RS-18-035, License Amendment Request to Revise Technical Specification 3.2.3, Axial Flux Difference

License Amendment Request to Revise Technical Specification 3.2.3, Axial Flux Difference

ML18092B081
Person / Time
Site:	Byron, Braidwood
Issue date:	04/02/2018
From:	Gullott D Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-18-035, TAC MA8252, TAC MA8253, TAC MA8254, TAC MA8255
Download: ML18092B081 (40)

Text

Exelon Generation 4300 Winfield Road Warrenville , IL 60555 www.exeloncorp.com RS-18-035 10 CFR 50.90 Aprit 2, 2018 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. STN 50-456 and STN 50-457 Byron Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455

Subject:

License Amendment Request to Revise Technical Specification 3.2.3, "Axial Flux Difference"

Reference:

Letter from G. F. Dick, Jr. (NRC) to 0. D. Kingsley (Exelon Generation Company),

"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 Number ML010510325)

In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit or early site permit," Exelon Generation Company, LLC, (EGC) requests amendments to Facility Operating License Nos. NPF-72 and NPF-77for Braidwood Station, Units 1 and 2, and Facility Operating License Nos. NPF-37 and NPF-66 for Byron Station, Units 1 and 2. This amendment request proposes to revise Technical Specification (TS) 3.2.3, "Axial Flux Difference (AFD)."

In the referenced letter, the NRC approved amendments for Braidwood Station and Byron Station to implement the Best Estimate Analyzer for Core Operations - Nuclear (BEACON') Power Distribution Monitoring System (PDMS). One of the changes in these amendments was a revision to TS 3.2.3 to incorporate the use of PDMS. TS 3.2.3 currently requires that AFD shall be maintained within the limits specified in the Core Operating Limits Report (COLR) when in MODE 1 with Thermal Power ~ 50% Reactor Thermal Power (RTP) when PDMS is inoperable.

Conversely, when PDMS is operable, there are no limitations on AFD for any RTP level. In October 2016, a nonconformance (i.e., a modeling error) was discovered in the Fuel Rod Design (FRO) performance criteria verification analysis. Specifically, while modeling certain American

April 2, 2018 U. S. Nuclear Regulatory Commission Page2 Nuclear Society (ANS) Condition 11 events, it was discovered that not all FRO performance criteria were satisfied for all values of AFD when~ 50% RTP. However, all FRO performance criteria remain satisfied at~ 50% RTP when AFD is maintained within the AFD limits specified in the COLR. A compensatory action to maintain AFD within the COLR AFD limits at-all-times when

~ 50% RTP was immediately implemented. Maintaining AFD within the COLR limits at-all-times when ~ 50% RTP was/is the normal operating practice as specified in plant procedures; however, this practice was re-emphasized in a Standing Order at both Braidwood Station and Byron Station. These Standing Orders remain in place pending approval of this proposed amendment.

This proposed amendment would revise TS 3.2.3 to require that AFD be maintained within the limits specified in the COLR during MODE 1 with Thermal Power ~ 50% RTP (regardless of the status of PDMS). An associated change would also be made to the NOTE modifying Surveillance Requirement (SR) 3.2.3.1.

The attached request is subdivided as follows:

- Attachment 1 provides a description and evaluation of the proposed changes.

- Attachment 2A provides the markup of the affected Braidwood Station TS page.

- Attachment 2B provides the markup of the affected Byron Station TS page.

- Attachment 3A provides the markup of the affected Braidwood Station TS Bases pages (for information only).

- Attachment 3B provides the markup of the affected Byron Station TS Bases pages (for information only).

The proposed amendment has been reviewed by the Braidwood Station and Byron Station Plant Operations Review Committees in accordance with the requirements of the EGC Quality Assurance Program.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," paragraph (b),

EGC is notifying the State of Illinois of this application for a license amendment by transmitting a copy of this letter and its attachments to the designated State of Illinois official.

EGC requests approval of the proposed license amendment request within one year of this submittal date; i.e., by April 2, 2019. Once approved, the amendment shall be implemented within 30 days.

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

April 2, 2018 U.S. Nuclear Regulatory Commission Page 3 I declare under penalty of perjury that the foregoing is true and correct. Executed on the 2nd day of April 2018.

Respectfully, David M. Gullatt Manager - Licensing Exelon Generation Company, LLC Attachments:

1. Evaluation of Proposed Changes 2A. Markup of Technical Specifications Page - Braidwood Station 2B. Markup of Technical Specifications Page - Byron Station 3A. Markup of Technical Specifications Bases Pages - Braidwood Station 3B. Markup of Technical Specifications Bases Pages- Byron Station cc: NRC Regional Administrator, Region 111 NRC Senior Resident Inspector, Braidwood Station NRC Senior Resident Inspector, Byron Station Illinois Emergency Management Agency- Division of Nuclear Safety

ATTACHMENT 1 Evaluation of Proposed Changes

Subject:

License Amendment Request to Revise Technical Specification 3.2.3, "Axial Flux Difference" 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 Changes to Technical Specification Requirements 2.2 Reason for the Proposed Changes

3.0 TECHNICAL EVALUATION

3.1 Current BEACON' I PDMS Design and Licensing Basis 3.2 Nonconformance Resolution Technical Justification

4.0 REGULATORY EVALUATION

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

6.0 REFERENCES

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ATTACHMENT 1 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," Exelon Generation Company, LLC, (EGC) requests amendments to Facility Operating License Nos. NPF-72 and NPF-77 for Braidwood Station, Units 1 and 2, and Facility Operating License Nos. NPF-37 and NPF-66 for Byron Station, Units 1 and 2. This amendment request proposes to revise Technical Specification (TS) 3.2.3, "Axial Flux Difference (AFD)."

In the Reference 1, the NRC approved amendments for Braidwood Station and Byron Station to implement the Best Estimate Analyzer for Core Operations - Nuclear (BEACON') Power Distribution Monitoring System (PDMS). One of the changes in these amendments was a revision to TS 3.2.3 to incorporate the use of PDMS. TS 3.2.3 currently requires that AFD shall be maintained within the limits specified in the Core Operating Limits Report (COLR) when in MODE 1 with Thermal Power ;;:: 50% Reactor Thermal Power (RTP) when PDMS is inoperable.

Conversely, when PDMS is operable, there are no limitations on AFD for any RTP level. In October 2016, a nonconformance (i.e., a modeling error) was discovered in the Fuel Rod Design (FRO) performance criteria verification analysis. This issue was entered into the Corrective Actions Program at both sites. Specifically, while modeling certain American Nuclear Society (ANS) Condition 11 events it was discovered that not all FRO performance criteria were satisfied for all values of AFD when ;;:: 50% RTP. However, all FRO performance criteria remain satisfied at ;;:: 50% RTP when AFD is maintained within the AFD limits specified in the COLR. A compensatory action to maintain AFD within the COLR AFD limits at-all-times when;;:: 50% RTP was immediately implemented. Maintaining AFD within the COLR limits at-all-times when

50% RTP was/is the normal operating practice as specified in plant procedures; however, this practice was re-emphasized in a Standing Order at both Braidwood Station and Byron Station.

These Standing Orders remain in place pending approval of this proposed amendment.

This proposed amendment would revise TS 3.2.3 to require that AFD be maintained within the limits specified in the COLR during MODE 1 with Thermal Power;;:: 50% RTP (regardless of the status of PDMS (described in Section 3.1 below)). An associated change would also be made to the NOTE modifying Surveillance Requirement (SR) 3.2.3.1.

2.0 DETAILED DESCRIPTION 2.1 Changes to Technical Specification Requirements This amendment request proposes to revise Technical Specification (TS) 3.2.3, "Axial Flux Difference (AFD)." Specifically, TS 3.2.3 would be revised to delete "when Power Distribution Monitoring System (PDMS) is inoperable" from the Limiting Condition for Operation (LCO)

Applicability statement and to delete the Note above Surveillance Requirement (SR) 3.2.3.1 that currently reads: "Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after declaring PDMS inoperable."

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ATTACHMENT 1 Evaluation of Proposed Changes The following specific TS changes are proposed:

TS 3.2.3, "Axial Flux Difference (AFD)"

The TS 3.2.3 LCO Applicability currently states the following:

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

The proposed amendment would revise this LCO Applicability statement to delete the reference to the PDMS and read as follows:

APPLICABILITY: MODE 1 with THERMAL POWER;::; 50% RTP.

Surveillance Requirement (SR) 3.2.3.1 Surveillance Requirement (SR) 3.2.3.1 specifies to "Verify AFD is within limits for each OPERABLE excore channel." This SR is currently modified by a Note that states the following:

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

The proposed amendment would delete this Note.

The marked-up TS pages showing the proposed changes are provided in Attachments 2A and 2B for Braidwood Station and Byron Station, respectively.

Conforming TS Bases change markups, provided for information only, are included in Attachments 3A and 3B for Braidwood Station and Byron Station, respectively. The TS Bases changes will be incorporated following NRC approval of the amendment request in accordance with TS 5.5.14, "Technical Specifications (TS) Bases Control Program." Associated changes to the Technical Requirements Manual (TRM), Core Operating Limits Report (COLR) and Updated Final Safety Analysis Report (UFSAR) will also be made upon approval of the proposed amendment.

The detailed justification for these changes is presented below in Section 3.0, "Technical Evaluation."

2.2 Reason for the Proposed Changes As noted above, a nonconformance (i.e., a modeling error) was discovered in the Fuel Rod Design (FRO) performance criteria verification analysis. As part of the normal cycle reload analysis, a confirmatory Condition II transient evaluation is performed to verify that the Over-Temperature Delta-Temperature (OT~T) and Over-Power Delta-Temperature (OP~T) reactor trip functions are sufficient to protect the TS 2.1.1 Safety Limits (i.e., Departure from Nucleate Boiling Ratio (DNBR) and peak fuel centerline temperature). The FRO performance criteria (e.g., cladding stress/strain discussed in Reference 5) are also evaluated as part of this transient evaluation.

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ATTACHMENT 1 Evaluation of Proposed Changes Note that Condition II events are defined as "Faults of Moderate Frequency." These faults, at worst, result in a reactor trip with the plant being capable of returning to operation. These faults (or events) do not propagate to cause a more serious fault, i.e., Condition Ill or IV events. In addition, Condition II events are not expected to result in fuel rod failures or reactor coolant system or secondary system overpressurization.

The FRO performance criteria verification analysis Condition II transient evaluation simulates the OTLiT trip function; however, the K2 and K3 constant terms in the OTLiT setpoint equation (shown below in Section 3.2) were not modeled. The Braidwood Station and Byron Station OTLiT actual, installed setpoints remain correctly implemented and uncompromised with this modeling nonconformance. This FRO verification modeling error resulted in a non-conservative trip setpoint in the analysis. This error created the potential for scenarios where the actual plant-installed OTLiT trip function would not have generated a reactor trip signal for some Condition II transients for which one of the FRO performance limits would be exceeded. These potential scenarios resulted in exceeding only the Westinghouse-specific fuel cladding offset yield stress limit of 0.2% at specific AFD values outside of the COLR AFD limits at~ 50% RTP.

Note that the 0.2% offset yield stress limit is specified in WCAP-12610-P-A (Reference 5).

Additional information regarding the consequences of this error are provided below in Section 3.0, "Technical Evaluation."

We again note that all FRO performance criteria remain satisfied at ~ 50% RTP when AFD is maintained within the AFD limits specified in the COLR. A compensatory action to maintain AFD within the COLR AFD limits at-all-times when ~ 50% RTP was immediately implemented.

Maintaining AFD within the COLR limits at-all-times when ~ 50% RTP was/is the normal operating practice as specified in plant procedures; however, this practice was re-emphasized in a Standing Order at both Braidwood Station and Byron Station. These Standing Orders remain in place pending approval of this proposed amendment as detailed above in Section 2.0.

Approval and implementation of the proposed amendment will formally eliminate the concern created by the subject nonconformance. The PDMS (described in Section 3.1 below) and the OT.llT trip function (i.e., DNBR protection) are unaffected by this modeling error and remain operable.

3.0 TECHNICAL EVALUATION

3.1 Current BEACON' I PDMS Design and Licensing Basis In Reference 1, Braidwood Station and Byron Station requested NRC approval to implement the Best Estimate Analyzer for Core Operations - Nuclear (BEACON') Power Distribution Monitoring System (PDMS). The NRC approved this request in Reference 2.

The BEACON' PDMS is an advanced core monitoring and operations support package which uses current instrumentation in conjunction with an analytical methodology for online generation of three-dimensional (3-D) core power distributions.

BEACON is a trademark of Westinghouse Electric Company LLC, its Affiliates and/or its Subsidiaries in the United States of America and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited.

Other names may be trademarks of their respective owners.

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ATTACHMENT 1 Evaluation of Proposed Changes The PDMS provides continuous core monitoring, core measurement reduction, core analysis, core follow, and core predictions. The PDMS maintains an on-line 3-D nodal model which is continuously updated to reflect the current plant operation conditions. The nodal solution method used by the PDMS is the same as the NRG-approved Westinghouse Advanced Nodal Code (ANC) core design code (Reference 7).

The Core Exit Thermocouples (CETs) and excore neutron flux detectors are used with the reference 3-D power distribution to determine the measured power distribution. By coupling the measured 3-D power distribution with an on-line evaluation, actual power distributions and core margins are better understood. The PDMS provides an indication of operating and design margins to address strategic fuel cycle operation changes.

The BEACON' methodology described in WCAP-12472-P-A (Reference 3) allows for changes in the core design methods and provides for more optimized core loading patterns. Additionally, the BEACON' methodology significantly improves the quality of the Technical Specification surveillance process since it uses a depleted model (i.e., a model that accounts for core burn up history) to match the actual operational profile. The PDMS continuously monitors the limiting Heat Flux Hot Channel Factor (Fa(Z)), Nuclear Enthalpy Rise Hot Channel Factor (FN~H), and DNBR; and enhances operational flexibility. The NRG-approved version of WCAP-12472-P-A was issued in August 1994.

Implementation of the PDMS at Braidwood Station and Byron Station did not replace, eliminate, or modify existing plant instrumentation. The PDMS software runs on a workstation connected to the plant process computer. The PDMS combines input from currently installed plant instrumentation and design data generated each fuel cycle. Together, this data provides a means to monitor power distribution limits continuously and to alert the operator when limits are being approached.

As stated above, when PDMS is operable, PDMS directly and continuously monitors the key power distribution parameters of Fa(z), FN~H, and DNBR. This direct monitoring capability eliminates the need to monitor the indirect indicators of AFD and Quadrant Power Tilt Ratio (QPTR). Technical Specification (TS) 3.2.1, "Heat Flux Hot Channel Factor (Fa(z))," and TS 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor (FN~H)," allow direct margin monitoring for demonstrating compliance with these LCOs. The requirements specified in TS 3.2.3, "Axial Flux Difference (AFD)," and TS 3.2.4, "Quadrant Power Tilt Ratio (QPTR)," are not applicable when the PDMS is operable.

As part of the implementation of BEACON', Braidwood Station and Byron Station use the Relaxed Axial Offset Control (RAOC) analysis approach for determining AFD limits when PDMS is inoperable. This analytical method is discussed in WCAP-10216-P-A, Revision 1 (Reference 4). RAOC was developed for relaxing the constraints on axial power distribution control. This was achieved by examination of a wide range of possible xenon distributions and the possible range of axial power distributions associated with each xenon distribution in both normal operation and accident conditions. Each power shape generated was examined to determine if LOCA limits were met or exceeded. The result of this examination was the development of an AFD band displayed as a function of power which meets the LOCA limits.

The power shapes within this range are then examined each cycle to ascertain whether they 5of13

ATTACHMENT 1 Evaluation of Proposed Changes meet the thermal-hydraulic constraints associated with DNB acceptance criteria as well as meet the LOCA limits. The AFD limits are then revised accordingly if necessary.

3.2 Nonconformance Resolution Technical Justification In WCAP-12472-P-A (i.e., BEACON') (Reference 3), it was shown that the precondition power distribution for the Condition II transients defined in WCAP-10216-P-A, Revision 1 (i.e., RAOC)

(Reference 4) was no longer constrained by the AFD limits. This topic was specifically addressed in WCAP 12472-P, Section 3.1, "BEACON Methodology," attached to the NRC Safety Evaluation for WCAP-12472-P-A.

The Condition II transients considered in Reference 4 (i.e., specifically addressed in Westinghouse Letter NS-EPR-2649, Section 11.D, "Condition II Analysis," included in WCAP-10216-P-A, Revision 1) are the following: 1) cooldown accident (manual rod control mode); 2) control rod withdrawal; and 3) boration I dilution (automatic rod control mode).

WCAP-12472-P-A (Reference 3) noted that since the requirements specified in TS 3.2.3, "Axial Flux Difference (AFD)," and TS 3.2.4, "Quadrant Power Tilt Ratio (QPTR)" are not applicable when the PDMS is operable, the transient precondition power distribution must only be constrained by operating within the COLR limits for rod insertion, power peaking, and DNBR.

As previously noted in Section 2.2 above, the Condition II transient confirmatory evaluation verifies that the protection provided by the OT~T and OP~T reactor trip functions is sufficient to protect the TS 2.1.1 Safety Limits for DNBR and peak fuel centerline temperature; and also verifies the FRO performance criteria (e.g. cladding stress/strain). It was discovered that, as currently modeled in the subject FRO confirmatory analysis, the OT~T reactor trip function simulated a non-conservative trip setpoint with respect to the FRO performance criteria. This error is specifically attributed to the fact that the K2 and K3 constant terms in the OT~T setpoint equation were not modeled in the FRO confirmatory analysis.

The OT~T reactor trip setpoint equation, as shown in TS 3.3.1, Table 3.3.1-1, is provided below for reference.

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ATTACHMENT 1 Evaluation of Proposed Changes 5

AT {l+fi ) [

1

• -- ]

( l+f2 s) l+f3 s

~ A To {K1 - Kz ( l+f4 s)

( l+fs s)

[r l

( l+f6 s)

- T1 ] + K3 ( P - P'l - f1 (A I>}

Where: AT is measured Reactor Coo1ant System (RCS) AT, °F.

AT0 is the indicated AT at RTP, f. 0 s is the Laplace tran sform operator, sec-1

• T is the rreasured RCS average temperature, °F.

r' i s the nominal Tavg at RTP, :s; *.

P is the rreasured pressurizer pressure, psig .

p' is the oomi na l RCS operating pressure, = *.

Ki=* Kz = * ~=*

Ti=* Tz =

• T3 ~

• L, =

• Ts=* T6 ~

• f 1 (AI} = *{* + (Ch: - ~)} when Ch: - q, <

• RTP 0% of RTP when

• RTP :s; ~ - q, ~

• RTP

• HC4t - q) - *J when <lt - q, >

• RTP Where Cit and q, are percent RTP in the upper and lower halv*es of the core, respective1y, and Cit + ~ is the tota 1 THERMAL POWER in percent RTP.

• As specified in the COLR.

In the FRO performance criteria verification analyses, the incorrectly modeled OTAT reactor trip function would not have appropriately terminated some Condition II transients causing the FRO fuel cladding offset yield stress limit of 0.2% to be exceeded (for specific AFD values outside the COLR AFD limits at ~ 50% RTP). It is important to note that this issue is not a nuclear safety concern, as exceeding the yield stress may potentially cause plastic deformation of the cladding; however, does not cause fuel failure in the accident scenarios considered for the OTfl T trip. As previously noted, the subject Condition II transients resulted in only exceeding the fuel cladding 0.2% offset yield stress limit at specific AFD values outside the COLR AFD limits at ~ 50% RTP.

It is also noted that the 0.2% offset yield stress limit is a Westinghouse-specific design constraint specified in WCAP-12610-P-A (Reference 5). NUREG-0800, "Standard Review Plan" (SRP), Section 4.2, "Fuel System Design," does not specify any cladding stress limit; rather, instead of transient cladding stress, SRP 4.2 specifies that transient cladding strain must be less than 1% from the pre-transient condition. This SRP 4.2 transient strain limit is met for Braidwood Station and Byron Station when considering the OTAT and OPAT actual nominal trip setpoints currently implemented at the stations.

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ATTACHMENT 1 Evaluation of Proposed Changes All FRO performance criteria remain satisfied at ~ 50% RTP when AFD is maintained within the AFD limits specified in the COLR. It was specifically confirmed that the AFD limits in the COLR, which were developed using the RAOC methodology, remain valid. Therefore, to eliminate the concerns created by this modeling error, the proposed amendment would revise TS 3.2.3 to require that AFD be maintained within the limits specified in the COLR during MODE 1 with Thermal Power ~ 50% RTP (regardless of the status of PDMS). Note that there continues to be no limitation on AFD when < 50% RTP.

It is again noted that BEACON'/PDMS and the OT~T trip function remain operable and uncompromised by the subject modeling nonconformance. This issue impacts only the FRO performance criteria verification of the operational space allowed when PDMS is operable, and does not suggest that the OT~T trip setpoint is incorrectly implemented at the plant. The on-line monitoring function of PDMS remains unaffected by this issue. The PDMS continues to calculate Fa(z), FN8H and DNBR margins correctly and perform its intended function. The PDMS margin alarm functions also remain unaffected.

The Westinghouse topical reports discussed in this submittal (i.e., References 3, 4, 5 and 6) are also listed in TS 5.6.5, "Core Operating Limits Report (COLR)." These topical reports remain valid and continue to support all aspects of BEACON'/PDMS (as currently implemented) with the exception that the transient precondition power distribution will now also (after approval of the proposed amendment) be constrained by the COLR limits for AFD to meet the FRO performance criteria (in addition to the COLR limits on rod insertion, peaking factors and DNBR).

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The proposed amendment has been evaluated to determine whether the applicable regulations and requirements, noted below, continue to be met.

General Design Criterion 10, "Reactor Design," states that the reactor core and associated coolant, control, and protection systems shall be designed with an appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

General Design Criterion 26, "Reactivity Control System Redundancy and Capability," states that two independent reactivity control systems of different design principles shall be provided.

One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure that the acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions.

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ATTACHMENT 1 Evaluation of Proposed Changes EGC has determined that the proposed changes remain in conformance with the above regulatory requirements.

4.2 No Significant Hazards Consideration In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit or early site permit," Exelon Generation Company, LLC, (EGC) requests amendments to Facility Operating License Nos. NPF-72 and NPF-77 for Braidwood Station, Units 1 and 2, and Facility Operating License Nos. NPF-37 and NPF-66 for Byron Station, Units 1 and 2. This amendment request proposes to revise Technical Specification (TS) 3.2.3, "Axial Flux Difference (AFD)."

TS 3.2.3 currently requires that AFD shall be maintained within the limits specified in the Core Operating Limits Report (COLR) when in MODE 1 with Thermal Power;::: 50% Reactor Thermal Power (RTP) when the Power Distribution Monitoring System (PDMS) is inoperable.

Conversely, when PDMS is operable, there are no limitations on AFD for any RTP level. In October 2016, a nonconformance (i.e., a modeling error) was discovered in the Fuel Rod Design (FRO) performance criteria verification analysis. This issue was entered into the Corrective Actions Program at both sites. Specifically, while modeling certain American Nuclear Society (ANS) Condition 11 events it was discovered that not all FRO performance criteria were satisfied for all values of AFD when ;::: 50% RTP. However, all FRO performance criteria remain satisfied at ;::: 50% RTP when AFD is maintained within the AFD limits specified in the COLR. A compensatory action to maintain AFD within the COLR AFD limits at-all-times when ;::: 50% RTP was immediately implemented. Maintaining AFD within the COLR limits at-all-times when

50% RTP was/is the normal operating practice as specified in plant procedures; however, this practice was re-emphasized in a Standing Order at both Braidwood Station and Byron Station.

These Standing Orders remain in place pending approval of this proposed amendment.

This proposed amendment would revise TS 3.2.3 to require that AFD be maintained within the limits specified in the COLR during MODE 1 with Thermal Power;::: 50% RTP (regardless of the status of PDMS). An associated change would also be made to the NOTE modifying Surveillance Requirement (SR) 3.2.3.1.

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

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

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ATTACHMENT 1 Evaluation of Proposed Changes EGC has evaluated the proposed amendment for Braidwood Station and Byron Station, using the criteria in 10 CFR 50.92, and has determined that the proposed amendment does not involve a significant hazards consideration. The following information is provided to support a finding of no significant hazards consideration.

Criteria

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

Response: No.

The proposed amendment requires that the AFD be maintained within the limits specified in the COLR at-all-times during MODE 1 when reactor power is ~ 50% RTP. This requirement will ensure that all FRO performance criteria remain satisfied during ANS Condition 11 events (i.e., Faults of Moderate Frequency); thus, ensuring the integrity of the fuel rod cladding. It is noted that maintaining AFD within the COLR limits at-all-times when ~ 50% RTP is the normal operating practice as specified in plant procedures.

The proposed change will have no impact on accident initiators or precursors; does not alter accident analysis assumptions; does not involve any physical plant modifications that would alter the design or configuration of the facility, or the manner in which the plant is maintained; and does not impact the probability of operator error.

The proposed amendment will not impact the ability of structures, systems, and components (SSCs) from performing their intended functions to mitigate the consequences of an accident. All accident analysis acceptance criteria will continue to be met as the proposed change will not affect the source term, containment isolation function, or radiological release assumptions for any accident previously evaluated.

Based on the above discussion, the proposed amendment 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 kind of accident from any accident previously evaluated?

Response: No.

The proposed change formalizes the existing operating practice of maintaining the AFD within the limits specified in the COLR at-all-times during MODE 1 when reactor power is

~ 50% RTP. This change ensures that all FRO performance criteria remain satisfied during ANS Condition II events. The ANS Condition II events have all been previously evaluated in the Updated Final Safety Analysis Report.

The proposed change does not involve a design change or other changes that would impact safety-related SSCs from performing their specified safety functions.

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ATTACHMENT 1 Evaluation of Proposed Changes The proposed change does not result in the creation of any new accident precursors; does not result in changes to any existing accident scenarios; and does not introduce any operational changes or mechanisms that would create the possibility of a new or different kind of accident.

Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The proposed change to maintain the AFD within the limits specified in the COLR at-all-times during MODE 1 when reactor power is ~ 50% RTP ensures that all FRO performance criteria remain satisfied during ANS Condition II events; and thus, will maintain the existing margin of safety related to FRO performance criteria and ensure the integrity of the fuel rod cladding. The AFD limits specified in the COLR have been established in accordance with the analysis approach described in NRG-approved Westinghouse Topical Reports.

In addition, this change will have no impact on the margin of safety associated with other reactor core safety parameters such as fuel hot channel factors, core power tilt ratios, loss of coolant accident peak cladding temperature and peak local power density.

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

Based on the above, EGC concludes that the proposed amendments 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 Based on the evaluation presented above, the proposed revision to TS 3.2.3 requiring that AFD be maintained within the limits specified in the COLR when in MODE 1 with reactor power

~ 50% RTP will safely resolve the modeling error identified in the FRO performance criteria verification analysis. This change formalizes the existing normal operating practice specified in plant procedures and will have no impact on the results of the existing transient and accident analysis.

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 site licensing basis and 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.

11 of 13

ATTACHMENT 1 Evaluation of Proposed Changes

5.0 ENVIRONMENTAL CONSIDERATION

EGC has evaluated this proposed operating license amendment consistent with the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21, "Criteria for and identification of licensing and regulatory actions requiring environmental assessments." EGC has determined that the proposed amendment to revise Technical Specification (TS) 3.2.3, "Axial Flux Difference (AFD)," requiring that AFD be maintained within the limits specified in the Core Operating Limits Report when in MODE 1 with reactor power 2:!: 50% rated thermal power, meets the criteria for a categorical exclusion set forth in paragraph (c)(9) of 10 CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review," and as such, has determined that no irreversible consequences exist in accordance with paragraph (b) of 10 CFR 50.92, "Issuance of amendment." This determination is based on the fact that the proposed amendment is being proposed as an amendment to the license issued pursuant to 10 CFR 50, "Domestic Licensing of Production and Utilization Facilities," which changes a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, "Standards for Protection Against Radiation," or which changes an inspection or a surveillance requirement, and the amendment meets the following specific criteria:

(i) The amendment involves no significant hazards consideration.

As demonstrated in Section 4.2, "No Significant Hazards Consideration," the proposed amendment does not involve any significant hazards consideration.

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

The proposed amendment does not result in an increase in power level, does not increase the production nor alter the flow path or method of disposal of radioactive waste or byproducts. It is expected that all plant equipment would operate as designed in the event of an accident to minimize the potential for any leakage of radioactive effluents.

The proposed amendment will have no impact on the amounts of radiological effluents released offsite during normal at-power operations or during the accident scenarios.

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

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

There is no change in individual or cumulative occupational radiation exposure due to the proposed amendment to revise TS 3.2.3 as stated above. Specifically, the proposed amendment has no impact on any radiation monitoring system setpoints. The proposed action will not change the level of controls or methodology used for processing of radioactive effluents or handling of solid radioactive waste, nor will the proposed action result in any change in the normal radiation levels within the plant.

12of13

ATTACHMENT 1 Evaluation of Proposed Changes Therefore, in accordance with 10 CFR 51.22, paragraph (b), no environmental impact statement or environmental assessment need be prepared in support of the proposed amendment.

6.0 REFERENCES

1. Letter from R. M. Krich (ComEd; now Exelon Generation Company) to NRC, "Request for Amendment to Technical Specifications for Byron and Braidwood Stations to Implement the Best Estimate Analyzer for Core Operations Nuclear Power Distribution Monitoring System,"

dated February 15, 2000.

2. Letter from G. F. Dick, Jr. (NRC) to 0. D. Kingsley (Exelon Generation Company), "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 Number ML010510325).

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

4. WCAP-10216-P-A, Revision 1, "Relaxation of Constant Axial Offset Control - Fa Surveillance Technical Specification," February 1994.

5. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," April 1995.

6. WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO'," July 2006.

7. WCAP-10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code," September 1986.

13of13

ATTACHMENT 2A Markup of Technical Specifications Page BRAIDWOOD STATION UNITS 1AND2 Docket Nos. 50-456 and 50-457 Facility Operating License Nos. NPF-72 and NPF-77 REVISED TS PAGE 3.2.3-1

AFD 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 AXIAL FLUX DIFFERENCE CAFD)

LCO 3.2.3 The AFD shall be maintained within the limits specified in the COLR.

---

------------NO TE------------ ----------------

The AFD shall be considered outside limits when two or more OPERABLE excore channels indicate AFD to be outside limits.

APPLICABILITY: MODE 1 with THERMAL POWER~ 50% RTP when Power Distribution Monitoring System CPDMS) is inoperable.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. AFD not within limits. A.l Reduce THERMAL POWER 30 minutes to < 50% RTP.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Verify AFD is within limits for each In accordance OPERABLE excore channel. with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 & 2 3.2.3 - 1 Amendment 165/165

ATTACHMENT 28 Markup of Technical Specifications Page BYRON STATION UNITS 1AND2 Docket Nos. 50-454 and 50-455 Facility Operating License Nos. NPF-37 and NPF-66 REVISED TS PAGE 3.2.3-1

AFD 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 AXIAL FLUX DIFFERENCE CAFD)

LCO 3.2.3 The AFD shall be maintained within the limits specified in the COLR.

---

------------NO TE------------ ----------------

The AFD shall be considered outside limits when two or more OPERABLE excore channels indicate AFD to be outside limits.

APPLICABILITY: MODE 1 with THERMAL POWER~ 50% RTP 1.'hen Pov,'er Distribution 1

Monitoring System CPDMS) is inoperable.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. AFD not within limits. A.1 Reduce THERMAL POWER 30 minutes to< 50% RTP.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Verify AFD is within limits for each In accordance OPERABLE excore channel . with the Surveillance Frequency Control Program BYRON - UNITS 1 &2 3.2.3 - 1 Amendment 171/171

ATTACHMENT 3A Markup of Technical Specifications Bases Pages BRAIDWOOD STATION UNITS 1AND2 Docket Nos. 50-454 and 50-455 Facility Operating License Nos. NPF-37 and NPF-66 REVISED BASES PAGES B 3.2.1-1 B 3.2.1-2 B 3.2.2-1 B 3.2.3-1 B 3.2.3-2 B 3.2.3-3 B 3.2.3-4 B 3.2.5-1 B 3.2.5-2

FaC Z)

B 3.2.1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 Heat Flux Hot Channel Factor CF0(Z))

BASES BACKGROUND The purpose of the limits on the values of F0(Z) is to limit the local Ci .e., pellet) peak power density. The value of F0(Z) varies along the axial height CZ) of the core.

F0(Z) is defined as the maximum local fuel rod linear power density (i.e., Peak Linear Heat Rate CPLHR)) divided by the average fuel rod linear power density, assuming nominal fuel pellet and fuel rod dimensions. Therefore, F0(Z) is a measure of the peak fuel pellet power within the reactor core.

During power operation when Power Distribution Monitoring System CPDMS) is inoperable, the global power distribution is limited by LCO 3.2.3, "AXIAL FLUX DIFFERENCE CAFD)," and LCO 3.2.4, "QUADRANT POWER TILT RATIO CQPTR)," which are directly and continuously measured process variables. These LCOs, along with LCO 3.1. 6, "Contra l 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). II F0(Z) varies with fuel loading patterns, control bank insertion, fuel burnup, and changes in axial power distribution.

F0(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, F0(Z) is determined continuously.

Using the measured three dimensional power distributions, it is possible to derive a measured value for F0(Z). However, because this value represents an equilibrium condition, it does not include the variations in the value of F0(Z) which are present during nonequilibrium situations, such as load following or power ascension.

To account for these possible variations, the equilibrium value of F0(Z) is adjusted as F~(Z) by an elevation dependent factor that accounts for the calculated worst case transient conditions.

BRAIDWOOD - UNITS 1 &2 B 3.2.1 - 1 Revision~

~(Z)

B 3.2.1 BASES BACKGROUND (continued)

When PDMS is inoperable, cCore 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 i noperable), and control rod insertion.

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

a. During a small break Loss Of Coolant Accident CLOCA) the peak cladding temperature must not exceed 2200°F and during a large break LOCA there must be a high level of probability that the peak cladding temperature does not exceed 2200°F (Ref. 1);

b. During a loss of forced reactor coolant flow accident, there must be at least 95% probability at the 95%

confidence level (the 95/95 Departure from Nucleate Boiling CDNB) 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 SOM with the highest worth control rod stuck fully withdrawn (Ref. 3).

Limits on F0(Z) ensure that the value of the initial total peaking factor assumed in the accident analyses remains valid. Other criteria must also be met (e.g., maximum cladding oxidation, maximum hydrogen generation, coolable geometry, and long term cooling). However, the peak cladding temperature is typically most limiting.

F0(Z) limits assumed in the LOCA analysis are typically limiting relative to Ci .e., lower than) the F0(Z) limit assumed in safety analyses for other postulated accidents.

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

F0(Z) satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

BRAIDWOOD - UNITS 1 &2 B 3.2.1 - 2 Revision~

F~H B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F~H)

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.

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

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

When Power Distribution Monitoring System CPDMS) is inoperable, F~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~H. However, during power operation when PDMS is inoperable, the global power distribution is monitored by LCO 3.2.3, "AXIAL FLUX DIFFERENCE CAFD)," and LCO 3.2.4, "QUADRANT POWER TILT RATIO CQPTR)," 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~H is determined continuously, and gl obal power di stributi on conti nues t o be mon itored by LCO 3.2.3, "AX IAL FLUX DIFFERENCE (AFO ). II BRAIDWOOD - UNITS 1 &2 B 3.2.2 - 1 Revision .g§

AFD B 3.2.3 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.3 AXIAL FLUX DIFFERENCE CAFD)

BASES BACKGROUND The purpose of this LCO is to establish limits on the values of the AFD in order to limit the amount of axial power distribution skewing to either the top or bottom of the core

'IY'hen Pm,'er Di stri buti on Mani tori ng System CPDMS) is inoperable. By limiting the amount of power distribution skewing, core peaking factors are consistent with the assumptions used in the safety analyses. Limiting power distribution skewing over time also minimizes the xenon distribution skewing, which is a significant factor in axial power distribution control. vJhen PDMS is OPERABLE, Peak Linear Heat Rate is measured continuously.

Relaxed Axial Offset Control CRAOC) (Ref. 2) is a calculational procedure that defines the allowed operational space of the AFD versus THERMAL POWER. The AFD limits are selected by considering a range of axial xenon distributions that may occur as a result of large variations of the AFD.

Subsequently, power peaking factors and power distributions are examined to ensure that the Loss of Coolant Accident CLOCA), loss of flow accident, and anticipated transient limits are met. Violation of the AFD limits invalidate the conclusions of the accident and transient analyses with regard to fuel cladding integrity.

The AFD is monitored on an automatic basis using the plant process computer, which has an AFD monitor alarm. The computer determines the 1 minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels is outside its specified limits.

BRAIDWOOD - UNITS 1 &2 B 3.2.3 - 1 Revision~

AFD B 3.2.3 BASES APPLICABLE The AFD is a measure of the axial power distribution skewing SAFETY ANALYSES to either the top or bottom half of the core. The AFD is sensitive to many core related parameters such as control bank positions, core power level, axial burnup, axial xenon distribution, and, to a lesser extent, reactor coolant temperature and boron concentration. The allowed range of the AFD is used in the nuclear design process to confirm that operation within these limits produces core peaking factors and axial power distributions that meet safety analysis requirements.

The RAOC methodology (Ref. 2) establishes a xenon distribution library with tentatively wide AFD limits . .QA-dimensional aAxial power distribution calculations are then performed to demonstrate that normal operation power shapes are acceptable for the LOCA and loss of flow accident, and for initial conditions of anticipated transients. The tentative limits are adjusted as necessary to meet the safety analysis requirements.

The limits on the AFD ensure that the Heat Flux Hot Channel Factor (F0(Z)) is not exceeded during either normal operation or in the event of xenon redistribution following power changes. The limits on the AFD also restrict the range of power distributions that are used as initial conditions in the analyses of Condition 2, 3, or 4 events.

This ensures that the fuel cladding integrity is maintained for these postulated accidents. The most important Condition 4 event is the LOCA. The most important Condition 3 event is the loss of flow accident. The most important Condition 2 events are uncontrolled bank withdrawal and boration or dilution accidents. Condition 2 accidents simulated to begin from within the AFD limits are used to confirm the adequacy of the Overpower AT and Overtemperature AT trip setpoints.

The limits on the AFD provide assurance that the thermal limits assumed in the accident analysis <F:H and F0(Z)) are met. Thereby, the AFD satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

BRAIDWOOD - UNITS 1 &2 B 3.2.3 - 2 Revision~

AFD B 3.2.3 BASES LCO The shape of the power profile in the axial Ci .e., the vertical) direction is largely under the control of the operator through the manual operation of the control banks or automatic motion of control banks. The automatic motion of the control banks is in response to temperature deviations resulting from manual operation of the Chemical and Volume Control System to change boron concentration or from power level changes.

Signals are available to the operator from the Nuclear Instrumentation System CNIS) excore neutron detectors (Ref. 3). Separate signals are taken from the top and bottom detectors. The AFD is defined as the difference in normalized flux signals between the top and bottom excore detectors in each detector well. For convenience, this flux difference is converted to provide flux difference units expressed as a percentage and labeled as %~flux or %~I.

A Note modifies the LCO by stating the AFD shall be considered outside limits when two or more OPERABLE excore channels indicate AFD to be outside limits.

The AFD limits are provided in the COLR. The AFD limits for RAOC do not depend on the target flux difference. However, the target flux difference may be used to minimize changes in the axial power distribution.

Violating this LCO on the AFD could produce unacceptable consequences if a Condition 2, 3, or 4 event occurs while the AFD is outside its specified limits.

APPLICABILITY The AFD requirements are applicable in MODE 1 with THERMAL POWER~ 50% RTP Ci .e., when the combination of THERMAL POWER and core peaking factors are of primary importance in safety analysis) when PDMS is inoperable.

For AFD limits developed using RAOC methodology, the value of the AFD does not affect the limiting accident consequences with THERMAL POWER< 50% RTP and for lower operating power MODES.

BRAIDWOOD - UNITS 1 &2 B 3.2.3 - 3 Revision~

AFD B 3.2.3 BASES ACTIONS As an alternative to restoring the AFD to within its specified limits, Required Action A.1 requires a THERMAL POWER reduction to< 50% RTP. This places the core in a condition for which the value of the AFD is not important in the applicable safety analyses. A Completion Time of 30 minutes is reasonable, based on operating experience, to reach 50% RTP without challenging plant systems.

SURVEILLANCE SR 3.2.3.1 REQUIREMENTS This Surveillance verifies that the AFD as indicated by the NIS excore channels is within limits. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

The AFD should be monitored and logged more frequently in periods of operation for which the power level or control bank positions are changing to allow corrective measures when the AFD is more likely to move outside limits.

/\ 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after declaring PDMS inoperable.

If SR 3.2.3.1 were not performed within its specified Frequency, this Note all 0 11s 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to verify /\FD is within 1

limits.

REFERENCES 1. WCAP-8403 Cnonproprietary), "Power Distribution Contra l and Load Fo 11 owing Procedures," Westinghouse Electric Corporation, September 1974.

2. R. W. Miller et al., "Relaxation of Constant Axial Offset Control: F0 Surveillance Technical Specification," WCAP-10217CNP), June 1983.

3. UFSAR, Section 7.7.1.3.1.

BRAIDWOOD - UNITS 1 &2 B 3.2.3 - 4 Revision .ge

DNBR B 3.2.5 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.5 Departure from Nucleate Boiling Ratio CDNBR)

BASES BACKGROUND The purpose of the limits on the value of DNBR determined by Power Distribution Monitoring System CPDMS) 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 CDNB~L) 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 CDNB) to the actual channel heat flux for given conditions.

During power operation '1.'hen PDMS is inoperable, the global power distribution is limited by LCO 3.2.3, "AXIAL FLUX DIFFERENCE CAFD)," 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, "Contra1 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 DNB~L value specified in the COLR.

DNB~L 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.

BRAIDWOOD- UNITS 1 &2 B 3.2.5 - 1 Revision~

DNBR B 3.2.5 BASES APPLICABLE This LCD 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 CCHF) correlation. This value provides a high degree of assurance that the hottest fuel rod in the core does not experience DNB. Maintaining the DNB~L 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:

LCD 3.1.6, LCD 3.2.1, "Heat Flux Hot Channel Factor CF0CZ))," and LCD 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor(F~H ). When PDMS is inoperable, the following LCOs ensure this: LCD 3.1.6, LCD 3.2.1, LCD 3.2.2, LCO 3.2.3, and LCD 3.2.4. In addi ti on, LCD 3.2.3 ensures that the i nitial cond itions assumed in the safety and accident ana lyses remai n val id rega rdl ess of PDMS operabi li ty.

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

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

Maintaining DNBR ~ DNB~L ensures the core operates within the limits assumed in the safety analyses. The DNB~L 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." LCD 3.2.5 represents the initial conditions of the safety analysis which are far more restrictive than the Safety Limit CSL). Should a violation of this LCD occur, the operator must check whether or not an SL may have been exceeded.

BRAIDWOOD- UNITS 1 &2 B 3.2.5 - 2 Revision~

ATTACHMENT 3B Markup of Technical Specifications Bases Pages BYRON STATION UNITS 1AND2 Docket Nos. 50-454 and 50-455 Facility Operating License Nos. NPF-37 and NPF-66 REVISED BASES PAGES B 3.2.1-1 B 3.2.1-2 B 3.2.2-1 B 3.2.3-1 B 3.2.3-2 B 3.2.3-3 B 3.2.3-4 B 3.2.5-1 B 3.2.5-2

F0(Z)

B 3.2.1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 Heat Flux Hot Channel Factor CF0(Z))

BASES BACKGROUND The purpose of the limits on the va l ues of F0(Z) is to limit the local Ci .e., pellet) peak power density. The value of F0(Z) varies along the axial height CZ) of the core.

F0(Z) is defined as the maximum local fuel rod linear power density Ci .e., Peak Linear Heat Rate CPLHR)) divided by the average fuel rod linear power density, assuming nominal fuel pellet and fuel rod dimensions. Therefore, F0(Z) is a measure of the peak fuel pellet power within the reactor core.

During power operation when Power Distribution Monitoring System CPDMS) is inoperable, the global power distribution is limited by LCO 3.2.3, "AXIAL FLUX DIFFERENCE CAFD)," 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 distri bution continues t o be l imited by LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFO ). II F0(Z) varies with fuel loading patterns, control bank insertion, fuel burnup, and changes in axial power distribution.

F0(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, F0(Z) is determined continuously.

Using the measured three dimensional power distributions, it is possible to derive a measured value for F0(Z). However, because this value represents an equilibrium condition, it does not include the variations in the value of F0(Z) which are present during nonequilibrium situations, such as load following or power ascension.

To account for these possible variations, the equilibrium value of F0(Z) is adjusted as F~(Z) by an elevation dependent factor that accounts for the calculated worst case transient conditions.

BYRON - UNITS 1 &2 B 3.2.1 - 1 Revision +e

Fa( Z)

B 3.2.1 BASES BACKGROUND (continued)

~Jhen PDMS is inoperable, cCore 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 small break Loss Of Coolant Accident (LOCA) the peak cladding temperature must not exceed 2200°F and during a large break LOCA there must be a high level of probability that the peak cladding temperature does not exceed 2200°F (Ref. 1);

b. During a loss of forced reactor coolant flow accident, there must be at least 95% probability at the 95%

confidence level (the 95/95 Departure from Nucleate Boiling CDNB) 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 SOM with the highest worth control rod stuck fully withdrawn (Ref. 3).

Limits on F0(Z) ensure that the value of the initial total peaking factor assumed in the accident analyses remains valid. Other criteria must also be met (e.g., maximum cladding oxidation, maximum hydrogen generation, coolable geometry, and long term cooling). However, the peak cladding temperature is typically most limiting.

F0(Z) limits assumed in the LOCA analysis are typically limiting relative to (i.e., lower than) the F0(Z) limit assumed in safety analyses for other postulated accidents.

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

F0(Z) satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

BYRON - UNITS 1 & 2 B 3.2.1 - 2 Revision .+/-g

F~H B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F~H)

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.

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

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

When Power Distribution Monitoring System CPDMS) is inoperable, F~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~H. However, during power operation when PDMS is inoperable, the global power distribution is monitored by LCO 3.2.3, "AXIAL FLUX DIFFERENCE CAFD), and 11 LCO 3.2.4, "QUADRANT POWER TILT RATIO CQPTR), which address 11 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~H is determined continuously ,

and global power distributi on continues t o be monitored by LCO 3.2. 3, "AXIAL FLUX DIFFERENCE (A FO ). II BYRON - UNITS 1 &2 B 3.2.2 - 1 Revision +e

AFD B 3.2.3 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.3 AXIAL FLUX DIFFERENCE CAFD)

BASES BACKGROUND The purpose of this LCO is to establish limits on the values of the AFD in order to limit the amount of axial power distribution skewing to either the top or bottom of the core

'*\'hen Power Di stri buti on Monitoring System ( PDMS) is inoperable. By limiting the amount of power distribution skewing, core peaking factors are consistent with the assumptions used in the safety analyses. Limiting power distribution skewing over time also minimizes the xenon distribution skewing, which is a significant factor in axial power di stri buti on control . When PDMS is OPER/\BLE, Peak Linear Heat Rate is measured continuously.

Relaxed Axial Offset Control CRAOC) (Ref. 2) is a calculational procedure that defines the allowed operational space of the AFD versus THERMAL POWER. The AFD limits are selected by considering a range of axial xenon distributions that may occur as a result of large variations of the AFD.

Subsequently, power peaking factors and power distributions are examined to ensure that the Loss of Coolant Accident CLOCA), loss of flow accident, and anticipated transient limits are met. Violation of the AFD-limits invalidate the conclusions of the accident and transient analyses with regard to fuel cladding integrity.

The AFD is monitored on an automatic basis using the plant process computer, which has an AFD monitor alarm. The computer determines the 1 minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels is outside its specified limits.

BYRON - UNITS 1 &2 B 3.2.3 - 1 Revision -le

AFD B 3.2.3 BASES APPLICABLE The AFD is a measure of the axial power distribution skewing SAFETY ANALYSES to either the top or bottom half of the core. The AFD is sensitive to many core related parameters such as control bank positions, core power level, axial burnup, axial xenon distribution, and, to a lesser extent, reactor coolant temperature and boron concentration. The allowed range of the AFD is used in the nuclear design process to confirm that operation within these limits produces core peaking factors and axial power distributions that meet safety analysis requirements.

The RAOC methodology CRef. 2) establishes a xenon distribution library with tentatively wide AFD limits . .Qfte dimensional aAxial power distribution calculations are then performed to demonstrate that normal operation power shapes are acceptable for the LOCA and loss of flow accident, and for initial conditions of anticipated transients. The tentative limits are adjusted as necessary to meet the safety analysis requirements.

The limits on the AFD ensure that the Heat Flux Hot Channel Factor CF0(Z)) is not exceeded during either normal operation or in the event of xenon redistribution following power changes. The limits on the AFD also restrict the range of power distributions that are used as initial conditions in the analyses of Condition 2, 3, or 4 events.

This ensures that the fuel cladding integrity is maintained for these postulated accidents. The most important Condition 4 event is the LOCA. The most important Condition 3 event is the loss of flow accident. The most important Condition 2 events are uncontrolled bank withdrawal and boration or dilution accidents. Condition 2 accidents simulated to begin from within the AFD limits are used to confirm the adequacy of the Overpower AT and Overtemperature AT trip setpoints.

The limits on the AFD provide assurance that the thermal limits assumed in the accident analysis CF~H and F0(Z)) are met. Thereby, the AFD satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

BYRON - UNITS 1 &2 B 3.2.3 - 2 Revision -+/-a

AFD B 3.2.3 BASES LCO The shape of the power profile in the axial (i.e., the vertical) direction is largely under the control of the operator through the manual operation of the control banks or automatic motion of control banks. The automatic motion of the control banks is in response to temperature deviations resulting from manual operation of the Chemical and Volume Control System to change boron concentration or from power level changes.

Signals are available to the operator from the Nuclear Instrumentation System CNIS) excore neutron detectors (Ref. 3). Separate signals are taken from the top and bottom detectors. The AFD is defined as the difference in normalized flux signals between the top and bottom excore detectors in each detector well. For convenience, this flux difference is converted to provide flux difference units expressed as a percentage and labeled as %~flux or %~I.

A Note modifies the LCO by stating the AFD shall be considered outside limits when two or more OPERABLE excore channels indicate AFD to be outside limits.

The AFD limits are provided in the COLR. The AFD limits for RAOC do not depend on the target flux difference. However, the target flux difference may be used to minimize changes in the axial power distribution.

Violating this LCO on the AFD could produce unacceptable consequences if a Condition 2, 3, or 4 event occurs while the AFD is outside its specified limits.

APPLICABILITY The AFD requirements are applicable in MODE 1 with THERMAL POWER~ 50% RTP Ci .e., when the combination of THERMAL POWER and core peaking factors are of primary importance in safety analysis) when PDMS is inoperable.

For AFD limits developed using RAOC methodology, the value of the AFD does not affect the limiting accident consequences with THERMAL POWER< 50% RTP and for lower operating power MODES.

BYRON - UNITS 1 &2 B 3.2.3 - 3 Revision +/-9

AFD B 3.2.3 BASES ACTIONS A.1 As an alternative to restoring the AFD to within its specified limits, Required Action A.1 requires a THERMAL POWER reduction to< 50% RTP. This places the core in a condition for which the value of the AFD is not important in the applicable safety analyses. A Completion Time of 30 minutes is reasonable, based on operating experience, to reach 50% RTP without challenging plant systems.

SURVEILLANCE SR 3.2.3.1 REQUIREMENTS This Surveillance verifies that the AFD as indicated by the NIS excore channels is within limits. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

The AFD should be monitored and logged more frequently in periods of operation for which the power level or control bank positions are changing to allow corrective measures when the AFD is more likely to move outside limits.

A 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after declaring PDMS inoperable.

If SR 3.2.3.1 were not performed within its specified Frequency, this Note al 1O'A'S 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to verify /\FD is vii thin limits.

REFERENCES 1. WCAP-8403 Cnonproprietary), "Power Distribution Control and Load Following Procedures," Westinghouse Electric Corporation, September 1974.

2. R. W. Miller et al., "Relaxation of Constant Axial Offset Control: F0 Surveillance Technical Specification," WCAP-10217CNP), June 1983.

3. UFSAR, Section 7.7.1.3.1.

BYRON - UNITS 1 &2 B 3.2.3 - 4 Revision+.§

DNBR B 3.2.5 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.5 Departure from Nucleate Boiling Ratio CDNBR) .

BASES BACKGROUND The purpose of the limits on the value of DNBR determined by Power Distribution Monitoring System CPDMS) 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 CDNB~L) 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 CDNB) to the actual channel heat flux for given conditions.

During power operation when PDMS is inoperable, the global power distribution is limited by LCO 3.2.3, "AXIAL FLUX DIFFERENCE CAFD)," and when PDMS is inoperabl e, LCO 3.2.4, "QUADRANT POWER TI LT RATIO CQPTR)," 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 DNB~L value specified in the COLR.

DNB~L 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.

BYRON - UNITS 1 &2 B 3.2.5 - 1 Revision -+/-e

DNBR B 3.2.5 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 CCHF) correlation. This value provides a high degree of assurance that the hottest fuel rod in the core does not experience DNB. Maintaining the DNBRwsL va 1ue 2 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 CF0(Z))," and LCO 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor(F~H ). When PDMS is inoperable, the following LCOs ensure this: LCO 3.1.6, LCO 3.2.1, LCO 3.2.2, LCO 3.2.3, and LCO 3.2.4. In additi on, LCO 3.2.3 ensures t hat the initial cond iti ons assumed i n t he safety and acci dent analyses remain valid rega rdless of PDMS operabi lity.

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

Another 1i mi t 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 CSL). Should a violation of this LCO occur, the operator must check whether or not an SL may have been exceeded.

BYRON - UNITS 1 &2 B 3.2.5 - 2 Revision -+/-9