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Emergency License Amendment Request - One-Time Change to Calvert Cliffs Unit 2 Control Element Assembly Group Definitions
	ML25290A335
Person / Time
Site:	Calvert Cliffs
Issue date:	10/17/2025
From:	Para W; Constellation Energy Generation, Constellation Energy Group
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
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ML25290A334	List: ML25290A335;
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200 Energy Way Kennett Square, PA 19348 www.constellation.com Proprietary Information - Withhold from Public Disclosure Under 10 CFR 2.390 contains Proprietary Information. Withhold from public disclosure under 10 CFR 2.390. When separated from Attachment 6, this document is decontrolled.

10 CFR 50.90 10 CFR 50.91(a)(5)

October 17, 2025 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Calvert Cliffs Nuclear Power Plant, Unit 2 Renewed Facility Operating License No. DPR-69 NRC Docket No. 50-318

SUBJECT:

Emergency License Amendment Request - One-Time Change to Calvert Cliffs Unit 2 Control Element Assembly Group Definitions Pursuant to 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit," Constellation Energy Generation, LLC (CEG), requests Nuclear Regulatory Commission (NRC) approval for a new additional condition of the Renewed Facility Operating License (FOL) No. DPR-69 for Calvert Cliffs Nuclear Power Plant (CCNPP), Unit 2. This approval is being requested on an emergency basis pursuant to 10 CFR 50.91(a)(5).

The proposed change is a swap of the assigned group for 2 Control Element Assemblies (CEAs) during the current operating cycle, Calvert Cliffs Unit 2 Cycle 26 (CC2C26). This amendment is necessary due to a degraded Control Element Drive Mechanism (CEDM) upper gripper coil (UGC) that is inhibiting operator demanded motion for CEA 36. The degraded CEDM does not impact the ability for CEA 36 to fully insert when tripped.

The proposed deterministic change request will remain in effect until Unit 2 next enters MODES 3, 4, or 5. This will occur no later than the end of CC2C26, currently planned for February 2027.

provides an evaluation of the proposed change. Attachments 2 and 3 provide marked-up and clean pages, respectively, of the FOL Appendix C, Additional Conditions, indicating the proposed change.

contains information proprietary to Framatome. The affidavit, provided in, sets forth the basis on which Framatomes information may be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390, "Public inspections, exemptions, requests for withholding." Accordingly, it is respectfully requested that the information that is proprietary to Framatome be withheld from public disclosure in accordance with 10 CFR 2.390.

Attachments 5 and 6 provide non-proprietary and proprietary copies, respectively, of the Framatome safety analysis report supporting the proposed change.

Emergency License Amendment Request One-Time Change to Calvert Cliffs Unit 2 Control Element Assembly Group Definitions Docket No. 50-318 October 17, 2025 Page 2 contains Proprietary Information. Withhold from public disclosure under 10 CFR 2.390. When separated from Attachment 6, this document is decontrolled.

CEG requests approval of the proposed change as soon as possible based on emergency circumstances at CCNPP in accordance with the provisions of 10 CFR 50.91(a)(5) but no later than November 3, 2025. This change is necessary to support a planned power reduction to repair a damaged Communication Control Measuring (CCM) card for the Unit 2 Main Generator Automatic Voltage Regulator (AVR). The proposed change, if approved, will be implemented within 14 days of issuance.

The proposed change has been approved by the CCNPP Plant Operations Review Committee in accordance with the requirements of the CEG Quality Assurance Program.

There are no regulatory commitments contained in this submittal.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," paragraph (b), CEG is notifying the State of Maryland of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

Should you have any questions concerning this submittal, please contact Adam Donell at (267) 533-5156.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 17th day of October 2025.

Respectfully, Wendi Para Sr. Manager - Licensing Constellation Energy Generation, LLC Attachments:

1. Evaluation of Proposed Change 2.

Proposed Facility Operating License, Appendix C, Additional Conditions Page (Markup) 3.

Proposed Facility Operating License, Appendix C, Additional Conditions Page (Clean) 4.

Affidavit from Framatome Supporting the Withholding of Information in from Public Disclosure 5.

Framatome Report ANP-4173NP Rev. 0, Calvert Cliffs Unit 2 Cycle 26 CEA Rod Swap Redesign Safety Analysis Report (Non-Proprietary) 6.

Framatome Report ANP-4173P Rev. 0, Calvert Cliffs Unit 2 Cycle 26 CEA Rod Swap Redesign Safety Analysis Report (Proprietary) cc:

USNRC Region I, Regional Administrator USNRC Senior Resident Inspector, Calvert Cliffs Nuclear Power Plant USNRC Project Manager, NRR - Calvert Cliffs Nuclear Power Plant Z. Barthel, State of Maryland

ATTACHMENT 1 Emergency License Amendment Request Calvert Cliffs Nuclear Power Plant, Unit 2 Renewed Facility Operating License No. DPR-69 NRC Docket No. 50-318 EVALUATION OF PROPOSED CHANGE

Subject:

Emergency License Amendment Request - One-Time Change to Calvert Cliffs Unit 2 Control Element Assembly Group Definitions 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 Control Element Assembly Description 2.2 Control Element Drive System Description 2.3 Current Facility Operating License and Technical Specification Requirements 2.4 Emergency Condition 2.5 Proposed Change 2.6 Additional Work Required to Change Control Element Assembly Group Definitions

3.0 TECHNICAL EVALUATION

3.1 Safety Analysis 3.2 Cycle-specific Restrictions

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration 4.4 Conclusions

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

Emergency License Amendment Request One-Time Change to Calvert Cliffs Unit 2 Control Element Assembly Group Definitions Page 1 of 18 1.0

SUMMARY

DESCRIPTION Pursuant to 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit," Constellation Energy Generation, LLC (CEG), requests Nuclear Regulatory Commission (NRC) approval for a new additional condition of the Renewed Facility Operating License (FOL) No. DPR-69 for Calvert Cliffs Nuclear Power Plant (CCNPP), Unit 2. This approval is being requested on an emergency basis pursuant to 10 CFR 50.91(a)(5).

The proposed change is a swap of the assigned group for 2 Control Element Assemblies (CEAs) during the current operating cycle, Calvert Cliffs Unit 2 Cycle 26 (CC2C26). This amendment is necessary due to a degraded Control Element Drive Mechanism (CEDM) upper gripper coil (UGC) that is inhibiting operator demanded motion for CEA 36. The degraded CEDM does not impact the ability for CEA 36 to fully insert when tripped.

2.0 DETAILED DESCRIPTION 2.1 Control Element Assembly Description 2.1.1 General System Description The reactor core control element configuration consists of 37 single CEAs and 20 dual CEAs.

Each dual CEA consists of two single CEAs connected to a single extension shaft and carried by a single CEDM. Considering the 20 dual CEAs as 40 single CEAs gives an overall equivalent of 77 single CEAs in the core.

During normal operation all of the CEAs are in the fully withdrawn position (all rods out).

Operator demanded CEA movement is accomplished via a digital control system that sequences CEDM coils to change or hold CEA position. Mechanical reactivity control is achieved by vertically maneuvering the positions of the CEA groups by the magnetic jack CEDMs. CEAs are moved in groups and with a defined overlap in order to satisfy the requirements of shutdown margin and provide a more uniform impact on reactivity. For conditions that require a rapid shutdown of the reactor, the Reactor Protection System (RPS) opens reactor trip breakers to deenergize CEDM holding coils, allowing the CEAs to drop into the core.

2.1.2 CEA Group Description The CEAs are designated as either regulating CEAs or shutdown CEAs. The regulating CEAs are divided into five groups (1 through 5). The shutdown CEAs are divided into three groups (A, B, and C). The locations of all the CEAs in one of four symmetrical core quadrants, and their assigned original group, are shown in the figure below.

Emergency License Amendment Request One-Time Change to Calvert Cliffs Unit 2 Control Element Assembly Group Definitions Page 2 of 18 The Power Dependent Insertion Limits (PDILs) specify the maximum permitted CEA insertion.

The PDIL curve for the current cycle of Unit 2 is in the Core Operating Limits Report (COLR)

(Reference 6.2). The PDIL curve illustrates the regulating CEA groups insertion order (5-4-3 1) with overlap between successive groups. The typical CEA withdrawal procedure is as follows:

a. With the reactor subcritical, shutdown group A is fully withdrawn, followed by shutdown group B and then shutdown group C.

b. Withdrawal of regulating CEAs commences starting with regulating group No. 1. CEA withdrawal continues with the prescribed overlap of regulating groups 2 through 5.

All CEAs are inserted for cold shutdown conditions and are typically fully withdrawn during full power steady state operation. All CEAs within a particular group are designated to be withdrawn or inserted nearly simultaneously. The allowable CEA misalignment within any CEA group is specified in Technical Specifications and intentional misalignment of CEAs within a group for the purpose of power shaping is not allowed.

Emergency License Amendment Request One-Time Change to Calvert Cliffs Unit 2 Control Element Assembly Group Definitions Page 3 of 18 2.2 Control Element Drive System Description 2.2.1 General System Description Control element assembly movement is driven by the CEDMs. The Digital Control Element Drive Control System (DCEDCS) transmits signals from the Control Element Drive System (CEDS) control panel to the Mag-Jack Pulse Generators & Monitor (M-PGM) Modules, which develop the current pulses for magnetic jack operation.

The upper and lower CEA group stops for the regulating and shutdown groups are provided by the CEA supervisory function to prevent the reactor from reaching undesirable conditions. The DCEDCS contains design features that ensure the following actions:

a. Insertion of the regulating CEAs within a predetermined permitted range before the shutdown groups are inserted;

b. Simultaneous withdrawal of no more than two groups of CEAs;

c. Proper sequential withdrawal of CEAs; and,

d. Withdrawal of the shutdown CEAs to a predetermined permitted range before the regulating groups are withdrawn.

The shutdown CEAs may be moved either individually or in defined groups. Movement is controlled manually or via a Semi-Automatic (SA) option, where a destination CEA position can be selected, and hands-off movement of the CEAs can be initiated. Movement of more than one shutdown group at any time is not possible. The shutdown CEAs must be withdrawn to within a predetermined permissible range of the upper limit before regulating group withdrawal is possible. A limit prevents group insertion of shutdown CEAs unless all regulating CEAs are within a predetermined permissible range of the lower limit.

The regulating CEAs may be moved in manual group, individual or sequential modes.

Movement is controlled manually or via a SA option, where a destination CEA position can be selected, and hands-off movement of the CEAs can be initiated. Sequential group movement functions such that when the moving group reaches a programmed low (high) position, the next group begins inserting (withdrawing); the initial group stops upon reaching its lower (upper) limit.

This procedure, applied successively to all regulating groups, allows a smooth and continuous rate-of-change of reactivity.

2.2.2 CEDM Description The CEDMs are of the magnetic jack-type drive. Each CEDM is capable of withdrawing, inserting, holding, or tripping the CEA from any point within its operating range. The CEDM drives the CEA and indicates the position of the CEA with respect to the core. For conditions that require a rapid shutdown of the reactor, the CEDM coils are deenergized by the RPS, allowing the CEA to drop into the core by gravity. The reactivity is reduced during such a drop at a rate sufficient to control the core under any operating transient or accident condition.

The magnetic jack motor assembly fits into the CEDM pressure tube housing within the reactor coolant system (RCS) pressure boundary. This integral unit carries the motor tube, lift and hold pawls, and magnets. DC electrical coils positioned externally around the CEDM housing supply the drive power. The CEA lifting operation consists of a series of magnetically-operated step

Emergency License Amendment Request One-Time Change to Calvert Cliffs Unit 2 Control Element Assembly Group Definitions Page 4 of 18 movements. Two sets of mechanical latches engage a notched drive shaft. The magnetic force is obtained from large DC magnet coils mounted on the outside of the motor tube. Power for the electromagnets is obtained from an M-PGM. The M-PGM actuates the stepping cycle and positions the CEA by energizing the various coils in a forward or reverse stepping sequence or holds the CEA stationary by energizing one coil at a reduced current while all other coils are deenergized. The CEAs are tripped upon interruption of electrical power to all coils.

2.3 Current Facility Operating License and Technical Specification Requirements The CCNPP Unit 2 FOL Appendix C, Additional Conditions, includes a requirement for NRC review and approval of any changes to COLR Figures 3.1.6, 3.2.3, and 3.2.5. COLR Figure 3.1.6 specifies CEA group insertion limits vs. the fraction of rated thermal power (RTP) and was developed based on CEA group assignments shown in Section 2.1.2 above. This license condition was implemented per Unit 2 Amendment No. 273 License Condition 4 (Reference 6.1).

Additionally, the Bases for TS 3.1.4, Control Element Assembly (CEA) Alignment, state that the CEAs are arranged into groups that are radially symmetric and that movement of the CEA groups do not introduce radial asymmetries in the core power distribution. CEA inoperability or misalignment may cause increased power peaking, due to the asymmetric reactivity distribution and a reduction in the total available CEA worth for reactor shutdown. Therefore, CEA alignment and operability are related to core operation in design power peaking limits and the core design requirement of a minimum shutdown margin (SDM).

2.4 Emergency Condition Following startup of CC2C26, in April 2025, the upper gripper coil for CEA 36 (core location L-19, regulating group 5) was determined to be failed, inhibiting demanded CEA 36 motion. CEA 36 is fully withdrawn and within 7.5 inches of other regulating group 5 CEAs. CEA 36 remains OPERABLE although it is not capable of moving by operator demand to control power. The issue does not impact the ability for CEA 36 to fully insert when tripped.

Because CEA 36 is in regulating group 5 (first group inserted to lower power), group 5 insertion is not possible due to TS 3.1.4 alignment requirements. CC2C26 is currently being operated with all CEAs fully withdrawn and only boration/dilution available to make reactivity changes.

Using only boration/dilution is sufficient for reactor control early in the fuel cycle and for reactivity control to maintain full power operation. However, as the Hot Full Power (HFP)

Moderator Temperature Coefficient (MTC) becomes more negative through the fuel cycle, CEA motion availability is needed for optimal plant control, Axial Shape Index (ASI) control, and for the TS required 2/3 cycle MTC measurement due in June 2026.

Additionally, it was recently identified that Communication Control Measuring (CCM) card #2 for the Unit 2 Main Generator Automatic Voltage Regulator (AVR) needs to be replaced, as the #1 card is a single point vulnerability in the current condition. A power reduction to approximately 14% RTP will be required to replace the damaged CCM card. Modeling has shown that with boron as the only means for reactivity control, there is reduced margin to ASI limits. Margin is further reduced as the cycle progresses and inadequate margin is available after approximately 14,000 megawatt-days per metric ton of uranium (MWd/MTU) burnup. The current cycle exposure is approximately 6,000 MWd/MTU. At the current cycle exposure, the maneuver is

Emergency License Amendment Request One-Time Change to Calvert Cliffs Unit 2 Control Element Assembly Group Definitions Page 5 of 18 feasible, with reduced margin, from a power distribution standpoint. However, without control rods available to control reactivity, power cannot be ramped quickly enough to reliably traverse an "AVOID OPERATION" region in less than ten minutes at applicable condenser back pressure levels to avoid a turbine and reactor trip. This maneuver is planned to commence on November 21, 2025, to minimize the time the single point vulnerability exists.

The proposed one-time license amendment is required to prevent the unnecessary shutdown of CCNPP, Unit 2 and is emergent due to the unforeseen challenges with both the degraded CEDM UGC for CEA 36 and the damaged AVR CCM card. Further, NRC approval is needed sooner than can be provided under exigent circumstances, and this license amendment request is timely considering the unplanned nature of the issues and time required to develop the technical justification supporting the proposed change. Therefore, CEG has determined that emergency circumstances exist and that CCNPP did not knowingly cause the emergent situation. Further, CEG affirms a best effort has been made for a timely license amendment application. Accordingly, CEG requests an expedited review of the proposed emergency license amendment in accordance with the provisions of 10 CFR 50.91(a)(5) to avoid the unnecessary reduction of margin and potential shutdown of CCNPP, Unit 2 during the planned AVR CCM card repair downpower.

2.5 Proposed Change For Unit 2 Cycle 26 only, CEG is proposing that CEA 36 (core location L-19) and CEA 60 (core location J-20) group assignments be swapped such that CEA 36 is part of regulating group 1 and CEA 60 is part of regulating group 5. This will allow operator demanded movement of regulating groups to control core reactivity.

The CCNPP Unit 2 FOL Appendix C, Additional Conditions, includes a requirement for NRC review and approval of any changes to COLR Figures 3.1.6, 3.2.3, and 3.2.5 (Reference 6.1, License Condition 4). Although COLR Figures 3.1.6, 3.2.3, and 3.2.5 are not expected to be updated as part of the proposed change, CEA group assignments impact the analyses that resulted in the need for this additional condition, specifically for COLR Figure 3.1.6. Therefore, Constellation is seeking approval for the proposed CEA group change consistent with this license condition.

Based on the analyses documented in Section 3.1, a new CCNPP Unit 2 FOL Appendix C, additional condition is proposed to be updated as shown in Attachments 2 (markup) and 3 (clean) as follows:

For Unit 2 Cycle 26 only:

CEA 36 (core location L-19) and CEA 60 (core location J-20) group assignments may be swapped such that CEA 36 is part of regulating group 1 and CEA 60 is part of regulating group

5. If CEA 36 and CEA 60 group assignments are swapped and core thermal power is lowered to less than or equal to 1% of rated thermal power, Unit 2 shall continue to MODE 3. If CEA 36 and CEA 60 group assignments are swapped and Unit 2 enters MODE 3, 4, or 5, the original CEA group assignments shall be restored prior to entering MODE 2. NRC prior review and approval is not required to restore the original CEA group assignments.

Aside from the FOL Appendix C additional condition described, the Facility Operating License

Emergency License Amendment Request One-Time Change to Calvert Cliffs Unit 2 Control Element Assembly Group Definitions Page 6 of 18 and Technical Specifications are unaffected by this change.

Other CCNPP Unit 2 FOL Appendix C, Additional Conditions were reviewed. No other additional conditions are impacted by the proposed change.

Surveillance Requirements (SRs) associated with this proposed change were reviewed. Any impacts due to the proposed change will be managed by the Surveillance Frequency Control Program.

The proposed change will result in radially asymmetric CEA groups for regulating groups 1 and

5. The impact of these asymmetric CEA groups on the core power distribution, power peaking, and SDM are evaluated in Section 3.1 2.6 Additional Work Required to Change Control Element Assembly Group Definitions In order to implement the proposed change in CEA group definitions for CEAs 36 and 60, the following work items will be evaluated and performed in accordance with appropriate CEG procedures:

Update DCEDCS subcomponents (human machine interface, position display system, and motion logic controller) to reflect the reassignment of CEA groups.

Modify M-PGM configurations to swap the timing of the swapped CEAs and to categorize the swapped CEAs in the reassigned group.

Modify the plant process computer (PPC) configuration, including displays and reports, to reflect the reassignment of CEA groups.

Modify configuration files and inputs to the core monitoring system (POWERTRAX),

including the neutronics code (PRISM) and incore monitoring code (INPAX).

Operations training and procedure revisions will be completed in accordance with modification process requirements.

The modifications described for the DCEDCS, M-PGM, PPC, POWERTRAX, PRISM, and INPAX will not affect any other safety-related systems and will not affect the ability of the reactor protection system to safely shut down the reactor.

3.0 TECHNICAL EVALUATION

The Technical Evaluation examined the impacts to the Plant Safety Analysis for the proposed change to the assigned group for 2 CEAs during the current operating cycle, CC2C26, due to a degraded CEDM for CEA 36. The degraded CEDM does not impact the ability for CEA 36 to fully insert when tripped.

3.1 Safety Analysis As described in Section 2.5, the proposed change would swap the assigned group for CEA 36 and adjacent CEA 60 such that CEA 36 is part of regulating group 1 and CEA 60 is part of regulating group 5. This change would remain in place through the end of CC2C26 operation or until repairs of CEA 36 are completed, whichever occurs first. This change impacts the nuclear design and safety analysis characteristics for this reload core design. The reload safety

Emergency License Amendment Request One-Time Change to Calvert Cliffs Unit 2 Control Element Assembly Group Definitions Page 7 of 18 evaluation process, which is used for each new fuel cycle, has been re-performed as appropriate to determine the nuclear design parameter changes, impact to core and fuel performance, and impact to the accident analyses described in Updated Final Safety Analysis Report (UFSAR)

Chapter 14 for the proposed CEA group reassignments. The nuclear design parameter changes associated with core operation under the proposed CEA group assignments were evaluated against a set of bounding values contained in the pertinent accident and transient analyses for the plant. The results of those evaluations are discussed in this section.

NRC-approved reload safety analysis codes and methods were used to determine if the changes in nuclear design parameters remain bounded by the key safety parameters assumed in the UFSAR Chapter 14 safety analysis. Additionally, impacts on margins to fuel thermal and power peaking limits related to departure from nucleate boiling (DNB) and centerline fuel melt (CFM) safety criteria due to the change in power distribution attributable to operation with the control rod insertion under these CEA group assignments were evaluated.

An evaluation of impacts to core and fuel performance, as well as the impact to the safety analyses described in UFSAR Chapter 14 and safety analysis parameters, is summarized in the cycle-specific reload safety evaluation documentation (Attachments 5 and 6) to confirm the acceptability of reactor operation with the new core configuration. There were no changes in analytical methods, computer codes, or safety analysis limits used to perform the core reload safety evaluation for CC2C26 with these new CEA group assignments. The CC2C26 core design, with updated CEA group assignments, was performed with full core models due to the asymmetrical impact the CEA group reassignment could impart. Results of the safety analysis impact evaluation are described in the following subsections.

The TS LCOs and associated SRs of CEAs are not impacted by the proposed change. As described in Section 2, all CEAs remain available to perform their safety related trip functions as described in TS.

3.1.1 Shutdown Margin The proposed change has minimal impacts to the available shutdown margin (SDM). TS LCO 3.1.1 states that the required SDM shall be above the COLR limit of 3,500 percent milli-k (pcm).

Maintaining the SDM above this limit ensures the safety analysis described in UFSAR Chapter 14 remains bounding. An evaluation of the SDM with the proposed CEA group reassignments determined the impact to be an increase of SDM by <40 pcm. Overall, the CC2C26 SDM with the proposed CEA group reassignments remains above the required SDM of 3,500 pcm as specified in the COLR.

3.1.2 Other impacted parameters A review of other major neutronic parameters such as Delayed Neutron Fraction, Boron Concentration, MTC, Doppler Temperature Coefficient (DTC), and maximum fuel assembly and fuel rod burnups occurred with the different CEA group assignments. All parameters changed by less than 1%, primarily attributed to code rounding and a more accurate modeling of the previous outage duration.

Emergency License Amendment Request One-Time Change to Calvert Cliffs Unit 2 Control Element Assembly Group Definitions Page 8 of 18 3.1.3 UFSAR Chapter 14 Accident Analyses Impacts The proposed CEA group swap for the remainder of CC2C26 has an impact on several UFSAR Chapter 14 accident analysis parameters routinely evaluated as part of the reload design process. Cycle-specific evaluations (Attachments 5 and 6) were performed to determine if the change in core design adversely impacts bounding key safety parameters assumed in the UFSAR Chapter 14 safety analyses and assess impacts on DNB and fuel thermal limits due to the change in power distribution. The bounding key safety parameters are developed in UFSAR Chapter 14 accident analyses of record (AORs) to ensure expected reactivity parameters and peaking conditions for various accident conditions are bounded. Therefore, if the cycle-specific evaluation meets the bounding parameters, the AOR remains satisfied. Results of the cycle-specific evaluations confirm that the limits assumed in the safety analysis remain bounding with incorporation of the additional cycle-specific restrictions specified in Section 3.2. Therefore, the proposed CEA group swap for CC2C26 does not impact the results presented in the UFSAR Chapter 14 accident analyses. Results and discussion of the UFSAR Chapter 14 accident analyses for CC2C26 with the CEA group reassignments are provided below.

3.1.3.1 Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power Startup Condition (UFSAR Section 14.2)

A CEA Withdrawal Accident at Low Power or While Subcritical is defined as an uncontrolled addition of reactivity to the reactor core caused by withdrawal of CEAs while in low power or subcritical conditions, resulting in a power excursion. The maximum reactivity insertion rate in the detailed plant analysis is calculated to compare to limits assumed in UFSAR Chapter 14 accident analyses. The proposed group reassignments of CEA 36 and CEA 60 will impact the order and combination of CEA withdrawal, which in turn will impact the localized reactor core power distribution for events where a power excursion occurs. The limiting transient power excursion for this event occurs at a power significantly less than 1% RTP. These new CEA group assignments challenged the cladding strain acceptance criteria and were unable to produce acceptable results at the most limiting conditions. Increasing the lower bound of allowed operation to greater than 1% RTP allowed the clad strain results to meet their acceptance criteria.

The minimum DNB Ratio (DNBR) performance and Fuel Centerline Melt performance for this event remains bounded by the Uncontrolled Control Rod Bank Withdrawal at Power event discussed in Section 3.1.3.2.

Due to the inability to gain successful clad strain results for the entirety of this event, operating restrictions must be established to prevent the Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power Startup Condition from occurring during those non-passing conditions. Specifically, any operation at or below 1% RTP or entry into Modes 3, 4, or 5 requires a repair of CEA 36 and re-establishment of the original CEA group assignments. These restrictions are described in Section 3.2 and enforced via the proposed change described in Section 2.5.

3.1.3.2 Uncontrolled Control Rod Bank Withdrawal at Power (UFSAR Section 14.2)

A CEA Withdrawal Accident at Power is defined as an uncontrolled addition of reactivity to the reactor core caused by withdrawal of CEAs while at part power or full power conditions, resulting

Emergency License Amendment Request One-Time Change to Calvert Cliffs Unit 2 Control Element Assembly Group Definitions Page 9 of 18 in a power excursion. The existing licensing basis is based predominantly on limiting ranges of DTC, MTC, and reactivity insertion rates. For this proposed CEA group re-assignment, those ranges of DTC, MTC and reactivity insertion rates remain unchanged. Therefore, the transient response of the licensing basis remains bounding, and no re-analysis of this event is required.

The minimum DNBR and peak Linear Heat Generation Rate (LHGR) continue to meet their acceptance criteria.

3.1.3.3 Control Rod Misoperation (UFSAR Section 14.11)

A Control Rod Misoperation is defined as a failure in the control rod system which causes a CEA to drop into the core. The existing licensing basis is based predominantly on a limiting range of dropped CEA worth and the most negative MTC for the cycle. For this CEA group re-assignment, the limiting MTC and range of dropped CEA worth remain unchanged. Therefore, the transient response of the licensing basis remains bounding, and no re-analysis of this event is required. The minimum DNBR and peak LHGR continue to meet their acceptance criteria.

3.1.3.4 Spectrum of Control Rod Ejection Accidents (UFSAR Section 14.13)

A Control Rod Ejection Accident is defined as a postulated rupture of a control rod drive mechanism allowing for full system pressure to act on the drive shaft, which ejects its control rod from the core. The existing licensing basis is based predominantly on a maximum rod motion based reactivity insertion rate and limiting ranges of the DTC and MTC. For the proposed CEA group re-assignment, the limiting DTC, MTC, and ejected CEA worth remain unchanged.

Therefore, the transient response of the licensing basis remains bounding, and no re-analysis of this event is required. The minimum DNBR and peak LHGR continue to meet their acceptance criteria.

3.1.3.5 Remaining UFSAR Chapter 14 Accident Analyses A review of all other UFSAR Chapter 14 accident analyses, including the Loss of Coolant Accident, was performed. These events were determined to remain bounded or unaffected with the new CEA group reassignments.

3.1.3.6 UFSAR Chapter 14 Accident Analyses Evaluation Summary The impact of the CEA group reassignments in CC2C26 on the nuclear design and safety analysis, including the UFSAR Chapter 14 events accident analyses, has been evaluated using the NRC-approved methods described in TS 5.6.5. These NRC-approved reload safety evaluation methods were used to determine if the proposed change in core configuration adversely impacts the bounding key safety parameters assumed in the UFSAR Chapter 14 Safety Analysis and to assess impacts on DNB and CFM due to the change in power distribution. Cycle-specific parameter evaluations for UFSAR Chapter 14 Safety Analysis parameters confirm that the values assumed in the safety analysis remain bounding for all UFSAR Chapter 14 Safety Analysis accidents with the exception of the Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power Startup Condition. However, after the incorporation of the operating restrictions specified in Section 3.2 and enforced via the proposed change described in Section 2.5, there is no change to the UFSAR Chapter 14 accident analyses. Therefore, the proposed CEA group reassignments for CC2C26 do not impact the results presented in UFSAR Chapter 14.

Emergency License Amendment Request One-Time Change to Calvert Cliffs Unit 2 Control Element Assembly Group Definitions Page 10 of 18 3.1.4 Trip Setpoint Verification - LCO Monitoring The DNB specified acceptable fuel design limit (SAFDL) is protected by way of the DNB LCO (SR 3.2.5.1) against two categories of anticipated operational occurrences (AOOs): (1) those requiring initial steady-state margin to the SAFDL, and (2) those requiring a combination of initial steady-state margin to the SAFDL and a RPS function. The limits on LHR (SR 3.2.1.2) represent limits within which the LHR algorithms are valid. When using excore detectors for LHR monitoring, LHR LCO COLR Figure 3.2.1-2 applies. With the proposed CEA group reassignments, additional margin in the negative ASI range was needed in the LHR LCO COLR Figure. The following revised COLR Figure 3.2.1-2 was created to support the proposed CEA group reassignments.

With this revised figure, limiting LHR calculation results demonstrate that margin exists and the adequacy of the LHR LCO function is maintained. This revised COLR Figure will be implemented as part of the proposed change.

Emergency License Amendment Request One-Time Change to Calvert Cliffs Unit 2 Control Element Assembly Group Definitions Page 11 of 18 3.2 Cycle-specific Restrictions To ensure safety analyses performed remain bounding, the following plant operating restrictions are required for CC2C26 if the proposed change is implemented:

1. Operation is limited to core thermal power > 1% RTP. If core thermal power drops to 1%

RTP or lower, Unit 2 must continue to MODE 3.

2. If Unit 2 enters MODE 3, 4, or 5, the original CEA group assignments must be restored prior to entering MODE 2.

These restrictions are imposed based on the results of Uncontrolled Control Rod Assembly Withdrawal from Subcritical or Lower Power Startup Condition event analyses as discussed in Section 3.1.3.1. If, following implementation of the proposed change, the CEA group assignments are restored during CC2C26, reassessment of the reload licensing basis will be performed based on the actual operating history of the plant using the same NRC-approved reload safety analysis codes and methods used for this proposed change.

Reactivity guidance for power maneuvers will continue to be provided to the operating crews in accordance with existing procedures with additional predicted power distribution information provided, as required, including Tq (Azimuthal Power Tilt) and FrT (Total Integrated Radial Peaking Factor).

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be met. CCNPP was not licensed to the General Design Criteria (GDCs) listed in 10 CFR Part 50, Appendix A, but was licensed based on the Atomic Energy Commission (AEC) proposed Principal Design Criteria (PDCs) published on July 10, 1967. The relevant principal design criteria for CCNPP are described in Appendix 1C of the CCNPP UFSAR. CEG has determined that the proposed change does not require any exemptions or relief from the following current applicable regulations and regulatory requirements, which were reviewed in making this determination:

10 CFR 50.36, Technical Specifications 10 CFR 50.36(c) provides that TS will include Limiting Conditions for Operation (LCOs) which are "the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met." In addition, 10 CFR 50.36 requires that a licensee's TS be derived from the analyses and evaluation included in the safety analysis report.

The proposed change does not involve any modifications to TS, including LCOs, Surveillance Requirements, required remedial actions, or shutdown requirements.

The proposed change does not affect CCNPP's compliance with the intent of 10 CFR 50.36.

Emergency License Amendment Request One-Time Change to Calvert Cliffs Unit 2 Control Element Assembly Group Definitions Page 12 of 18 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 50.62(c)(2) requires that each pressurized water reactor manufactured by Combustion Engineering or by Babcock and Wilcox must have a diverse scram system from the sensor output to interruption of power to the control rods. This scram system must be designed to perform its function in a reliable manner and be independent from the existing reactor trip system (from sensor output to interruption of power to the control rods).

The proposed change does not impact CCNPPs scram system function or reliability.

The degraded CEDM for CEA 36 does not impact the ability for CEA 36 to fully insert when tripped.

The proposed change does not affect CCNPP's compliance with the intent of 10 CFR 50.62.

UFSAR Appendix 1C Criterion 6, Reactor Core Design (Similar to GDC 10, Reactor Design)

UFSAR Appendix 1C Criterion 6 requires that the reactor core be designed to function throughout its design lifetime, without exceeding acceptable fuel damage limits which have been stipulated and justified. The core design, together with reliable process and decay heat removal systems, shall provide for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and for transient situations which can be anticipated, including the effects of the loss of power to recirculation pumps, tripping out of a turbine generator set, isolation of the reactor from its primary heat sink, and loss of all off-site power.

Analyses documented in Attachments 5 and 6 and summarized in Section 3 demonstrate that the proposed change maintains appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including uncertainties and for transient situations which can be anticipated.

The proposed change does not affect CCNPP's compliance with the intent of UFSAR Appendix 1C Criterion 6.

UFSAR Appendix 1C Criterion 7, Suppression of Power Oscillations (Similar to GDC 12, Suppression of Reactor Power Oscillations)

UFSAR Appendix 1C Criterion 7 requires that the core design, together with reliable controls, ensures that power oscillations which could cause damage in excess of acceptable fuel damage limits are not possible or can be readily suppressed.

The proposed change does not significantly impact the control rod worth of the CEAs.

Therefore, the ability of the CEAs to readily suppress power oscillations is not impacted by the proposed change.

The proposed change does not affect CCNPP's compliance with the intent of UFSAR Appendix 1C Criterion 7.

Emergency License Amendment Request One-Time Change to Calvert Cliffs Unit 2 Control Element Assembly Group Definitions Page 13 of 18 UFSAR Appendix 1C Criterion 8, Overall Power Coefficient (Similar to GDC 11, Reactor Inherent Protection)

UFSAR Appendix 1C Criterion 8 requires that the reactor be designed so that the overall power coefficient in the power operating range shall not be positive.

Analyses documented in Attachments 5 and 6 and summarized in Section 3 demonstrate that the proposed change impacts reactivity feedback coefficients by less than 1% and that reactivity feedback coefficients are not positive in the power operating range.

The proposed change does not affect CCNPP's compliance with the intent of UFSAR Appendix 1C Criterion 8.

UFSAR Appendix 1C Criterion 14, Core Protection Systems, and Criterion 15, Engineered Safety Features Protection Systems (Similar to GDC 20, Protection system functions)

UFSAR Appendix 1C Criterion 14 requires that core protection systems together with associated equipment be designed to act automatically to prevent or to suppress conditions that could result in exceeding acceptable fuel damage limits.

UFSAR Appendix 1C Criterion 15 requires that protection systems be provided for sensing accident situations and initiating the operation of necessary engineered safety features.

The proposed change does not impact core protection systems, engineered safety features protection systems, or their ability to sense accident situations and act automatically to prevent or to suppress conditions that could result in exceeding acceptable fuel damage limits and operate engineered safety features.

The proposed change does not affect CCNPP's compliance with the intent of UFSAR Appendix 1C Criteria 14 and 15.

UFSAR Appendix 1C Criterion 19, Protection System Reliability, and Criterion 20, Protection Systems Redundancy and Independence (Similar to GDC 21, Protection system reliability and testability)

UFSAR Appendix 1C Criterion 19 requires that protection systems be designed for high functional reliability and in-service testability commensurate with the safety functions to be performed.

UFSAR Appendix 1C Criterion 20 requires that redundancy and independence designed into protection systems be sufficient to assure that no single failure or removal from service of any component or channel of a system will result in loss of the protection function. The redundancy provided shall include, as a minimum, two channels of protection for each protection function to be served. Different principles shall be used where necessary to achieve true independence of redundant instrumentation components.

The proposed change is necessary due to a degraded CEDM UGC that is inhibiting operator demanded motion for CEA 36. The degraded CEDM does not impact the ability

Emergency License Amendment Request One-Time Change to Calvert Cliffs Unit 2 Control Element Assembly Group Definitions Page 14 of 18 for CEA 36 to fully insert when tripped. The proposed change does not impact the reliability, redundancy, or independence of protection systems.

The proposed change does not affect CCNPP's compliance with the intent of UFSAR Appendix 1C Criteria 19 and 20.

UFSAR Appendix 1C Criterion 27, Redundancy of Reactivity Control, Criterion 28, Reactivity Hot Shutdown Capability, Criterion 29, Reactivity Shutdown Capability, and Criterion 30, Reactivity Holddown Capability (Similar to GDC 26, Reactivity control system redundancy and capability, and GDC 27, Combined reactivity control systems capability)

UFSAR Appendix 1C Criterion 27 requires that at least two independent reactivity control systems, preferably of different principles, be provided.

UFSAR Appendix 1C Criterion 28 requires that at least two of the reactivity control systems provided shall independently be capable of making and holding the core subcritical from any hot standby or hot operating condition, including those resulting from power changes, sufficiently fast to prevent exceeding acceptable fuel damage limits.

UFSAR Appendix 1C Criterion 29 requires that at least one of the reactivity control systems provided be capable of making the core subcritical under any condition (including anticipated operational transients) sufficiently fast to prevent exceeding acceptable fuel damage limits. Shutdown margins greater than the maximum worth of the most effective control rod when fully withdrawn shall be provided.

UFSAR Appendix 1C Criterion 30 requires that at least one of the reactivity control systems provided be capable of making and holding the core subcritical under any conditions with appropriate margins for contingencies.

The proposed change maintains two independent reactivity control systems of different principles (i.e., CEAs and boration). Both reactivity control systems will continue to independently be capable of making and holding the core subcritical from any hot standby or hot operating condition, including those resulting from power changes, sufficiently fast to prevent exceeding acceptable fuel damage limits.

Analyses documented in Attachments 5 and 6 and summarized in Section 3 demonstrate that, after the incorporation of the operating restrictions specified in Section 3.2 and enforced via the proposed changes described in Section 2.5, with the proposed CEA group reassignments for CC2C26:

the reactivity control systems are capable of making the core subcritical under any condition (including anticipated operational transients) sufficiently fast to prevent exceeding acceptable fuel damage limits, shutdown margin remains greater than the maximum worth of the most effective control rod when fully withdrawn, and the reactivity control systems are capable of making and holding the core subcritical under any conditions with appropriate margins for contingencies.

Emergency License Amendment Request One-Time Change to Calvert Cliffs Unit 2 Control Element Assembly Group Definitions Page 15 of 18 The proposed change does not affect CCNPP's compliance with the intent of UFSAR Appendix 1C Criteria 27, 28, 29, and 30.

UFSAR Appendix 1C Criterion 31, Reactivity Control Systems Malfunction (Similar to GDC 25, Protection system requirements for reactivity control malfunctions)

UFSAR Appendix 1C Criterion 31 requires that the reactivity control systems be capable of sustaining any single malfunction, such as unplanned continuous withdrawal (not ejection) of a control rod, without causing a reactivity transient which could result in exceeding acceptable fuel damage limits.

Analyses documented in Attachments 5 and 6 and summarized in Section 3 demonstrate that, after the incorporation of the operating restrictions specified in Section 3.2 and enforced via the proposed changes described in Section 2.5, there is no change to the UFSAR Chapter 14 accident analyses and the reactivity control systems remain capable of sustaining any single malfunction. The proposed CEA group reassignments for CC2C26 do not impact the results presented in UFSAR Chapter 14.

The proposed change does not affect CCNPP's compliance with the intent of UFSAR Appendix 1C Criterion 31.

UFSAR Appendix 1C Criterion 32, Maximum Reactivity Worth of Control Rods (Similar to GDC 28, Reactivity limits)

UFSAR Appendix 1C requires that limits, which include considerable margin, shall be placed on the maximum reactivity worth of control rods or elements and on rates at which reactivity can be increased to ensure that the potential effects of a sudden or large change of reactivity cannot (a) rupture the reactor coolant pressure boundary or (b) disrupt the core, its support structures or other vessel internals sufficiently to impair the effectiveness of emergency core cooling.

Analyses documented in Attachments 5 and 6 and summarized in Section 3 demonstrate that, after the incorporation of the operating restrictions specified in Section 3.2 and enforced via the proposed changes described in Section 2.5, there is no change to the UFSAR Chapter 14 accident analyses, including those evaluating control rod withdrawal, misoperation, and ejection.

The proposed change does not affect CCNPP's compliance with the intent of UFSAR Appendix 1C Criterion 32.

4.2 Precedent The following precedents are applicable to this submittal. In precedent 4.2.1, the U.S. NRC granted Callaway Plant Unit 1 an amendment allowing operation for one cycle with no control rod assembly installed in one core location. The Callaway change and its impacts are similar to the proposed change for CC2C26, and a similar evaluation of safety analyses and other impacts was required. In precedent 4.2.2, the U.S. NRC granted Calvert Cliffs Unit 1 an amendment allowing operation with the center CEA excluded from TS requirements for operability and alignment for one cycle. The previous Calvert Cliffs change is similar to the proposed change for CC2C26 since both address a CEA unable to be moved by operator demand.

Emergency License Amendment Request One-Time Change to Calvert Cliffs Unit 2 Control Element Assembly Group Definitions Page 16 of 18

1. Letter from Mahesh L. Chawla (U.S. NRC) to Fadi Diya (Ameren Missouri), Callaway Plant, Unit No. 1 - Issuance of Amendment No. 239 RE: Revision to Technical Specifications to Allow Control Rod to be Removed for Cycle 28 (EPID L-2025-LLA-0016), dated April 10, 2025 (ADAMS Accession No. ML25084A354)

2. Letter from Daniel G. McDonald (U.S. NRC) to Mr. G. C. Creel (Baltimore Gas and Electric Company), Issuance of Amendment for Calvert Cliffs Nuclear Power Plant Unit 1 (TAC No. 79709), dated February 20, 1991 (ADAMS Accession No. ML010470133) 4.3 No Significant Hazards Consideration Constellation Energy Generation, LLC (CEG), proposes a swap of the assigned group for 2 Control Element Assemblies (CEAs) during the current operating cycle, Calvert Cliffs Unit 2 Cycle 26 (CC2C26). This amendment is necessary due to a degraded Control Element Drive Mechanism (CEDM) upper gripper coil (UGC) that is inhibiting operator demanded motion for CEA 36. The degraded CEDM does not impact the ability for CEA 36 to fully insert when tripped.

Constellation Energy Generation, LLC (CEG) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

1.

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

Response: No.

The Control Element Assemblies (CEAs) and the analyses associated with them are not accident initiators. An evaluation of shutdown margin (SDM) with the proposed change remains above the required SDM. The response of the CEAs to an accident and their impact on the core is analyzed using conservative techniques and NRC-approved reload safety analysis codes and methods. Analysis results were compared to approved acceptance criteria. These evaluation results show that the response to an accident, after implementation of the proposed change, is within approved acceptance criteria.

The proposed change will not alter or prevent the ability of the CEAs from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. Therefore, the proposed change does not affect accident or transient initiation or consequences.

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

2.

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

Response: No.

The proposed change does not adversely affect any fission product barrier, nor does it alter the safety function of safety systems, structures, or components, or their roles in

Emergency License Amendment Request One-Time Change to Calvert Cliffs Unit 2 Control Element Assembly Group Definitions Page 17 of 18 accident prevention or mitigation. The proposed change involves no physical changes beyond those required to ensure the affected CEAs move with their reassigned groups, their movement is correctly displayed to operators, and analyses performed by the core monitoring system reflect the reassigned groups. The proposed change was determined to have no impact on previously evaluated accident analyses. The proposed change results in operation within fuel design limits.

Therefore, 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 proposed change does not adversely affect any fission product barrier, nor does it alter the safety function of safety systems, structures, or components, or their roles in accident prevention or mitigation. The proposed change has an impact on several accident analysis parameters routinely evaluated as part of the reload design process.

Cycle-specific evaluations were performed to determine if the change in core design adversely impacts bounding key safety parameters assumed in the safety analyses and assess impacts on DNB and fuel thermal limits due to the change in power distribution.

The proposed change, including operational restrictions for restoring the original CEA group assignments, was determined to have no impact on previously evaluated accident analyses. The proposed change does not result in exceeding design basis limits. All licensed safety margins are maintained.

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

Based on the above, CEG concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

4.4 Conclusions In conclusion, based on the considerations discussed 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 Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

CEG has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, "Standards for Protection Against Radiation." However, the proposed amendment does not involve: (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the

Emergency License Amendment Request One-Time Change to Calvert Cliffs Unit 2 Control Element Assembly Group Definitions Page 18 of 18 proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review," paragraph (c)(9). Therefore, pursuant to 10 CFR 51.22, paragraph (b), no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment.

6.0 REFERENCES

6.1 Letter from Douglas V. Pickett (U.S. NRC) to Mr. George H. Gellrich (Calvert Cliffs Nuclear Power Plant, LLC), Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 -

Amendment RE: Transition from Westinghouse Nuclear Fuel to Areva Nuclear Fuel (TAC Nos. ME2831 and ME2832), dated February 18, 2011 (ADAMS Accession No. ML110390224).

6.2 Letter from Christopher J. Smith (Constellation) to U.S. Nuclear Regulatory Commission, Core Operating Limits Report for Unit 2, Cycle 26, Revision 0, dated March 11, 2025 (ADAMS Accession No. ML25072A033 and ML25072A034).

ATTACHMENT 2 Emergency License Amendment Request Calvert Cliffs Nuclear Power Plant, Unit 2 Renewed Facility Operating License No. DPR-69 NRC Docket No. 50-318 Proposed Facility Operating License, Appendix C, Additional Conditions Page (Markup) 11

11 Appendix C (Contd)

Additional Conditions Facility Operating License No. DPR-69 Amendment No. 323317 Amendment No.

Additional Condition Implementation Date 317 Up to two Framatome PROtect' Lead Test Assemblies utilizing Chromium-coated M5 cladding and chromia doped pellets may be placed in limiting regions of the core for up to 3 cycles commencing with the implementation of Amendment 317.

With implementation of this Amendment 317 The safety limits specified in TS 2.1.1.2 regarding fuel centerline melt temperature for Framatome fuel, < 5081°F, decreasing by 58°F per 10,000 MWD/MTU and adjusted for burnable poison per XN-NF-79-56(P)(A),

Revision 1, Supplement 1 is not applicable for the Framatome PROtect' Lead Test Assemblies utilizing Chromium-coated M5 cladding and chromia doped pellets for up to 3 cycles commencing with the implementation of Amendment 317.

With implementation of this Amendment 317 The requirement that the RODEX2 predicted rod internal pressure shall remain below the steady state system pressure is not applicable for the Framatome PROtect' Lead Test Assemblies utilizing Chromium coated M5 cladding and chromia doped pellets for up to 3 cycles commencing with the implementation of Amendment 317.

With implementation of this Amendment Insert:

For Unit 2 Cycle 26 only:

CEA 36 (core location L-19) and CEA 60 (core location J-20) group assignments may be swapped such that CEA 36 is part of regulating group 1 and CEA 60 is part of regulating group 5. If CEA 36 and CEA 60 group assignments are swapped and core thermal power is lowered to less than or equal to 1% of rated thermal power, Unit 2 shall continue to MODE 3. If CEA 36 and CEA 60 group assignments are swapped and Unit 2 enters MODE 3, 4, or 5, the original CEA group assignments shall be restored prior to entering MODE 2. NRC prior review and approval is not required to restore the original CEA group assignments.

Insert:

This amendment is effective immediately and shall be implemented within 14 days of issuance.

ATTACHMENT 3 Emergency License Amendment Request Calvert Cliffs Nuclear Power Plant, Unit 2 Renewed Facility Operating License No. DPR-69 NRC Docket No. 50-318 Proposed Facility Operating License, Appendix C, Additional Conditions Page (Clean) 11

11 Appendix C (Contd)

Additional Conditions Facility Operating License No. DPR-69 Amendment No. 330317 Amendment No.

Additional Condition Implementation Date 317 With implementation of this Amendment 317 With implementation of this Amendment 317 With implementation of this Amendment This amendment is effective immediately and shall be implemented within 14 days of issuance.

Up to two Framatome PROtect' Lead Test Assemblies utilizing Chromium-coated M5 cladding and chromia doped pellets may be placed in limiting regions of the core for up to 3 cycles commencing with the implementation of Amendment 317.

The safety limits specified in TS 2.1.1.2 regarding fuel centerline melt temperature for Framatome fuel, < 5081°F, decreasing by 58°F per 10,000 MWD/MTU and adjusted for burnable poison per XN-NF-79-56(P)(A),

Revision 1, Supplement 1 is not applicable for the Framatome PROtect' Lead Test Assemblies utilizing Chromium-coated M5 cladding and chromia doped pellets for up to 3 cycles commencing with the implementation of Amendment 317.

The requirement that the RODEX2 predicted rod internal pressure shall remain below the steady state system pressure is not applicable for the Framatome PROtect' Lead Test Assemblies utilizing Chromium coated M5 cladding and chromia doped pellets for up to 3 cycles commencing with the implementation of Amendment 317.

For Unit 2 Cycle 26 only:

CEA 36 (core location L-19) and CEA 60 (core location J-20) group assignments may be swapped such that CEA 36 is part of regulating group 1 and CEA 60 is part of regulating group

5. If CEA 36 and CEA 60 group assignments are swapped and core thermal power is lowered to less than or equal to 1% of rated thermal power, Unit 2 shall continue to MODE 3. If CEA 36 and CEA 60 group assignments are swapped and Unit 2 enters MODE 3, 4, or 5, the original CEA group assignments shall be restored prior to entering MODE 2. NRC prior review and approval is not required to restore the original CEA group assignments.

330

ATTACHMENT 4 Emergency License Amendment Request Calvert Cliffs Nuclear Power Plant, Unit 2 Renewed Facility Operating License No. DPR-69 NRC Docket No. 50-318 Affidavit from Framatome Supporting the Withholding of Information in Attachment 6 from Public Disclosure

A F F I D A V I T 1.

My name is Philip A. Opsal. I am Manager, Product Licensing for Framatome Inc. (formally known as AREVA Inc.), and as such I am authorized to execute this Affidavit.

2.

I am familiar with the criteria applied by Framatome to determine whether certain Framatome information is proprietary. I am familiar with the policies established by Framatome to ensure the proper application of these criteria.

3.

I am familiar with the Framatome information contained in Framatome Document No. ANP-4173P Revision 0, Calvert Cliffs Unit 2 Cycle 26 CEA Rod Swap Redesign Safety Analysis Report. Information contained in this Document has been classified by Framatome as proprietary in accordance with the policies established by Framatome for the control and protection of proprietary and confidential information.

4.

This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by Framatome and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5.

This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) Trade secrets and commercial or financial information.

6.

The following criteria are customarily applied by Framatome to determine whether information should be classified as proprietary:

(a)

The information reveals details of Framatomes research and development plans and programs or their results.

For Information Only

(b)

Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c)

The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for Framatome.

(d)

The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for Framatome in product optimization or marketability.

(e)

The information is vital to a competitive advantage held by Framatome, would be helpful to competitors to Framatome, and would likely cause substantial harm to the competitive position of Framatome.

The information in this Document is considered proprietary for the reasons set forth in paragraphs 6(b), 6(c), 6(d), and 6(e) above.

7.

In accordance with Framatomes policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside Framatome only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8.

Framatome policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9.

The foregoing statements are true and correct to the best of my knowledge, information, and belief.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on October 15, 2025.

Philip A. Opsal Manager, Product Licensing Framatome Inc.

Philip A. Opsal For Information Only

ATTACHMENT 5 Emergency License Amendment Request Calvert Cliffs Nuclear Power Plant, Unit 2 Renewed Facility Operating License No. DPR-69 NRC Docket No. 50-318 Framatome Report ANP-4173NP Rev. 0, Calvert Cliffs Unit 2 Cycle 26 CEA Rod Swap Redesign Safety Analysis Report (Non-Proprietary)

For Information Only

v t e Letter Sequence Request
EPID:L-2025-LLA-0158, Emergency License Amendment Request - One-Time Change to Calvert Cliffs Unit 2 Control Element Assembly Group Definitions (Open)
Initiation Request, Request, Request Acceptance... Administration Withholding Request Acceptance Questions Answers Results Approval Other:
MONTHYEAR2025-10-21 [Table View]

ML25290A335

Text

SUBJECT:

Subject:

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

1.0 INTRODUCTION

SUMMARY

SUMMARY

9.0 REFERENCES

1.0 INTRODUCTION

SUMMARY

SUMMARY

9.0 REFERENCES

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