS) or by the movable   position (if sufficient peaking factor margin exists, the power level is incore detector system     greater than 85 percent of rated power, and bank D demand is greater than 215 steps withdrawn), and c) +/- 24 steps of the bank demand (MIDS).
position (if the power level is less than or equal to 85 percent of rated power). Above 85 percent of rated power, sufficient peaking factor margin is demonstrated by satisfying the requirements of Table 3.1.4-1, e.g., for an 18 step indicated misalignment and rods less than 215 steps withdrawn, the peak measured F0 (Z) from the latest iI1coIe flux map must be at least 5.0% less than the limit and the peak measured core powe r dist ribut ion          F~H from the latest inoore flux map must be at least 2.0% less than the inform ation                        limitin       e. For power levels less than or equal to 85 percent of r       power, the peaking factor margin does not have to be verified on an explicit basis. This is due to the rate of peaking factor margin core power distribution           increase (due to the peaking factor limit increasing) as the power level Inform ati on decreases being greater than the peaking factor margin loss (due to the increased control rod misalignment). This effect is described in WCAP-15432 Rev. 2. These limits are applicable to all shutdown and control rods (of all banks) over the range of Oto 230 steps withdrawn inclusive.
Control rods in a single bank move together with no individual rod insertion differing by more than 30 steps from the bank demand position (operation at greater than 85 percent of rated power and demand less than 215 steps), nor more than 36 steps (operation at less than or equal to 85 percent of rated power or operation at greater than 85 percent of rated power and demand position greater than or equal to 215 steps withdrawn). An indicated misalignment limit of 18 steps precludes a rod misalignment of greater than 30 steps with consideration of instrumentation error; 24 steps indicated misalignment corresponds to 36 steps with instrumentation error.
Point Beach                         B 3.1.4-3                Unit 1 - Amendment No. 212 Unit 2 - Amendment No. 217

Rod Group Alignment Limits B 3.1.4 BASES LCO                preserve the design peaking factors , and F8(Z), F~(Z) and F~H must be verified directly      inooro A9apping . Bases Section 3.2 (Power core power distribution D' *
* ImI ) contains more complete discussions of the relation information             of F8(Z), F~(Z), and F~H to the operating limits.
Shutdown and control rod OPERABILITY and alignment are directly related to power distributions and SOM, which are initial conditions assumed in safety analyses. Therefore they satisfy Criterion 2 of 10 CFR 50 .36(c)(2)(ii).
The limits on shutdown or control rod alignments ensure that the assumptions in the safety analysis will remain valid . The requirements on control rod OPERABILITY ensure that upon reactor trip, the assumed reactivity will be available and will be inserted . The control rod OPERABILITY requirement is satisfied provided the control rod will fully insert within the required rod drop time assumed in the safety analysis.
Control rod malfunctions that result in the inability to move a control rod (e.g. lift coil and rod control system logic failures), but do not impact the control rod trippability, do not result in control rod inoperability. The LCO requirements also ensure that the RCCAs and banks maintain the correct power distribution and rod alignment.
The requirement to maintain the rod alignment to within +/- 18 steps (for power operation above 85% and bank demand position less than 215 steps) or within+/- 24 steps (for power operation greater than 85% and bank demand position greater than or equal to 215 steps) is conservative . The minimum misalignment assumed in safety analysis is 36 steps, and in some cases a total misalignment from fully withdrawn to fully inserted is assumed. 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 neutron (or fission) power is generated, 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 bottomed and the reactor is shut down and not producing fission power. In the shutdown MODES, the OPERABILITY of the shutdown and control rods has the potential to affect the required SOM, but this effect can be compensated for by an increase in the boron concentration of the RCS. See LCO 3.1.1, "SHUTDOWN MARGIN (SOM)" for SOM in MODE 2 with kett < 1.0, and MODES 3, 4, and 5 and LCO 3.9.1, "Boron Concentration," for boron concentration requirements during refueling.
Point Beach                                B 3.1.4-5               Unit 1 - Amendment No. 212 Unit 2 - Amendment No. 217

Rod Group Alignment Limits B 3.1.4 BASES ACTIONS (continued) B.2, 8.3, 8.4, and 8.5 For continued operation with a misaligned rod, RTP must be reduced ,
SOM must periodically be verified within limits, hot channel factors F8(Z), F~(Z) and F~H must be verified within limits, and the safety analyses must be re-evaluated to confirm continued operation ls perm issible.
Reduction of power to 75% RTP ensures that local LHR increases due to a misaligned RCCA will not cause the core design criteria to be exceeded (Ref. 4 ). The Completion Time of 2 hours gives the operator sufficient time to accomplish an orderly power reduction without challenging the Reactor Protection System.
When a rod is known to be misaligned, there is a potential to impact the SOM. Since the core conditions can change with time, periodic verification of SOM is required. A Frequency of 12 hours is sufficient to ensure this requirement continues to be met.
informa t ion Verifying that F8(Z), F~(Z) and F~H, are within the required limits ensures that current operation at 75% RTP with a rod misaligned is resulting in power distributions that may invalidate safety analysis assumptions at full power. The Completion Time of 72 hours all s sufficient time to obtain flux maps of tho core power distribution ~
tho inooro flw( R1apping system and to calculate F8(Z), F~(Z) and F~H' 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 . The accident analyses presented in the FSAR Chapter 14 (Ref. 4) that may be adversely affected will be evaluated to ensure that the analysis results remain valid for the duration of continued operation under these conditions. A Completion Time of 5 days is sufficient time to obtain the required input data and to perform the analysis .
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 MODE 3 within 6 hours, which obviates concerns about the development of undesirable xenon or power distributions. The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging the plant systems.
Point Beach                               B3.1.4-7                Unit 1 - Amendment No. 264 Unit 2 - Amendment No. 267

Fa(Z)
B 3.2.1 BASES B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 Heat Flux Hot Channel Factor (Fa(Z))
BASES BACKGROUND            The purpose of the limits on the values of Fa(Z) is to li mit the local (i.e. , pellet) peak power density. The value of F0 (Z) varies along the axial height (Z) of the core .
F0 (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 , F0 (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 limits on power distributions on a continuous basis.
F0 (Z) varies with fuel loading patterns, control bank insertion, fuel burnup , and changes in axial power distri bution .
F0 (Z) is measured periodically using the inoore Eleteetor s,istorn . These measurements are generally taken with the core at or near equilibrium conditions.
Using the rneasurod three dimensional power distributions, it is possible to derive a measured value for F0 (Z). However, because this value represents an equilibrium condition, it does not include the variations in the value of F0 (Z) which are present during non-equilibrium situations ,
such as load following or power ascension.
To account for these possible variations, the equilibrium value of F0 (Z) is adjusted as P U Z) by an elevation dependent factor (W(Z)) that accounts for the calculated worst case transient conditions.
Core monitoring and control under non-equilibrium conditions are accomplished by operating the core within the limits of the appropriate LCOs , including the limits on AFD , QPTR, and control rod insertion.
Point Beach                                 B 3.2.1 -1                Unit 1 Amendment No. 201 Unit 2 Amendment No. 206

Fo(Z)
B 3.2.1 BASES BACKGROUND             In the unlikely event tl'lst fficssurcfficAts iAElicstc that the limit for Fow(z)
(continued)            cou ld be exceeded du ring future non-equil ibrium operation, a more
. - - - - - - - - - - --. restrictive Constant Axia l Offset Control (CAOC) operating space Core power distribution      specified in the Core Operating Limits Report (COLR) may be information, as referenced   implemented      to restore marg in to the Fow(z) limit. A CAOC operating below, can be provided by    space    is a uniq ue combination of an al lowab le AFD band and Control the power distribution       Bank    Insertion  Limits. A more restrictive CAOC operating space would
                           ..--~ lay a narrower AFD band, shal lower Control Bank Insertion Limits, monitoring system           or a com
* the two . W(z) functions for each CAOC operating (PDMS) or by the movable space are specified in the '-'L. '*~ ~one of the CAOC operating incore detector system      spaces provide adequate marg in to the Fa                       n TH ERMAL (M IDS).                     POWER must be limited to less than RATED THERMAL P APPLICABLE            This LCO precludes core power distributions that violate the following SAFETY ANALYSES       fuel design criteria:
: a. During a large break loss of coolant accident (LOCA), the peak cladding temperature must not exceed 2200&deg;F (Ref. 1 );
: b. During a loss of forced reactor coolant flow accident, there must be at least 95% probability at the 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience a departure from nucleate boiling (DNB) condition;
: c. During an ejected rod accident, the energy deposition to the fuel must not exceed 225 cal/gm for unirradiated fuel and 200 cal/gm for irradiated fuel (Ref. 2); and
: d. The control rods must be capable of shutting down the reactor with a minimum required SOM with the highest worth control rod stuck fully withdrawn (Ref. 3).
Limits on F0 (Z) ensure that the value of the initial total peaking factor assumed in the accident analyses remains valid. Other criteria must also be met (e.g., maximum cladding oxidation, maximum hydrogen generation, coolable geometry, and long term cooling). However, the peak cladding temperature is typically most limiting.
Fo(Z) limits assumed in the LOCA analysis are typically limiting relative to (i.e., lower than) the F0 (Z) limit assumed in safety analyses for other postulated accidents. Therefore, this LCO provides conservative limits for other postulated accidents.
F0 (Z) satisfies Criterion 2 of 10 CFR 50 .36(c)(2)(ii).
Point Beach                                   B 3.2.1-2                  Unit 1 Amendment No. 269 Unit 2 Amendment No. 271

Fo(Z)
B 3.2 .1 BASES LCO                      The Heat Flux Hot Channel Factor, F0 (Z) , shall be limited by the following relationships:
Fa (z) s CFo K (Z)           for P > 0.5 p
Fo(Z) :s; CFo K(z)           forP s 0.5

===0.5 where===
CF 0 is the F0 (Z) limit at RTP provided in the COLR, K(Z) is the normalized F0 (Z) limit as a function of core height provided in the COLR, and p = THERMAL POWER RTP For this facility , the actual values of CF 0 and K(Z) are given in the COLR; however, CF 0 is normally a number on the order of 2.60, and K(Z) is a function that looks like the one provided in Figure B 3.2.1-1.
For Constant Axial Offset Control operation , F0 (Z) is approximated by Fca(Z) and Fwa(Z) . Thus, both f C0 (Z) and f W0 (Z) must meet the preceding limits on Fa(Z).                                    core power distribution Information An Fca(Z) eva luation requires obtaining an inooFe flu)( l'l'lop in MODE 1.
  ! wh ich a From the ir?core flu)c 1'19ap results we obtain the measured value (Frt.i(Z))
of F0 (Z). Then ,
is obta ined .
FS(Z) = F~ (Z) 1.08 If t he core power distribution informat ion Is obta ined with t he Power Distribution Mon itoring             e 1.08 is a factor that accounts for fuel manufacturing tolerances System (P DMS); t hen:
flux map measurement uncertainty.
Fcu(Z} = FMo(Z) X 1.03 X [ 1.0 +
(Uo/100]                                                cellent approximation for F 0 (Z) when the reactor is at the ower at which the
* was taken .
where 1.03 accounts fo r fuel ma nufacturing t olerances and U0 acco unts for PDMS uncertainty as    The expression for F~ (Z) is: F~(Z) = F~ (Z) (W(Z)foLR Rj described in Equation 5-19 of p
Reference 7.
If the core powe r distri bution information is obtained from a fl ux map us ing the movable incore detector syste m; Point Beach                                       B 3.2.1-3                 Unit 1 - Amendment No. 269 Unit 2 - Amendment No . 271

Fo(Z)
B 3.2.1 BASES SURVEILLANCE            SR 3.2.1.1 REQUIREMENTS Verification that FS(Z) is within its specified limits involves increasing FM 0 (Z) to allow for manufacturing tolerance and measurement uncertainties in order to obtain Fca . Speeifioally , rM    0 (Z) is the as described in the preceedi ng LCO sect ion.
r ~(Z)     FtAfZ) 1.08 (Re( 4). FS(Z) is then compared to its specified limits.
The limit with which Fc0 (Z) is compared varies inversely with power above 50% RTP and directly with a function called K(Z) provided in the COLR.
Performing this Surveillance in MODE 1 prior to exceeding 75% RTP following a refueling outage ensures that some determination of Foc(z) is made prior to achievi ng a significant power level where the peak linear heat rate could approach the limits assumed in the safety analyses .
If THERMAL POWER has been increased by ~ 10% RTP since the initial or most recent determination of Fca(z), another evaluation of this factor is required 24 hours after achieving equilibrium conditions at this higher power level (to ensure that Fc0 (Z) values are being reduced sufficiently with power increase to stay within the LCO limits).
Equilibrium cond itions are achieved when the core is sufficiently stable at the intended operating conditions requ ired to perform the Surveillance.
The allowance of up to 24 hours after achieving equilibrium conditions at the increased THERMAL POWER level to complete the next Foc(z) surveillance applies to situations where the Foc(z) has already been measured at least once at a reduced THERMAL POWER level. The observed margin in the previous surve illance will provide assurance that increasing power up to the next plateau will not exceed the Fo limit, and that the core is behaving as designed.
This Frequency condition is not intended to require verification of these parameters after every 10% increase in RTP above the THERMAL POWER at which the last verification was performed . It only requires verification after a power level is achieved for extended operation that is 10% higher than the TH ERMAL POWER at which F0 C(Z) was last measured .
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Point Beach                                   B 3.2.1-10               Unit 1 - Amendment No. 269 Unit 2 - Amendment No. 271

Fo(Z)
B 3.2.1 BASES SURVEILLANCE                 SR 3.2.1.2 REQUIREMENTS (continued)                  The nuclear design process includes calculations performed to determine that the core can be operated within the F0 (Z) limits.
Bee se f;lwc R'l  ps me ta lrnn in steady state conditions, the variations co re power distribu tion information ls ta ken at or near 1-----~
in po er distribution resulting from normal operational maneuvers are not gresent in th_ flu>c map data . These variations are , however, cor stffe f\Jativel y ~ culated by considering a wide range of unit
!measureme nts 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)COLR_ Multiplying the measured total peaking factor, FC0 (Z), by W(Z)COLR gives the maximum F0 (Z) calculated to occur in normal operation, F~(Z).
The limit with which Fw0 (Z) is compared varies inversely with power above 50% RTP and directly with the function K(Z) provided in the COLR.
The W(Z)C 0 LR factors are provided in the COLR for discrete core elevations. Flux map data are typically taken for 30 to 75 core elevations. F~(Z) evaluations are not applicable for Westinghouse WCAP-17661-P-A, Revision 1, Improved RAOC and CAOC Fo Surveillance Technical Specifications, February 2019.
These regions of the core are 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 (usually the top and bottom 10% or 15%).
SR 3.2.1.2 requires a surveillance of Fow(z) during the initial startup following each refueling with in 24 hours after exceeding 70% RTP .
THERMAL POWER levels below 70% are typically non-limiting with respect to the limit for Fow(z) . Furthermore, startup physics testing and flux symmetry measurements, also performed at low power, provide confi rmation that the core is operating as expected . Th is frequency ensures that verification of F0 W(Z) is performed prior to extended operation at power levels where the maximum permitted peak LHR cou ld be challenged and that the first required performance of SR 3.2 .1.2 after a refue li ng is performed at a power level high enough to provide a high level of confidence in the accuracy of the Surveillance result.
Point Beach                                           B 3.2.1-11              Unit 1 - Amendment No. 269 Unit 2 - Amendment No. 271

Fo(Z)
B 3.2 .1 BASES SURVEILLANCE Equilibrium conditions are ach ieved when the core is sufficiently stable REQUIREMENTS at the intended operating conditions requ ired to perform the (continued)  Surveillance.
If a previous Surveillance of Fow(z) was performed at part power conditions, SR 3.2.1.2 also requ ires that F0 W(Z) be verified at power levels > 10% RTP above the TH ERMAL POWER of its last verification with in 24 hours after achieving equil ibrium cond itions. This ensures that Fow(z) is with in its limit using radial pea king factors measured at the higher power level. The allowance of up to 24 hours after achieving equi librium conditions wi ll provide a more accu rate measurement of F0 W(Z) by allowing sufficient time to achieve equi librium conditions and obtain the power distri bution measurement.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES   1. 10CFR50.46,1974.
: 2. FSAR, Section 14.2.6.
: 3. FSAR, Chapter 3.
: 4. WCAP - 7308-L-P-A, "Evaluation of Nuclear Hot Channel Factor Uncertainties, " June 1988.
: 5. WCAP-8403 (nonproprietary), Power Distribution Control and Load Following Procedures, Westinghouse Electric Corporation ,
September 1974.
: 6. Westinghouse WCAP-17661 -P-A, Revision 1, Improved RAOC and CAOC Fa Surveillance Technica l Specifications, February 2019.
: 7. WCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System," August 1994.
Point Beach                           B 3.2.1-12                       Unit 1 - Amendment No. 269 Unit 2 - Amendment No. 271

F~H B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FiH)
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 Core power distribution                                     on local (pellet) and integrated fuel rod peak power density are information , as referenced                                 expressed in terms of hot channel factors. Control of the core power below, can be provided by                                   distribution with respect to these factors ensures that local conditions in the power distribution                                       the fuel rods and coolant channels do not challenge core integrity at monitoring system
* during either normal operation or a postulated accident (PDMS) or by the movable                                    analyzed in the safe y a incore detector system FiH is defined as the ratio of the integral of the linear power along the (MIDS).
fuel rod with the highest integrated power to the average integrated fuel rod power. Therefore, FiH is a measure of the maximum total power produced in a fuel rod .
FiH is sensitive to fuel loading patterns, bank insertion, and fuel burnup. FiH typically increases with control bank insertion and typically decreases with fuel burnup.
                                                                                                                    ~                    informati on is FiH is not directly measurable but is inferred from a power distrib ,on i.:.:.nli::.::fo::..:.
rm   :..:::a:::;tia:::._:n_ _ _ ___J----:)~ map obtained with the mo*,*able inoore detester system . Specifi      ly, the results of tho three dimensional power distribution map are analyzed by a computer to determine FiH. 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 basis value of the departure from nucleate boiling (DNB) is met for normal operation, operational transients, and any transient condition arising from events of moderate frequency. The DNB design basis precludes DNB and is met by limiting the minimum local DNB heat flux ratio to 1.3 using the WRB-1 or W-3 CHF correlations. All DNB limited transient events are assumed to begin with an FiH value that satisfies the LCO requirements. Operation outside the LCO limits may produce unacceptable consequences if a DNB limiting event occurs. The DNB design basis ensures that there is no overheating of the fuel that results Point Beach                                      B 3.2.2-1                             December, 2019

F~H B 3.2.2 BASES APPLICABLE         The fuel is protected in part by Technical Specifications , which ensure SAFETY ANALYSES    that the initial conditions assumed in the safety and accident analyses (continued)        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 3.1.6, "Control Bank Insertion Limits,"
LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor (FiH)," and core power distribution LCO 3.2 .1, "Heat Flux Hot Channel Factor (F0 (Z))."
information FiH and F0 (Z) are measured periodically using the A9ovable ineore detector Sylsteffl . Measurements are generally taken with the core at, or near, steady state cond itions. 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 .
Fi H satisfies Criterion 2 of the NRC Policy Statement.
LCO                FiH shall be maintained within the limits of the relationship provided in the COLR.
The FiH 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 a DNB.
The limiting value of Fi H, described by the equation contained in the COLR, is the design radial peaking factor used in the unit safety analyses.
A power multiplication factor in this equation includes an additional margin for higher radial peaking from reduced thermal feedback and greater control rod insertion at low power levels. The limiting value of FiH is allowed to increase 0.3% for every 1% RTP reduction in THERMAL POWER.
APPLICABILITY       The FiH limits must be maintained in MODE 1 to preclude core power distributions from exceeding the fuel design limits for DNBR and PCT.
Applicability in other modes is not required because there is either insufficient stored energy in the fuel or insufficient energy being transferred to the coolant to require a limit on the distribution of core power. Specifically, the design bases events that are sensitive to FiH in other modes (MODES 2 through 5) have significant margin to DNB, and therefore , there is no need to restrict FiH in these modes.
Point Beach                               B 3.2 .2-3               Unit 1 - Amendment No. 201 Unit 2 - Amendment No. 206

F~H B 3.2.2 BASES ACTIONS (continued) The allowed Completion Time of 72 hours 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.
perfomed Once the power level has been reduced to< 50%                per Required Action A.1.2.1, an inoore f lux map (SR 3.2.2.1 1 must 60 obta ined and the measured value of FiH verified not to exceed the allowed limit at the lower power level. The unit is provided 20 additional hours to perform this task over and above the 4 hours allowed by either Action A.1.1 or Action A.1.2.1. The Completion Time of 24 hours is acceptable because of the increase in the DNB margin, which is obtained at lower power levels, and the low probability of having a DNB core power distribution information                 limiting event within this 24 hour period. Additionally, operating ex
* e has indicated that this Completion Time is sufficient to obtain t e ineore f lu)( map , perform the required calculations, and evaluate FiH .
Verification that FiH is within its specified limits after an out of limit occurrence ensures that the cause that led to the FiH exceeding its limit is corrected, and that subsequent operation proceeds within the LCO limit. This Action demonstrates that the FiH limit is within the LCO limits prior to exceeding 50% RTP, again prior to exceeding 75% RTP, and within 24 hours after THERMAL POWER is ~ 95% RTP.
This Required Action is modified by a Note that states that THERMAL POWER does not have to be reduced prior to performing this Action .
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 MODE 2 within 6 hours. The allowed Completion Time of 6 hours 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.
Point Beach                               B 3.2.2-5                 Unit 1 - Amendment No. 201 Unit 2 - Amendment No. 206

F~H B 3.2.2 BASES obtaining core powe r distribution information SURVEILLANCE               SR 3.2.2.1 REQUIREMENTS The value of F~H is determined by using the movable ineore detector system to obtain a fl u>< distribution map . A data reduction computer information                         rogram then calculates the maximum value of F~H from the measured i.:.:.::;=::.::.::.:.:_ _ _ ___J-~1'1~
w~cS d;-::
is;;tr;1;u~1~0..::;,:: +Re measured value of Fi H must be multiplied by 1.0             ccount for measurement uncertainty before making The measured value of ~N l!H must             parisons to the FiH limit.
be mult iplied by a measurement uncerta inty factor.
After each refueling, FiH must be determined in MODE 1 prior to If t he core power distribution         exceed ing 75% RTP. This requirement ensures that FiH limits are met Informa tion Is obtained wit h th e at the beginning of each fuel cycle.
Power Distribut ion Mon itoring System (P DMS), t he measured val ue of FNAH must be mult iplied      The Surveillance Frequency is controlled under the Surveillance by the factor [1.0 + (ULH/100)]         Frequency Control Program .
where ULH accounts for PDMS uncertain t y determined as described in Equation 5-19 of Refe rnece 4.
If the power distribut ion
: 1. FSAR, Section 14.2.6.
information is obta ined from a fl ux map using the movable             2. FSAR, Chapter 3.
incore detector syste m, t he
: 3. 10 CFR 50.46 .
: 4. WCAP-12472-P-A, "BEACON Co re Monitoring and Operations Support Syste m," August 1994.
Point Beach                                                   B 3.2.2-6                    Unit 1 - Amendment No. 253 Unit 2 - Amendment No. 257

AFD B 3.2.3 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.3 AXIAL FLUX DIFFERENCE (AFD)
BASES BACKGROUND            The purpose of this LCO is to establish limits on the values of the AFD in order to limit the amount of axial power distribution skewing to either the top or bottom of the core. By limiting the amount of power Core power distribution      distribution skewing, core peaking factors are consistent with the information , as referenced  assumptions used in the safety analyses. Limiting power distribution below, can be provided by skewing over time also minimizes the xenon distribution skewing, which the power distribution      L
                            ~i:s~a~s=ig~n~if:ic:a~n~t ~
f a~c:to:r~in~ax~i~
al'.2p~o~w~e~r.;:d~is~tr~ib::!!:u:!!!ti~o:J..U,.u.u.~"7' monitoring system            The operating scheme used to control the axial power distribution, (PDMS) or by the movable     Constant Axial Offset Control (CAOC), involves maintaining the AFD incore detector system       within a tolerance band around a burnup dependent target, known as (MIDS).                      the target flux difference, to minimize the variation of the axial peaking factor and axial xenon distribution during unit maneuvers.
The target flux difference is determined at equilibrium xenon conditions.
The control banks must be positioned within the core in accordance with their insertion limits and Control Bank D should be inserted near its normal position (i.e.,~ 220 steps withdrawn) for steady state operation at high power levels. The power level should be as near RTP as practical. The value of the target flux difference obtained under these conditions divided by the Fraction of RTP is the target flux difference at RTP for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RTP value by the appropriate fractional THERMAL POWER level.
The AFD is monitored on an automatic basis using the unit process computer that has an AFD monitor alarm. The frequency of monitoring the AFD by the computer is once per minute providing an essentially continuous accumulation of penalty deviation time that allows the operator to assess the status of the penalty deviation time. The computer determines the 1 minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFDs for two or more OPERABLE excore channels are outside the target band and the THERMAL POWER IS> 90% RTP. During operation at THERMAL POWER levels< 90% RTP but> 15% RTP, the computer sends an alarm message when the cumulative penalty deviation time is > 1 hour in the previous 24 hours.
Periodic updating of the target flux difference value is necessary to follow the change of the flux difference at steady state conditions with burnup. The Nuclear Enthalpy Rise Hot Channel Factor F11H and QPTR LOCs limit the radial component of the peaking factors.
Point Beach                                            B 3.2.3-1                                       Unit 1 - Amendment No. 241 Unit 2 - Amendment No. 245

AFD B 3.2.3 BASES LCO (continued) Accordingly, while THERMAL POWER is :2:: 50% RTP and < 90% RTP (i.e., Part b of the LCO), a 1 hour cumulative penalty deviation time limit, cumulative during the preceding 24 hours, is allowed during which the unit may be operated outside of the target band but within the acceptable operation limits provided in the Unit 2 COLR. Note 2 allows this penalty time to accumulate at the rate of 1 minute for each 1 minute of operating time within the power range of Part b of this LCO (i.e.,
THERMAL POWER;:;: 50% RTP). The cumulative penalty time is the sum of penalty times from Parts b and c of this LCO.
For THERMAL POWER levels > 15% RTP and < 50% RTP (i.e., Part c of this LCO), deviations of the AFD outside of the target band are less significant. Note 3 allows the accumulation of 0.5 minute penalty deviation time per 1 minute of actual time outside the target band and reflect this reduced significance. With THERMAL POWER < 15% RTP, AFD is not a significant parameter in the assumptions used in the safety analysis and, therefore, requires no limits. Because the xenon distribution produced at THERMAL POWER levels less than RTP does affect the power distribution as power is increased, unanalyzed xenon and power distribution is prevented by limiting the accumulated penalty deviation time.
For surveillance of the power range channels performed according to SR 3.3.1.6, Note 4 allows deviation outside the target band for 16 hours and no penalty deviation time accumulated. Some deviation in the AFD is required for doing the NIS calibration with the incore detector system.
This calibration is performed every 92 days.
core powe r distribution information .
APPLICABILITY   The AFD requirements are applicable in MODE 1 greater than 15% RTP. Above 50% RTP, the combination of THERMAL POWER and core peaking factors are the core parameters of primary importance in safety analyses (Ref 1).
Between 15% RTP and 90% RTP, this LCO is applicable to ensure that the distributions of xenon are consistent with safety analysis assumptions.
At or below 15% RTP and for lower operating MODES, the stored energy in the fuel and the energy being transferred to the reactor coolant are low. The value of the AFD in these conditions does not affect the consequences of the design basis events.
Low signal levels in the excore channels may preclude obtaining valid AFD signals below 15% RTP.
Point Beach                            B 3.2.3-4              Unit 1 - Amendment No. 241 Unit 2 - Amendment No. 245

AFD B 3.2.3 BASES w it h core pow er dist ribution SURVEILLANCE SR 3.2.3.3                               information REQUIREMENTS (continued)  Measurement of the target flux difference is accomplished 1::W-i:i::ffillffi.l--i::1-flux ma!:) when the core is at equilibrium xenon conditions, preferably at high power levels with the control banks nearly withdrawn. This ~
ma-1:)-provides the equilibrium xenon axial power distribution from which the target value can be determined. The target flux difference varies slowly with core burnup.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
A Note modifies this SR to allow the predicted end of cycle AFD from the cycle nuclear design to be used to determine the initial target flux difference after each refueling.
REFERENCES   1. WCAP-8403 (nonproprietary), Power Distribution Control and Load Following Procedures, Westinghouse Electric Corporation, September 1974.
: 2. NS-TMA-2198, Westinghouse to NRC Letter, Operation and Safety Analysis Aspects of Improved Load Follow Package, January 31, 1980.
: 3. NS-CE-687, Westinghouse to NRC Letter, Power Distribution Control Analysis, July 16, 1975.
: 4. FSAR, Chapter 14.
Point Beach                       B 3.2.3-7                    Unit 1 -Amendment No. 253 Unit 2 - Amendment No. 257

QPTR B 3.2.4 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.4 QUADRANT POWER TILT RATIO (QPTR)
BASES BACKGROUND        The QPTR limit ensures that the gross radial power distribution remains consistent with the design values used in the safety analyses. Precise radial power distribution measurements are made during startup Core power distribution    testing, after refueling, and periodically during power dperation.
information, as referenced The power density at any point in the core must be limited so that the below, can be provided by fuel design criteria are maintained . Together, LCO 3.2.3, "AXIAL FLUX the power distribution     DIFFERENCE (AFD)," LCO 3.2.4, and LCO 3.1.6, "Control Rod monitoring system                 ion Limits," provide limits on process variables that characterize (PDMS) or by the movable   and con       the three dimensional power distribution of the reactor core .
incore detector system     Control of thes       riables ensures that the core operates within the fuel (MIDS).                    design criteria and tn       e P.OWer distribution remains within the bounds used in the safety analyse .
APPLICABLE         This LCO precludes core power distributions that violate the following SAFETY ANALYSES    fuel design criteria:
: a. During a large break loss of coolant accident, the peak cladding temperature must not exceed 2200&deg; F (Ref. 1);
: b. During a loss of forced reactor coolant flow accident, there must be at least 95% probability at the 95% confidence level (the 95/95 departure from nucleate boiling (DNB) criterion) that the hot fuel rod in the core does not experience a DNB condition;
: c. During an ejected rod accident, the energy deposition to the fuel must not exceed 225 cal/gm for unirradiated and 200 cal/gm for irradiated fuel (Ref. 2); and
: d. The control rods must be capable of shutting down the reactor with a minimum required SOM with the highest worth control rod stuck fully withdrawn (Ref. 3).
The LCO limits on the AFD, the QPTR, the Heat Flux Hot Channel Factor (F 0 (Z)), the Nuclear Enthalpy Rise Hot Channel Factor (F~H),
and control bank insertion are established to preclude core power distributions that exceed the safety analyses limits.
Point Beach                              B 3.2.4-1                  Unit 1 - Amendment No. 201 Unit 2 - Amendment No. 206

QPTR B 3.2.4 BASES ACTIONS (continued) A.2 After completion of Required Actlon A.1 , the QPTR alarm may still be in its alarmed state. As such , any additional changes in the QPTR are detected by requiring a check of the QPTR once per 12 hours thereafter. A 12 hour Completion Time is sufficient because any additional change in QPTR would be relatively slow.
A.3 The peaking factors F~H and F0 (Z) , as approximated by F?:i(Z) and F~(Z), are of primary importance in ensurlng that the power distribution remains consistent with the initial conditions used in the safety obtain core power              analyses. Performing SRs on F~H and F0 (Z) within the Completion distribution information t---   Jime of 24 hours after achieving equilibrium conditions from a thermal powe          tion per Required Action A.1 ensures that these primary indicators of pow
* ibution are within their respective limits.
Equilibrium conditions are ac
* hen the core is sufficiently stable at intended operating conditions to support flu x Fl'lBppiRg. A Completion Time of 24 hours after achieving equilibrium conditions from a thermal power reduction per Required Action A.1 takes into consideration the obtain core power               rate at which peaking factors are likely to change , and the time required distribution information        to stabilize t e pan a perfo rm a flu x 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 F~H and F0 (Z) with changes in power distribution . Relatively small changes are expected due to either burn up and xenon redistribution or correction of the cause for exceeding the QPTR limit.
A.4 Although F~H and F0 (Z) are of primary importance as initial conditions in the safety analyses, other changes in the power distribution may occur as the QPTR limit is exceeded and may have an impact on the validity of the safety analysis. A change in the power distribution can affect such reactor parameters as bank worths and peaking factors for rod malfunction accidents. 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 gross radial power distribution that requires an investigation and evaluation that is accomplished by examining the -iRcore power distribution. Specifically , the core peaking factors and the quadrant tilt must be evaluated because they are the factors that best characterize the core power distribution.
Point Beach                               B 3.2.4-3               Unit 1 - Amendment No. 201 Unit 2 - Amendment No. 206

QPTR B 3.2.4 BASES ACTIONS (continued) This re-evaluation is required to ensure that, before increasing THERMAL POWER to above the limit of Required Action A.1, the reactor core conditions are consistent with the assumptions in the safety analyses.
A.5 If the QPTR has exceeded the 1 .02 limit and a re-evaluation of the safety analysis is completed and shows that safety requirements are met, the excore detectors are normalized to restore QPTR to within limits prior to increasing THERMAL POWER to above the limit of Required Action A.1. Normalization is accomplished in such a manner that the indicated QPTR following normalization is near 1.00. This is done to detect any subsequent significant changes in QPTR.
Required Action A.5 is modified by two Notes. Note 1 states that the OPT is not restored to within limits until after the re-evaluation of the safety analysis has determined that core conditions at RTP are within the safety analysis assumptions (i.e., Required Action A.4). Note 2 states that if required Action A.5 is performed, then required Action A.6 shall be performed. Required Action A.5 normalizes the excore detectors to restore QPTR to within limits, which restores compliance with LCO 3.2.4. Thus, Note 2 prevents exiting the Actions prior to I verifying   ----)     oomp leting J:lu:>< mapping to &#xa5;erify peaking factors, per Required Action A.6 . These Notes are intended to prevent any ambiguity about the required sequence of actions.
Once the flux tilt is restored to within limits (i.e., Required Action A.5 is performed), it is acceptable to return to full power operation. However, as an added check that the core power distribution is consistent with the safety analysis assumptions, Required Action A.6 requires verification that F0 (Z), as approximated by FS(Z) and Fi(z), and F~H are within their specified limits within 24 hours of achieving equilibrium conditions at RTP. As an added precaution, if the core power does not reach equilibrium conditions at RTP within 24 hours, but is increased slowly, then the peaking factor surveillances must be performed within 48 hours after increasing thermal power above the limit of Required Action A.1. These Completion Times are intended to allow adequate time to increase THERMAL POWER to above the limit of Required Action A.1, while not permitting the core to remain with unconfirmed power distributions for extended periods of time.
Point Beach                               B 3.2.4-4                  Unit 1 - Amendment No. 201 Unit 2 - Amendment No. 206

QPTR B 3.2.4 BASES SURVEILLANCE        With an NIS power range channel inoperable, tilt monitoring for a REQUIREMENTS        portion of the reactor core becomes degraded. Large tilts are likely (contlnued)         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.
For purposes of monitoring the QPTR when one power range channel either the Power Distribution Is inoperab te, the moveable incore detectors are used to confirm that System (PDMS) or              the normalized symmetric power distribution is consistent with the indicated QPTR and any previous data indicating a tilt. The incore detector monitoring is performed with a full incore flux map.
With one NIS channel inoperable, the indicated tilt may be changed from the value indicated with all four channels OPERABLE. To confirm that no change in tilt has actually occurred, which might cause the QPTR limit to be exceeded , the incore result may be compared against previous flux maps. Nominally, quadrant tilt from the Surveillance should be within 2% of the tilt shown by the most recent flux map data.
REFERENCES          1. 10 CFR 50.46.
: 2. FSAR, Section 14.2.6.
: 3. FSAR, Chapter 3.
Point Beach                              B 3.2.4-6               Unit 1 - Amendment No. 253 Unit 2 - Amendment No. 257

RPS Instrumentation B 3.3.1 BASES SURVEILLANCE          The SRs for each RPS Function are identified by the SRs column of REQUIREMENTS          Table 3.3.1-1 for that Function.
A Note has been added to the SR Table stating that Table 3.3 .1-1 determines which SRs apply to which RPS Functions .
Note that each channel of process protection supplies both trains of the RPS. When testing Channel I, Train A and Train B must be examined.
~ - - - - - - - - - - Similarly, Train A and Train B must be examined when testing Core power distribution     Channel 11, Channel Ill, and Channel IV (if applicable). The CHANNEL information , as referenced CALIBRATION and COTs are performed in a manner that is consistent below, can be provided by with the assumptions used in analytically calculating the required the power distribution      channel accuracies .
morn'tonng
        . sys tem SR    3311 (PDMS) or by the movable           * *
* incore detector system      Performance of the CHANNEL CHECK ensures that gross failure of (MIDS).                     instrumentation has not occurred. A CHANNEL CHECK is normally a
- - - - - - - - - - - comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the unit staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program .
SR 3.3.1.2 SR 3.3.1.2 compares the calorimetric heat balance calculation to the NIS channel output. If the calorimetric exceeds the NIS channel output by> 2% RTP, the NIS is not declared inoperable, but must be adjusted.
If the NIS channel output cannot be properly adjusted, the channel is declared inoperable.
Point Beach                                B 3.3.1-42               Unit 1 - Amendment No. 253 Unit 2 - Amendment No. 257

RPS Instrumentation B 3.3 .1 BASES SURVEILLANCE            Two Notes mod ify SR 3.3 .1.2. The first Note indicates that th e NIS REQUIREMENTS            channel output shall be adjusted consistent with the calorimetric results
( continued)            if the absolute difference between the N IS channel output and the calorimetric is > 2% RTP . The second Note clarifies that this Surveillance is requ ired only if reactor power is z 15% RTP and that 12 hour 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 .
    !core power distribution information SR 3.3.1 .3 compares the inoore system to the NIS channel output.
a.a.1.a is performed by means of the mm,ceable ineore detection system . If the absolute difference is z 3%, the NIS channel is still OPERABLE , but must be readjusted.          core power distribution information If the NIS channel cannot be properly readjusted, the chann I is declared inoperable. This Surveillance is performed to verify e f(.6.I) input to the overtemperature .6.T Function.
Two Notes modify SR 3.3.1 .3. Note 1 indicates that the excore channel shall be adjusted if the absolute difference between the i-Reere and excore AFD is z 3% .
Note 2 clarifies that the Surveillance is required only if reactor power is z 50%    RTP and that 24 hours is allowed for performing the first Surveillance after reaching 50% RTP.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3 .1.4 SR 3.3.1.4 is the performance of a TADOT. This test shall verify OPERABILITY by actuation of the end devices.
Point Beach                                 B 3.3.1-43                Un it 1 - Amendment No. 253 Unit 2 - Amendment No. 257

RPS Instrumentation B 3.3.1 BASES SURVEILLANCE The RTB test shall include separate verification of the undervoltage and REQUIREMENTS shunt trip mechanisms. The independent test for bypass breakers is
{continued)  included in SR 3.3.1.13. The bypass breaker test shall include an undervoltage trip. A Note has been added to SR 3.3.1.4 to indicate that this test must be performed on the bypass breaker prior to placing it in service.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.1.5 SR 3.3.1.5 is the performance of an ACTUATION LOGIC TEST. The train being tested is placed in the bypass condition , thus preventing inadvertent actuation. All possible logic combinations, with and without applicable permissives, are tested for each protection function. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3 .1.5 is modified by two Notes. Note 1 provides an 8 hour delay in the requirement to perform this Surveillance for the Source Range Neutron Flux trip function instrumentation when power is reduced to below P-6. This Note allows a normal shutdown to proceed without a delay for testing in MODE 2 and for a short time in MODE 3 until the RTBs are open and SR 3.3.1 .5 is no longer requ ired to be performed . If the unit is to be in MODE 2 below P-6 for > 8 hours, this Surveillance must be performed prior to 8 hours after reducing power below P-6.
Note 2 excludes the RCP Breaker Position (Two Loop), Reactor Coolant Flow-Low (Two Loop) and Underfrequency Bus A01 and A02 Trip Functions, and the P-6, P-7, P-8, P-9 and P-10 Interlocks. These functions/interlocks are tested via SR 3.3.1.15.
SR 3.3.1.6          core power distribution information SR 3.3.1.6 is a calibration of the excore channels to the inoore ohannels. If the measurements do not agree , the excore channels are not declared inoperable but must be calibrated to agree with the ffi69fe deteotor measurements . If the excore channels cannot be adjuste , the channels are declared inoperable. This Surveillance is performed t verify the f(6.I) input to the overtemperature 6-T Function.
core power distribution information Point Beach                        B 3.3.1-44              Unit 1 - Amendment No. 253 Unit 2 - Amendment No. 257

RPS Instrumentation B 3.3.1 BASES SR 3.3.1.10 is modified by two Notes as identified in Table 3.3.1-1. The SURVEILLANCE     first Note requires evaluation of channel performance for the condition REQUIREMENTS     where the as-found setting for the channel setpoint is outside its (continued)      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. The performance of these channels will be evaluated under the station's Corrective Action Program. Entry into the Corrective Action Program will ensure required review and documentation of the condition to establish a reasonable expectation for continued OPERABILITY. 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, 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.
co re power distribution SR 3.3.1.11 information
                               . 1.11 is the performance of a CHANNEL CALIBRATION, as describe         R 3.3.1.10. This SR is modified by a Note stating that neutron detector         excluded from the CHANNEL CALIBRATION .
The CHANNEL CALIB                         power range neutron detectors consists of a n                                          a power calorimetric and Hl::H~l-at3~~H'fffee-ar::~e-.:i~~H-&#xa5;. The CHANNEL CALIBRATION for the source range and intermediate range neutron detectors consists of obtaining the detector plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to the manufacturer's data. This Surveillance is not required for the NIS power range detectors for entry into MODE 2 or 1, and is not required for the NIS intermediate range detectors for entry into MODE 2, because the unit must be in at least MODE 2 to perform the test for the intermediate range detectors and MODE 1 for the power range detectors. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Point Beach                           B 3.3.1-48               Unit 1 - Amendment No. 253 Unit 2 - Amendment No. 257

Point Beach Nuclear Plant, Units 1 and 2                                    NRC 2022-0025 Docket Nos. 50-266 and 50-301                                                   Enclosure Page 48 of 53 Attachment 5 Marked-up Technical Requirements Manual Pages (Information only)
(5 Pages Follow)

POINT BEACH NUCLEAR PLANT                                                             TRMINDEX TECHNI_CAL REQUIREMENTS MANUAL                                                        Revision 133 April 11 , 2022 TABLE OF CONTENTS TABLE OF CONTENTS SECTION         TITLE                                                                                                        REVISION TRM 1.0         Technical Requirements Manual Content and Organization .. ...... ........... 3 TRM 2.0         Reports TRM 2.1      Core Operating Limits Report (COLR) Unit 1 Cycle 41 ......................... ..... .. 23 TRM 2 .1     Core Operating Limits Report (COLR) Unit 2 Cycle 39 ....... ....... .. .... ..... ....... 22 TRM 2.2      Pressure Temperature Limits Report .. .... .. .. .. ... ....... .......... ...................... .. ... 12 TRM 2.3      Previous Cycle COLR Unit 1 Cycle 28 ...... ..... .... .......... .. .. ...... .... .... ...... CANCELED TRM 3.0         Use, Application, TLCO Applicability and TSR Applicability ........ ... ..... .. 1 TRM 3.1.1    Control Rod Worth .... ... ........ ..................... ... ...................... .... .. .... ................. 1 TRM 3.2.1    Movable In-Core Instrumentation ...... ... .... ....... .......... ...... ...... .... ....... .... .. ...... 1 ITRM 3.2.2 Power Distribution Monitoring System              I TRM 3.3.1    Instrumentation .... : ... ........ .............. ......... .................. ......... ....... .. .. ............... 3 TRM 3.3.2    Leading Edge Flow Meter (LEFM) (Unit 1 and Unit 2) ........................... .... ... 6 TRM 3.4.1    Reactor Coolant Activity ... ... ............... ... ........ ....... ... .. .. ..... .. ............. ... .......... 2 TRM 3.4.2    Pressurizer Temperature Limits ........... ........... ... .. ..... ... ... .............................. O TRM 3.4.3    Primary System Integrity .................. ... .... .......... ... ............................... CANCELED TRM 3.4.4    Reactor Coolant Gas Vent System ... ...... .... .............. ................ ......... .. ..... .. .. 3 TRM 3.4.5    Reactor Coolant Oxygen, Chloride And Fluoride Concentrations ................. 1 TRM 3.5.1    Chemical And Volume Control System (Unit 1 and Unit 2) ...... .. ... .... ....... .. .. 12 TRM 3.7.1    Fuel Storage .......... ................. .. ..... ............ .............. ............ .... ....... .... ... .. ..... O TRM 3.7.2    Shock Suppressors (Snubbers) .. ..... ...... .................. ............. .. .............. ........ 2 TRM 3.7.3    Steam Generator Pressure And Temperature (PIT) Limits ..... .. .................... 2 TRM 3.7.4    Sealed Radioactive Sources ............................... .. ....... ... ..... ... .................... .0 TRM 3.7.5    Auxiliary Building Crane Lifting Devices ............ ... .. .. .. ........... .. .. ................ .. . 1 Page 1 of 3

Movable In-Core Instrumentation TRM 3.2.1 BASES BACKGROUND      Custom Technical Specification (CTS) 15.3.11, Movable In-core Instrumentation, was identified for relocation to the Technical Requirements Manual (TRM) during the conversion to Improved Technical Specifications (Reference 1).
The Movable In-Core Instrumentation System (Reference 2) has four Refer to TRM 3.2.2       drives, four detectors, and 36 thimbles in the core. The A and B for movable incore       detectors can be routed to eighteen thimbles. The C and D detectors detector                 can be routed to twenty-seven thimbles. Consequently, the full system requirements for         has a great deal more capability than would be needed for the the BEACON               calibration of the ex-core detectors.
Power Distributio n Monitoring System        To calibrate the excore detectors channels, it is only necessary that the (PDMS) .                 Movable In-Core System be used to determine the gross power distribution in the core as indicated by the power balance between the top and bottom halves of the core.
APPLICABILITY   After the excore system is calibrated initially, recalibration is needed only infrequently to compensate for changes in the core , due for example to fuel depletion, and for changes in the detectors.
ACTIONS If the recalibration is not performed , the mandated power reduction assures safe operation of the reactor since it will compensate for an error of 10% in the excore protection system. Experience at Beznau (Switzerland) and Ginna has shown that drift due to changes in the core or instrument channels is very slight. Thus, the 10% reduction is considered to be very conservative.
SURVEILLANCE    None REQUIREMENTS REFERENCE       1. Improved Technical Specifications Conversion, Section 3.2.2, Description of Change R.1.
: 2. FSAR, Section 7.4.
POINT BEACH TRM                        3.2.1-2                           December 19, 2013

Power Distribution Monitoring System (PDMS)
TRM 3.2.2 TRM 3.2.2 Movable In-core Instrumentation TLCO 3.2.2       The PDMS shall be Functional with the minimum required channels shown in Table 3.2.2-1 .
APPLICABILI TY:  In MODE 1 > 25% RTP.
ACTIONS CONDITION                    REQUIRED ACTION                COMPLETION TIME A. TLCO 3.2.2 not met.         A.1    Suspend the use of the         Immed iately PDMS SURVEILLANCE REQUIREMENTS SURVEILLANCE                                      FREQUENCY TSR 3.2 .2.1    Perform a CHANNEL CHECK.                              31 days TSR 3.2.2 .2   Perform calibration of the PDMS using the              Once after each movable incore detector system with at least           refueling prior to 75% of the detector thimbles and at least 2            THERMAL detector thimbles per quadrant , using the             POWER minimum thermocouple coverage , and with               exceeding THERMAL POWER> 25% RTP .                               75% RTP TSR 3 .2.2 .3  Perform cal ibration of the PDMS using the             31 EFPD with movable incore detector system with at least           minimum 50% of the detector thimbles and at least 2            thermocouple detector thimbles per quadrant , using the             coverage minimum thermocouple coverage , and with THERMAL POWER > 25 % RTP.                              OR 180 EFPD with opt imum thermocouple coverage POINT BEACH TRM                          3.2.2-1                                     Rev. O

Power Distribution Monitoring System (PDMS)
TRM 3.2.2 BASES SURVEILLANCE    For PDMS calibration, the quantity and the coverage distribution of core REQUIREMENT     exit thermocouples used as data input must meet certain criteria. With TSR 3.2.2.3     respect to thermocouple coverage , the available core exit thermocouple coverage can be "optimum" or "minimum" as described below. This criterion affects the TSR Frequency:
Optimum thermocouple coverage satisfies the minimum thermocouple Functionality requirement in Table 3.2.2-1 with the added requirement that the Functional pattern covers all internal fuel assemblies (no face along a baffle) within a chessboard "kn ight move" i. e., a face adjacent and one diagonal square away, or otherwise defined as two locations vertically and one location horizontally OR two locations horizontally and one location vertically .
Minimum thermocouple coverage satisfies thermocouple minimum Functionality requirements of Table 3.2.2-1 but does not meet the "knight move" pattern discussed above.
POINT BEACH TRM                       3.2.2-2                                        Rev. O

Power Distribution Monitoring System (PDMS)
TRM 3.2.2 Table 3.2 .2-1 Power Distribution Monitoring System FUNCTION                                      MINIMUM REQUIRED INPUTS
: 1. Control Bank Position                      4 control banks     (a)
: 2. RCS Cold Leg Temperature T-cold            1 RCS loops     (bl
: 3. Reactor Power Level                        1 (c)
: 4. Power Range Excore Detector Signals
: 5. Core Exit Thermocouple                     10 with _::: 2 per quadrant Temperatures (a)   Determined from either valid demand position or the average of individual IRPI indications. A maximum of 1 rod position indicator may be inoperable when RCCA position indications are being used as input to the PDMS .
(b)   Either narrow range or wide range RTDs.
(c)   Either valid secondary calorimetric, average power range neutron flux power, or average RCS loop !J..T.
(d)   An input is a channel which consists of corresponding upper and lower detector sections.
POINT BEACH TRM                                3.2.2-3                                        Rev. 0}}