RS-24-026, License Amendment to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Technical Specifications 5.6.5, Core Operating Limits Report (COLR)

License Amendment to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Technical Specifications 5.6.5, Core Operating Limits Report (COLR)

ML24116A112
Person / Time
Site:	Braidwood, Byron
Issue date:	04/25/2024
From:	Steinman R Constellation Energy Generation
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
RS-24-026
Download: ML24116A112 (1)

Text

4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office

RS-24-026 10 CFR 50.90 April 25, 202 4

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 to Braidwood Station, U nits 1 and 2, and Byron Station, Units 1 and 2, Technical Specifications 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)"

In accordance with 10 CFR 50.90, " Application for amendment of license, construction permit, or early site permit," Constellation Energy Generation, LLC (CEG ) is submitting a request for amendments to 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).

This license amendment request (LAR) proposes to revise the Technical Specifications (TS) to remove COLR analytical method 5.6.5.b.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."

provides the evaluation of the proposed Technical Specification change, including the description of the proposed change, the no significant hazards consideration determination, and the existing TS pages marked up to reflect the proposed change.

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

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

CEG requests approval of the proposed license amendment by April 30, 202 5. Once approved, the amendment will be implemented within 90 days.

The proposed amendment has been reviewed and approved by the Braidwo od and Byron Plant Operations Review Committees in accordance with the requirements of the CEG Quality Assurance Program.

April 25, 2024 U.S. Nuclear R egulatory Commission Page 2

There are n o regulatory commitments co ntained in this l etter. Should yo u have any questions concerning this l etter, please contact Ms. Lisa Zurawski at ( 779) 231-6796.

I declare under penal ty of perjury that t he foregoing is t rue and co rrect. Executed on t he 25 th day of April 2024.

Respectfully,

Rebecca L. Steinman Sr. M anager-Licensing Constellation Energy Generation, LLC

Attachments:

1) Evaluation of Proposed Change

2) Braidwood Mark-up of Technical Specifications Pages

3) Byron Mark-up of Technical Specifications Pages

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 ATTACHMENT 1 Evaluation of Proposed Change

Subject:

License Amendment to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Technical Specifications 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)"

1.0

SUMMARY

DESCRIPTION

2.0 DETAILED DESCRIPTION 2.1 Proposed Changes

3.0 TECHNICAL EVALUATION

3.1 Safety Analysis Results 3.2 Probabilistic Risk Analysis Results 3.3 Defense in Depth 3.4 Safety Margin 3.5 Define Implementation and Monitoring Program 3.6 Configuration Risk Management Program

4.0 REGULATORY EVALUATION

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

6.0 REFERENCES

Page 1 of 28 ATTACHMENT 1 Evaluation of Proposed Change

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) is submitting a request for amendments to 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).

This license amendment request (LAR) proposes to revise the Technical Specifications (TS) to remove COLR analytical method 5.6.5.b.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."

CEG proposes to eliminate the Anticipated Transients without SCRAM (ATWS) Moderator Temperature Limit from the Braidwood and Byron Technical Specifications.

The removal of this limit has no effect on any design or licensing basis evaluation other than ATWS.

The CEG Probabilistic Risk Analysis (PRA) demonstrates that the elimination of this limit results in a small, acceptable increase in the risk of Core Damage Frequency (CDF). The changes are within "Region III" (Very Small Changes) of Figure 4 ("Acceptance Guidelines for Core Damage Frequency") in NRC Regulatory Guide (RG) 1.174, Revision 3.

The CEG PRA demonstrates that the elimination of this limit results in a small, acceptable increase in the risk of Large Early Release Frequency (LERF). The changes are within " Region III" (Very Small Changes) of Figure 5 ("Acceptance Guidelines for Large Early Release Frequency") in NRC RG 1.174, Revision 3.

Defense in depth is sufficient. No modifications or other limitations are needed. Additional defense in depth does not provide significant addition safety value.

A configuration risk management program is not needed. The ATWS risks are acceptably low without any restrictions on Rod Control, Auxiliary Feedwater (AF), Pressurizer Power Operated Relief Valves (PORVs), or any other functions.

A monitoring plan is not needed. The proposed change does not change the design, operating characteristics, or reliability of any system, structure, or component (SSC).

2.0 DETAILED DESCRIPTION

2.1 Proposed Changes

CEG proposes to delete Technical Specification 5.6.5.b.5. This specification is one of the Core Operating Limits Report (COLR) references. This Technical Specification is one of many Page 2 of 28 ATTACHMENT 1 Evaluation of Proposed Change

analytical methods and limitations on the parameter " Moderator Temperature Coefficient (MTC)," Technical Specification 3.1.3. The MTC limit generated by the methods described in Technical Specification 5.6.5.b.5 only apply to the ATWS transient described in the Updated Final Safety Analysis Report (UFSAR) Chapter 15.8.

2.1.1 Proposed Licensing Basis Changes

CEG proposes to delete Technical Specification 5.6.5.b.5. This specification is one of the Core Operating Limits Report (COLR) references. Technical Specification 5.6.5.b.5 is an analytical method which CEG uses to confirm that a cycle specific core design can meet licensing basis criteria associated with Anticipated Transients W ithout SCRAM (ATWS) events (UFSAR Chapter 15.8). This specification affects the range of values for the Limiting Conditions for Operation of Technical Specification 3.1.3, Moderator Temperature Coefficient. The MTCs most positive value is limited by Technical Specification Figure 3.1.3-1 and cycle specific nuclear design constraints as described in the UFSAR and all other analytical methods listed in Technical Specification 5.6.5.b, including reference 5.6.5.b.5.

After implementing this change, the cycle specific limits and confirmations associated with COLR Reference 5.6.5.b.5 will no longer be applied by CEG. The MTC will continue to be limited by Technical Specification Figure 3.1.3-1 and will be confirmed to be acceptable with all other COLR references listed in Technical Specification 5.6.5.b.

The proposed licensing basis change meets the current regulations. After implementing the proposed change, Braidwood and Byron will continue to meet the ATWS Rule, 10 CFR 50.62, "Requirements for reduction of risk from ATWS events for light-water-cooled nuclear power plants" and continue to meet all other licensing basis commitments related to ATWS including those from NRC Generic Letter 83-28, "Required Actions Based on Generic Implications of Salem ATWS Events".

The proposed change will allow designs with higher h ot full power critical boron concentrations and less negative MTCs at the beginning of the fuel cycle. The effects of those changes are within the limits of existing regulations and licensing bases.

2.1.2 Proposed Changes to Systems, Structures, and Components

This licensing basis change will result in no changes to any systems, structures, or components.

The proposed change affects cycle specific attributes of nuclear fuel loaded into the core. Each fuel cycle optimizes the number of new fuel assemblies, enrichment, burnable poison loading, selection of once-and twice-burned fuel, loading pattern, etc., of nuclear fuel loaded into the core. After the proposed change is implemented, CEG will optimize the design without applying any limitations that would have been required by the analytical methodology and limits described in Technical Specification 5.6.5.b.5. In all aspects other than MTC, the fuel and burnable poisons will be identical to what is described in the UFSAR C hapter 4.0.

The optimized designs after this change is implemented may also have a higher hot full power critical boron concentration over some fraction of the operating cycle. These higher boron concentrations are constrained by other design and licensing basis conditions, and those constraints remain unchanged.

Page 3 of 28 ATTACHMENT 1 Evaluation of Proposed Change

No other system, structure, or component change is proposed. The result is that some reactor cores may be operating with a less negative MTC over a small fraction of the fuel cycle.

2.1.3 Industry Experience

Both the NRC and the industry have prepared several studies over the years which assess the probabilistic and deterministic consequences of an ATWS event. Most of the older studies are not mentioned below as the newer studies supersede them.

2.1.3.1 WCAP-11992 (Reference 6.2): This report is referenced by the current licensing basis and Technical Specification 5.6.5.b.5. CEG follows the methodology described in this document to confirm that the least negative MTC is constrained in order to limit the risk and consequences of an ATWS.

This study from 1988 combines both probabilistic and deterministic methods to show how MTC can be constrained in a way to reduce the Core Damage Frequency (CDF) from an ATWS. The PRA methodology presented in WCAP-11992 is a predecessor of the CEG PRA used for Braidwood and Byron. The CEG PRA for Braidwood and Byron use different definitions for "Unfavorable Exposure Time (UET), " and these are reconciled in Section 2.2.3.

2.1.3.2 WCAP-15831-P-A (Reference 6.4): This study from 2007 combines both probabilistic and deterministic methods to evaluate ATWS risk under a range of MTCs and system configurations.

WCAP-15831-P-A uses r isk informed methods (RG 1.174) to demonstrate that with a positive MTC design, ATWS r isk is still acceptable. This report is a more modern approach than what was presented in WCAP-11992.

Braidwood and Byron have not referenced this report or its conclusions in any licensing bases.

This report does contain several plant specific evaluations and conclusions that applied to Braidwood and Byron at the time the report was written.

Below is a list of conclusions from WCAP-15831-P-A. Descriptions of how those conclusions are reflected in this request are provided below and elsewhere in this request.

a) Core Damage would occur when Reactor Coolant System ( RCS) pressure exceeds 3200 psig.

b) RCS Overpressurization will not likely lead to missiles or other catastrophic failures that would mechanically challenge containment.

c) Steam generator tubes will fail due to overpressure and pre-existing flaws when the tube differential pressure exceeds 3584 psid.

d) Benefit of Rod Control System: WCAP-15831-P-A includes a generic evaluation that the reactivity insertion rate from Rod Control would successfully mitigate an ATWS in conjunction with the least negative MTC at Hot Full Power. Table 4-37 Case 1 demonstrates "No UET" for this condition. It is possible that other design features or off Page 4 of 28 ATTACHMENT 1 Evaluation of Proposed Change

nominal initial conditions could lead to overpressure. The Rod Control system is normally in "Automatic" mode, and that mode provides a faster insertion rate than "Manual. " The WCAP demonstrates that the manual mode is sufficient. Plant and cycle specific analyses are not warranted on the basis of the Risk Analysis results in Section 3.2.1.

e) Configuration Risk Management Program (CMP): WCAP-15831-P-A proposes a configuration risk management program to limit the unavailability of key ATWS mitigating functions. This request does not include a CMP: the risk reduction of the CMP is too low to warrant the actions.

f) Large Early Release: WCAP-15831-P-A very conservatively assumes that all core damage events are Large Early Release events. This request will not apply that overly-conservative end state; instead, Section 3.1.3 of this request will describe how most of the core damage events do NOT result in a large early release.

2.1.3.3 NUREG-0460, "Anticipated Transients Without SCRAM for Light Water Reactors," April, 1978: This early NRC study reports a safety objective of an ATWS core melt frequency of 1.0E-06 or less, and contains early insights on event initiating frequencies and estimated failure rates of SSCs that are required for a successful SCRAM.

2.1.3.4 NUREG-1000, "Generic Implications of ATWS Events at the Salem Nuclear Power Plant": This study provides the basis for the actions required by NRC Generic Letter 83-28. The corrective actions are described in the "NRC Generic Letter 83-28" section below. This NUREG describes the important role of MTC in the PWR response to ATWS.

2.1.3.5 SECY-83-293, "Amendments to 10 CFR 50 Related to Anticipated Transients Without SCRAM (ATWS) Events": This NRC document contains the rationale for the ATWS Rule,10CFR50.62. This rule created the requirement for Braidwood and Byron to add the ATWS Mitigation System Actuation Circuit (AMSAC) to the plants design. This AMSAC circuit detects an ATWS (low steam generator level) and actuates both a turbine trip and AF.

2.1.3.6 NUREG/CR-7110 Vol 2, "State-of-the-Art Reactor Consequence Analyses Project Volume 2: Surry Integrated Analysis": This Sandia National Laborat ories document examines severe reactor accident phenomena and offsite consequences using Surry (Westinghouse PWR) as model. Section 5.3.1.2 discusses the radionuclide releases from a postulated steam generator tube rupture. In this case, the release is caused by a thermally induced steam generator tube failure. This section evaluates the effectiveness of the scrubbing of radionuclides from the release point (ruptured tubes) to offsite. The scrubbing credited is from the steam generator and steamline surfaces.

2.1.3.7 NRC Generic Letter 83-28 (and Supplement 1), "Required Actions Based on Generic Implications of Salem ATWS Events": This Generic Letter describes the corrective actions from NUREG-1000. The corrective actions from this Generic Letter include the actions below:

• Eliminate vulnerabilities associated with inadequate preventive maintenance practices and vulnerable replacement parts for Westinghouse Reactor Trip Breaker Undervoltage trip coils.

Page 5 of 28 ATTACHMENT 1 Evaluation of Proposed Change

• Modifications to Westinghouse Reactor Protection Systems to add the Shunt Trip to the automatic SCRAM function.

2.1.3.8 NUREG-1780 Regulatory Effectiveness of the Anticipated Transient Without Scram Rule

This NRC study was issued in 2001. This study reflects on the success of the ATWS rule and the actions taken by licensees. The NUREG further identifies other new areas of risk including large uncertainties in the reliability of the Reactor Protection System and the adverse impact of MTC due to longer PWR fuel cycles. The Executive Summary states the following:

During the ATWS rulemaking the NRC staff set a goal that probability of an ATWS should be no more than 1.0E-05/reactor-year. Probability of an ATWS was defined as the annual frequency of an ATWS leading to plant conditions that exceed certain design parameters that can result in core melt, containment failure, and the release of radioactivity and can be viewed as the expected CDF of an unmitigated ATWS

Mitigative functions are considered by the ATWS rule regulatory basis to be non-viable if the ATWS peak pressure exceeds 3200 psig; and a sufficiently negative MTC will limit the ATWS peak pressure. Fuel design to achieve longer cycles and higher power ratings may result in less negative MTCs at full power for a larger fraction of the cycle time, during which time ATWS mitigation may be less effective. Further fuel cycle changes and power upgrades that could affect the ATWS risk may require compensatory measures (e.g., hardware or procedural), consistent with the underlying regulatory basis behind the ATWS rule.

The Safety and Probabilistic Analyses presented in this request demonstrate that the removal of the ATWS MTC Limit results in acceptable risk without hardware or procedural compensatory measures.

Section 2.1.3 of NUREG-1780 provides the basis for the 10-minute delay between an ATWS and actions taken outside of the control room to locally trip the reactor at the Motor Generator (MG) set output breakers.

2.2 Background

2.2.1 ATWS Rule and CEG Implementation of AMSAC

Below is a key summary of the design and maintenance changes implemented at Braidwood and Byron in response to NRC regulations:

Reactor Trip Breaker Shunt Trip Modifications: In 1989, Braidwood and Byron completed modifications to the Reactor Trip Breakers in response to NRC Generic Letter 85-09, Technical Specification for Generic Letter 83-28, Item 4.3. The original design included two manual Reactor Trip Breaker controls in the main control room. The original manual trip switch design would trip both the undervoltage and shunt trips on both trains of Reactor Trip Breakers. The original Solid State Protection System (SSPS) d esign would only actuate (de-energize) the undervoltage trip coil on the associated trains trip breaker. The modification added the shunt trip to the automatic trip in SSPS.

Page 6 of 28 ATTACHMENT 1 Evaluation of Proposed Change

ATWS Mitigation System Actuation Circuit (AMSAC): In 1989, Braidwood and Byron completed modifications to incorporate AMSAC. This circuit uses the existing reactor protection sensors for Steam Generator Low Level, and using circuits independent from the Reactor Protection System, implements the logic and time delays for AMSAC. AMSAC trips the M ain Turbine and initiates the Auxiliary Feedwater System whenever three-out-of-four steam generator levels are greater than 3% below the Reactor Protection System (RPS) low-low setpoint, and the Turbine Impulse pressure is greater than C-20 setpoint (30% of nominal full power).

2.2.2 Moderator Temperature Coefficient (MTC)

Together, the fuel and the soluble boron in the reactor coolant provide reactivity feedback which mitigates the consequences of an ATWS event. The reactivity feedback, MTC, usually provides negative reactivity to the core when the moderator temperature increases.

MTC varies over core life. The variability of MTC is dominated by the trend of the h ot full power critical boron concentration over core life. This critical b oron concentration is highest near beginning of core life (at or greater than 1100 ppm in current designs, higher after the proposed change is implemented) and is near 0 ppm (typically less than 25 ppm) at end of life. Operators check and adjust as necessary the critical boron concentration several times per day to maintain the reactivity balance. The reactivity balance is affected by fuel burnup, burnable poison burnup, power changes and adjustments, control rod movement, and fission product buildup and decay.

The core designer can make the MTC more negative (more favorable for ATWS response) by increasing the burnable poisons in the core design (the fixed neutron absorbers reduce the need for soluble boron, and the peak h ot full power critical boron concentration is reduced). This action directly results in lower h ot full power critical boron concentrations at beginning of core life.

The highest h ot full p ower critical boron concentration (and least negative MTC) occurs near the beginning of life. Technical Specification 3.1.3 limits the h ot full power MTC to 0 pcm per °F, but the least negative MTC will be more negative, approximately -2 pcm per °F (1 pcm = 1E-5 k/k).

Technical Specification 3.1.3 limits the h ot zero power (HZP) MTC to +7 pcm per °F. As power increases from zero power to hot full power, operators dilute the RCS Boron concentration to compensate for power defect, Xenon, and other fission product poisons. This dilution results in a typical reduction in MTC of at least 9 pcm per °F between zero power and f ull power. By designing the core to meet the hot zero power MTC limit, there is normally at least 2 pcm per °F margin to the hot full power limit.

Section 3.2.1.2 ("Fuel and Soluble Boron in the Reactor Coolant") discusses the relationship between MTC and the successful mitigation of ATWS.

2.2.3 CEG Implementation of Positive MTC and UET

The initial d esign basis of Braidwood and Byron limited the MTC to no greater than 0 pcm per °F. In 1994, CEG requested to modify the MTC to the current limit as described in Technical Specification 3.1.3. The NRC accepted this change but required an additional limitation on the MTC as described by the limits and analytical methods of Technical Specification 5.6.5.b.5.

(ML020870197).

Page 7 of 28

ATTACHMENT 1 Evaluation of Proposed Change

The original reason for including the requirements of Technical Specification 5.6.5.b.5 was to preserve the r isk basis of the plant design including the plant changes described in section 2.2.2. When CEG requested the positive MTC, the PRA methods and assumptions were overly conservative and did not predict acceptably low risk. The current PRA, described in Section 3.2, is capable of demonstrating an acce ptably small risk without crediting the requirements of Technical Specification 5.6.5.b.5.

In the current licensing basis, the MTC at h ot full power (HFP) must be constrained (MTC less positive) in a way which results in a successful ATWS end state in the limiting licensing basis event. Success is defined by limiting ATWS transient to a peak RCS pressure of less than or equal to 3200 psi. This is the " Service Limit C" as defined by the ASME Section III Code 1971 Edition through Summer 1973 as applied to the Braidwood and Byron Reactor Pressure Vessels. The current licensing basis uses this pressure criteria to define the " core damage" end state of the PRA. This pressure limit applies to a specific limiting Licensing Basis ATWS Event evaluation. The current licensing basis allows this limit to be exceeded, but only for up to 5% of the operating cycle.

The Limiting Licensing Basis ATWS Event is defined as follows:

• The initiating events are (1) a simultaneous loss of load and loss of main feedwater (condition caused by a loss of the main condenser) and (2) a loss of main feedwater from HFP.

• The automatic r eactor trip function fails.

• If the initiating event does not include a turbine trip, the turbine trip occurs when the AMSAC setpoint and time delays have been met.

• The current licensing basis references WCAP-11992 (Reference 6.2), which generically evaluates the peak pressure under a variable combination of the conditions below:

o An MTC which varies over core life o {2, 1, 0} Pressurizer PORVs available o {2, 1} Auxiliary Feedwater Trains available o {Y, N} Automatic Control Rod Insertion available

• CEG and the NRC agreed to only evaluate MTC and UET for a single case: variable MTC, 2 Pressurizer PORVs available, 2 Auxiliary Feedwater Trains available, and NO Control Rod Insertion available (except for the local manual SCRAM which occurs after the peak RCS pressure occurs).

The consequences of the event described above results in a large mismatch between power generated by the reactor and power removed by the steam generators. The immediate result is that the primary coolant returning to the reactor heats up rapidly, resulting in (1) thermal expansion of the coolant and (2) negative reactivity due to the MTC. The magnitude of the MTC directly influences the magnitude of the heatup. For the limiting Licensing Basis ATWS Event,

an MTC more negative than approximately -8 pcm per °F limits the heatup such that the peak water level in the pressurizer is less than 100%, and the combined safety and relief valve capacity is sufficient to prevent excessive RCS overpressure (less than 3200 psi). Operator Page 8 of 28 ATTACHMENT 1 Evaluation of Proposed Change

actions outside of the control room (locally tripping the reactor) are required to terminate the event 10 minutes after the initiating event.

The current l icensing basis defines the " Unfavorable Exposure Time" (UET) to refer to a brief period in the fuel cycle (less than 5% of the fuel cycle) when the MTC is not negati ve enough for the limiting Licensing Basis ATWS Event to meet RCS pressure acceptance criteria. In the context of the limiting Licensing Basis ATWS Event analysis, the terms " favorable" and "unfavorable" MTC refer to MTC values which allow the event to pass or fail the event analysis.

The COLR limit for most positive MTC is set using the following:

• The new fuel is per the design, and the once-and twice-burned fuel is from the previous cycle with a burnup approximately 2000 MWD/MTU lower than the projected low-end burnup (carries over the maximum reactivity, maximizes MTC)

• The core designer verifies that this cores ATWS Critical Power Trajectories remain below bounding ATWS Critical Power Trajectory Limits (a more direct comparison than MTC).

• The maximum Hot Zero Power, All Rods Out N o Xenon MTC from the entire cycle is extracted from the model and is set as the most positive MTC COLR Limit.

The core designer must optimize the number of assemblies, enrichment, and burnable poisons of the core to meet this and all other design and licensing basis criteria. The ATWS constraints in the current licensing basis are the limiting factors on the least negative MTC of the fuel cycle.

3.0 Technical Evaluation

The proposed licensing basis change does not compromise the fundamental safety principles that are the basis of plant design and operation.

3.1 Safety Analysis Results

The UFSAR C hapter 15 results (excluding Chapter 15.8, "ANTICIPATED TRANSIENTS WITHOUT SCRAM (ATWS) ") are unaffected by the proposed change. The C hapter 15 Safety Analysis results already consider the full range of the MTC allowed by Technical Specification 3.1.3. The MTC limits for ATWS developed by the methods in Technical Specification 5.6.5.b.5 are bound by the limits in Technical Specification 3.1.3.

Deterministic safety analyses were not performed for this evaluation. Section 3.1.3 describes the basis for using a generic deterministic analysis. The PRA results presented in S ection 3.2 demonstrate that the MTC limits required by Technical Specification 5.6.5.b.5 are not required in order to maintain acceptable ATWS risk. A detailed description of the effect of the change is included in S ection 3.3.

The performance of systems, structures, and components which perform ATWS mitigating functions are described in S ection 3.3.1.

Page 9 of 28 ATTACHMENT 1 Evaluation of Proposed Change

The proposed change will not affect any d esign or l icensing basis analysis or result other than ATWS. All other d esign and licensing bases evaluations consider the entire range of MTC as defined by Technical Specification 3.1.3 and cycle specific COLR.

The proposed change does not affect the functional capability, reliability, and availability of affected equipment. The proposed change will result in a small, acceptable increase in the risk of exceeding the RCS o verpressure acceptance criteria as demonstrated in S ection 3.2.

3.1.1 Turbine Trip

Tripping the main turbine is an important action in ATWS. The value of the turbine trip is to reduce the rate of depletion of the steam generator secondary inventory, providing the operators more time to complete the remote trip of the reactor. At Braidwood and Byron, the following events will result in a turbine trip:

Automatic Turbine Trip: In a n ATWS, the AMSAC function will generate a turbine trip at a setpoint just 3% below the Reactor Protection S team Generator Low Level setpoint. The actuation has a 9 second time delay. The normal automatic turbine trip will not function during an ATWS. For Braidwood and Byron, the turbine trip that normally occurs with a reactor trip signal is interlocked with the reactor trip breaker position. In an ATWS, with the reactor trip breakers closed, the turbine trip from RPS will not occur following the ATWS, but the AMSAC trip will occur shortly afterwards.

Manual Turbine Trip: The ATWS procedure directs the operator to trip the turbine in the second step. In the event that the turbine trip is not successful, the operator is directed to initiate a Main Steam Isolation. In an ATWS, the main control room operator would likely trip the turbine in this step before the automatic reactor trip would occur.

The current licensing basis for Braidwood and Byron ATWS is based on an ATWS event in Section 2.2.3 above. In that event, the initiating event is a simultaneous loss of load and loss of main feedwater from a hot full power condition caused by a loss of the main condenser.

3.1.2 Generic and Plant Specific UET

No new plant specific analyses were performed to re-evaluate UETs. The plant specific UETs would determine the fraction of the fuel cycle where ATWS mitigation fails to prevent RCS overpressure beyond 3200 psi. This re-analysis would identify the MTC which results in 3200 psi "success" depending upon several combinations of Rod Insertion, Pressurizer PORVs available, and Aux iliary Feedwater trains available (12 cases).

The determination of UET in WCAP-15831-P-A (Reference 6.4) was based on the Braidwood and Byron stations. The UET values may change slightly if re-evaluated today, but the small changes to the UETs would not change the PRA or safety analysis conclusions that the management of MTC for ATWS is not warranted. This amendment request demonstrates an acceptably small increase in risk of CDF and LERF whether or not MTC and UET are restricted.

There is no MTC which corresponds to 100% ATWS success, and with core designs with the most unfavorable MTC, most ATWS events would be successfully terminated by operator action

Page 10 of 28 ATTACHMENT 1 Evaluation of Proposed Change

after 10 minutes. Additional MTC calculations (which then result in new UET values) would not significantly change the conclusion of the PRA.

3.1.3 Large Early Release Frequency (LERF)

This evaluation concludes that most ATWS events which lead to core damage will not lead to a large early release. This is contrary to several of the studies in S ection 2.1.3, Industry Experience. In the earlier studies, the reports conservatively concluded that the core damage end state resulted in a Large Early Release. The following factors influence whether an ATWS event would proceed to a Large Early R elease:

3.1.3.1 Pressure Induced Steam Generator Tube Rupture (Main and Auxiliary Feedwater):

WCAP-15831-P-A (Reference 6.4) establishes a 3584 psi (differential) criteria for the ATWS pressure induced steam generator tube failure. The Braidwood and Byron PRA predicts that 98% of Full Power Internal Events (FPIE) CDF ATWS risk events occur with some feedwater (Main or Auxiliary) supplied to the steam generator. This flow is sufficient to maintain the steam generators pressurized to at least 1175 psig, the setpoint of the first Main Steam Safety Relief valve. If the hypothetical tube differential pressure is greater than 3584 psi, then the RCS pressure would need to reach 4759 psi with the secondary side of the steam generator pressurized. It is far more likely that the RCS will fail (Hot Leg Safe End Weld or Head Closure Bolts, etc.) than a steam generator tube. Further, NUREG/CR-7110 Vol 2 (Reference 6.8) demonstrates that the scrubbing and decontamination provided by the steam generator and main steam line surfaces would mitigate the Large Early Release.

3.1.3.2 Thermally Induced Steam Generator Tube Rupture: In the very unlikely event that the core becomes uncovered and the steam generator tubes fail under extreme temperature, a Large Early Release is unlikely. NUREG/CR-7110 Vol 2 ( Reference 6.8) demonstrates that the scrubbing and decontamination provided by the steam generator and main steam line surfaces would mitigate the L arge Early Release.

3.1.3.3 Other Containment Bypass Events (RCS Catastrophic Failure due to Overpressure):

Other highly unlikely containment bypass scenarios are built into the Braidwood and Byron PRA, and some of these would predict a Large Early Release. Failure of containment isolation is modeled in the Braidwood and Byron PRA.

3.1.3.4 Inter System LOCA: Other highly unlikely inter system LOCA probabilities are built into the Braidwood and Byron PRA, and some of these would predict a Large Early Release.

The Braidwood and Byron PRAs contain the logic to determine the Large Early Release Frequency consistent with the end states listed above.

3.2 Probabilistic Risk Analysis Results

The proposed change results in a small, acceptable increase of risk consistent with the intent of the NRCs policy statement on safety goals for the operations of nuclear power plants.

Page 11 of 28 ATTACHMENT 1 Evaluation of Proposed Change

3.2.1 PRA Results Compared to Region III in RG 1.174

Table 3.2.1.A below displays the Braidwood and Byron PRA results for this change using the current PRA models for Internal Events (including Internal Flood), and Fire. The change in risk (CDF and LERF) reflect the condition when MTC is "unfavorable" 100% of the time.

The external events contribution to the ATWS risk was determined to be insignificant compared to the risk from internal events and fire. Therefore, the analysis results for this current application compare the change in CDF and LERF risk due to internal events and fire.

This (MTC is unfavorable 100% of the time) is conservative because the MTC is unfavorable for much less than 50% of the fuel cycle under most conditions. WCAP-15831-P-A (Reference 6.4) predicts UETs under several plant conditions (combinations of unavailable Aux Feed, PORVs, and Rod Insertion for High Reactivity Cores) with UETs in the range of 42-71%.

Table 3.2.1.A: Braidwood and Byron CDF and LERF with ATWS Unconstrained MTC

Site CDFBase CDF LERFBase LERF BWD-1 5.5E-5 4.6E-7 2.3E-6 6.3E-10

BWD-2 5.3E-5 4.6E-7 2.4E-6 4.5E-10

BYR-1 5.6E-5 4.7E-7 2.1E-6 7.5E-10

BYR-2 5.2E-5 4.7E-7 2.1E-6 7.1E-10 Note: CDF and LERF values are per year

The results above compare very favorably to the criteria in Figure 4 and 5 (Acceptance Guidelines for CDF and LERF) of RG 1.174. These criteria and the comparison to PRA results are summarized below in Table 3.2.1.B

Table 3.2.1.B: RG 1.174 Acceptance Guidelines

CDF CDF LERF LERF Region III 1E-3 1E-6 1E-4 1E-7 Ratio 5.7% 47.0% 2.4% 0.8%

The Ratio in T able 3.2.1.B is the ratio, expressed in percent, of the Region III criteria (Very small changes, the lowest risk criteria) to the plant specific PRA results. Only the highest percentage of the 4 units' results is presented. In all cases, the values are well within the Region III criteria.

3.2.2 Quality of the Current Braidwood and Byron PRA

The Braidwood and Byron Stations, Units 1 and 2 PRA models quality are acceptable for this application since they satisfy the technical acceptability guidance of RG 1.200 Revision 2, as confirmed by the NRC's approval of the Braidwood and Byron Station Risk-Informed Completion Time (RICT) License Amendment Request. The Braidwood and Byron Station PRA models have an acceptable periodic update and review process:

"Based on the NRC staffs review of the licensees submittal and assessments, the NRC staff concludes that the Byron and Braidwood PRA models for internal events, including Page 12 of 28 ATTACHMENT 1 Evaluation of Proposed Change

internal flooding, and for fire events [] satisfy the guidance of RG 1.200, Revision 2.

The NRC staff based this conclusion on the findings that the PRA models conform sufficiently to the applicable industry PRA standards for internal events, including internal flooding, and for fire events at an appropriate capability category, considering the licensees acceptable disposition of the peer review of F&Os [] and NRC staff review. "

And

"The licensee described its PRA model update process that ensures the PRA models

[] are maintained consistent with the as -built and as-operated and maintained plant. "

In addition, while there remain several open PRA model findings from the previously approved TSTF-505, "Provide Risk -Informed Extended Completion Times - RITSTF Initiative 4b" (ML20037B221) and 10 CFR 50.69, "Risk -informed categorization and treatment of structures, systems and components for nuclear power reactors" ( ML18264A092) LARs, all of these findings have been addressed in the current models supporting this LAR.

The ATWS model in the PRA includes the following:

• PRA results for ATWS CDF and LERF are the sum of results from the Full Power Internal Events (FPIE) PRA and the Fire PRA.

• Initiating events include all Initiating events in accordance with the ASME PRA Standard.

• Turbine trips on the AMSAC signal if not already tripped as part of the initial condition.

• Event tree used to model ATWS events is based on WCAP-11992. The event tree is structurally the same as Figure 4-1 in the WCAP. The difference is that the top event for the power level less than 40% has been deleted. WCAP-15831, Section 8, notes that power operation, including shutdown and power level >40%, is the largest contributor to ATWS CDF (at least 88% for the three core types).

• UET is modeled as part of the " Primary Pressure Relief" node. It is a complex node that includes conditions for Rod Insertion, Pressurizer Power Operated Relief Valve availability, and Auxiliary Feedwater Train availability.

The values of CDF and LERF were determined by a bounding sensitivity where the unfavorable exposure is assumed for the entire fuel cycle. That is, the events associated with "no pressure relief is enough" are set to TRUE and the other events where various number of PORVs and/or Safety Relief valves required are set to FALSE.

The Internal Events PRA and the Fire PRA were quantified for the unfavorable case as described above. For Internal Events, a truncation limit of 1.0E-11 per year was used for all CDF cases, and a limit of 1.0E-13 per year was used for all LERF cases. For Fire, a truncation limit of 1.0E-10 per year was used for all CDF cases, and a limit of 1.0E-11 per year was used for all LERF cases. These truncation limits are consistent with the base PRA model quantification, with the exception of Fire LERF which lowered truncation from 1.0E-10 to 1.0E-11 per year to capture lower LERF cutsets.

Page 13 of 28 ATTACHMENT 1 Evaluation of Proposed Change

3.2.3 Differences Between Current Braidwood and Byron PRA vs. WCAP-11992

The results of this evaluation differ from those in WCAP-11992. The current Braidwood and Byron PRAs use more sophisticated modeling and some different inputs. Below is a summary of some of the key input differences between the current Braidwood and Byron PRA vs.

WCAP-11992.

Table 3.2.3: Differences Between Current Braidwood and Byron PRA vs. WCAP-11992

Parameter WCAP-11992 BYR-1 BWD-1 Initiating Events 4.0E+0 3.7E-1 4.7E-1

/Reactor-year RX Trip Failure 6.0E-7 6.0E-7 4.7E-7 Probability PORV Fails to Open 5.0E-3 1.4E-2 8.6E-3 Probability Failure of AFW (power > 6.3E-2 5.7E-2 4.4E-2 40%, 100% flow)

Failure of AFW (power > 1.6E-2 1.6E-2 1.1E-3 40%, 50% flow)

3.2.4 Turbine Trip Modeling in the Current Braidwood and Byron PRA

The Braidwood and Byron PRA logic model for failure of AMSAC is represented by failure of AMSAC itself or failure of the low-level instrument to both normal and AMSAC logic. The unavailability of AMSAC has a failure probability of 1.00E-02, which is based on data from WCAP-11992. The unavailability of the low-level instrument to both normal and AMSAC has a failure probability of 8.16E-04, which is calculated based on generic data from NUREG/CR-6928 (2020), " Industry-Average Performance for Components and Initiating Events at U.S.

Commercial Nuclear Power Plants."

3.2.5 Large Early Release Modeling in the Current Braidwood and Byron PRA

The probability of pressure-induced steam generator tube rupture is calculated based on values in WCAP-16341-P, Table E-7. This table provides rupture probabilities for different numbers of depressurized Steam Generators (SGs) with different tube conditions. For an individual, depressurized steam generator with average tube conditions, the probability of pressure-induced steam generator tube rupture is at PI = 0.0304 (based on the average of the NRR and RES flaw distributions from NUREG-1570, NRR: 5.9E-3, RES: 5.5E-2). This is considered to be representative of Braidwood and Byron over the life cycle of the SGs.

3.3 Defense in Depth

The proposed licensing basis change is consistent with the defense-in-depth philosophy.

• The proposed licensing basis change does not compromise the fundamental safety principles that are the basis of plant design and operation.

Page 14 of 28 ATTACHMENT 1 Evaluation of Proposed Change

• The proposed licensing basis change does not significantly reduce the effectiveness of a layer of defense that exists in the plant design before the implementation of the proposed licensing basis change.

• The proposed change does not affect the plants design basis: ATWS is not a design basis event. Design basis events analyzed in C hapter 15 of the UFSAR are not affected by this change.

3.3.1 Preserve a reasonable balance among the layers of defense.

Below is a detailed description of the levels of defense for a nticipated transients. These layers include the following:

• SSCs that would normally SCRAM the unit in response to a nticipated transients. This includes those that are directly credited by the transient evaluations in Chapter 15 of t he UFSAR and includes other functions that are not credited but will reliably provide a SCRAM function.

• SSCs that provide other ATWS mitigating functions without operator action, and

• SSCs that perform ATWS mitigating functions which require operator action.

3.3.1.1 Systems, Structures, and Components which Provide SCRAM Functions

Several SSCs provide the primary mitigation of Anticipated Transients in the current Licensing Basis.

Reactor Protection System: The Reactor Protection System (RPS) consists of sensors and circuit cards which monitor key plant parameters and provides alarms in the control room and provides trip signals to both trains of the Solid-State Protection System. The RPS has either 3 or 4 sensors and associated alarms for every NSSS parameter that provides input to the reactor trip function (the 4th channel is for protection channels that also provide a control system input).

In most ATWS conditions, more than one parameter will exceed the alarm and trip setpoints.

For example, the " Inadvertent Closure of the Main Steam Isolation Valves " Event (UFSAR Chapter 15.2.4, leads to simultaneous loss of load and loss of feedwater) leads to the low steam generator level trip, and it also leads to high pressurizer pressure and Over Temperature Delta-T trip functions.

Solid State Protection System : The Solid-State Protection System (SSPS) receives trip signals from the RPS and applies the appropriate logic (2-out-of-3 or 2-out-of-4) and coincidence to process reactor trips and other e ngineered safeguards outputs. Table 3.3.1 -1 Functions 15.a and 15.b are inputs to SSPS without going through RPS.

For Reactor Trips, SSPS sends two different trip signals to the Reactor Trip Breakers: one trip is the undervoltage trip, and the other is the shunt trip. Each train of SSPS actuates the same train of Reactor Trip Breaker.

Page 15 of 28 ATTACHMENT 1 Evaluation of Proposed Change

Reactor Trip Breakers: The Reactor Trip Breakers are located electrically between the 240 VAC power supplies for the R od Control System (2 parallel Motor Generator Sets) and the Rod Control System p ower cabinets. When the Reactor Trip Breakers open, the Rod Control System power cabinets lose power, which trips the reactor as described below. The plant design includes two diverse and redundant means to trip the Reactor Trip Breakers electrically. One device is an undervoltage trip mechanism that is normally energized, and trips on a loss of power. This is normally powered by the SSPS undervoltage driver, and SSPS interrupts the undervoltage driver to trip the reactor. The other diverse trip is a shunt trip. This is a normally de-energized coil that trips the Reactor Trip Breaker when energized. The shunt trip coil is also actuated by SSPS when SSPS processes a trip signal. The shunt trip function was part of the original Reactor Trip Breaker design but was originally not part of the SSPS design. The shunt trip was part of the manual trip function in the main control room and added to SSPS as part of the response to the Salem ATWS event prior to the plants initial startup.

Control Rod Drive Mechanisms: The Control Rod Drive Mechanisms (CRDMs) are electro-mechanical jacking devices that provide fine control of the Control Rod position in the core. The CRDMs are located on the Reactor Vessel Head, and one CRDM controls one Rod Control Cluster Assembly. There are 53 CRDMs and associated Rod Control Cluster Assemblies in the core. The electrical power for the CRDMs is from the Reactor Trip Breakers and through the Rod Control System p ower cabinets. The Rod Control System p ower cabinets do not have an active role in the SCRAM function. When the Reactor Trip Breakers open (trip), the Rod Control System power cabinets lose power, resulting in a loss of power to the CRDMs. On a loss of power, the CRDMs release the Rod Control Cluster Assembly drive shaft, and the Rod Control Cluster Assembly falls into the core.

Rod Control Cluster Assemblies: The Rod Control Cluster Assemblies (RCCAs) are described in UFSAR C hapter 4.2.1.6. There are 53 RCCAs in the reactor. The RCCAs contain a neutron absorbing material (Ag-In-Cd) that can be used for fine reactivity and axial power distribution control. When the reactor is at power, nearly all RCCAs are fully withdrawn. Technical Specifications 3.1.5 and 3.1.6 limits the amount of rod insertion allowed when the reactor is at power. The RCCAs provide the reactivity of the SCRAM function. The SCRAM function is considered successful when 52 of the 53 RCCAs fully insert. The RCCAs are qualified to insert under all a nticipated transient conditions including any seismic or thermohydraulic core conditions.

3.3.1.2 Systems, Structures, and Components which Mitigate ATWS Consequences

In the ATWS event, some very unlikely combination of conditions leads to the failure of the normal SCRAM f unction described above. The SSCs below are design features that automatically or inherently mitigate the consequences of the ATWS event.

Fuel and Soluble Boron in the Reactor Coolant: Together, the fuel and the soluble boron in the reactor coolant provide reactivity feedback which mitigates the consequences of an ATWS event. The reactivity feedback, MTC, varies over core life, and is described in detail in Section 2.2.1, above.

Early in the limiting Licensing Basis ATWS E vent, the loss of load results in a mismatch between heat generated by the fuel and heat removed by the steam generators. This mismatch

Page 16 of 28 ATTACHMENT 1 Evaluation of Proposed Change

causes the Reactor Coolant System a verage temperature to increase. The temperature increase provides negative reactivity due to the inherently negative MTC.

Generic studies (Reference 6.4) demonstrate that the MTC would need to be -8 pcm per °F for the current licensing basis case to pass. The MTC will meet this value for at least half of the fuel cycle if only constrained by Technical Specification Figure 3.1.3-1. The current licensing basis also requires MTC to be constrained by the ATWS UET defined within Technical Specification 5.6.5.b.5 (Reference 6.1), which generally constrains to design to meet a Hot Full Power MTC of -8 pcm per °F for at least 95% of the cycle.

The relationship between the " Unfavorable" MTC and failure to meet ATWS overpressure criteria is described in the discussions of the other SSCs below.

All design basis events described in the UFSAR Chapter 15 successfully meet all acceptance criteria with the full range of MTC, limited by Technical Specification F igure 3.1.3-1 and the cycle specific Core Operating Limits Report for the fuel cycle.

Automatic Control Rod Insertion: In the ATWS event, following the turbine trip, the automatic Rod Control System senses a large deviation in RCS Temperature -Average and the program Temperature-Reference. This causes the Automatic Rod Control System to demand control rod insertion at maximum speed (72 steps pe r minute). This is faster than the speed that the operators can insert manually (48 steps per minute). This inserts negative reactivity rapidly. The current licensing basis does not credit this function, but this function is normally available. The proposed change does not affect the design, maintenance, or reliability of this system to rapidly insert control rods in response to an ATWS.

Turbine Trip: The turbine trip function decreases the steam demand by the secondary. This reduces the rate of loss of steam generator secondary inventory and provides the operator with more time to complete actions outside of the control room to locally trip the reactor. The t urbine trip does not occur directly from the conditions that would cause the reactor to trip: the automatic turbine trip occurs only after one or both Reactor Trip Breakers open (the turbine trip is caused by the P-4 Permissive which is tied to the reactor trip breaker aux contacts). In an ATWS, the turbine trip is most likely caused by control room operator action (step 2 of the ATWS procedure). An automatic turbine trip occurs from AMSAC. Also, the turbine may trip as part of the initial event (loss of load event, spurious m ain steam isolation, loss of condenser).

Note, throughout this document, the "ATWS Procedure" generically refers to four procedures,

one for each unit entitled, "RESPONSE TO NUCLEAR POWER GENERATION/ATWS, which are besides numbering identical procedures for Braidwood and Byron Units 1 and 2.

ATWS Mitigation System Actuation Circuit: The current design of the plant incorporates AMSAC. This system is diverse and separate from the Reactor Protection System and Solid-State Protection system and is described in UFSAR C hapter 7.7.1.21. This system senses an ATWS when steam generator water levels fall below the ATWS setpoint, and after a brief time delay, actuates a trip of the Main Turbine and a start of both trains of Auxiliary Feedwater. The proposed change does not affect the design, maintenance, or reliability of this system to trip the turbine and initiate the Auxiliary Feedwater System.

Page 17 of 28 ATTACHMENT 1 Evaluation of Proposed Change

Pressurizer: In the limiting L icensing Basis ATWS E vent, the function of the pressurizer is to (1) maintain the RCS pressure boundary, (2) accommodate RCS coolant expansion due to the heatup, and (3) pass steam (and potentially liquid) to the inlet of Safety and Power Operated Relief Valves. In the limiting Licensing Basis ATWS E vent, the RCS heats up due to the loss of load and loss of feedwater. The thermal expansion of the Reactor Coolant results in the Pressurizer water level to increase. The level change compresses the steam space, resulting in a pressure increase that will exceed the capacity of the pressurizer spray control function. The increase in pressure is mitigated by Power Operated and Safety Relief Valves described below.

The steam space combined with the relief capacity of the relief valves provides overpressure protection to the RCS. For most conditions, this steam space is large enough to accommodate the rapid heatup and expansion of the RCS in the ATWS event. In the event of an " unfavorable MTC", the RCS heatup and thermal expansion may result in a condition that passes water (instead of steam) to the Pressurizer Safety and Power Operated Relief Valves. The proposed change does not affect the design, maintenance, or reliability of the Pressurizer to perform its design functions in response to an ATWS, but the passage of water to the Pressurizer Safety and Relief Valves will affect the performance of these valves and affect the overall RCS overpressure systemic response as described.

Pressurizer Safety Relief Valves: In the limiting L icensing Basis ATWS E vent, the function of the pressurizer Safety Relief Valves is (1) maintain the RCS pressure Boundary and (2) pass steam (and potentially liquid) to the Pressurizer Relief Tank when the safety valve setpoint is reached.

The Pressurizer Safety Relief Valves are 3 high-capacity steam relief valves. These safety relief valves have sufficient capacity to relieve steam and keep the RCS pressure below the safety limit for all d esign basis events. In the limiting Licensing Basis ATWS E vent, when the MTC is favorable (more negative than -8 pcm per °F), the RCS heatup will peak, and the resulting water level in the pressurizer will be less than 100%. The inlet conditions of the Pressurizer Safety and Power Operated Relief valves will be steam (except for the initial volume of water in the loop seal of the safety relief valves), and the relief valves will perform as designed. The proposed change does not affect the design, maintenance, or reliability of the Pressurizer to perform its design functions in response to an ATWS, but the passage of water to the Pressurizer Safety and Relief Valves will affect the performance of these valves and affect the overall RCS overpressure systemic response as described below. The passage of water from these valves to the Pressurizer relief tank is within the capability of the valves (at a significantly reduced relief capacity) and attached piping.

Pressurizer Power Operated Relief Valves: In the limiting Licensing Basis ATWS Event, the function of the P ressurizer PORVs is (1) maintain the RCS pressure b oundary and (2) pass steam (and potentially liquid) to the Pressurizer Relief Tank when the PORV setpoint is reached. The Pressurizer PORVs are 2 low-capacity steam relief valves, and they (1) supplement the available relief capacity of the safety relief valves and (2) reduce the challenges to the safety relief valves by mitigating smaller pressure transients.

The current l icensing basis credits the availability and the capacity of both PORVs. The Probabilistic Risk Analysis considers that the PORVs may be unavailable at the beginning of the ATWS transient: Technical Specification 3.4.11, "Pressurizer Power Operated Relief Valves (PORVs)," allows continued operation with one of the two PORVs inoperable, provided that the PORV is still capable of being manually cycled, and that this PORV will be isolated by its associated block valve, and that this block valve is energized and capable of being opened.

Page 18 of 28 ATTACHMENT 1 Evaluation of Proposed Change

The proposed change does not affect the design, maintenance, or reliability of the Pressurizer PORVs to perform its design functions in response to an ATWS, but the passage of water to the Pressurizer Relief Valves will affect the performance of these valves and affect the overall RCS overpressure systemic response as described below. The passage of water from these valves to the Pressurizer relief tank is within the capability of the valves (at a significantly reduced relief capacity) and attached piping.

Auxiliary Feedwater System: In the limiting Licensing Basis ATWS Event, the function of the AF System is to automatically actuate (from AMSAC), and supply feedwater to the steam generators. The Reactor Protection System and the Solid State Protection System can start the AF System earlier than would AMSAC, and this will likely be successful if the ATWS is caused by failures of the Reactor Trip Breakers. The Reactor Trip Breaker failures would not inhibit the RPS and SSPS to start the system.

The AF System function reduces the rate at which the steam generator secondary inventory is depleted by the ATWS event. The water provides heat removal from the RCS and water to maintain Steam Generator secondary water inventory. The current l icensing basis credits the availability and the capacity of both trains of AF. The PRA considers that one train may be unavailable at the beginning of the ATWS transient. Technical Specification 3.7.5, "Auxiliary Feedwater (AF) System," allows one train to be removed from service for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The proposed change does not affect the design, maintenance, or reliability of this system to actuate and deliver water in response to an ATWS.

Steam Generators, Main Steam Lines, and Main Steam Safety Relief Valves: In the limiting Licensing Basis ATWS Event, the function of these SSCs is (1) maintain the RCS pressure Boundary (Steam Generator Tubes) and (2) remove heat from the RCS (converting secondary water inventory to steam), (3) transport steam from the steam generator to the Main Steam Safety Relief Valves, (4) and in the event of a steam generator tube failure, mitigate the transport of any RCS coolant activity to the environment by scrubbing in the secondary inventory and plate out on the structural surfaces of the steam generator and main steamline.

The proposed change does not change the design, maintenance, or reliability of these SSCs to perform their d esign and licensing basis functions, in response to an ATWS.

Chemical and Volume Control System: The Chemical and Volume Control System (CVCS) provides several support functions for the reactor, including the ability to change the Boric Acid concentration in the RCS. Some anticipated transients credit the function for the CVCS to automatically realign the CVCS pump suction from the CVCS Volume Control Tank to the Refueling Water Storage Tank (RWST). The RWST is considered a safety related source for emergency boration of the RCS. This automatic emergency boration is not part of the limiting ATWS event but does occur in other events. The manual action for the operator to initiate emergency boration is discussed in S ection 3.3.1.3, below.

Page 19 of 28 ATTACHMENT 1 Evaluation of Proposed Change

3.3.1.3 Operator Actions which Provide Mitigation of ATWS Consequences

Operator actions are required to mitigate the consequences of an ATWS. The procedure to respond to an ATWS is based on the Westinghouse Owners Group Emergency Response Guidelines (Reference 6.6), called " RESPONSE TO NUCLEAR POWER GENERATION/ATWS," also knows as " Functional Restoration Procedure S.1." In this document, it will be called the ATWS Procedure.

In the hierarchy of e mergency procedures, this procedure is the highest priority e mergency procedure that could apply to an a nticipated transient. Most other emergency procedures would be suspended until the operators confirm the reactor is subcritical.

The operators are well trained and practiced in the control room actions for the ATWS procedure, and non-licensed operators are trained in the actions to locally trip the reactor and MG sets. Licensed Operators are required to commit the first steps of this procedure to memory.

Below are the manual operator actions in the ATWS Procedure that mitigate an ATWS (these operator actions are in addition to those which verify automatic actions have occurred):

Manual SCRAM: The very first step of the ATWS Procedure directs the operator to manually trip the reactor at both locations on the main control board. Both switches actuate both the undervoltage and shunt trip devices of the Reactor Trip Breakers. Both switches actuate trips for both Reactor Trip Breakers. Both of these functions are designed to be successful regardless of any malfunction of SSPS.

Manual Rod Insertion: The very first step of the ATWS Proc edure directs the operator in manually insert control rods. This step states " IF Reactor will NOT trip, THEN allow control rods to insert automatically until rod speed is less than 48 STEPS/MINUTE, THEN manually insert control rods." Normally, the automatic rod insertion is faster (72 steps per minute following turbine trip).

Manual Turbine Trip: Step 2 of the ATWS Procedure directs the operator to trip the M ain Turbine. If the turbine will not trip, the ATWS P rocedure directs the operator to drive the valves closed with the turbine controls, and if that does not work, actuate Main Steam Isolation.

Emergency Boration: Step 4 of the ATWS Procedure directs the operators to initiate an emergency boration of the reactor. For additional defense in depth, there are three possible success paths. The procedure first directs the operator to use the " Emergency Boration" flowpath from the Boric Acid Storage Tank (7000 ppm Boric Acid) via a motor operated valve,

_CV8104. This boration path provides the greatest negative reactivity addition rate. If that path is not available, the ATWS Procedure directs the operator to re-align the suction of the charging pump from the CVCS Volume Control Tank to the Refueling Water Storage Tank (23 00 ppm Boric Acid). If that path is not available, the ATWS Procedure directs the operator to borate using the normal boration flowpath at the maximum available rate.

Pressurizer Power Operated Relief Valves: Step 4 of the ATWS Procedure directs the operators to verify the P ressurizer PORVs are open if the RCS pressure is greater than 2335 psig. Also, if the associated PORV block valves are closed, this step directs the operator to un-block those valves. These actions not only provide additional relief to an RCS overpressure condition, it also Page 20 of 28 ATTACHMENT 1 Evaluation of Proposed Change

provides a lower back-pressure to the charging pumps, which improves emergency boration flow delivery. Note that the PORVs should have opened automatically under this condition, and the operator action is defense in depth.

Auxiliary Feedwater: Step 3 of the ATWS Procedure directs the operator to check if the AF pumps are running, and if not, provides direction to manually start the pumps. If that does not work, it directs the operator to dispatch an operator to locally start the diesel-driven AF pump.

Confirmation of Steam Generator Water Levels: Step 8 of the ATWS Procedure directs the operator to confirm that the water levels are sufficient (on scale in the narrow range), and if not, confirm that the total flow delivered to the steam generators is at least 900 gpm. If that condition is not met, the ATWS Procedure directs the operators to align pumps and valves as necessary to establish that flow. If the AF system is not able to deliver 900 gpm, then the procedure will be revised to direct the operator to check for the availability of the Electric Main Feedwater Pump and start that pump. That pump has more than the required capacity for this step. (This procedure change will be completed before the amendment is implemented).

Local Trip of the Reactor Trip Breakers, MG Set Generator Breakers, and MG set motor breakers: Step 6 of the ATWS Procedure directs the operator to locally trip the reactor. The control room team would direct a field operator to perform this task. There are several possible success paths:

• The field operator is first directed to locally trip the Reactor Trip and Bypass Breakers.

Either train would successfully trip the reactor. The bypass breakers are normally racked out, and only racked in and closed during the staggered testing for the reactor trip breakers.

• The second success path is for the field operator to locally trip the motor and generator breakers of the motor-generator sets (MG). A successful trip of the generator outputs would result in an instant loss of power to the CRDMs and SCRAM the reactor. A successful trip of the motors would result in a delayed SCRAM since the MG sets have a flywheel, and they continue to supply AC power at a reduced frequency. These controls are in the same room and area as the local actions for the reactor trip breakers.

• The third success path is for the field operator to locally trip the breakers at the substations which supply power to the MG set motors. These substations are in the turbine building.

Ultimately, local trip of the Reactor Trip Breakers terminates the ATWS and allows the operators to transition back into a normal post reactor trip procedure. This action is credited 10 minutes after the ATWS initiating event (Section 2.1.3 of NUREG-1780).

3.3.2 Preserve adequate capability of design features without an overreliance on programmatic activities as compensatory measures.

The proposed change preserves adequate capability of design features without an overreliance on programmatic activities as compensatory measures.

Page 21 of 28 ATTACHMENT 1 Evaluation of Proposed Change

The proposed change does not change any design feature, operating regime, or maintenance activity that would adversely affect any design function. All operating procedures and maintenance practices remain unchanged. After the proposed change, the h ot full power MTC may be more positive for some part of the operating cycle but still within the limits for all other design and licensing basis events. No other programmatic activities are affected.

3.3.3 Preserve system redundancy, independence, and diversity commensurate with the expected frequency and consequences of challenges to the system, including consideration of uncertainty.

The proposed change preserves system redundancy, independence, and diversity commensurate with the expected frequency and consequences of challenges to the system, including consideration of uncertainty.

The full range of MTC is as constrained by Technical Specification 3.1.3 and included as necessary in all C hapter 15 accident and transient analyses. The proposed change does not affect the redundancy, independence, and diversity of the SSCs credited in the analyses, and does not affect the uncertainties included in the analyses.

The proposed change does not affect any of the SSCs associated with AMSAC, other than how the fuel and soluble boron may result in an unfavorable MTC in the l imiting Licensing Basis ATWS Event analysis for a larger fraction of the fuel cycle.

3.3.4 Preserve adequate defense against potential Common Cause Failures.

The proposed change preserves adequate defense against potential common cause failures.

The proposed change will affect the MTC, and under some conditions early in core life, and ATWS event could lead to an event which exceeds the ATWS RCS p ressure limit of 3200 psi.

The PRA analysis in S ection 3.2 demonstrates a small, acceptable increase in probability of this failure mode.

The proposed change does not modify or reduce the functionality of any other component or system.

3.3.5 Maintain multiple fission product barriers.

The proposed change maintains multiple fission product barriers. The proposed change does not modify or significantly reduce the effectiveness of any of the fission product barriers.

Changes to the fuel and soluble boron concentrations do not reduce the effectiveness of any fission product barriers.

3.3.6 Preserve sufficient defense against human errors.

The proposed change preserves sufficient defense against human errors. The proposed change does not modify any SSCs in any way that results in changes to operations or maintenance of SSCs. The interaction between operators and the plant during an ATWS event is not changed or adversely effected.

Page 22 of 28 ATTACHMENT 1 Evaluation of Proposed Change

3.3.7 Continue to meet the intent of the plants design criteria.

The proposed change continues to meet the intent of the plants design criteria. The proposed change affects the MTC for a portion of the operating cycle as it relates to the l imiting Licensing Basis ATWS Event, but the MTC will remain within the limits of Technical Specification 3.1.3 and will continue to be within the limits used in all affected Design Basis Events. The changes to MTC do not adversely affect any of the General Design Criteria i n 10 CFR 50, Appendix A, General Design Criterion 27, "Combined reactivity control systems capability" and Criterion 29, "Protection against anticipated operational occurrences."

3.3.8 Risk Reduction Value of Additional Defense in Depth

CEG has evaluated the risk worth for a potential plant modification for additional ATWS risk mitigation. The evaluation confirms that additional defense in depth is not valuable enough to warrant a significant plant modification.

The hypothetical modification would add a control room function (operator action at step 1 of the ATWS Procedure) to insert all control rods, terminating the ATWS. The hypothetical modification would permit the Rod Control Logic Cabinets to send commands to the power cabinets to de-energize both the stationary and moveable gripper coils, resulting in all rods dropping into the core.

Table 3.3.8, below, summarizes the potential risk elimination value.These sensitivities were run on the Unit 1 plants only (Unit 2 results would provide the same conclusion). The risk values represent the total CDF and LERF of all events in the Full Power Internal Events (FPIE) PRA Model and Fire Model. The CDF and LERF values represent the condition with the ATWS UET Limits removed (unfavorable 100% of the time with the modification installed).

Table 3.3.8: Risk Reduction Value of Hypothetical ATWS Mitigation Modification

Site BWD-1 BYR-1 CDF 5.5E-5 5.7E-5 CDF -2.5E-7 -2.6E-7 LERF 2.3E-6 2.1E-6 LERF -6.8E-10 -4.8E-10 Note: CDF and LERF values are per year.

The "" is the maximum change in risk of this hypothetical modification (negative number). The actual decrease in risk would be lower (less negative) because the results in Table 3.3.8 assumes the modification is always successful. The modification would reduce ATWS related risk, but the total risk reduction is not significant.

There is no current conceptual design for the hypothetical modification. One possible alternative would include installing a diverse SCRAM system. The cost estimate in the NRCs SECY 293 is $1 Million per unit in 1983 dollars. The total risk reduction value would not likely result in any modification, especially one with a price tag greater than $1 Million per unit.

Page 23 of 28 ATTACHMENT 1 Evaluation of Proposed Change

3.4 Safety Margin

The proposed licensing basis change maintains sufficient safety margins. The proposed change does not adversely affect any of the Safety Limits for design basis events. The MTC, after the proposed change, will continue to meet Technical Specification 3.1.3, and the design basis events in C hapter 15 are evaluated using the full range of MTC.

3.5 Define Implementation and Monitoring Program

CEG is not proposing a monitoring plan. No actions are proposed.

The primary goal of a monitoring plan is to ensure that no unexpected adverse safety degradation occurs because of the change(s) to the licensing basis. The principal concern is the possibility that the aggregate impact of changes that could lead to an unacceptable increase in the number of failures from unanticipated degradation, including possible increases in common-cause mechanisms.

The proposed change will not result in any changes to the configuration or reliability of any system, structure, or component. There are no equipment related parameters to trend.

3.6 Configuration Risk Management Program

CEG is not proposing to use configuration risk management to reduce ATWS risk. No new configuration risk management actions are proposed.

In WCAP-15831-P-A (Reference 6.4), the Pressurized Water Reactor Owners Group proposed to limit discretionary preventive maintenance of AF at times in the cycle when it would result in unfavorable ATWS results. For Byron Unit 1 and Braidwood Unit 1, the CDF contribution for this condition (100% unfavorable UET) is 3.5E-6 and 2.3E-6 respectively. For the nominal conditions (base case), the CDF contribution is 3.5E-6 and 2.3E-6 respectively. This is a negligible change.

Therefore no new configuration risk management actions are needed to reduce ATWS risk.

Page 24 of 28 ATTACHMENT 1 Evaluation of Proposed Change

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria

The regulatory requirements associated with this amendment application include the following:

10 CFR 50.36, "Technical specifications," details the content and information that must be included in a station's Technical Specifications (TS). In accordance with 10 CFR 50.36, TSs 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 changes to the Braidwood and Byron TS would ensure that the administrative controls relied upon in the COLR are properly described in the TSs. The proposed TS 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.

The following regulatory requirements pertinent to this amendment application are unaffected by the proposed changes:

• 10 CFR 50.62, "Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants"

• 10 CFR 50 Appendix A, GDC Criterion 1, "Quality Standards and Records"

• 10 CFR 50 Appendix A, GDC Criterion 27, " Combined Reactivity Control System Capability"

• 10 CFR 50 Appendix A, GDC Criterion 28, "Reactivity Limits"

• 10 CFR 50 Appendix A, GDC Criterion 29, "Protection Against Anticipated Operational Occurrences"

• 10 CFR 50 Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants"

• NRC Generic Letter 83-28 dated July 8, 1983 (and Supplement 1 dated October 7, 1992), "Required Actions Based on Generic Implications of Salem ATWS Events"

Page 25 of 28 ATTACHMENT 1 Evaluation of Proposed Change

4.2 No Significant Hazards Consideration Determination

In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Constellation Energy Generation, LLC (CEG) is submitting a request for amendments to 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).

This license amendment request (LAR) proposes to revise the Technical Specifications (TS) to remove COLR analytical method 5.6.5.b.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-11998/11993 and ComEd application of the UET methodology addressed in "Additional Information Regarding Application for Amendment to Facility Operating Licenses-Reactivity Control Systems."

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;

2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or

3) Involve a significant reduction in a margin of safety.

In support of this determination, an evaluation of each of the three criteria set forth in 10 CFR 50.92 is provided below:

1. Does the Proposed Change Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated?

Response: No

The proposed license amendment would revise Braidwood and Byron Technical Specifications to remove COLR analytical method 5.6.5.b.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." The removal of this analytical method will not change the Moderator Temperature Coefficient (MTC) as currently stated in TS 3.1.3 and Figure 3.1.3-1. Braidwood and Byron will continue to meet the ATWS Rule, 10 CFR 50.62.

Therefore, it is concluded that this 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 Kind of Accident from any Accident Previously Evaluated?

Response: No

Page 26 of 28 ATTACHMENT 1 Evaluation of Proposed Change

The proposed changes involve no change to any plant equipment, including how equipment is operated and maintained. Thus, no new accidents are required to be postulated.

Since the proposed changes do not involve changes to plant equipment there is no mechanism for creating a new or different kind of accident not previously evaluated. Therefore, it is concluded that the proposed change does not create the possibility of a new or different kind of 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 changes do not alter equipment design or the way in which the equipment is operated or maintained. The operating parameters are altered by the proposed change. However, the change to the operating parameters do not involve a significant reduction in a m argin of sa fety.

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

In consideration of all the above, CEG concludes that the proposed changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and on that basis, 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

The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change 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 change.

Page 27 of 28 ATTACHMENT 1 Evaluation of Proposed Change

6.0 REFERENCES

6.1 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" (This is the Core Operating Limit Analytical Method in Technical Specification 5.6.5.b.5 proposed for deletion.) (ADAMS Accession No. ML20078S658)

6.2 WCAP-11992, "Joint Westinghouse Owners Group / Westinghouse Program: ATWS Rule Administration Process, " December 1988

6.3 Regulatory Guide 1.174, Revision 3, "An Approach For Using Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-Specific Changes To The Licensing Basis"

6.4 WCAP-15831-P-A, "WOG Risk-Informed ATWS Assessment and Licensing Implementation Process, " August, 2007 (ADAMs Accession No. ML072550560)

6.5 NUREG-0460, "Anticipated Transients Without SCRAM for Light Water Reactors, "

April, 1978

6.6 NUREG-1000, "Generic Implications of ATWS Events at the Salem Nuclear Power Plant"

6.7 SECY-83-293, "Amendments to 10CFR50 Related to Anticipated Transients Without SCRAM (ATWS) Events"

6.8 NUREG/CR-7110 Vol 2, "State-of-the-Art Reactor Consequence Analyses Project Volume 2: Surry Integrated Analysis"

6.9 NRC Generic Letter 83-28 dated July 8, 1983 (and Supplement 1 dated October 7, 1992), "Required Actions Based on Generic Implications of Salem ATWS Events"

6.10 NUREG-1780, " Regulatory Effectiveness of the Anticipated Transient Without Scram Rule"

Page 28 of 28 ATTACHMENT 1

BRAIDWOOD STATION UNITS 1 and 2

Renewed Facility Operating License Nos. NPF-72 and NPF-77

Docket Nos. STN 456 and STN-50-457

Mark-up of Technical Specifications Pages (one page)

Reporting Requirements 5.6

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 ( NF "; 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. WCAP-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."

BRAIDWOOD UNITS 1 & 2 5.6 3 Amendment 190 ATTACHMENT 2

BYRON STATION UNITS 1 and 2

Renewed Facility Operating License Nos. NPF-37 and NPF-66

Docket Nos. STN 454 and STN-50-455

Mark-up of Technical Specifications Page (one page)

Reporting Requirements 5.6

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 ";

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. WCAP-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."

BYRON UNITS 1 & 2 5.6 3 Amendment 196