Text

Application to Remove Power Distribution Monitoring System (Pdms) Details from Technical Specifications
	ML24145A116
Person / Time
Site:	Braidwood, Byron
Issue date:	05/24/2024
From:	Steinman R; Constellation Energy Generation
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	RS-24-043
	Download: ML24145A116 (1)
	v • d • e

4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office RS-24-043 10 CFR 50.90 May 24, 2024 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. 50-456 and 50-457 Byron Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. 50-454 and 50-455

Subject:

Application to Remove Power Distribution Monitoring System (PDMS) Details from Braidwood Station and Byron Station Technical Specifications

References:

1. NRC Safety Evaluation Report, Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2 - Issuance of Amendments to Technical Specifications for Implementation of the Best Estimate Analyzer for Core Operations Nuclear Power Distribution Monitoring System (TAC Nos. MA8254, MA8255, MA8252, and MA8253), dated February 13, 2001 (ADAMS Accession No. ML010510325)

2. NRC Safety Evaluation Report, Braidwood Station, Units 1 and 2, and Byron Station, Unit Nos. 1 and 2 - Issuance of Amendments Regarding Axial Flux Difference Technical Specifications (EPID L-2018-LLA-0098), dated December 12, 2018 (ADAMS Accession No. ML18302A227)

In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Constellation Energy Generation, LLC (CEG) requests amendments to the Technical Specifications (TS) for Renewed Facility Operating License Nos. NPF-72 and NPF-77 for Braidwood Station, Units 1 and 2 (Braidwood); Renewed Facility Operating License Nos.

NPF-37 and NPF-66 for Byron Station, Units 1 and 2 (Byron).

Braidwood and Byron TS currently reference the Westinghouse power distribution monitoring system known as Best Estimate Analyzer for Core Operations Nuclear (BEACON). The BEACON related TS changes were approved in 2001 (Reference 1). CEG desires to remove extraneous detail related to the specific core monitoring system software and more closely align with the Standard Technical Specifications (STS), NUREG-1431 Revision 5.

May 24, 2024 U.S. Nuclear Regulatory Commission Page 2 The attached request is subdivided as follows:

provides a description and evaluation of the proposed changes.

provides the markup of the affected TS pages for Braidwood.

provides the markup of the affected TS pages for Byron.

provides the markup of the affected TS Bases pages for Braidwood as information only.

provides the markup of the affected TS Bases pages for Byron as information only.

The proposed changes have been reviewed by Plant Operations Review Committees at Braidwood and Byron in accordance with the requirements of the CEG Quality Assurance Program.

CEG requests approval of the proposed license amendment request by April 30, 2025. Once approved, the Braidwood amendments shall be implemented no later than the start of the 2025 fall outage and the Byron amendments shall be implemented no later than the start of the 2026 spring outage. This implementation period will provide adequate time for the affected station documents to be revised using the appropriate change control mechanisms.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (b), CEG is notifying the States of Illinois of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Brian Seawright at (779) 231-6151.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 24th day of May 2024.

Respectfully, Rebecca L. Steinman Senior Manager - Licensing Constellation Energy Generation, LLC

Steinman, Rebecca Lee Digitally signed by Steinman, Rebecca Lee Date: 2024.05.24 11:27:42

-05'00'

May 24, 2024 U.S. Nuclear Regulatory Commission Page 3 Attachments:

1) Evaluation of Proposed Changes

2) Proposed Technical Specifications Changes (Mark-Up) - Braidwood Station, Units 1 and 2

3) Proposed Technical Specifications Changes (Mark-Up) - Byron Station, Units 1 and 2

4) Proposed TS Bases Changes (Mark-Up) - Braidwood Station, Units 1 and 2 - For Information Only

5) Proposed TS Bases Changes (Mark-Up) - Byron Station, Units 1 and 2 - For Information Only cc:

NRC Regional Administrator, Region III NRC Senior Resident Inspector, Braidwood Station NRC Senior Resident Inspector, Byron Station Illinois Emergency Management Agency - Division of Nuclear Safety Evaluation of Proposed Changes Page 1 of 11

Subject:

Application to Remove Power Distribution Monitoring System (PDMS) Details from Braidwood Station and Byron Station Technical Specifications 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

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

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

Page 2 of 11 Evaluation of Proposed Changes 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit or early site permit," Constellation Energy Generation, LLC (CEG) requests amendments to the Technical Specifications (TS) for Renewed Facility Operating License Nos. NPF-72 and NPF-77 for Braidwood Station, Units 1 and 2 (Braidwood) and Renewed Facility Operating License Nos.

NPF-37 and NPF-66 for Byron Station, Units 1 and 2 (Byron).

Braidwood and Byron TS currently reference the Westinghouse power distribution monitoring system known as Best Estimate Analyzer for Core Operations Nuclear (BEACON). The BEACON related TS changes were approved in 2001 (Reference 1). CEG desires to remove extraneous detail related to the specific core monitoring system software and more closely align with the Standard Technical Specifications (STS), NUREG-1431 Revision 5.

2.0 DETAILED DESCRIPTION Braidwood and Byron incorporated BEACON PDMS in 2001 to take advantage of a continuously updated online 3-D nodal model to generate power distribution. Adoption of the BEACON methodology improved the ability to optimize core loading patterns. The PDMS continuously monitors the limiting FQ(Z), FNH, and Departure from Nucleate Boiling Ratio (DNBR) as a replacement to the Axial Flux Difference (AFD) and Quadrant Power Tilt Ratio (QPTR) limits. The implementation of PDMS did not replace, eliminate, or modify existing plant instrumentation. The PDMS software runs on a workstation connected to the plant process computer and combines input from currently installed plant instrumentation and design data generated each fuel cycle.

CEG proposes to remove extraneous detail related to specific core monitoring system software from the TS. Specifically, the proposed changes remove PDMS references in Braidwood and Byron TS in the following Limiting Conditions of Operations (LCO):

LCO 3.1.4 - Rod Group Alignment Limits LCO 3.2.1 - Heat Flux Hot Channel Factor LCO 3.2.2 - Nuclear Enthalpy Rise Hot Channel Factor LCO 3.2.4 - Quadrant Power Tilt Ratio (QPTR)

TS 5.6.5 - Core Operating Limits Report (COLR)

In addition to revising the LCO listed above, CEG proposes to delete LCO 3.2.5 - Departure from Nucleate Boiling Ratio (DNBR).

CEG acknowledges LCO 3.1.7 and LCO 3.2.3 were initially revised as part of the amendments that incorporated the PDMS (Reference 1). The historical LCO 3.1.7 change involved moving Required Action (RA) A.1 and RA B.1 requirements for using movable incore detectors to verify position of the rods with inoperable Digital Rod Position Indication (DRPI) to the TS Bases. This information is being retained in the TS Bases; however, the option to use PDMS as an option to verify rod positions is being removed as part of this proposed amendment.

Evaluation of Proposed Changes As part of the implementation of PDMS, Braidwood and Byron utilized the Relaxed Axial Offset Control (RAOC) methodology for determining AFD as described in WCAP-10216-P-A, Revision 1. Amendments were approved by the NRC in 2018 (Reference 2) to revise LCO 3.2.3 requiring that AFD be maintained within the limits specified in the COLR during MODE 1 with reactor thermal power (RTP) 50 percent (regardless of the status of PDMS).

Since PDMS is not currently referenced in LCO 3.2.3 no changes are necessary.

A detail description of the proposed changes is provided for each section of the TS below.

LCO 3.1.4 - Rod Group Alignment Limits As part of the implementation of the BEACON methodology, the Completion Time (CT) for TS RA B.2 was revised from "2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months " to "2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months from discovery of Condition B concurrent with inoperability of Power Distribution Monitoring System (PDMS)." Similarly, RA C.2 was revised from "6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months " to "6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months from discovery of Condition C concurrent with inoperability of PDMS."

CEG proposes to remove the exception to the normal "time zero" for beginning the allowed outage time from the CT associated with RAs B.2 and C.2 reverting to the previous CT of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , respectively, with no dependence on PDMS operability.

Additionally, RA B.4 is revised to replace the current determine action with a requirement to perform the Surveillance Requirements (SRs) that determine the referenced parameters of Heat Flux Hot Channel Factor and Nuclear Enthalpy Rise Hot Channel Factor. These SRs are SR 3.2.1.1, SR 3.2.1.2, and SR 3.2.2.1. This proposed change aligns the Braidwood and Byron TS with the STS in NUREG-1431.

RA C.3 to restore rod(s) to within alignment limit when PDMS is OPERABLE is deleted. This proposed change also includes the removal of reference to RA C.3 from Condition D.

LCO 3.2.1 - Heat Flux Hot Channel Factor The proposed change adds a Note to Conditions A and B to ensure the respective RAs to perform SRs 3.2.1.1 and 3.2.1.2 are performed during each entry into the Condition.

Proposed change to the CTs for Condition A include added clarification that the RAs need to be performed after each FQC(Z) determination. This is consistent with STS NUREG-1431, LCO 3.2.1B.

RAs B.3, B.4, and B.5 are revised to reflect the reduction of the maximum allowable power of the AFD limits, which is consistent with STS NUREG-1431, LCO 3.2.1B.

The proposed change deletes SR 3.2.1.1 Note 2 and SR 3.2.1.2 Note 3. SR 3.2.1.1 Note 1 and SR 3.2.1.2 Note 1 are moved to the beginning of the section as a single generic note to the SRs section of LCO 3.2.1. SR 3.2.1.2 Note 2 includes an option to increase FQW(Z) by a factor of 1.02. SR 3.2.1.3 and SR 3.2.1.4 are deleted from this LCO. These changes restore the SRs to pre-BEACON implementation wording.

Page 3 of 11 Evaluation of Proposed Changes Page 4 of 11 LCO 3.2.2 - Nuclear Enthalpy Rise Hot Channel Factor LCO 3.2.2 requires FNH to be within the limits specified in the COLR in MODE 1. The current RA A.1 is revised to reflect restoring FNH to within the COLR limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (proposed RA A.1.1), or to reduce thermal power to <50% RTP (proposed RA A.1.2.1) and reduce the Power Range Neutron Flux - High to 55% RTP (proposed RA A.1.2.2). RAs A.2 and A.3 are updated to reflect completion of SR 3.2.2.1.

SR 3.2.2.1 Note and SR 3.2.2.2 are deleted.

LCO 3.2.4 - Quadrant Power Tilt Ratio (QPTR)

Currently, when PDMS is OPERABLE precise radial power distribution measurements are obtained continuously and QPTR limits are not required and requires reverting to the QTPR TS when PDMS is inoperable. This amendment proposes to remove the PDMS dependence from the TS. Accordingly, reference to an inoperable PDMS is deleted from the LCO 3.2.4 Applicability section. RA A.2 is revised to delete the following action, and reduce thermal power 3% from RTP for each 1% of QPTR >1.00. CTs A.4, A.5, and A.6 contain editorial changes in specification of the limit of RA A.1. These changes are consistent with NUREG-1431.

SR 3.2.4.1 Note 3 and SR 3.2.4.2 Note 2 are deleted. SR 3.2.4.1 and 3.2.4.2 are revised to remove the explicit numerical QPTR limit since it is defined in the LCO. These proposed changes to the SR are consistent with NUREG-1431.

LCO 3.2.5 - Departure from Nucleate Boiling Ratio (DNBR)

LCO 3.2.5 is deleted from the TS. LCO 3.2.5 was initially incorporated with implementation of PDMS in 2001 (Reference 1); however, with the removal of PDMS from the TS this LCO is no longer needed. This proposed change is consistent with NUREG-1431.

TS 5.6.5 - Core Operating Limits Report (COLR)

LCO 3.2.5, Departure from Nucleate Boiling Ratio (DNBR) is deleted from the list of LCOs in Section 5.6.5.a. WCAP-12472-P-A is deleted from the list of analytical methods used in Section 5.6.5.b.

Attachments 2 and 3 to this amendment request provides the markup pages of the existing TS for Braidwood and Byron, respectively, showing the proposed changes. Attachments 4 and 5 provides the TS Bases markups for Braidwood and Byron, respectively, and are provided as information only.

3.0 TECHNICAL EVALUATION

The TS require key parameters to meet the fuel cycle design margins to assure safe core operation under steady state and transient conditions. The proposed changes remove the extraneous detail related to BEACON PDMS from the Braidwood and Byron TS as it relates to Evaluation of Proposed Changes Page 5 of 11 these key parameters. Specifically, the key parameters related to rod alignment, peak linear heat rate (FQ(Z)), Nuclear Enthalpy Rise Hot Channel Factor (FNH), and QPTR.

Rod alignment is directly related to power distributions and shutdown margin (SDM), which are initial conditions assumed in safety analyses. The operability of the control rods is an initial assumption in all safety analyses that assume rod insertion upon reactor trip.

The purpose of the limits on the values of FQ(Z) is to limit the local (i.e., fuel pellet) peak power density. The FQ(Z) limits define limiting values for core power peaking that ensure that the 10 CFR 50.46 acceptance criteria are met during a Loss of Coolant Accident (LOCA).

The FNH limit identifies the coolant flow channel with the maximum enthalpy rise. This channel has the least heat removal capability and thus the highest probability for DNB. The limiting value of FNH, described by the equation contained in the COLR is the design radial peaking factor used in the plant safety analyses.

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 testing, after refueling and periodically during power operation.

The proposed changes continue to maintain these key parameters within the Braidwood and Byron TS. The justification for the proposed changes to each individual TS are further discussed below.

LCO 3.1.4. Rod Group Alignment Limits Maximum rod misalignment is an initial assumption in the safety analysis that directly affects core power distributions and assumptions of available SDM. The limits on shutdown or control rod alignment ensure that the assumptions in the safety analysis remain valid. Failure to meet LCO 3.1.4 requirements may produce unacceptable power peaking factors and linear heat rates, or unacceptable SDMs, all of which may constitute initial conditions inconsistent with the safety analysis.

The proposed change to RA B.4 requires performance of SR 3.2.1.1, SR 3.2.1.2, and SR 3.2.2.1 to verify FQC(Z), FQW(Z), FNH are within limits specified in the COLR. Verifying that FQ(Z), as approximated by FQC(Z) and FQW(Z), and FNH, is within the required limits ensures that operation at 75% RTP with a rod misaligned does not result in power distributions that may invalidate safety analysis assumptions at full power. The CT of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is not impacted.

The proposed change to Condition C retains the requirement to bring the unit into an acceptable mode or condition in the event more than one rod is found to be misaligned or becomes misaligned because of bank movement. This action prevents conditions outside of the accident analysis assumptions. This status is achieved by bringing the unit to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months which remains unimpacted by the proposed change. The proposed change to Condition D reflects the changes to Condition C. Since the proposed change to RA C.2 applies when PDMS is operable, RA C.3 becomes redundant with a greater CT. As such, RA C.3 and its reference in Condition D is not required and removed from the LCO.

Evaluation of Proposed Changes Page 6 of 11 Proposed changes to CT B.2 and C.2 ensure the associated RA is completed regardless of operability of PDMS.

LCO 3.2.1. Heat Flux Hot Channel Factor (FQ(Z))

This LCO requires operation within the bounds assumed in the safety analyses. Calculations are performed in the core design process to confirm that the core can be controlled in such a manner during operation that it can stay within the LOCA FQ(Z) limits. If FQC(Z) cannot be maintained within the LCO limits, reduction of the core power is required and if FQW(Z) cannot be maintained within the LCO limits, reduction of the core power and AFD limits is required.

Performance of SR 3.2.1.1 and SR 3.2.1.2 are necessary to assure FQ(Z) is properly evaluated prior to increasing thermal power. To ensure SR 3.2.1.1 and SR 3.2.1.2 are performed, a Note is added to Condition A and Condition B to require the completion of RAs A.4 and B.5, respectively.

Clarification is provided by proposed changes to the CTs of Condition A to ensure the respective RA is performed after each FQC(Z) determination. The RAs A.1, A.2, and A.3 may be affected by subsequent determinations of FQC(Z) and would require the respective RA to be performed following each FQC(Z) determination to comply with the respective thermal power, power range neutron flux - high trip setpoints, or overpower T trip setpoints. Decreases in FQC(Z) would allow increasing the maximum allowable setpoints to this revised limit.

Removing of extraneous detail from the TS, the FQ(Z) monitoring requirements (verification that FQC(Z) and FQW(Z) are within their specified limits) are completed via SRs 3.2.1.1 and 3.2.1.2 only. LCO 3.2.1 surveillances are no longer dependent on an operable PDMS system, and this is reflected by the deletion of SR 3.2.1.3, SR 3.2.1.4, SR 3.2.1.1 Note 2, and SR 3.2.1.2 Note 3.

The change in location of SR 3.2.1.1 and SR 3.2.1.2 Note 1 to the beginning of the LCO 3.2.1 SRs section aligns with NUREG-1431 and does not impact the applicability of the Note to SR 3.2.1.1 or SR 3.2.1.2. Additionally, the SR 3.2.1.2 Note is revised to reflect an editorial change which aligns with NUREG-1431.

LCO 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor (FNH)

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

RA A.1 is modified to either restore FNH within its limit with RA A.1.1 or to reduce thermal power to <50% RTP with RA A.1.2.1 and reduce the Power Range Neutron Flux - High to 55% RTP with RA A.1.2.2. The addition to restore FNH within its limit may involve realigning any misaligned rods or reducing power enough to bring FNH within its power dependent limit. The allowed Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months provides an acceptable time to restore FNH to within its Evaluation of Proposed Changes Page 7 of 11 limits without allowing the plant to remain in an unacceptable condition for an extended period of time.

The proposed change to LCO 3.2.2 Condition A includes administrative changes to align format with NUREG-1431, including direct reference of SR 3.2.2.1 which requires verification of FNH.

With the removal of PDMS references in the LCO, the surveillance for FNH is no longer dependent on the operability status of PDMS and are noted to perform SR 3.2.2.1 to verify FNH.

SR 3.2.2.1 is the only surveillance required to verify FNH is within limits after each refueling prior to exceeding 75% RTP and in accordance with the Surveillance Frequency Control Program.

Applicability to perform the surveillance based on PDMS operability is not required, therefore, the Notes to SR 3.2.2.1 and SR 3.2.2.2 are deleted.

LCO 3.2.4, Quadrant Power Tilt Ratio (QPTR)

The QPTR limit of 1.02, at which corrective action is required, provides a margin of protection for both the DNB ratio and linear heat generation rate contributing to excessive power peaks resulting from X-Y plane power tilts. A limiting QPTR of 1.02 can be tolerated before the margin for uncertainty in FQ(Z) and FNH is possibly challenged.

PDMS directly monitors the key power distribution parameters (i.e., FQ(Z), FNH, and DNBR) continuously and previously removed the need to monitor the indirect indicators (i.e., QPTR).

The removal of PDMS from the TS requires adding back the verification of QPTR regardless of the status of PDMS. This is reflected by the removal of the language in the LCO Applicability section regarding when PDMS is inoperable, and removal of SR 3.2.4.1 Note 3 and SR 3.2.4.2 Note 2 which require performance of the respective surveillance 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after declaration of PDMS inoperable.

RA A.2 removes a redundancy for the performance of RA A.1 that reduces thermal power 3%

from RTP for each 1% of QPTR >1.00.

All other proposed changes align LCO 3.2.4 text with NUREG-1431 and are editorial. These changes do not impact the requirements of this LCO.

LCO 3.2.5 Departure from Nucleate Boiling Ratio (DNBR)

The applicability of LCO 3.2.5 is only required when PDMS is operable, as PDMS is capable of directly and continuously monitoring the DNBR parameters. With the removal of the extraneous detail from the TS, LCO 3.2.5 is not required and is deleted. Power distribution is limited by LCO 3.2.3 for AFD limits and the proposed changes to LCO 3.2.4 for QPTR limits. This is consistent with NUREG-1431 which does not include an explicit LCO for DNBR.

TS 5.6.5 CORE Operating Limits Report (COLR)

TS 5.6.5 list the analytical methods used to determine the core operating limits in the COLR previously reviewed and approved by the NRC. The analytical methods used to support PDMS is captured in WCAP-12472-P-A. This topical report is no longer required and deleted from the Evaluation of Proposed Changes Page 8 of 11 TS. LCO 3.2.5 is removed from the list to reflect changes made as part of this amendment request.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements / Criteria The regulatory requirements associated with the proposed change reflects 10 CFR 50.36, "Technical specifications." In accordance with 10 CFR 50.36, TS are required to include (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls.

The Core Operating Limits Report (COLR) is part of the administrative controls which is included in the Braidwood and Byron TS in accordance with 10 CFR 50.36.

The proposed change does not change a design feature and continues to require operating restrictions that are initial conditions assumed in safety analyses. Key parameters and process variables are covered by the proposed changes, and the proposed changes are consistent with the format, level of detail, and structure of NUREG-1431, Standard Technical Specifications Westinghouse Plants, Volume 1 Specifications, Revision 5.0 dated March 2021. Therefore, Criterion 2 of 10 CFR 50.36(c)(2)(ii) continues to be met.

4.2 No Significant Hazards Consideration In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit or early site permit," Constellation Energy Generation, LLC, (CEG) requests amendments to the Technical Specifications (TS) for Renewed Facility Operating License Nos. NPF-72 and NPF-77 for Braidwood Station, Units 1 and 2 (Braidwood) and Renewed Facility Operating License Nos.

NPF-37 and NPF-66 for Byron Station, Units 1 and 2 (Byron).

Braidwood and Byron TS currently reference the Westinghouse power distribution monitoring system known as Best Estimate Analyzer for Core Operations Nuclear (BEACON). The BEACON related TS changes were approved in 2001. CEG desires to remove extraneous detail related to the specific core monitoring system software and more closely align with the Standard Technical Specification, NUREG-1431 Revision 5.

According to 10 CFR 50.92, "Issuance of amendment," paragraph (c), a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

(1) Involve a significant increase in the probability or consequences of any accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

Evaluation of Proposed Changes Page 9 of 11 CEG has evaluated the proposed changes, using the criteria in 10 CFR 50.92, and has determined that the proposed changes do not involve a significant hazards consideration. The following information is provided to support a finding of no significant hazards consideration.

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

Response: No.

The proposed changes remove extraneous detail and aligns TS closer to the industry STS, NUREG-1431 Revision 5. Monitoring rod alignment, as well as verification of key parameters continue to be performed by the proposed changes. The proposed change does not impact existing plant instrumentation, as the incorporation of PDMS did not introduce new instrumentation or calculation system other than an interface systems and integration analysis. There are no changes to any direct protection or control functions.

SR and their frequencies are consistent with NUREG-1431, along with Braidwood and Byrons Surveillance Frequency Control Program. The proposed changes do not alter safety limit values or accident analyses supporting the Braidwood and Byron licensing basis.

The proposed changes do not impact any of the accident initiators.

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

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

Response: No.

No new or different accidents result from removal of extraneous detail from the Braidwood and Byron TS. Braidwood and Byron continue to require core monitoring to verify and monitor rod alignment and key parameters. The proposed changes do not affect or create any new or different accident initiator. There are no changes in the parameters within which the plant is normally operated, and the changes do not impose any new or different operating requirements. The proposed change does not impact existing plant instrumentation.

Therefore, the proposed changes do not create the possibility of a new or different accident from any accident previously evaluated.

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

Response: No.

The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. The proposed change continues to monitor and verify rod alignment and key parameters. SR and their frequencies are consistent with NUREG-1431, along with Braidwood and Byrons Surveillance Frequency Control Program. There are no changes to the required limits set forth in the Braidwood and Evaluation of Proposed Changes Page 10 of 11 Byron TS, and there are no changes in the accident analyses. The reactor will continue to be operated within its analyzed operating and design envelope.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above evaluation, CEG concludes that the proposed changes do 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.

4.3 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

CEG has evaluated the proposed amendment and has determined that the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(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.

Evaluation of Proposed Changes Page 11 of 11

6.0 REFERENCES

1. NRC Safety Evaluation Report, Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2 - Issuance of Amendments to Technical Specifications for Implementation of the Best Estimate Analyzer for Core Operations Nuclear Power Distribution Monitoring System (TAC Nos. MA8254, MA8255, MA8252, and MA8253), dated February 13, 2001 (ADAMS Accession No. ML010510325)

2. NRC Safety Evaluation Report, Braidwood Station, Units 1 and 2, and Byron Station, Unit Nos. 1 and 2 -Issuance of Amendments Regarding Axial Flux Difference Technical Specifications (EPID L-2018-LLA-0098), dated December 12, 2018 (ADAMS Accession No. ML18302A227)

3. NRC NUREG-1431, Revision 5, Standard Technical Specifications Westinghouse Plants, Volume 1, Specifications dated September 2021 (ADAMS Accession No. ML21259A155)

4. NRC NUREG-1431, Revision 5, Standard Technical Specifications Westinghouse Plants, Volume 2, Bases dated September 2021 (ADAMS Accession No. ML21259A159)

Proposed Technical Specifications Changes (Mark-Up)

Braidwood Station, Units 1 and 2 NRC Docket Nos. 50-456 and 50-457 TS Pages 3.1.4-2 through 3.1.4-4 TS Pages 3.2.1-1 through 3.2.1-7 TS Pages 3.2.2-1 through 3.2.2-3 TS Pages 3.2.4-1 through 3.2.4-4 TS Page 3.2.5-1 TS Page 5.6-3

Rod Group Alignment Limits 3.1.4 BRAIDWOOD UNITS 1 & 2 3.1.4 2 Amendment ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.

One rod not within alignment limit.

B.1.1 Verify SDM is within the limits specified in the COLR.

OR B.1.2 Initiate boration to restore SDM to within limit.

AND B.2 AND B.3 AND B.4 Reduce THERMAL POWER to 75% RTP.

Verify SDM is within the limits specified in the COLR.

Perform SR 3.2.1.1, SR 3.2.1.2, and SR 3.2.2.1. Determine Heat Flux Hot Channel Factor (FQ(Z)) and Nuclear Enthalpy Rise Hot Channel Factor (FN H).

AND 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 1 hour 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months from discovery of Condition B concurrent with inoperability of Power Distribution Monitoring System (PDMS)

Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 72 hours (continued)

Rod Group Alignment Limits 3.1.4 BRAIDWOOD UNITS 1 & 2 3.1.4 3 Amendment ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B.

(continued)

B.5 Re-evaluate safety analyses and confirm results remain valid for duration of operation under these conditions.

5 days C.

More than one rod not within alignment limit.

C.1.1 Verify SDM is within the limits specified in the COLR.

OR C.1.2 Initiate boration to restore required SDM to within limit.

AND C.2 Be in MODE 3.

AND C.3

NOTE---------

Only required to be performed when PDMS is OPERABLE.

Restore rod(s) to within alignment limit.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 1 hour 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months from discovery of Condition C concurrent with inoperability of PDMS 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (continued)

Rod Group Alignment Limits 3.1.4 BRAIDWOOD UNITS 1 & 2 3.1.4 4 Amendment ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.

Required Action and associated Completion Time of Condition B or Required Action C.3 not met.

D.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1

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

Not required to be performed for rods associated with inoperable rod position indicator or demand position indicator.

Verify position of individual rods within alignment limit.

In accordance with the Surveillance Frequency Control Program SR 3.1.4.2 Verify rod freedom of movement (trippability) by moving each rod not fully inserted in the core 10 steps in either direction.

In accordance with the Surveillance Frequency Control Program SR 3.1.4.3 Verify rod drop time of each rod, from the fully withdrawn position, is 2.7 seconds from the beginning of decay of stationary gripper coil voltage to dashpot entry, with:

a.

Tavg 550°F; and b.

All reactor coolant pumps operating.

Prior to criticality after each removal of the reactor head

FQ(Z) 3.2.1 BRAIDWOOD UNITS 1 & 2 3.2.1 1 Amendment 3.2 POWER DISTRIBUTION LIMITS 3.2.1 Heat Flux Hot Channel Factor (FQ(Z))

LCO 3.2.1 FQ(Z), as approximated by

)

Z

(

FC Q

and

)

Z

(

FW Q

, shall be within the limit specified in the COLR.

APPLICABILITY:

MODE 1.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. A.

NOTE---------

Required Action A.4 shall be completed whenever this Condition is entered.

)

Z

(

FC Q

not within limit.

A.1 Reduce THERMAL POWER 1% RTP for each 1%

)

Z

(

FC Q

exceeds limit.

AND Q

15 minutes after each FC(Z) determination A.2 Reduce Power Range Neutron Flux-High trip setpoints 1%

for each 1%

)

Z

(

FC Q

exceeds limit.

AND Q

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each FC(Z) determination A.3 Reduce Overpower T trip setpoints 1%

for each 1%

)

Z

(

FC Q

exceeds limit.

Q 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each FC(Z) determination AND A.4 Perform SR 3.2.1.1 and SR 3.2.1.2.

Prior to increasing THERMAL POWER above the limit of Required Action A.1 (continued)

FQ(Z) 3.2.1 BRAIDWOOD UNITS 1 & 2 3.2.1 2 Amendment ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B.

NOTE---------

Required Action B.5 shall be completed whenever this Condition is entered.

)

Z

(

FW Q

not within limit.

B.1 Reduce THERMAL POWER as specified in the COLR.

AND 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B.2 Reduce AFD limits as specified in the COLR.

AND 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B.3 Reduce Power Range Neutron Flux-High trip setpoints 1%

for each 1% that THERMAL POWER is limited below RATED THERMAL POWER by Required Action B.1the maximum allowable power of the AFD limits is reduced.

AND 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months B.4 Reduce Overpower T trip setpoints 1%

for each 1% that the maximum allowable power of the AFD limits is reducedTHERMAL POWER is limited below RATED THERMAL POWER by Required Action B.1.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months AND B.5 Perform SR 3.2.1.1 and SR 3.2.1.2.

Prior to increasing THERMAL POWER above

FQ(Z) 3.2.1 BRAIDWOOD UNITS 1 & 2 3.2.1 3 Amendment CONDITION REQUIRED ACTION COMPLETION TIME the maximum allowable power of the AFD limitsand AFD limits above the limits of Required Actions B.1 and B.2 C.

Required Action and associated Completion Time not met.

C.1 Be in MODE 2.

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

FQ(Z) 3.2.1 BRAIDWOOD UNITS 1 & 2 3.2.1 4 Amendment SURVEILLANCE REQUIREMENTS

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

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

SURVEILLANCE FREQUENCY SR 3.2.1.1

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

1.

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

2.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after declaring Power Distribution Monitoring System (PDMS) inoperable.

Performance of SR 3.2.1.3 satisfies the initial performance of this SR after declaring PDMS inoperable.

Verify

)

Z

(

FC Q

is within limit specified in the COLR.

Prior to exceeding 75% RTP after each refueling AND Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which

)

Z

(

FC Q

was last verified AND In accordance with the Surveillance Frequency Control Program

FQ(Z) 3.2.1 BRAIDWOOD UNITS 1 & 2 3.2.1 5 Amendment SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.2.1.2

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

1.

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

2. If

)

Z

(

FW Q

measurements indicate that either the maximum over z [

)

Z

(

FC Q

/ K(Z)]

OR maximum over z [

)

Z

(

FW Q

/ K(Z)]

has increased since the previous evaluation of

)

Z

(

FC Q

or if

)

Z

(

FW Q

is expected to increase prior to the next evaluation of

)

Z

(

FC Q

a.

Increase

)

Z

(

FW Q

by the greater of a factor of [1.02] or the appropriate factor specified in the COLR and reverify

)

Z

(

FW Q

is within limits specified in the COLR; or b.

Repeat SR 3.2.1.2 once per 7 EFPD until either a. above is met or two successive flux maps indicate that the maximum over z [

)

Z

(

FC Q

/ K(Z)]

and maximum over z [

)

Z

(

FW Q

/ K(Z)]

has not increased.

(continued)

FQ(Z) 3.2.1 BRAIDWOOD UNITS 1 & 2 3.2.1 6 Amendment SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.2 (continued)

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

3.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after declaring PDMS inoperable. Performance of SR 3.2.1.4 satisfies the initial performance of this SR after declaring PDMS inoperable.

Verify

)

Z

(

FW Q

is within limit specified in the COLR.

Prior to exceeding 75% RTP after each refueling AND Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which

)

Z

(

FW Q

was last verified AND In accordance with the Surveillance Frequency Control Program (continued)

FQ(Z) 3.2.1 BRAIDWOOD UNITS 1 & 2 3.2.1 7 Amendment SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.2.1.3

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

Only required to be performed when PDMS is OPERABLE.

Verify FC(Z)

Q is within limit specified in the COLR.

In accordance with the Surveillance Frequency Control Program SR 3.2.1.4

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

Only required to be performed when PDMS is OPERABLE.

Verify FW(Z)

Q is within limit specified in the COLR.

In accordance with the Surveillance Frequency Control Program

N H

F 3.2.2 BRAIDWOOD UNITS 1 & 2 3.2.2 1 Amendment 3.2 POWER DISTRIBUTION LIMITS 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor

)

F

(

N H

LCO 3.2.2 N

H F shall be within the limit specified in the COLR.

APPLICABILITY:

MODE 1.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

NOTE---------

Required Actions A.2 and A.34 must be completed whenever Condition A is entered.

N H

F not within limit.

A.1.1 Restore to within limit.

OR A.1.2.1 Reduce THERMAL POWER to < 50% RTP.

AND A.1.2.2 Reduce Power Range Neutron Flux-High trip setpoints to 55% RTP.

AND A.2 Perform Determine F N

H.

SR 3.2.2.1.

AND A.3 Reduce Power Range Neutron Flux-High trip setpoints to 55% RTP.

AND 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 4 hours 2472 hours0.0286 days 0.687 hours 0.00409 weeks 9.40596e-4 months 7224 hours (continued)

N H

F 3.2.2 BRAIDWOOD UNITS 1 & 2 3.2.2 2 Amendment ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

(continued)

A.43

NOTE---------

THERMAL POWER does not have to be reduced to comply with this Required Action.

Perform SR 3.2.2.1.

Determine F N

H.

Prior to exceeding 50% RTP AND Prior to exceeding 75% RTP AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reaching 95% RTP B.

Required Action and associated Completion Time not met.

B.1 Be in MODE 2.

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

N H

F 3.2.2 BRAIDWOOD UNITS 1 & 2 3.2.2 3 Amendment SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1

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

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after declaring PDMS inoperable. Performance of SR 3.2.2.2 satisfies the initial performance of this SR after declaring PDMS inoperable.

Verify N

H F is within limits specified in the COLR.

Prior to exceeding 75% RTP after each refueling AND In accordance with the Surveillance Frequency Control Program SR 3.2.2.2

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

Only required to be performed when PDMS is OPERABLE.

Verify H

F N is within limit specified in the COLR.

In accordance with the Surveillance Frequency Control Program

QPTR 3.2.4 BRAIDWOOD UNITS 1 & 2 3.2.4 1 Amendment 3.2 POWER DISTRIBUTION LIMITS 3.2.4 QUADRANT POWER TILT RATIO (QPTR)

LCO 3.2.4 The QPTR shall be 1.02.

APPLICABILITY:

MODE 1 with THERMAL POWER > 50% RTP when Power Distribution Monitoring System (PDMS) is inoperable.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

QPTR not within limit.

A.1 Reduce THERMAL POWER 3% from RTP for each 1% of QPTR

> 1.00.

AND A.2 Determine QPTR and reduce THERMAL POWER 3% from RTP for each 1% of QPTR

> 1.00.

AND A.3 Perform SR 3.2.1.1, SR 3.2.1.2, and SR 3.2.2.1.

AND 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after each QPTR determination Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 24 hours after achieving equilibrium conditions from a THERMAL POWER reduction per Required Action A.1 AND Once per 7 days thereafter (continued)

QPTR 3.2.4 BRAIDWOOD UNITS 1 & 2 3.2.4 2 Amendment ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

(continued)

A.4 Re-evaluate safety analyses and confirm results remain valid for duration of operation under this condition.

AND A.5

NOTES--------

1.

Perform Required Action A.5 only after Required Action A.4 is completed.

2.

Required Action A.6 shall be completed whenever Required Action A.5 is performed.

Normalize excore detectors to restore QPTR to within limits.

AND Prior to increasing exceeding the THERMAL POWER above the limit of Required Action A.1 Prior to increasing exceeding the THERMAL POWER above the limits of Required Action A.1 (continued)

QPTR 3.2.4 BRAIDWOOD UNITS 1 & 2 3.2.4 3 Amendment ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

(continued)

A.6

NOTE---------

Perform Required Action A.6 only after Required Action A.5 is completed.

Perform SR 3.2.1.1, SR 3.2.1.2, and SR 3.2.2.1.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after achieving equilibrium conditions at RTP not to exceed 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after exceeding the THERMAL POWER limit of Required Action A.1 B.

Required Action and associated Completion Time not met.

B.1 Reduce THERMAL POWER to 50% RTP.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

QPTR 3.2.4 BRAIDWOOD UNITS 1 & 2 3.2.4 4 Amendment 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 channel inputs can be used for calculating QPTR.

2.

SR 3.2.4.2 may be performed in lieu of this Surveillance.

3.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after declaring PDMS inoperable.

Verify QPTR is within limit 1.02 by calculation.

In accordance with the Surveillance Frequency Control Program SR 3.2.4.2

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

1. Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after input from one Power Range Neutron Flux channel is inoperable with THERMAL POWER > 75% RTP.

2.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after declaring PDMS inoperable.

Verify QPTR is within limit 1.02 using the movable incore detectors.

In accordance with the Surveillance Frequency Control Program

DNBR 3.2.5 BRAIDWOOD UNITS 1 & 2 3.2.5 1 Amendment 3.2 POWER DISTRIBUTION LIMITS 3.2.5 DELETED Departure from Nucleate Boiling Ratio (DNBR)

LCO 3.2.5 DNBR shall be within the limit specified in the COLR.

APPLICABILITY:

MODE 1 with THERMAL POWER 50% RTP when Power Distribution Monitoring System (PDMS) is OPERABLE.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

DNBR not within limit.

A.1 Restore DNBR to within limit.

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months B.

Required Action and associated Completion Time not met.

B.1 Reduce THERMAL POWER to

< 50% RTP.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.5.1 Verify DNBR is within limit specified in the COLR.

In accordance with the Surveillance Frequency Control ProgramIn accordance with the Surveillance Frequency Control Program

Reporting Requirements 5.6 BRAIDWOOD UNITS 1 & 2 5.6 3 Amendment 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

SL 2.1.1, "Reactor Core SLs";

LCO 3.1.1, "SHUTDOWN MARGIN (SDM)";

LCO 3.1.3, "Moderator Temperature Coefficient";

LCO 3.1.5, "Shutdown Bank Insertion Limits";

LCO 3.1.6, "Control Bank Insertion Limits";

LCO 3.1.8, "PHYSICS TESTS Exceptions - MODE 2";

LCO 3.2.1, "Heat Flux Hot Channel Factor (FQ(Z))";

LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor

)

F

( N H

LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)";

LCO 3.2.5, "Departure from Nucleate Boiling Ratio (DNBR)";

LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits"; and LCO 3.9.1, "Boron Concentration";

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

WCAP-9272-P-A, "Westinghouse Reload Safety Evaluations Methodology," July 1985.

2.

Not UsedWCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System," August 1994.

3.

NFSR-0016, "Commonwealth Edison Company Topical Report on Benchmark of PWR Nuclear Design Methods," July 1983.

4.

NFSR-0081, "Commonwealth Edison Company Topical Report on Benchmark of PWR Nuclear Design Methods Using the Phoenix-P and ANC Computer Codes," July 1990.

5.

ComEd letter from D. Saccomando to the Office of Nuclear Reactor Regulation dated December 21, 1994, transmitting an attachment that documents applicable sections of WCAP-11992/11993 and ComEd application of the UET methodology addressed in "Additional Information Regarding Application for Amendment to Facility Operating Licenses-Reactivity Control Systems."

Proposed Technical Specifications Changes (Mark-Up)

Byron Station, Units 1 and 2 NRC Docket Nos. 50-454 and 50-455 TS Pages 3.1.4-2 through 3.1.4-4 TS Pages 3.2.1-1 through 3.2.1-7 TS Pages 3.2.2-1 through 3.2.2-3 TS Pages 3.2.4-1 through 3.2.4-4 TS Page 3.2.5-1 TS Page 5.6-3

Rod Group Alignment Limits 3.1.4 BYRON UNITS 1 & 2 3.1.4 2 Amendment ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.

One rod not within alignment limit.

B.1.1 Verify SDM is within the limits specified in the COLR.

OR B.1.2 Initiate boration to restore SDM to within limit.

AND B.2 AND B.3 AND B.4 Reduce THERMAL POWER to 75% RTP.

Verify SDM is within the limits specified in the COLR.

Perform SR 3.2.1.1, SR 3.2.1.2, and SR 3.2.2.1. Determine Heat Flux Hot Channel Factor (FQ(Z)) and Nuclear Enthalpy Rise Hot Channel Factor (FN H).

AND 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 1 hour 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months from discovery of Condition B concurrent with inoperability of Power Distribution Monitoring System (PDMS)

Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 72 hours (continued)

Rod Group Alignment Limits 3.1.4 BYRON UNITS 1 & 2 3.1.4 3 Amendment ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued)

B.5 Re-evaluate safety analyses and confirm results remain valid for duration of operation under these conditions.

5 days C. More than one rod not within alignment limit.

C.1.1 Verify SDM is within the limits specified in the COLR.

OR C.1.2 Initiate boration to restore required SDM to within limit.

AND C.2 Be in MODE 3.

AND C.3

NOTE---------

Only required to be performed when PDMS is OPERABLE.

Restore rod(s) to within alignment limit.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 1 hour 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months from discovery of Condition C concurrent with inoperability of PDMS 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (continued)

Rod Group Alignment Limits 3.1.4 BYRON UNITS 1 & 2 3.1.4 4 Amendment ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and associated Completion Time of Condition B or Required Action C.3 not met.

D.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1

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

Not required to be performed for rods associated with inoperable rod position indicator or demand position indictor.

Verify position of individual rods within alignment limit.

In accordance with the Surveillance Frequency Control Program SR 3.1.4.2 Verify rod freedom of movement (trippability) by moving each rod not fully inserted in the core 10 steps in either direction.

In accordance with the Surveillance Frequency Control Program SR 3.1.4.3 Verify rod drop time of each rod, from the fully withdrawn position, is 2.7 seconds from the beginning of decay of stationary gripper coil voltage to dashpot entry, with:

a.

Tavg 550°F; and

b.

All reactor coolant pumps operating.

Prior to criticality after each removal of the reactor head

FQ(Z) 3.2.1 BYRON UNITS 1 & 2 3.2.1 1 Amendment 3.2 POWER DISTRIBUTION LIMITS 3.2.1 Heat Flux Hot Channel Factor (FQ(Z))

LCO 3.2.1 FQ(Z), as approximated by

)

Z

(

FC Q

and

)

Z

(

FW Q

, shall be within the limit specified in the COLR.

APPLICABILITY:

MODE 1.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

NOTE---------

Required Action A.4 shall be completed whenever this Condition is entered.

)

Z

(

FC Q

not within limit.

A.1 Reduce THERMAL POWER 1% RTP for each 1%

)

Z

(

FC Q

exceeds limit.

AND Q

15 minutes after each FC(Z) determination A.2 Reduce Power Range Neutron Flux-High trip setpoints 1%

for each 1%

)

Z

(

FC Q

exceeds limit.

AND Q

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each FC(Z) determination A.3 Reduce Overpower T trip setpoints 1%

for each 1%

)

Z

(

FC Q

exceeds limit.

AND Q

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each FC(Z) determination A.4 Perform SR 3.2.1.1 and SR 3.2.1.2.

Prior to increasing THERMAL POWER above the limit of Required Action A.1 (continued)

FQ(Z) 3.2.1 BYRON UNITS 1 & 2 3.2.1 2 Amendment ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. B.

NOTE---------

Required Action B.5 shall be completed whenever this Condition is entered.

)

Z

(

FW Q

not within limit.

B.1 Reduce THERMAL POWER as specified in the COLR.

AND 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B.2 Reduce AFD limits as specified in the COLR.

AND 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B.3 Reduce Power Range Neutron Flux-High trip setpoints 1%

for each 1% that the maximum allowable power of the AFD limits is reduced.THERMAL POWER is limited below RATED THERMAL POWER by Required Action B.1.

AND 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months B.4 Reduce Overpower T trip setpoints 1%

for each 1% that the maximum allowable power of the AFD limits is reduced.THERMAL POWER is limited below RATED THERMAL POWER by Required Action B.1.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months AND

FQ(Z) 3.2.1 BYRON UNITS 1 & 2 3.2.1 3 Amendment CONDITION REQUIRED ACTION COMPLETION TIME B.5 Perform SR 3.2.1.1 and SR 3.2.1.2.

Prior to increasing THERMAL POWER above the maximum allowable power of the AFD limitsand AFD limits above the limits of Required Actions B.1 and B.2 C.

Required Action and associated Completion Time not met.

C.1 Be in MODE 2.

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

FQ(Z) 3.2.1 BYRON UNITS 1 & 2 3.2.1 4 Amendment SURVEILLANCE REQUIREMENTS

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

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

SURVEILLANCE FREQUENCY SR 3.2.1.1

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

1.

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

2.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after declaring Power Distribution Monitoring System (PDMS) inoperable.

Performance of SR 3.2.1.3 satisfies the initial performance of this SR after declaring PDMS inoperable.

Verify

)

Z

(

FC Q

is within limit specified in the COLR.

Prior to exceeding 75% RTP after each refueling AND Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which

)

Z

(

FC Q

was last verified AND In accordance with the Surveillance Frequency Control Program

FQ(Z) 3.2.1 BYRON UNITS 1 & 2 3.2.1 5 Amendment SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.2.1.2

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

1.

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

2. If

)

Z

(

FW Q

measurements indicate that either the maximum over z [

)

Z

(

FC Q

/ K(Z)]

OR maximum over z [

)

Z

(

FW Q

/ K(Z)]

has increased since the previous evaluation of

)

Z

(

FC Q

or if

)

Z

(

FW Q

is expected to increase prior to the next evaluation of

)

Z

(

FC Q

a.

Increase

)

Z

(

FW Q

by the greater of a factor of [1.02] or the appropriate factor specified in the COLR and reverify

)

Z

(

FW Q

is within limits specified in the COLR; or b.

Repeat SR 3.2.1.2 once per 7 EFPD until either a. above is met or two successive flux maps indicate that the maximum over z [

)

Z

(

FC Q

/ K(Z)]

and maximum over z [

)

Z

(

FW Q

/ K(Z)]

has not increased.

(continued)

FQ(Z) 3.2.1 BYRON UNITS 1 & 2 3.2.1 6 Amendment SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.2 (continued)

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

3.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after declaring PDMS inoperable. Performance of SR 3.2.1.4 satisfies the initial performance of this SR after declaring PDMS inoperable.

Verify

)

Z

(

FW Q

is within limit specified in the COLR.

Prior to exceeding 75% RTP after each refueling AND Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which

)

Z

(

FW Q

was last verified AND In accordance with the Surveillance Frequency Control Program (continued)

FQ(Z) 3.2.1 BYRON UNITS 1 & 2 3.2.1 7 Amendment SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.2.1.3

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

Only required to be performed when PDMS is OPERABLE.

Verify FC(Z)

Q is within limit specified in the COLR.

In accordance with the Surveillance Frequency Control Program SR 3.2.1.4

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

Only required to be performed when PDMS is OPERABLE.

Verify FW(Z)

Q is within limit specified in the COLR.

In accordance with the Surveillance Frequency Control Program

N H

F 3.2.2 BYRON UNITS 1 & 2 3.2.2 1 Amendment 3.2 POWER DISTRIBUTION LIMITS 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor

)

F

(

N H

LCO 3.2.2 N

H F shall be within the limit specified in the COLR.

APPLICABILITY:

MODE 1.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

NOTE---------

Required Actions A.2 and A.34 must be completed whenever Condition A is entered.

N H

F not within limit.

A.1.1 Restore to within limit.

OR A.1.2.1 Reduce THERMAL POWER to < 50% RTP.

AND A.1.2.2 Reduce Power Range Neutron Flux-High trip setpoints to 55% RTP.

AND A.2 Perform SR 3.2.2.1.

Determine F N

H.

AND A.3 Reduce Power Range Neutron Flux-High trip setpoints to 55% RTP.

AND 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 4 hours 2472 hours0.0286 days 0.687 hours 0.00409 weeks 9.40596e-4 months 7224 hours (continued)

N H

F 3.2.2 BYRON UNITS 1 & 2 3.2.2 2 Amendment ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

(continued)

A.34

NOTE---------

THERMAL POWER does not have to be reduced to comply with this Required Action.

Perform SR 3.2.2.1.

Determine F N

H.

Prior to exceeding 50% RTP AND Prior to exceeding 75% RTP AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reaching 95% RTP B.

Required Action and associated Completion Time not met.

B.1 Be in MODE 2.

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

N H

F 3.2.2 BYRON UNITS 1 & 2 3.2.2 3 Amendment SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1

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

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after declaring PDMS inoperable. Performance of SR 3.2.2.2 satisfies the initial performance of this SR after declaring PDMS inoperable.

Verify N

H F is within limits specified in the COLR.

Prior to exceeding 75% RTP after each refueling AND In accordance with the Surveillance Frequency Control Program SR 3.2.2.2

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

Only required to be performed when PDMS is OPERABLE.

Verify H

F N is within limit specified in the COLR.

In accordance with the Surveillance Frequency Control Program

QPTR 3.2.4 BYRON UNITS 1 & 2 3.2.4 1 Amendment 3.2 POWER DISTRIBUTION LIMITS 3.2.4 QUADRANT POWER TILT RATIO (QPTR)

LCO 3.2.4 The QPTR shall be 1.02.

APPLICABILITY:

MODE 1 with THERMAL POWER > 50% RTP when Power Distribution Monitoring System (PDMS) is inoperable.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

QPTR not within limit.

A.1 Reduce THERMAL POWER 3% from RTP for each 1% of QPTR

> 1.00.

AND A.2 Determine QPTR and reduce THERMAL POWER 3% from RTP for each 1% of QPTR

> 1.00.

AND A.3 Perform SR 3.2.1.1, SR 3.2.1.2, and SR 3.2.2.1.

AND 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after each QPTR determination Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 24 hours after achieving equilibrium conditions from a THERMAL POWER reduction per Required Action A.1 AND Once per 7 days thereafter (continued)

QPTR 3.2.4 BYRON UNITS 1 & 2 3.2.4 2 Amendment ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

(continued)

A.4 Re-evaluate safety analyses and confirm results remain valid for duration of operation under this condition.

AND A.5

NOTES--------

1.

Perform Required Action A.5 only after Required Action A.4 is completed.

2.

Required Action A.6 shall be completed whenever Required Action A.5 is performed.

Normalize excore detectors to restore QPTR to within limits.

AND Prior to exceeding increasing the THERMAL POWER above the limit of Required Action A.1 Prior to exceeding increasing the THERMAL POWER above the limits of Required Action A.1 (continued)

QPTR 3.2.4 BYRON UNITS 1 & 2 3.2.4 3 Amendment ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

(continued)

A.6

NOTE---------

Perform Required Action A.6 only after Required Action A.5 is completed.

Perform SR 3.2.1.1, SR 3.2.1.2, and SR 3.2.2.1.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after achieving equilibrium conditions at RTP not to exceed 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after exceeding the THERMAL POWER limit of Required Action A.1 B.

Required Action and associated Completion Time not met.

B.1 Reduce THERMAL POWER to 50% RTP.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

QPTR 3.2.4 BYRON UNITS 1 & 2 3.2.4 4 Amendment 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 channel inputs can be used for calculating QPTR.

2.

SR 3.2.4.2 may be performed in lieu of this Surveillance.

3.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after declaring PDMS inoperable.

Verify QPTR is within limit 1.02 by calculation.

In accordance with the Surveillance Frequency Control Program SR 3.2.4.2

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

1. Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after input from one Power Range Neutron Flux channel is inoperable with THERMAL POWER > 75% RTP.

2.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after declaring PDMS inoperable.

Verify QPTR is within limit 1.02 using the movable incore detectors.

In accordance with the Surveillance Frequency Control Program

DNBR 3.2.5 BYRON UNITS 1 & 2 3.2.5 1 Amendment 3.2 POWER DISTRIBUTION LIMITS 3.2.5 DELETEDDeparture from Nucleate Boiling Ratio (DNBR)

LCO 3.2.5 DNBR shall be within the limit specified in the COLR.

APPLICABILITY:

MODE 1 with THERMAL POWER 50% RTP when Power Distribution Monitoring System (PDMS) is OPERABLE.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

DNBR not within limit.

A.1 Restore DNBR to within limit.

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months B.

Required Action and associated Completion Time not met.

B.1 Reduce THERMAL POWER to

< 50% RTP.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.5.1 Verify DNBR is within limit specified in the COLR.

In accordance with the Surveillance Frequency Control Program

Reporting Requirements 5.6 BYRON UNITS 1 & 2 5.6 3 Amendment 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

SL 2.1.1, "Reactor Core SLs";

LCO 3.1.1, "SHUTDOWN MARGIN (SDM)";

LCO 3.1.3, "Moderator Temperature Coefficient";

LCO 3.1.5, "Shutdown Bank Insertion Limits";

LCO 3.1.6, "Control Bank Insertion Limits";

LCO 3.1.8, "PHYSICS TESTS Exceptions - MODE 2";

LCO 3.2.1, "Heat Flux Hot Channel Factor (FQ(Z))";

LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor

)

F

( N H

LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)";

LCO 3.2.5, "Departure from Nucleate Boiling Ratio (DNBR)";

LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits"; and LCO 3.9.1, "Boron Concentration";

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

WCAP-9272-P-A, "Westinghouse Reload Safety Evaluations Methodology," July 1985.

2.

Not UsedWCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System," August 1994.

3.

NFSR-0016, "Commonwealth Edison Company Topical Report on Benchmark of PWR Nuclear Design Methods," July 1983.

4.

NFSR-0081, "Commonwealth Edison Company Topical Report on Benchmark of PWR Nuclear Design Methods Using the Phoenix-P and ANC Computer Codes," July 1990.

5.

ComEd letter from D. Saccomando to the Office of Nuclear Reactor Regulation dated December 21, 1994, transmitting an attachment that documents applicable sections of WCAP-11992/11993 and ComEd application of the UET methodology addressed in "Additional Information Regarding Application for Amendment to Facility Operating Licenses-Reactivity Control Systems."

Proposed Bases Changes (Mark-Up)

Braidwood Station, Units 1 and 2 NRC Docket Nos. 50-456 and 50-457 (For Information Only)

Rod Group Alignment Limits B 3.1.4 BRAIDWOOD UNITS 1 & 2 B 3.1.4 8 Revision BASES ACTIONS (continued)

However, in many cases, realigning the remainder of the group to the misaligned rod may not be desirable. For example, realigning control bank B to a rod that is misaligned 15 steps from the top of the core would require a significant power reduction, since control bank D must be fully inserted and control bank C must be partially inserted.

With a misaligned rod, SDM must be verified to be within limit (specified in the COLR) or boration must be initiated to restore SDM to within limit.

Power operation may continue with one RCCA trippable but misaligned, provided that SDM is verified within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

The Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months represents the time necessary for determining the actual unit SDM and, if necessary, aligning and starting the necessary systems and components to initiate boration to restore SDM to within limit.

B.2, B.3, B.4, and B.5 For continued operation with a misaligned rod, THERMAL POWER must be reduced when Power Distribution Monitoring System (PDMS) is inoperable, SDM must periodically be verified within limits (specified in the COLR), hot channel factors (FQ(Z) and

)

F

( N H

must be verified within limits, and the safety analyses must be re-evaluated to confirm continued operation is permissible.

Reduction of power to 75% RTP when PDMS is inoperable, 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 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months gives the operator sufficient time to accomplish an orderly power reduction without challenging the Reactor Protection System. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."

In this Required Action, the Completion Time only begins on discovery that both:

a.

One rod is not within alignment limit; and b.

PDMS is inoperable.

Rod Group Alignment Limits B 3.1.4 BRAIDWOOD UNITS 1 & 2 B 3.1.4 9 Revision BASES ACTIONS (continued)

Discovering one rod not within alignment limit coincident with PDMS inoperable results in starting the Completion Time for the Required Action. During power operation when PDMS is OPERABLE, LHR is measured continuously. Therefore, a reduction of power to 75% RTP is not necessary to ensure that local LHR increases due to a misaligned RCCA will not cause the core design criteria to be exceeded.

When a rod is known to be misaligned, there is a potential to impact the SDM. Since the core conditions can change with time, periodic verification of SDM is required. A Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient to ensure this requirement continues to be met.

Verifying that FQ(Z) and N

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

H F.

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

Accident analyses (Ref. 3) requiring re-evaluation for continued operation with a misaligned rod include:

1.

Increase in heat removal by the secondary system:

a.

Excessive increase in secondary steam flow, b.

Inadvertent opening of a steam generator power operated relief or safety valve, and c.

Steam system piping failure;

Rod Group Alignment Limits B 3.1.4 BRAIDWOOD UNITS 1 & 2 B 3.1.4 11 Revision BASES ACTIONS (continued)

C.2 If more than one rod is found to be misaligned or becomes misaligned because of bank movement when PDMS is inoperable, the unit conditions may fall outside of the accident analysis assumptions. Since automatic bank sequencing would continue to cause misalignment, 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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." In this Required Action, the Completion Time only begins on discovery that both:

a.

More than one rod is not within alignment limit; and b.

PDMS is inoperable.

Discovering more than one rod not within alignment limit coincident with PDMS inoperable results in starting the Completion Time for the Required Action.

C.3 If more than one rod is found to be misaligned or becomes misaligned because of bank movement when PDMS is OPERABLE, operation may continue in Condition C for a period that should not exceed 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The allowed Completion Time is reasonable, based on the available information on power distributions (Ref. 6). This Required Action is modified by a Note that requires the performance of Required Action C.3 only when PDMS is OPERABLE.

Rod Group Alignment Limits B 3.1.4 BRAIDWOOD UNITS 1 & 2 B 3.1.4 12 Revision BASES ACTIONS (continued)

D.1 When Required Actions of Condition B or C.3 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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , which obviates concerns about the development of undesirable xenon or power distributions. The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging the plant systems.

SURVEILLANCE SR 3.1.4.1 REQUIREMENTS Verification that the position of individual rods is within alignment limits provides a history that allows the operator to detect a rod that is beginning to deviate from its expected position. When a rods alignment cannot be verified due to a DRPI failure, the position of the rod can be determined by use of the movable incore detectors and/or PDMS. The position of the rod may be determined from the difference between the measured core power distribution and the core power distribution expected to exist based on the position of the rod indicated by the group step counter demand position.

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

The SR is modified by a note that permits it to not be performed for rods associated with an inoperable demand position indicator or an inoperable rod position indicator.

The alignment limit is based on the demand position indicator which is not available if the indicator is inoperable. LCO 3.1.7, "Rod Position Indication," provides Actions to verify the rods are in alignment when one or more rod position indicators are inoperable.

Rod Group Alignment Limits B 3.1.4 BRAIDWOOD UNITS 1 & 2 B 3.1.4 14 Revision BASES REFERENCES 1.

10 CFR 50, Appendix A, GDC 10 and GDC 26.

2.

10 CFR 50.46.

3.

UFSAR, Chapter 15.

4.

UFSAR, Section 15.4.3.

5.

UFSAR, Section 15.1.5.

6.

DeletedWCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System," August 1994.

Rod Position Indication B 3.1.7 BRAIDWOOD UNITS 1 & 2 B 3.1.7 6 Revision BASES ACTIONS (continued)

A.1 and A.2 When one DRPI per group in one or more groups fails, (i.e.,

one rod position per group can not be determined by the DRPI System) the position of the rod can still be determined by use of the movable incore detectors or Power Distribution Monitoring System (PDMS). The Required Action may also be satisfied by ensuring at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months that FQ satisfies LCO 3.2.1, FNH satisfies LCO 3.2.2, and SHUTDOWN MARGIN is within the limits provided in the COLR, provided the nonindicating rods have not been moved. When PDMS is OPERABLE, the position of the rod may be determined from the difference between the measured core power distribution and the core power distribution expected to exist based on the position of the rod indicated by the group step counter demand position. Based on experience, normal power operation does not require excessive movement of banks. If a bank has been significantly moved, the Required Action of BC.1 or BC.2 below is required. Therefore, verification of RCCA position within the Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is adequate for allowing continued full power operation, since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small.

Required Action A.1 requires verification of the position of a rod with an inoperable DRPI once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months which may put excessive wear and tear on the moveable incore detector system when PDMS is inoperable; Required Action A.2 provides an alternative. Required Action A.2 requires verification of rod position every 31 EFPD, which coincides with the normal surveillance frequency for verification of core power distribution.

Required Action A.2 includes six distinct requirements for verification of the position of rods associated with an inoperable DRPI:

a.

Initial verification within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of the inoperability of the DRPI; b.

Re-verification once every 31 Effective Full Power Days (EFPD) thereafter;

FQ(Z)

B 3.2.1 BRAIDWOOD UNITS 1 & 2 B 3.2.1 1 Revision B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 Heat Flux Hot Channel Factor (FQ(Z))

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

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

During power operation when Power Distribution Monitoring System (PDMS) is inoperable, the global power distribution is limited by LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)," which are directly and continuously measured process variables. These LCOs, along with LCO 3.1.6, "Control Bank Insertion Limits,"

maintain the core within power distribution limits on a continuous basis. During power operation when PDMS is OPERABLE, PLHR is measured continuously, and global power distribution continues to be limited by LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)."

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

FQ(Z) is measured periodically using the incore detector system when PDMS is inoperable. These measurements are generally taken with the core at or near equilibrium conditions. When PDMS is OPERABLE, FQ(Z) is determined continuously.

Using the measured three dimensional power distributions, it is possible to derive a measured value for FQ(Z). However, because this value represents an equilibrium condition, it does not include the variations in the value of FQ(Z) which are present during nonequilibrium situations, such as load following or power ascension.

To account for these possible variations, the equilibrium value of FQ(Z) is adjusted as

)

Z

(

FW Q

by an elevation dependent factor that accounts for the calculated worst case transient conditions.

FQ(Z)

B 3.2.1 BRAIDWOOD UNITS 1 & 2 B 3.2.1 2 Revision BASES BACKGROUND (continued)

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 (only when PDMS is inoperable), and control rod insertion.

APPLICABLE This LCO precludes core power distributions that violate SAFETY ANALYSES the following fuel design criteria:

a.

During a Loss Of Coolant Accident (LOCA) the 10 CFR 50.46 acceptance criteria must be met (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 prompt energy deposition to the fuel must not exceed 200 cal/gm (Ref. 2); and d.

The control rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn (Ref. 3).

Limits on FQ(Z) ensure that the value of the initial total peaking factor assumed in the accident analyses remains valid.

FQ(Z) limits assumed in the LOCA analysis are typically limiting relative to (i.e., lower than) the FQ(Z) limit assumed in safety analyses for other postulated accidents.

Therefore, this LCO provides conservative limits for other postulated accidents.

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

FQ(Z)

B 3.2.1 BRAIDWOOD UNITS 1 & 2 B 3.2.1 3 Revision BASES LCO The Heat Flux Hot Channel Factor, FQ(Z), shall be limited by the following relationships:

)

Z

(

K P

F

)

Z

(

F RTP Q

Q

for P > 0.5

)

Z

(

K 5

0 F

)

Z

(

F RTP Q

Q

for P 0.5 where:

RTP Q

F is the FQ(Z) limit at RTP provided in the

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

For this facility, the actual values of RTP Q

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

F 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.

FQ(Z) is approximated by

)

Z

(

FC Q

and

)

Z

(

FW Q

. Thus, both

)

Z

(

FC Q

and

)

Z

(

FW Q

must meet the preceding limits on FQ(Z).

When PDMS is inoperable, aAn

)

Z

(

FC Q

evaluation requires obtaining an incore flux map in MODE 1. From the incore flux map results we obtain the measured value

))

Z

(

F

( M Q

of FQ(Z). Then,

)

0815 1

(

)

Z

(

F

)

Z

(

F M

Q C

Q

=

where 1.0815 is a factor that accounts for fuel manufacturing tolerances and flux map measurement uncertainty.

)

Z

(

FC Q

is an excellent approximation for FQ(Z) when the reactor is at the steady state power at which the incore flux map was taken.

FQ(Z)

B 3.2.1 BRAIDWOOD UNITS 1 & 2 B 3.2.1 4 Revision BASES LCO (continued)

When PDMS is OPERABLE, FQ(Z) is determined continuously.

Then, FQ M

Q C

Q U

)

Z

(

F

)

Z

(

F

=

where UFQ is a factor that accounts for measurement uncertainty (Ref. 4) and engineering uncertainty defined in the COLR.

The expression for

)

Z

(

FW Q

is:

)

Z

(

W

)

Z

(

F

)

Z

(

F C

Q W

Q

=

where W(Z) is a cycle dependent function that accounts for power distribution transients encountered during normal operation. W(Z) is included in the COLR. When PDMS is inoperable, tThe

)

Z

(

FC Q

is calculated at equilibrium conditions.

The FQ(Z) limits define limiting values for core power peaking that ensure that the 10 CFR 50.46 acceptance criteria are met during a LOCA (Ref. 1).

This LCO requires operation within the bounds assumed in the safety analyses. Calculations are performed in the core design process to confirm that the core can be controlled in such a manner during operation that it can stay within the LOCA FQ(Z) limits. If

)

Z

(

FC Q

cannot be maintained within the LCO limits, reduction of the core power is required and if

)

Z

(

FW Q

cannot be maintained within the LCO limits, reduction of the AFD limits is required. Note that sufficient reduction of the AFD limits will also result in a reduction of the core power.

Violating the LCO limits for FQ(Z) may produce unacceptable consequences if a design basis event occurs while FQ(Z) is outside its specified limits.

APPLICABILITY The FQ(Z) limits must be maintained in MODE 1 to prevent core power distributions from exceeding the limits assumed in the safety analyses. Applicability in other MODES is not required because there is either insufficient stored energy in the fuel or insufficient energy being transferred to the reactor coolant to require a limit on the distribution of core power.

FQ(Z)

B 3.2.1 BRAIDWOOD UNITS 1 & 2 B 3.2.1 5 Revision BASES ACTIONS A.1, A.2, and A.3 Reducing THERMAL POWER by 1% RTP for each 1% by which

)

Z

(

FC Q

exceeds its limit, maintains an acceptable absolute power density. The Completion Time of 15 minutes provides an acceptable time to reduce power in an orderly manner and without allowing the unit to remain in an unacceptable condition for an extended period of time. The maximum allowable power level initially determined by Required Action A.1 may be affected by subsequent determinations of Q

FC(Z) and would require power reductions within 15 minutes of Q

the FC(Z) determination, if necessary to comply with the Q

decreased maximum allowable power level. Decreases in FC(Z) would allow increasing the maximum allowable power level and increasing power up to this revised limit.

A reduction of the Power Range Neutron Flux-High trip setpoints by 1% for each 1% by which

)

Z

(

FC Q

exceeds its limit, is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is sufficient considering the small likelihood of a severe transient in this time period and the preceding prompt reduction in THERMAL POWER in accordance with Required Action A.1. The maximum allowable Power Range Neutron Flux-High trip setpoints initially determined by Required Action A.2 may be affected by subsequent determinations of FC(Z)

Q and Q

would require Power Range Neutron Flux-High trip setpoint reductions within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the FC(Z) determination, if Q

necessary to comply with the decreased maximum allowable Power Range Neutron Flux-High trip setpoints. Decreases in FC(Z) would allow increasing the maximum allowable Power Range Neutron Flux-High trip setpoints.

Reduction in the Overpower T trip setpoints (value of K4) by 1% for each 1% by which

)

Z

(

FC Q

exceeds its limit, is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions.

The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is sufficient considering the small likelihood of a severe transient in this time period, and the preceding prompt reduction in THERMAL POWER in accordance with Required Action A.1. The maximum allowable Overpower T trip setpoints initially determined by Required Action A.3 may be affected by subsequent Q

determinations of FC(Z) and would require Overpower T trip C

Q setpoint reductions within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the F (Z) determination, if necessary to comply with the decreased

FQ(Z)

B 3.2.1 BRAIDWOOD UNITS 1 & 2 B 3.2.1 6 Revision BASES ACTIONS (continued) maximum allowable Overpower T trip setpoints. Decreases in

)

Z

(

FC Q

would allow increasing the maximum allowable Overpower T trip setpoints.

A.4 Verification that

)

Z

(

FC Q

has been restored to within its limit, by performing SR 3.2.1.1 and SR 3.2.1.2 prior to increasing THERMAL POWER above the limit imposed by Required Action A.1, ensures that core conditions during operation at higher power levels and future operation are consistent with safety analyses assumptions.

Condition A is modified by a Note that requires Required Action A.4 to be performed whenever the Condition is entered. This ensures that SR 3.2.1.1 and SR 3.2.1.2 will be performed prior to increasing THERMAL POWER above the limit of Required Action A.1, even when Condition A is exited prior to performing Required Action A.4. Performance of SR 3.2.1.1 and SR 3.2.1.2 are necessary to assure FQ(Z) is properly evaluated prior to increasing THERMAL POWER.

B.1 If it is found that the maximum calculated value of FQ(Z) that can occur during normal maneuvers,

)

Z

(

FW Q

, exceeds its specified limits, there exists a potential for

)

Z

(

FC Q

to become excessively high if a normal operational transient occurs. Reducing THERMAL POWER as specified in the COLR within the allowed Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , maintains an acceptable absolute power density such that even if a transient occurred, core peaking factors are not exceeded.

B.2 If it is found that the maximum calculated value of FQ(Z) that can occur during normal maneuvers,

)

Z

(

FW Q

, exceeds its specified limits, there exists a potential for

)

Z

(

FC Q

to become excessively high if a normal operational transient occurs. Reducing the AFD limits as specified in the COLR within the allowed Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months restricts the axial flux distribution such that even if a transient occurred, core peaking factors are not exceeded.

FQ(Z)

B 3.2.1 BRAIDWOOD UNITS 1 & 2 B 3.2.1 7 Revision BASES ACTIONS (continued)

B.3 A reduction of the Power Range Neutron Flux-High trip setpoints by 1% for each 1% by which the maximum allowable THERMAL POWER is reduced is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is sufficient considering the small likelihood of a severe transient in this time period and the preceding prompt reduction in THERMAL POWER and AFD limits in accordance with Required Actions B.1 and B.2.

B.4 Reduction in the Overpower T trip setpoints (value of K4) by 1% for each 1% by which the maximum allowable THERMAL POWER is reduced is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is sufficient considering the small likelihood of a severe transient in this time period and the preceding prompt reduction in THERMAL POWER and AFD limits in accordance with Required Actions B.1 and B.2.

B.5 Verification that

)

Z

(

FW Q

has been restored to within its limit, by performing SR 3.2.1.1 and SR 3.2.1.2 prior to increasing THERMAL POWER and AFD limits above the limit imposed by Required Actions B.1 and B.2, ensures that core conditions during operations at higher power levels and future operation are consistent with safety analyses assumptions.

Condition B is modified by a Note that requires Required Action B.5 to be performed whenever the Condition is entered. This ensures that SR 3.2.1.1 and SR 3.2.1.2 will be performed prior to increasing THERMAL POWER above the limit of Required Action B.1, even when Condition B is exited prior to performing Required Action B.5. Performance of SR 3.2.1.1 and SR 3.2.1.2 are necessary to assure FQ(Z) is properly evaluated prior to increasing THERMAL POWER.

FQ(Z)

B 3.2.1 BRAIDWOOD UNITS 1 & 2 B 3.2.1 8 Revision BASES ACTIONS (continued)

C.1 If the Required Actions of A.1 through A.4, or B.1 through B.5, are not met within their associated Completion Times, the unit must be placed in a MODE or condition in which the LCO requirements are not applicable. This is done by placing the unit in at least MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The allowed Completion Time is reasonable based on operating experience regarding the amount of time it takes to reach MODE 2 from full power operation in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.1.1 and SR 3.2.1.2 are modified by a Note (i.e.,

REQUIREMENTS Note 1) that applies during the first power ascension after a refueling. It states that THERMAL POWER may be increased until an equilibrium power level has been achieved at which a meaningful power distribution map can be obtained. These SRs are normally performed at > 40% RTP to provide core conditions as much like the full power conditions as possible (Ref. 5). This allowance is modified, however, by one of the Frequency conditions that requires verification that

)

Z

(

FC Q

and

)

Z

(

FW Q

are within their specified limits after a power rise of more than 10% RTP (and establishing equilibrium conditions) over the THERMAL POWER at which they were last verified to be within specified limits. Because

)

Z

(

FC Q

and

)

Z

(

FW Q

could not have previously been measured in this reload core, there is a second Frequency condition, applicable only for reload cores, that requires determination of these parameters before exceeding 75% RTP.

This ensures that some determination of

)

Z

(

FC Q

and

)

Z

(

FW Q

are made at a lower power level at which adequate margin is available before going to 100% RTP. Also, this Frequency condition, together with the Frequency condition requiring verification of

)

Z

(

FC Q

and

)

Z

(

FW Q

following a power increase of more than 10%, ensures that FQ(Z) is verified as soon as RTP (or any other level for extended operation) is achieved. In the absence of these Frequency conditions, it is possible to increase power to RTP and operate for 31 days without verification of

)

Z

(

FC Q

and

)

Z

(

FW Q

. The Frequency condition is not intended to require verification of these parameters after every 10% increase in power level above the last verification. It only requires verification after a power level is achieved for extended operation that is 10% higher than that power at which FQ(Z) was last measured.

FQ(Z)

B 3.2.1 BRAIDWOOD UNITS 1 & 2 B 3.2.1 10 Revision BASES SURVEILLANCE REQUIREMENTS (continued)

This Surveillance has been modified by two Notes. Note 2 requires the measured value of FC(Z)

Q be obtained from incore flux map results only when PDMS is inoperable. Note 2 modifies the required performance of the Surveillance and states that this Surveillance is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after declaring PDMS inoperable, and that the last performance of SR 3.2.1.3 prior to declaring PDMS inoperable satisfies the initial performance of this SR after declaring PDMS inoperable. If SR 3.2.1.1 was not performed within its specified Frequency, this Note allows 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to verify FC(Z)

Q is within limit using either the incore flux map results or by taking credit for the last performance of SR 3.2.1.3 when PDMS was OPERABLE.

SR 3.2.1.2 The nuclear design process includes calculations performed to determine that the core can be operated within the FQ(Z) limits. Because flux maps are taken in steady state conditions, the variations in power distribution resulting from normal operational maneuvers are not present in the flux map data. These variations are, however, conservatively calculated by considering a wide range of unit maneuvers in normal operation. The maximum peaking factor increase over steady state values, calculated as a function of core elevation, Z, is called W(Z). Multiplying the measured total peaking factor,

)

Z

(

FC Q

, by W(Z) gives the maximum FQ(Z) calculated to occur in normal operation,

)

Z

(

FW Q

The limit with which

)

Z

(

FW Q

is compared varies inversely with power above 50% RTP and directly with the function K(Z) provided in the COLR.

The W(Z) curve is provided in the COLR for discrete core elevations. Flux map data are typically taken for 61 core elevations.

FQ(Z)

B 3.2.1 BRAIDWOOD UNITS 1 & 2 B 3.2.1 11 Revision BASES SURVEILLANCE REQUIREMENTS (continued)

)

Z

(

FW Q

evaluations are not applicable for the following axial core regions, measured in percent of core height:

a.

Lower core region, from 0 to 15% inclusive; and b.

Upper core region, from 85 to 100% inclusive.

If the top and bottom exclusion zones are reduced to 8%,

then

)

Z

(

FW Q

evaluations are not applicable for the following axial core regions, measured in percent of core height:

a.

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

Upper core region, from 92 to 100% inclusive.

Typically, the top and bottom 15% 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. However, the top and bottom exclusion zones can be reduced to 8% if the predicted transient peak FQ(Z) is located within the top and bottom 8%

to 15% of the core. The reduction of the top and bottom exclusion zones from 15% to 8% of the core still meets the

)

Z

(

FC Q

measurement uncertainty of 5%.

This Surveillance has been modified by a three Notes. Note 2 that may require that more frequent surveillances be performed. If

)

Z

(

FW Q

is evaluated, an evaluation of the expression below is required to account for any increase to

)

Z

(

FM Q

that may occur and cause the FQ(Z) limit to be exceeded before the next required FQ(Z) evaluation.

If the two most recent FQ(Z) evaluations show an increase in any of the expressions Increase of maximum over z [

)

Z

(

FC Q

/ K(Z)] since the previous evaluation, Increase of maximum over z [

)

Z

(

FW Q

/ K(Z)] since the previous evaluation, Expected increase of maximum over z [

)

Z

(

FW Q

/ K(Z)] prior to the next evaluation,

FQ(Z)

B 3.2.1 BYRON UNITS 1 & 2 B 3.2.1 12 Revision BASES SURVEILLANCE REQUIREMENTS (continued) it is required to meet the FQ(Z) limit with the last

)

Z

(

FW Q

increased by an appropriate factor specified in the COLR (Ref. 7), or to evaluate FQ(Z) more frequently, each 7 EFPD.

The last condition is met if:

max [(

W(Z, Bn+1)) / K(Z)] >

max [(

W(Z, Bn)) / K(Z)]

Where Bn is burnup when the surveillance is performed and Bn+1 is the burnup when the next surveillance will be performed.

These alternative requirements prevent FQ(Z) from exceeding its limit for any significant period of time without detection.

Note 3 requires the measured value of FW(Z)

Q be obtained from Q

incore flux map results only when PDMS is inoperable.

Note 3 modifies the required performance of the Surveillance and states that this Surveillance is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after declaring PDMS inoperable, and that the last performance of SR 3.2.1.4 prior to declaring PDMS inoperable satisfies the initial performance of this SR after declaring PDMS inoperable. If SR 3.2.1.2 were not performed within its specified Frequency, this Note allows 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to verify FW(Z) is within limit using either the incore flux map results or by taking credit for the last performance of SR 3.2.1.4 when PDMS was OPERABLE.

Performing the Surveillance in MODE 1 prior to exceeding 75% RTP ensures that the FQ(Z) limit is met when RTP is achieved, because peaking factors are generally decreased as power level is increased.

FQ(Z) is verified at power levels 10% RTP above the THERMAL POWER of its last verification, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving equilibrium conditions to ensure that FQ(Z) is within its limit at higher power levels.

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

PDMS is OPERABLE.

SR 3.2.1.4 The confirmation of the power distribution parameter, FW(Z) is an additional verification over the automated monitoring performed by PDMS. This assures that PDMS is functioning properly and that the core limits are met. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This Surveillance is modified by a Note that requires the performance of SR 3.2.1.3 for determining FC(Z) only when FQ(Z)

B 3.2.1 BRAIDWOOD UNITS 1 & 2 B 3.2.1 13 Revision BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.2.1.3 The confirmation of the power distribution parameter, FC(Z)

Q Q

is an additional verification over the automated monitoring performed by PDMS. This assures that PDMS is functioning properly and that the core limits are met.

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

This Surveillance is modified by a Note that requires the performance of SR 3.2.1.4 for determining FW(Z) only when PDMS is OPERABLE.

REFERENCES 1.

10 CFR 50.46.

2.

UFSAR, Section 15.4.8.

3.

10 CFR 50, Appendix A, GDC 26.

4.

DeletedWCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System," August 1994.

5.

ANSI/ANS-19.6.1-1985, "Reload Startup Physics Test for Pressurized Water Reactors," December 13, 1985.

6.

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

7.

WCAP-10216-P-A, Revision 1A, "Relaxation of Constant Axial Offset Control (and) FQ Surveillance Technical Specification," February 1994.

N H

F B 3.2.2 BRAIDWOOD UNITS 1 & 2 B 3.2.2 1 Revision B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (

N H

F)

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

N H

F is defined as the ratio of the integral of the linear power along the fuel rod with the highest integrated power to the average integrated fuel rod power. Therefore, N

H F is a measure of the maximum total power produced in a fuel rod.

N H

F is sensitive to fuel loading patterns, control bank insertion, and fuel burnup.

N H

F typically increases with control bank insertion and typically decreases with fuel burnup.

When Power Distribution Monitoring System (PDMS) is inoperable, F N

H is not directly measurable but is inferred from a power distribution map obtained with the movable incore detector system. Specifically, the results of the three dimensional power distribution map are analyzed by a computer to determine F N

H. This factor is calculated at least every 31 EFPD. However, during power operation, the global power distribution is monitored by LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)," which address directly and continuously measured process variables.However, during power operation when PDMS is inoperable, 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.

During power operation when PDMS is OPERABLE, the linear power along the fuel rod with the highest integrated power is measured continuously and F N

H is determined continuously, and global power distribution continues to be monitored by LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)."

N H

F B 3.2.2 BRAIDWOOD UNITS 1 & 2 B 3.2.2 3 Revision BASES APPLICABLE SAFETY ANALYSES (continued)

The allowable F N

H limit increases with decreasing power level. This functionality in F N

H is included in the analyses that provide the Reactor Core Safety Limits (SLs) of SL 2.1.1. Therefore, any DNB events in which the calculation of the core limits is modeled implicitly use this variable value of F N

H in the analyses. Likewise, all transients that may be DNB limited are assumed to begin with an initial F N

H as a function of power level defined by the COLR limit equation.

The Nuclear Enthalpy Rise Hot Channel Factor (F N

H), the Nuclear Heat Flux Hot Channel Factor (FQ(Z)), and the axial peaking factors are supported by the LOCA safety analyses that verify compliance with the 10 CFR 50.46 acceptance criteria (Ref. 3).

The fuel is protected in part by Technical Specifications, which ensure that the initial conditions assumed in the safety and accident analyses remain valid. The following LCOs ensure this: LCO 3.1.6, "Control Bank Insertion Limits," LCO 3.2.1, "Heat Flux Hot Channel Factor (FQ(Z)),"

LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor (F N

H),"

LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR).", and LCO 3.2.5, "Departure from Nucleate Boiling Ratio (DNBR)."

N FH and FQ(Z) are measured periodically using the movable incore detector system when PDMS is inoperable.

Measurements are generally taken with the core at, or near, steady state conditions. Core monitoring and control under transient conditions (Condition 1 events) are accomplished by operating the core within the limits of the LCOs on AFD, QPTR, and Control Bank Insertion Limits. When PDMS is OPERABLE, F N

H and FQ(Z) are determined continuously. Core monitoring and control under transient conditions (Condition 1 events) are accomplished by operating the core within the limits of the LCOs on DNBR and Control Bank Insertion Limits.

N H

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

N H

F B 3.2.2 BRAIDWOOD UNITS 1 & 2 B 3.2.2 5 Revision BASES LCO N

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

The N

H F 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 FN H

described by the equation contained in the COLR, is the design radial peaking factor used in the plant safety analyses.

The power multiplication factor in this equation provides margin for higher radial peaking from reduced thermal feedback and greater control rod insertion at low power levels. The limiting value of N

H F is allowed to increase 0.3% for every 1% RTP reduction in THERMAL POWER.

APPLICABILITY The N

H F limits must be maintained in MODE 1 to prevent core power distributions from exceeding the fuel design limits for DNBR and the 10 CFR 50.46 acceptance criteria (Ref.3).

Applicability in other modes is not required because there is either insufficient stored energy in the fuel or insufficient energy being transferred to the reactor coolant to require a limit on the distribution of core power. Specifically, the design bases events that are sensitive to N

H F in other modes (MODES 2 through 5) have significant margin to DNB, and therefore, there is no need to restrict N

H F in these modes.

ACTIONS A.1, A.2, and A.3, and A.4 With N

H F exceeding its limit, Condition A is entered.

N H

F may be restored to within its limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , through, for example, realigning any misaligned rods or reducing power enough to bring N

H F within its power dependent limit. If the value of N

H F is not restored to within its specified limit, THERMAL POWER must be reduced to

< 50% RTP in accordance with Required Action A.1.2.1. When the N

H F limit is exceeded, the DNBR limit is not likely violated in steady state operation, because events that could significantly perturb the N

H F value (e.g., static control rod misalignment) are considered in the safety analyses.

N H

F B 3.2.2 BRAIDWOOD UNITS 1 & 2 B 3.2.2 6 Revision BASES ACTIONS (continued)

However, the DNBR limit may be violated if a DNB limiting event occurs. Reducing THERMAL POWER to < 50% RTP increases the DNB margin and is not likely to cause the DNBR limit to be violated in steady state operation. Thus, the allowed Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months provides an acceptable time to restore N

H F to within its limits without allowing the unit to remain in an unacceptable condition for an extended period of time.

Condition A is modified by a Note that requires that Required Actions A.2 and A.34 must be completed whenever Condition A is entered. Thus, even if N

H F is restored within the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months time period of Required Action A.1.1, Required Action A.2 would nevertheless require another measurement and calculation of N

H F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in accordance with SR 3.2.2.1. Required Action A.34 requires that another determination of N

H F must be done prior to exceeding 50% RTP, prior to exceeding 75% RTP, and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reaching or exceeding 95% RTP.

Required Action A.2 requires the measured value of N

H F

verified not to exceed the allowed limit at the lower power level once the power level has been reduced to < 50% RTP per Required Action A.1.2.1. The unit is provided 20 additional hours to perform this task over and above the 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allowed by either Action A.1.1 or Action A.1.2.1 The Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is acceptable because of the increase in the DNB margin, which is obtained at lower power levels, and the low probability of having a DNB limiting event within this 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period. Additionally, operating experience has indicated that this Completion Time is sufficient to obtain the incore flux map when PDMS is inoperable, perform the required calculations, and evaluate N

H F.

If the value of N

H F is not restored to within its specified limit either by adjusting a misaligned rod or by reducing THERMAL POWER, Required Action A.1.2.23 requires the Power Range Neutron Flux-High trip setpoints be reduced to 55% RTP. The reduction in trip setpoints ensures that continuing operation remains at an acceptable low power level with adequate DNBR margin.

N H

F B 3.2.2 BRAIDWOOD UNITS 1 & 2 B 3.2.2 7 Revision BASES ACTIONS (continued)

The allowed Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to reset the trip setpoints per Required Action A.1.2.23 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.

Required Action A.34 requires verification that N

H Fis within its specified limits after an out of limit occurrence. This ensures that the cause that led to the N

H F exceeding its limit is corrected, and that subsequent operation proceeds within the LCO limit. This Action demonstrates that the N

H F

limit is within the LCO limits prior to exceeding 50% RTP, again prior to exceeding 75% RTP, and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is 95% RTP.

This Required Action is modified by a Note that states that THERMAL POWER does not have to be reduced to comply with this Required Action.

B.1 If the Required Actions of A.1.1 through A.4 are not met within their associated Completion Times, the unit must be placed in a MODE in which the LCO requirements are not applicable. This is done by placing the unit in at least MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The allowed Completion Time is reasonable, based on operating experience regarding the amount of time it takes to reach MODE 2 from full power operation in an orderly manner and without challenging plant systems.

N H

F B 3.2.2 BRAIDWOOD UNITS 1 & 2 B 3.2.2 8 Revision BASES SURVEILLANCE SR 3.2.2.1 REQUIREMENTS The value of N

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

H F from the measured flux distributions. The measured value of N

H F must be multiplied by 1.04 to account for measurement uncertainty before making comparisons to the N

H Flimit.

After each refueling, N

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

H F

limits are met at the beginning of each fuel cycle.

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

This Surveillance has been modified by a Note. The Note requires the measured value of F N

H be obtained from incore flux map results only when PDMS is inoperable. The Note modifies the required performance of the Surveillance and states that this Surveillance is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after declaring PDMS inoperable, and that the last performance of SR 3.2.2.2 prior to declaring PDMS inoperable satisfies the initial performance of this SR after declaring PDMS inoperable. If SR 3.2.2.1 were not performed within its specified Frequency, this Note allows 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to verify F N

H is within limit using either the incore flux map results or by taking credit for the last performance of SR 3.2.2.2 when PDMS was OPERABLE.

N H

F B 3.2.2 BRAIDWOOD UNITS 1 & 2 B 3.2.2 9 Revision BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.2.2.2 The confirmation of the power distribution parameter, F N

H, is an additional verification over the automated monitoring performed by PDMS. This assures that PDMS is functioning properly and that the core limits are met.

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

This Surveillance is modified by a Note that requires the performance of SR 3.2.2.2 for determining F N

H only when PDMS is OPERABLE.

REFERENCES 1.

UFSAR, Section 15.4.8.

2.

10 CFR 50, Appendix A, GDC 26.

3.

10 CFR 50.46.

QPTR B 3.2.4 BRAIDWOOD UNITS 1 & 2 B 3.2.4 1 Revision 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 testing, after refueling, and periodically during power operation.

The power density at any point in the core must be limited so that the fuel design criteria are maintained. Together, LCO 3.1.6, "Control Bank Insertion Limits," LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, provide limits on process variables that characterize and control the three dimensional power distribution of the reactor core. Control of these variables ensures that the core operates within the fuel design criteria and that the power distribution remains within the bounds used in the safety analyses. When Power Distribution Monitoring System (PDMS) is OPERABLE, Peak Linear Heat Rate and the linear power along the fuel rod with the highest integrated power are measured continuously.

APPLICABLE Limits on QPTR preclude core power distributions that SAFETY ANALYSES violate the following fuel design criteria:

a.

During a Loss Of Coolant Accident (LOCA) the 10 CFR 50.46 acceptance criteria must be met (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 prompt energy deposition to the fuel must not exceed 200 cal/gm (Ref. 2); and

QPTR B 3.2.4 BRAIDWOOD UNITS 1 & 2 B 3.2.4 2 Revision BASES APPLICABLE SAFETY ANALYSES (continued) d.

The control rods must be capable of shutting down the reactor with a minimum required Shutdown Margin 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 (FQ(Z)), the Nuclear Enthalpy Rise Hot Channel Factor

),

F

( N H

and control bank insertion, sequence and overlap limits are established to preclude core power distributions that exceed the safety analyses limits.

The QPTR limits ensure that N H

F and FQ(Z) remain below their limiting values by preventing an undetected change in the gross radial power distribution.

In MODE 1, the N H

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

The limits on the QPTR provide assurance that the thermal limits assumed in the accident analysis ( N H

F and FQ(Z)) are met. Thereby, the QPTR satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The QPTR limit of 1.02, at which corrective action is required, provides a margin of protection for both the DNB ratio and linear heat generation rate contributing to excessive power peaks resulting from X-Y plane power tilts.

A limiting QPTR of 1.02 can be tolerated before the margin for uncertainty in FQ(Z) and N H

F is possibly challenged.

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

Applicability in MODE 1 50% RTP and in other MODES is not required because there is neither sufficient stored energy in the fuel nor sufficient energy being transferred to the reactor coolant to require the implementation of a QPTR limit on the distribution of core power. The QPTR limit in these conditions is, therefore, not important. Note that the N H

F and FQ(Z) LCOs still apply below 50% RTP, but allow

QPTR B 3.2.4 BRAIDWOOD UNITS 1 & 2 B 3.2.4 3 Revision BASES APPLICABILITY (continued) progressively higher peaking factors as THERMAL POWER decreases below 50% RTP.

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

The maximum allowable power level initially determined by Required Action A.1 may be affected by subsequent determinations of QPTR. Increases in QPTR would require power reductions within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of QPTR determination, if necessary to comply with the decreased maximum allowable power level. Decreases in QPTR would allow increasing the maximum allowable power level and increasing power up to this revised limit.

A.2 After completion of Required Action 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 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter. A 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Completion Time is sufficient because any additional change in QPTR would be relatively slow. periodic monitoring provides a basis for maintaining the appropriate reduced power level. As such, a check of the QPTR is required once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . If the QPTR continues to increase, THERMAL POWER has to be reduced accordingly, such that it is maintained at a reduced power level of 3% from RTP for each 1% by which QPTR exceeds 1.00.

Any of the Surveillance methods for determining QPTR may be used within the constraints for acceptability of the Surveillance (i.e., if the excore detectors are available, they should be used; if the excore detectors are not available, the moveable incore detectors may be used). A 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Completion Time is sufficient because any additional change in QPTR should be relatively slow. Further, this Completion Time is consistent with the Frequency required for the Surveillances with an inoperable alarm or instrumentation.

QPTR B 3.2.4 BRAIDWOOD UNITS 1 & 2 B 3.2.4 6 Revision BASES ACTIONS (continued)

Required Action A.6 is modified by a Note that states that the peaking factor surveillances may only be done after the excore detectors have been normalized to restore QPTR to within limits (i.e., Required Action A.5). The intent of this Note is to have the peaking factor surveillances performed at operating power levels, which can only be accomplished after the excore detectors are normalized to restore QPTR to within limits and the core returned to power.

B.1 If Required Actions A.1 through A.6 are not completed within their associated Completion Times, the unit must be brought to a MODE or condition in which the requirements do not apply. To achieve this status, THERMAL POWER must be reduced to 50% RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The allowed Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is reasonable, based on operating experience regarding the amount of time required to reach the reduced power level without challenging plant systems.

SURVEILLANCE SR 3.2.4.1 REQUIREMENTS This Surveillance verifies that the QPTR, as indicated by the Nuclear Instrumentation System (NIS) excore channels, is within its limits. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR is modified by three two Notes. Note 1 allows QPTR to be calculated with three power range channels if THERMAL POWER is 75% RTP and the input from one Power Range Neutron Flux channel is inoperable. Note 2 allows performance of SR 3.2.4.2 in lieu of SR 3.2.4.1. Note 3 modifies the required performance of the Surveillance and states that this Surveillance is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after declaring PDMS inoperable.

If SR 3.2.4.1 were not performed within its specified Frequency, this Note allows 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to verify QPTR is within limits.

For those causes of QPT that occur quickly (e.g., a dropped rod), there typically are other indications of abnormality that prompt a verification of core power tilt.

QPTR B 3.2.4 BRAIDWOOD UNITS 1 & 2 B 3.2.4 8 Revision BASES SURVEILLANCE REQUIREMENTS (continued)

This Surveillance is modified by two a Notes. Note 1which states that it is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the input from one Power Range Neutron Flux channel is inoperable and the THERMAL POWER is > 75% RTP.

Note 2 modifies the required performance of the Surveillance and states that this Surveillance is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after declaring PDMS inoperable.

If SR 3.2.4.2 were not performed within its specified Frequency, this Note allows 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to verify QPTR is 1.02 using the movable incore detectors.

REFERENCES 1.

10 CFR 50.46.

2.

UFSAR, Section 15.4.8.

3.

10 CFR 50, Appendix A, GDC 26.

DNBR B 3.2.5 BRAIDWOOD UNITS 1 & 2 B 3.2.5 1 Revision B 3.2 POWER DISTRIBUTION LIMITS B 3.2.5 DeletedDeparture from Nucleate Boiling Ratio (DNBR)

BASES BACKGROUND The purpose of the limits on the value of DNBR determined by Power Distribution Monitoring System (PDMS) is to provide assurance of fuel integrity during Condition I (Normal Operation and Operational Transients) and Condition II (Faults of Moderate Frequency) events by providing the reactor operator with the information required to avoid exceeding the minimum Axial Power Shape Limiting DNBR (DNBRAPSL) in the core during normal operation and in short-term transients.

DNBR is defined as the ratio of the heat flux required to cause Departure from Nucleate Boiling (DNB) to the actual channel heat flux for given conditions.

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

During power operation when PDMS is OPERABLE, DNBR is determined continuously.

Continuously monitoring the operation of the core significantly limits the adverse nature of power distribution initial conditions for transients. The core depletion status, xenon distribution, and soluble boron concentration restrict the possible power and reactivity transients. Continuously monitoring the power distribution allows the actual DNBR value to be maintained the DNBRAPSL value specified in the COLR. DNBRAPSL is the DNBR value determined to be the most sensitive to the core axial power distribution at the initial conditions of the limiting accident during the cycle-specific core reload design accident analysis process.

DNBR B 3.2.5 BRAIDWOOD UNITS 1 & 2 B 3.2.5 2 Revision BASES APPLICABLE This LCO precludes core power distributions that violate SAFETY ANALYSES the following fuel design criteria:

During a loss of forced reactor coolant flow accident, there must be at least 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB.

The DNB safety analysis limit for a loss of forced reactor coolant flow accident (Ref. 1) is met by limiting DNBR to the 95/95 DNB design criterion of 1.4 using the WRB-2 Critical Heat Flux (CHF) correlation. This value provides a high degree of assurance that the hottest fuel rod in the core does not experience DNB.

Maintaining the DNBRAPSL value the DNBR value assumed in the safety and accident analyses ensures that the 95/95 DNB design criterion of 1.4 is met.

The fuel is protected in part by Technical Specifications, which ensure that the initial conditions assumed in the safety and accident analyses remain valid. When PDMS is OPERABLE, this LCO and the following LCOs ensure this:

LCO 3.1.6, LCO 3.2.1, "Heat Flux Hot Channel Factor (FQ(Z))," and LCO 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor(

H F

N ). When PDMS is inoperable, the following LCOs ensure this: LCO 3.1.6, LCO 3.2.1, LCO 3.2.2, and LCO 3.2.4.

In addition, LCO 3.2.3 ensures that the initial conditions assumed in the safety and accident analyses remain valid regardless of PDMS operability.

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

LCO DNBR shall be maintained within the limit of the relationship specified in the COLR.

Maintaining DNBR DNBRAPSL ensures the core operates within the limits assumed in the safety analyses. The DNBRAPSL limit must be maintained to prevent core power distributions from exceeding the fuel design limits for DNBR.

Another limit on DNBR is provided in SL 2.1.1, "Reactor Core SLs." LCO 3.2.5 represents the initial conditions of the safety analysis which are far more restrictive than the Safety Limit (SL). Should a violation of this LCO occur, the operator must check whether or not an SL may have been exceeded.

DNBR B 3.2.5 BRAIDWOOD UNITS 1 & 2 B 3.2.5 3 Revision BASES APPLICABILITY The DNBR limit must be maintained in MODE 1 with THERMAL POWER 50% RTP when PDMS is OPERABLE to ensure DNB design criteria will be met in the event of an unplanned loss of forced coolant flow transient.

ACTIONS A.1 Parameters affecting DNBR include Reactor Coolant System (RCS) pressure, RCS average temperature, RCS total flow rate, and Thermal Power. RCS pressure and RCS average temperature are controllable and measurable parameters. RCS total flow rate is not a controllable parameter and is not expected to vary during steady state operation. With DNBR not within limit due to RCS pressure or RCS average temperature, action must be taken to restore these parameter(s). With DNBR not within limit due to the indicated RCS total flow rate, power must be reduced, as required by Required Action B.1, to restore DNB margin and eliminate the potential for violation of the accident analysis bounds.

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time for restoration of DNBR provides sufficient time to adjust unit parameters, to determine the cause for the off normal condition, and to restore the readings within limits, and is based on plant operating experience.

B.1 If the value of DNBR is not restored to within its specified limit, THERMAL POWER must be reduced to < 50% RTP in accordance with Required Action B.1.

Reducing THERMAL POWER to < 50% RTP increases the DNB margin and is not likely to cause the DNBR limit to be violated in steady state operation. Thus, the allowed Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months provides an acceptable time to restore DNBR to within its limits without allowing the unit to remain in an unacceptable condition for an extended period of time.

DNBR B 3.2.5 BRAIDWOOD UNITS 1 & 2 B 3.2.5 4 Revision BASES SURVEILLANCE SR 3.2.5.1 REQUIREMENTS The confirmation of the power distribution parameter, DNBR, is an additional verification over the automated monitoring performed by PDMS. This assures that PDMS is functioning properly and that the core limits are met.

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

REFERENCES 1.

UFSAR, Chapter 15.

Proposed Bases Changes (Mark-Up)

Byron Station, Units 1 and 2 NRC Docket Nos. 50-454 and 50-455 (For Information Only)

Rod Group Alignment Limits B 3.1.4 BYRON UNITS 1 & 2 B 3.1.4 8 Revision BASES ACTIONS (continued)

However, in many cases, realigning the remainder of the group to the misaligned rod may not be desirable. For example, realigning control bank B to a rod that is misaligned 15 steps from the top of the core would require a significant power reduction, since control bank D must be fully inserted and control bank C must be partially inserted.

With a misaligned rod, SDM must be verified to be within limit (specified in the COLR) or boration must be initiated to restore SDM to within limit.

Power operation may continue with one RCCA trippable but misaligned, provided that SDM is verified within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

The Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months represents the time necessary for determining the actual unit SDM and, if necessary, aligning and starting the necessary systems and components to initiate boration to restore SDM to within limit.

B.2, B.3, B.4, and B.5 For continued operation with a misaligned rod, THERMAL POWER must be reduced when Power Distribution Monitoring System (PDMS) is inoperable, SDM must periodically be verified within limits (specified in the COLR), hot channel factors (FQ(Z) and

)

F

( N H

must be verified within limits, and the safety analyses must be re-evaluated to confirm continued operation is permissible.

Reduction of power to 75% RTP when PDMS is inoperable, 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 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months gives the operator sufficient time to accomplish an orderly power reduction without challenging the Reactor Protection System. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."

In this Required Action, the Completion Time only begins on discovery that both:

a.

One rod is not within alignment limit; and b.

PDMS is inoperable.

Rod Group Alignment Limits B 3.1.4 BYRON UNITS 1 & 2 B 3.1.4 9 Revision BASES ACTIONS (continued)

Discovering one rod not within alignment limit coincident with PDMS inoperable results in starting the Completion Time for the Required Action. During power operation when PDMS is OPERABLE, LHR is measured continuously. Therefore, a reduction of power to 75% RTP is not necessary to ensure that local LHR increases due to a misaligned RCCA will not cause the core design criteria to be exceeded.

When a rod is known to be misaligned, there is a potential to impact the SDM. Since the core conditions can change with time, periodic verification of SDM is required. A Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient to ensure this requirement continues to be met.

Verifying that FQ(Z) and N

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

H F.

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

Accident analyses (Ref. 3) requiring re-evaluation for continued operation with a misaligned rod include:

1.

Increase in heat removal by the secondary system:

a.

Excessive increase in secondary steam flow, b.

Inadvertent opening of a steam generator power operated relief or safety valve, and c.

Steam system piping failure;

Rod Group Alignment Limits B 3.1.4 BYRON UNITS 1 & 2 B 3.1.4 11 Revision BASES ACTIONS (continued)

C.2 If more than one rod is found to be misaligned or becomes misaligned because of bank movement when PDMS is inoperable, the unit conditions may fall outside of the accident analysis assumptions. Since automatic bank sequencing would continue to cause misalignment, 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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." In this Required Action, the Completion Time only begins on discovery that both:

a.

More than one rod is not within alignment limit; and b.

PDMS is inoperable.

Discovering more than one rod not within alignment limit coincident with PDMS inoperable results in starting the Completion Time for the Required Action.

C.3 If more than one rod is found to be misaligned or becomes misaligned because of bank movement when PDMS is OPERABLE, operation may continue in Condition C for a period that should not exceed 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The allowed Completion Time is reasonable, based on the available information on power distributions (Ref. 6). This Required Action is modified by a Note that requires the performance of Required Action C.3 only when PDMS is OPERABLE.

Rod Group Alignment Limits B 3.1.4 BYRON UNITS 1 & 2 B 3.1.4 12 Revision BASES ACTIONS (continued)

D.1 When Required Actions of Condition B or C.3 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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , which obviates concerns about the development of undesirable xenon or power distributions. The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging the plant systems.

SURVEILLANCE SR 3.1.4.1 REQUIREMENTS Verification that the position of individual rods is within alignment limits at a Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months provides a history that allows the operator to detect a rod that is beginning to deviate from its expected position. When a rods alignment cannot be verified due to a DRPI failure, the position of the rod can be determined by use of the movable incore detectors and/or PDMS. The position of the rod may be determined from the difference between the measured core power distribution and the core power distribution expected to exist based on the position of the rod indicated by the group step counter demand position.

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

The SR is modified by a note that permits it to not be performed for rods associated with an inoperable demand position indicator or an inoperable rod position indicator.

The alignment limit is based on the demand position indicator which is not available if the indicator is inoperable. LCO 3.1.7, "Rod Position Indication," provides Actions to verify the rods are in alignment when one or more rod position indicators are inoperable.

Rod Group Alignment Limits B 3.1.4 BYRON UNITS 1 & 2 B 3.1.4 14 Revision BASES REFERENCES 1.

10 CFR 50, Appendix A, GDC 10 and GDC 26.

2.

10 CFR 50.46.

3.

UFSAR, Chapter 15.

4.

UFSAR, Section 15.4.3.

5.

UFSAR, Section 15.1.5.

6.

DeletedWCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System," August 1994.

Rod Position Indication B 3.1.7 BYRON UNITS 1 & 2 B 3.1.7 6 Revision BASES ACTIONS (continued)

A.1 and A.2 When one DRPI per group in one or more groups fails, (i.e.,

one rod position per group can not be determined by the DRPI System) the position of the rod can still be determined by use of the movable incore detectors or Power Distribution Monitoring System (PDMS). The Required Action may also be satisfied by ensuring at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months that FQ satisfies LCO 3.2.1, FNH satisfies LCO 3.2.2, and SHUTDOWN MARGIN is within the limits provided in the COLR, provided the nonindicating rods have not been moved. When PDMS is OPERABLE, the position of the rod may be determined from the difference between the measured core power distribution and the core power distribution expected to exist based on the position of the rod indicated by the group step counter demand position. Based on experience, normal power operation does not require excessive movement of banks. If a bank has been significantly moved, the Required Action of BC.1 or BC.2 below is required. Therefore, verification of RCCA position within the Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is adequate for allowing continued full power operation, since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small.

Required Action A.1 requires verification of the position of a rod with an inoperable DRPI once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months which may put excessive wear and tear on the moveable incore detector system when PDMS is inoperable; Required Action A.2 provides an alternative. Required Action A.2 requires verification of rod position every 31 EFPD, which coincides with the normal surveillance frequency for verification of core power distribution.

Required Action A.2 includes six distinct requirements for verification of the position of rods associated with an inoperable DRPI:

a.

Initial verification within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of the inoperability of the DRPI; b.

Re-verification once every 31 Effective Full Power days (EFPD) thereafter;

FQ(Z)

B 3.2.1 BYRON UNITS 1 & 2 B 3.2.1 1 Revision B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 Heat Flux Hot Channel Factor (FQ(Z))