ML25304A004

Issuance of Exigent Amendment for One-Time Change to Control Element Assembly Group Definitions (Exigent Circumstances)

ML25304A004
Person / Time
Site:	Calvert Cliffs
Issue date:	11/03/2025
From:	Geoffrey Miller NRC/NRR/DORL/LPL2-1
To:	Mudrick C Constellation Energy Generation
References
EPID L-2025-LLA-0158
Download: ML25304A004 (1)

Text

November 3, 2025 Mr. Christopher Mudrick Senior Vice President Constellation Energy Generation, LLC President and Chief Nuclear Officer Constellation Nuclear 4300 Winfield Rd Warrenville, IL 60555

SUBJECT:

CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT 2 - ISSUANCE OF EXIGENT AMENDMENT FOR ONE-TIME CHANGE TO CONTROL ELEMENT ASSEMBLY GROUP DEFINITIONS (EXIGENT CIRCUMSTANCES)

(EPID L-2025-LLA-0158)

Dear Mr. Mudrick:

In response to your application dated October 17, 2025 (Agencywide Documents Access and Management System Accession No. ML25290A335), the U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 330 to Renewed Facility Operating License No. DPR-69 for Calvert Cliffs Nuclear Power Plant (CCNPP), Unit 2. The amendment adds a license condition to allow a one-time change to the CCNPP control element assembly (CEA) group definitions.

The amendment is issued under exigent circumstances as described in the provisions of paragraph 50.91(a)(6) ),Notice for Public Comment: State Consultation, of Title 10 of the Code of Federal Regulations due to the time critical nature of the amendment.

In this instance, an exigent situation exists because the amendment is required to support a repair in November that requires changing power.. This is because CEA 36 is currently unable to be moved with its assembly regulating group (though CEA 36 is still OPERABLE if a plant trip is experienced). During upcoming needed plant maneuvers, movement of CEA 36, along with other CEAs in its current assembly group will be required. The licensee stated that 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.

A copy of the related safety evaluation is also enclosed. The safety evaluation describes the exigent circumstances under which the amendment was issued and the final no significant hazards consideration determination. A Notice of Issuance addressing the final no significant hazards consideration determination and opportunity for a hearing associated with the exigent circumstances will be included in the Commissions monthly Federal Register notice.

If you have questions, please contact me at 301-415-2481 or ed.miller@nrc.gov.

Sincerely,

/RA/

G. Edward Miller, Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No.: 50-318

Enclosures:

1. Amendment No. 330 to DPR-69

2. Safety Evaluation cc: Listserv

CALVERT CLIFFS NUCLEAR POWER PLANT, LLC CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. 50-318 CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 330 License No. DPR-69

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Constellation Energy Generation, LLC dated October 17, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to Appendix C, Additional Conditions, of the Renewed Facility Operating License, and paragraph 2.C.(5) of Renewed Facility Operating License No. DPR-69 is hereby amended to read as follows:

(5)

Additional Conditions The Additional Conditions contained in Appendix C as revised through Amendment No. 330 are hereby incorporated into this license. Constellation Energy Generation, LLC shall operate the facility in accordance with the Additional Conditions.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 14 days.

FOR THE NUCLEAR REGULATORY COMMISSION:

David Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Page 4 of the Renewed Facility Operating License and affected page of Appendix C of the Renewed Facility Operating License Date of Issuance: November 3, 2025 DAVID WRONA Digitally signed by DAVID WRONA Date: 2025.11.03 09:48:25 -05'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 330 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-69 DOCKET NO. 50-318 Replace the following page of Renewed Facility Operating License No. DPR-69 with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the area of change.

REMOVE INSERT 4

4 Replace the following page of Appendix C to Renewed Facility Operating License No. DPR-69 with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Appendix C to Renewed Facility Operating License No. DPR-69 REMOVE INSERT 11 11 Amendment No. 330 (4)

Environmental Monitoring Program If harmful effects or evidence of irreversible damage are detected by the biological monitoring program, hydrological monitoring program, and the radiological monitoring program specified in the Appendix B Technical Specifications, Constellation Energy Generation, LLC (the licensee) will provide to the staff a detailed analysis of the problem and a program of remedial action to be taken to eliminate or significantly reduce the detrimental effects or damage.

(5)

Additional Conditions The Additional Conditions contained in Appendix C as revised through Amendment No. 330 are hereby incorporated into this license.

Constellation Energy Generation, LLC shall operate the facility in accordance with the Additional Conditions.

(6)

Secondary Water Chemistry Monitoring Program Constellation Energy Generation, LLC shall implement a secondary water chemistry monitoring program to inhibit steam generator tube degradation. This program shall include:

a.

Identification of a sampling schedule for the critical parameters and control points for these parameters;

b.

Identification of the procedures used to quantify parameters that are critical to control points;

c.

Identification of process sampling points;

d.

Procedure for recording and management of data;

e.

Procedures defining corrective actions for off control point chemistry conditions; and

f.

A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events required to initiate corrective action.

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

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 330 TO THE RENEWED FACILITY OPERATING LICENSE NO. DPR-69 CONSTELLATION ENERGY GENERATION, LLC CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT 2 DOCKET NO. 50-318

1.0 INTRODUCTION

By letter dated October 17, 2025 (Agencywide Documents Access and Management System Accession No. ML25290A335), Constellation Energy Generation, LLC (CEG, the licensee) requested that the U.S. Nuclear Regulatory Commission (NRC, the Commission) amend the Calvert Cliffs Nuclear Power Plant (CCNPP), Unit 2, Renewed Facility Operating License (RFOL) No. DPR-69 on an emergency basis pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.91(a)(5),Notice for Public Comment: State Consultation. The NRC staff determined that using the exigent procedures in 10 CFR 50.91(a)(6)(i)(B) to provide reasonable notice and opportunity to comment to the public in the area surrounding a licensee's facility of the licensee's amendment and of the NRCs proposed determination that no significant hazards consideration is involved in the requested amendment would be sufficient to meet the requested review schedule.

The license amendment request (LAR) proposed to add a license condition to allow a one-time change to the CCNPP control element assembly (CEA) group definitions. The proposed change would swap the assigned group for 2 CEAs during the current operating cycle, Calvert Cliffs, Unit 2, Cycle 26 (CC2C26). The licensee stated that the amendment is necessary due to a degraded Control Element Drive Mechanism (CEDM) upper gripper coil (UGC) which inhibits the operator demanded motion for CEA 36 and that the degraded CEDM does not impact the ability for CEA 36 to fully insert when tripped. The proposed change will remain in effect for CCNPP, Unit 2 until the end of the CC2C26 cycle which is currently planned for February 2027.

1.1 System Description and Operation The LAR states that there are 37 single CEAs and 20 dual CEAs in the reactor core with each dual CEA consisting of 2 single CEAs. The dual CEAs are connected by an extension shaft and carried out by a single CEDM. The operator-demanded CEA movement from a fully withdrawn position during normal operation is performed through a digital control system that sequences

the CEDM coils. The CEAs are located in one of the four symmetrical core quadrants and are designated as either a regulating CEA or a shutdown CEA. The CEAs are generally in fully withdrawn position during full power operation with movement for power regulation and shutdown driven by the CEDMs. The shutdown CEAs can be moved either individually or in defined groups while the regulating CEAs can be moved in manual group, individual or sequential modes.

The CEA that cannot be moved on operator demand due to a degraded CEDM grid coil is CEA 36 and is part of regulating group 5. The LAR states that group 5 is the first group inserted to lower power during operation.

1.2 Proposed Change The licensee proposed to swap the group assignments for CEA 36 (core location L-19) and CEA 60 (core location J-20) such that CEA 60 will be changed to being in regulating group 5 and CEA 36 will be changed to being in regulating group 1.

License condition 4 in the CCNPP, Unit 2, RFOL Appendix C, Additional Conditions (ML110390224), states that Figures 3.1.6, 3.2.3, and 3.2.5 of the Core Operating Limits Report (COLR) shall not be changed without prior NRC review and approval until a basis is developed for analyzing the CEA rod bank movement at full power operations. The licensee stated in the LAR that updating COLR Figures 3.1.6, 3.2.3, and 3.2.5 is not part of the LAR. However, as described in section 3.1.4 Trip Setpoint Verification - LCO [Limiting Condition of Operation]

Monitoring of the LAR, the licensee created a revised COLR Figure 3.2.1-2, Linear Heat Rate to support the proposed CEA group reassignments. The CEA group assignments impact the basis for Figure 3.1.6 which resulted in a need for an additional license condition. Accordingly, the licensee proposes adding a new condition to CCNPP, Unit 2, RFOL Appendix C.

Specifically, the licensee proposes adding the following:

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 The LAR did not propose any changes to Technical Specifications (TSs).

2.0 REGULATORY EVALUATION

Regulations in 10 CFR Part 50.90, Application for amendment of license, construction permit, or early site permit, requires that whenever a holder of a license wishes to amend the license, including TSs in the license, an application for amendment must be filed, fully describing the changes desired. Under 10 CFR 50.92(a), determinations on whether to grant an applied-for

license amendment are to be guided by the considerations that govern the issuance of initial licenses or construction permits to the extent applicable and appropriate.

The regulations in 10 CFR 50.91(a)(5) state, in part, that where the Commission finds that an emergency situation exists, in that failure to act in a timely way would result in derating or shutdown of a nuclear power plant, or in prevention of either resumption of operation or of increase in power output up to the plants licensed power level, it may issue a license amendment involving no significant hazards consideration without prior notice and opportunity for hearing or for public comments.

The regulations in 10 CFR 50.91(a)(6) state, in part, that exigent circumstances exist when: (1) a licensee and the NRC must act quickly; (2) time does not permit the NRC to publish a Federal Register notice allowing 30 days for prior public comment; and (3) the NRC determines that the amendment involves no significant hazards consideration. Under the provisions in 10 CFR 50.91(a)(6), the NRC notifies the public in one of two ways: (1) by issuing a Federal Register notice providing an opportunity for hearing and allowing at least 2 weeks from the date of the notice for prior public comments; or (2) by using local media to provide reasonable notice to the public in the area surrounding the licensees facility.

In 10 CFR 50.36, Technical specifications, the NRC established its regulatory requirements related to TSs. Pursuant to 10 CFR 50.36(c)(1)-(5), TSs are required to include items in the following five specific categories related to station operation: (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 rule does not specify the particular requirements to be included in a plants TSs.

Regulations in 10 CFR 50.62(c)(2) require 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).

CCNPP, Units 1 and 2, were not licensed to the 10 CFR Part 50, Appendix A General Design Criteria (GDC). The Atomic Energy Commission (AEC) published the final rule that added Appendix A to 10 CFR Part 50 in the Federal Register (FR) (36 FR 3255) on February 20, 1971, with the rule effective on May 21, 1971. The Commission did not apply the final GDC to plants with construction permits issued prior to May 21, 1971, which includes Calvert Cliffs, Units 1 and

2.

Appendix 1C, AEC Proposed General Design Criteria for Nuclear Power Plants, of the CCNPP Units, 1 and 2, Updated Final Safety Analysis Report (UFSAR), Revision 53 (ML23250A368)

lists the 1967 AEC draft proposed General Design Criteria. The applicable 1967 AEC draft proposed GDCs are as follows:

AEC GDC 6, Reactor Core Design AEC GDC 7, Suppression of Power Oscillations AEC GDC 8, Overall Power Coefficient AEC GDC 14, Core Protection Systems AEC GDC 15, Engineered Safety Features Protection Systems AEC GDC 19, Protection System Reliability AEC GDC 20, Protection Systems Redundancy and Independence AEC GDC 27, Redundancy of Reactivity Control AEC GDC 28, Reactivity Hot Shutdown Capability AEC GDC 29, Reactivity Shutdown Capability AEC GDC 30, Reactivity Holddown AEC GDC 31, Reactivity Control Systems Malfunction AEC GDC 32, Maximum Reactivity Worth of Control Rods 2.2 Regulatory Guidance The following guidance was used in the review of the proposed amendment.

NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition (SRP) Section 4.6, Functional Design of Control Rod Drive System, Revision 2, March 2007 (ML070540139), informed the regulatory requirements and areas of review for the proposed change NUREG-0800, Chapter 15, Transient and Accident Analysis, Revision 3, March 2007 (ML070710376) informed the regulatory requirements and areas of review for the proposed change.

3.0 TECHNICAL EVALUATION

The NRC staff evaluated the LAR to determine if the proposed changes are consistent with the applicable regulations and guidance discussed in Section 2.0 of this safety evaluation (SE). The NRC staff reviewed the LAR along with the relevant sections of the CCNPP, Units 1 and 2, UFSAR (ML23250A368) and the applicable TS. The NRC staff compared the impacted design parameters in the LAR with the corresponding key safety parameters to verify that they were bounded by the safety analysis performed in Chapter 14 of the UFSAR.

3.1 Impact on Safety Analysis The proposed swap of CEA 36 to regulating group 1 and the CEA 60 to regulating group 5 impacts the nuclear design and safety analysis characteristics for the reload core design. The licensee stated in the LAR that analyses related to mechanical design, neutronics, thermal hydraulics, setpoints verification, and safety analyses were evaluated to support the proposed swap. The LAR states that the safety limits and the NRC-approved safety analysis codes and analytical methods used to perform the evaluation remain unchanged from the existing core reload SE performed for the CC2C26.

Evaluation of the impacts of purposed new CEA group reassignments on departure from nucleate boiling (DNB) and centerline fuel melt (CFM) limits due to the potential change in power distribution was performed. Attachments 5 and 6 to the LAR summarize the evaluations performed to confirm the acceptability of reactor operation with the new CEA group configuration in the core. All analyses performed to support the reload evaluation for the proposed new CEA group assignments were evaluated with a full-core geometry to ensure that any core imbalance introduced by the CEA rod swap is fully incorporated.

3.1.1 Impact on Shutdown Margin Evaluations performed to support the new CEA group assignments show that the proposed change has a minimal impact on the Shutdown Margin and that the overall CC2C26 SDM for the CEA swap remains above the required SDM specified in the COLR.

Based on the review of the information presented in the LAR and associated attachments 5 and 6, the NRC staff finds that the SDM limits are maintained above the TS LCO 3.1.1, which states that the required SDM shall be above the COLR limit of 3,500 percent milli-k (pcm).

3.1.2 Impact on Other Key Parameters Nuclear Design The licensee stated in the LAR that evaluation of impact of the CEA assignment swap on the major neutronics parameters such as Delayed Neutron Fraction, Boron concentration, Moderator Temperature Coefficient (MTC), Doppler Temperature Coefficient (DTC), and maximum fuel assembly and fuel rod burnups occurred shows that these parameters changed by less than 1 percent, which is primarily attributed to code rounding and a more precise modeling performed during the previous outage.

Based on the review of the information presented in the LAR and associated attachments 5 and 6, the NRC staff finds that neutronics characteristics for the CC2C26 SDM for CEA swap remain bounded by the values assumed in the safety analyses.

Mechanical Design The licensee stated in the attachment to the LAR that the evaluation of the mechanical design of all the fuel assemblies to be operated in the core for CC2C26 shows that there is no change due to the CEA rod swap redesign.

The NRC staff finds the licensee evaluation to be acceptable given structural or mechanical changes to the core design or geometry are not impacted due to the swapping of the CEA rod group assignments.

Thermal-Hydraulic Design The licensee stated in the attachment to the LAR that thermal-hydraulic characteristics of the fuel and the reload core for the CC2C26 CEA rod group assignment swap were evaluated utilizing NRC-approved methodologies to support the current TSs and the COLR limits.

Based on the review of the information provided, the NRC staff finds the key thermal-hydraulic parameters of linear heat generation limit (LHGR), fuel centerline melt (FCM), DNB and correlation limit remain unchanged and acceptable.

3.1.3 Impact on Accident Analysis The proposed CEA group swap for the CC2C26 has an impact on several accident analyses parameters which are evaluated in CCNPP, Units 1 and 2, UFSAR, Chapter 14. The cycle-specific evaluations are performed as part of reload design to ensure that the bounding key safety parameters used in UFSAR Chapter 14 accident analyses of record (AORs) remain bounding. The licensee performed cycle-specific calculations (attachments 5 and 6 of the LAR) to evaluate the impact of the change to the CEA rod group swap on the key safety parameters assumed in the safety analyses and further assess its impact on DNB and fuel thermal limits.

Such changes are expected from the change in power distribution at off-rated conditions due to the CEA rod swap reassignment.

A review of UFSAR Chapter 14 accident analyses for CC2C26 with the CEA group reassignments presented in the LAR is summarized in the subsection below.

Uncontrolled CEA Withdrawal from a Subcritical or Low Power Startup Condition This event is described in the CCNPP, Units 1 and 2, UFSAR, Section 14.2 as a power excursion event resulting from an uncontrolled addition of reactivity caused by withdrawal of CEAs while operating in low power or subcritical conditions. The proposed group reassignments of CEA 36 and CEA 60 impact the order and combination of CEA withdrawal and thus can change the localized reactor core power distribution for the event.

The licensee stated in the LAR that minimum DNB Ratio (MDNBR) performance and FCM performance for this event remains bounded by the Uncontrolled Control Rod Bank Withdrawal at Power event which is discussed in Section 3.1.3.2 of the LAR. The licensee further stated that the new CEA group assignments challenged the cladding strain acceptance criteria and were unable to produce acceptable results at the most limiting conditions, which occurs at power significantly less than 1 percent of the Rated Thermal Power (RTP).

Due to inability to meet the acceptance criteria for powers below 1 percent RTP, the licensee proposed an operating restriction which states that any operation at or below 1 percent RTP or entry into Modes 3, 4, or 5 requires re-establishment of the original CEA group assignments.

The LAR states that the clad strain results meet their acceptance criteria at above 1 percent RTP.

Based on the review of the information presented, the NRC staff finds the impact of the CEA rod group swap on the uncontrolled CEA withdrawal accident at low power or subcritical conditions to be acceptable because the acceptance criteria for cladding strain for the event are met at power levels above 1 percent of the RTP and the licensee imposed an operating restriction which states that any operation at or below 1 percent RTP or entry into Modes 3, 4, or 5 would require re-establishment of the original CEA group assignments.

Uncontrolled Rod Bank Withdrawal at Power This event is described in the CCNPP, Units1 and 2, UFSAR, Section 14.2 as power excursion caused due to uncontrolled reactivity addition to the core caused by withdrawal of CEAs while at

power conditions. Attachments 5 and 6 to the LAR state that the key parameters considered in the analysis of this event are the reactivity insertion rate due to rod motion, the Fuel Temperature Coefficient (FTC) and the MTC. The LAR further stated that the transient response in the AOR presented in the UFSAR bounds the core design from CEA rod swap redesign, and hence no reanalysis of this event is required. The MDNBR for this event is greater than the correlation limit and the peak LHGR during the event remains below the limit that precludes FCM. The MDNBR and the FCM performance for this event bounds the CEA withdrawal event at low power or subcritical conditions. No reanalysis was performed for this event.

Based on the review of the information presented in attachments 5 and 6 to the LAR, the NRC staff finds the licensee evaluation of impact of the CEA rod group swap on the uncontrolled CEA withdrawal accident at low power or subcritical conditions to be acceptable. The staffs findings of acceptability are based on the acceptance criteria for cladding strain for the event being met at power conditions and the transient response in the AOR presented in the UFSAR bounding the changes in the core design from the CEA rod group swap.

Control Rod Misoperation This event is described in the CCNPP, Units 1 and 2, UFSAR, Section 14.11 as failure of the control rod system which leads to dropping of CEA into the core causing a power redistribution.

Such redistribution of power can lead to a local augmentation of the peaking factor in the affected region of the core. The LAR stated that the limiting MTC and the range of dropped CEA worth remains unchanged due to swap in CEA rod group assignments and therefore the transient response documented in AOR bounds the core design from the proposed swap. No reanalysis was performed for this event.

Based on the review of the information presented in Attachments 5 and 6 to the LAR, the NRC staff finds the licensee evaluation of impact of control rod misoperation to be acceptable. The staffs finding of acceptability is based on fact that the key parameters of the FTC and the MTC for the event remain bounding, the specified acceptable fuel design limits are not violated since the initial steady-state DNB margin remains protected and the peak LHGR during the event was shown to remain below the FCM limit.

Spectrum of Control Rod Ejection Accidents This event is described in the CCNPP, Units 1 and 2, UFSAR, Section 14.13 as postulated rupture of a control rod drive mechanism housing which allows the full system pressure to act on the drive shaft and control rod from the core. This event could lead to a rapid positive reactivity insertion and an increase in radial power peaking leading to localized fuel damage.

The licensee stated in the LAR that the existing licensing basis is based on a maximum rod motion-based reactivity insertion and limiting ranges for the DTC and MTC. No reanalysis was performed for this event since the limiting DTC, MTC, and ejected CEA worth remain unchanged.

Based on the review of the information presented in attachments 5 and 6 to LAR, the NRC staff finds the licensee evaluation of impact of spectrum of control rod ejection accidents to be acceptable. The staffs finding of acceptability is based on the information presented that the limiting DTC, MTC, and ejected CEA worth remain unchanged from the CEA rod group swap and hence the MDNBR and peak LHGR for the event remain unchanged and below the safety analysis limit.

Other Chapter 14 Accident Analyses The LAR stated that a review of all other UFSAR Chapter 14 accident analysis, including loss of coolant accident (LOCA) was performed and these events were determined to be unaffected by the change. A review of Section 5.2 of the attachments 5 and 6 to the LAR shows that following category of events were evaluated and dispositioned with justification along with their corresponding event in the NUREG-0800 (SRP) Chapter 15 event category:

Increase in Heat Removal by the Secondary System (SRP 15.1)

Decrease in Heat Removal by the Secondary System (SRP 15.2)

Decrease in Reactor Coolant System Flow (SRP 15.3)

Reactivity and Power Distribution Anomalies (SRP 15.4)

Increases in Reactor Coolant System Inventory (SRP 15.5)

Decrease in Reactor Coolant Inventory (SRP 15.6)

Radioactive Releases from a Subsystem or Component (SRP 15.7)

Anticipated Transients Without Scram (SRP 15.8)

Based on the review of the information presented in attachments 5 and 6 to the LAR, the NRC staff finds the licensee identified the events from the CCNPP, Units 1 and 2, UFSAR, Chapter 14 corresponding to the SRP events listed above, that were applicable.

The NRC staff finds that the cycle-specific parameter evaluations for UFSAR Chapter 14 safety analysis performed using the NRC-approved methods confirm that the values assumed in the safety analysis remain bounding, except for the uncontrolled CEA withdrawal at low power or subcritical condition event. The licensee proposed an operating restriction in Section 3.2 of the LAR to address the event, and the results meet acceptance criteria with the proposed operating restriction. Hence, the NRC staff finds the proposed CEA group reassignments to be acceptable for the rest of CC2C26.

3.1.4 Impact on Trip Setpoints - LCO Monitoring The licensee does not propose changes to the TS allowable values, and states in attachments 5 and 6, Section 6.0, Trip Setpoint Verification, that Framatome Inc. setpoint analyses verify the validity of the current trip setpoints.

The licensee stated in the LAR that with the proposed CEA group reassignments additional margin in the negative Axial Shape Index (ASI) was needed and thus Figure 3.2.1-2 of the COLR was revised to support the limits on the Linear Hear Rate (LHR) LCO. The licensee stated that the limiting LHR calculation results, in conjunction with the revised figure, ensure that margin exists and the adequacy of the LHR LCO function is maintained. The NRC staff find the change acceptable since the change ensures margin to fuel design limit exists and the adequacy of the LHR LCO function is maintained.

3.2 Licensee Proposed Cycle-Specific Restrictions The licensee proposed following two operating restrictions in Section 3.2 of the LAR:

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.

Section 2.5 of the LAR has the proposed condition to be added to the CCNPP, Unit 2, RFOL Appendix C based on the operating restrictions. No TSs are impacted by the proposed operating restriction.

The proposed operating restrictions have been evaluated and found acceptable in Section 3.1.3 of this SE. Based on the evaluation presented in the LAR and adding the proposed conditions to the operating license, the NRC staff find the proposed operating restrictions to be acceptable.

3.3 Compliance with the GDC The NRC staff's review of the applicable GDC listed in Section 2.0 of this SE is summarized below.

AEC GDC 6, Reactor Core Design This GDC is analogous to GDC 10, Reactor Design of the Appendix A to 10 CFR Part 50. The licensee evaluated the existing reload analysis in light of the redesign and analyzed impacted events in accordance with the NRC-approved methods described in TS 5.6.5. The NRC staff finds that this meets the requirements of AEC GDC 6 since the analysis performed shows that the fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

AEC GDC 7, Suppression of Power Oscillations This GDC is analogous to GDC 12, Suppression of Reactor Power Oscillations of the Appendix A to 10 CFR Part 50. The licensee evaluation in that LAR showed that the proposed change does not significantly impact the control rod worth of the CEAs. Hence, the ability of the CEAs to readily suppress power oscillations is not impacted by the proposed swap in CEA rod group. The NRC staff finds that this criterion remains satisfied as the rod worths are not impacted and therefore the staff agrees the CEAs would remain capable of suppressing power oscillations, which could otherwise result in conditions exceeding specified acceptable fuel design limits.

AEC GDC 8, Overall Power Coefficient This GDC is analogous to GDC 11, Reactor Inherent Protection of the Appendix A to 10 CFR Part 50. The licensee evaluations performed in attachments 5 and 6 demonstrated that the proposed change impacts the reactivity feedback coefficients by less than 1 percent and that the reactivity feedback coefficients are not positive in the power operating range.

The NRC staff finds this criterion is satisfied since the reactivity feedback coefficients are non-positive for power operating conditions, thereby providing negative reactivity feedback characteristics.

AEC GDC 14, Core Protection Systems, and AEC GDC 15, Engineered Safety Features Protection Systems These GDCs are analogous to GDC 20, Protection System Functions, of the Appendix A to 10 CFR Part 50. The licensee evaluations showed that proposed change does not impact core protection systems and engineered safety features protection systems. Hence, 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 is not impacted by the change. The NRC staff finds this criterion is satisfied since the core protection systems and the engineered safety protection systems are not impacted by the change and their ability to provide required protection is unchanged.

AEC GDC 19, Protection System Reliability, and AEC GDC 20, Protection Systems Redundancy and Independence These GDCs are analogous to GDC 21, Protection System Reliability and Testability, of Appendix A to 10 CFR Part 50. The licensee performed evaluations and stated that the proposed change does not impact the ability of the CEA with degraded CEDM UGC (CEA 36) to fully insert when tripped. The NRC staff finds this criterion is satisfied since the proposed change does not impact reliability, redundancy, or independence of protection systems.

AEC GDC 27, Redundancy of Reactivity Control, AEC GDC 28, Reactivity Hot Shutdown Capability, AEC GDC 29, Reactivity Shutdown Capability, and AEC GDC 30, Reactivity Holddown These GDCs combined are analogous to GDC 26, Reactivity Control System Redundancy and Capability, and GDC 27, Combined Reactivity Control Systems Capability, of the Appendix A to 10 CFR Part 50. The evaluations that the licensee performed showed that the reactivity control systems are capable of keeping and maintaining the core subcritical under any condition sufficiently fast to prevent violation of any fuel design limits and that the core SDM remains greater than the rod worth of the most reactive rod fully withdrawn. Further, the proposed change does not impact the independence of the two different reactivity control systems: CEAs and boron injection. The NRC staff finds this criterion is satisfied since the independence of the reactivity control systems is not impacted and the systems remain capable of making the core subcritical from any hot standby or hot operating condition sufficiently fast to prevent exceeding acceptable fuel damage limits.

AEC GDC 31, Reactivity Control Systems Malfunction This GDC is analogous to GDC 11, Protection System Requirements for Reactivity Control Malfunctions of the Appendix A to 10 CFR Part 50. The evaluations that the licensee performed show that after the incorporation of the operating restrictions, there is no change to the accident analyses The licensee incorporated operating restrictions for the uncontrolled reactivity addition event below 1 percent of the RTP, which are enforced via the proposed changes described in Section 2.5 of the LAR. The reactivity control systems, after the incorporation of the operating restrictions, remain unchanged and capable of sustaining any single malfunction. The NRC staff finds that this criterion remains satisfied as the reactor trip function remains fully capable of

performing its function with the proposed CEA group swap, and fuel design limits are not exceeded for any malfunctions.

AEC GDC 32, Maximum Reactivity Worth of Control Rods This GDC is analogous to GDC 28, Reactivity Limits of the Appendix A to 10 CFR Part 50. The licensee performed evaluations show that after the incorporation of the operating restrictions, there is no change to the accident analyses, including the evaluations performed for control rod withdrawal, misoperation, and ejection. The NRC staff finds that this criterion remains satisfied, as the licensee evaluations demonstrated that the reactivity insertion rate, SDM, and the safety analysis limits remain met for rest of the entire fuel cycle with the proposed CEA group swap.

3.7 Conclusion The NRC staff concludes that the licensee proposed one-time change to Calvert Cliffs, Unit 2, CEA group definitions that would swap the group assignments by assigning CEA 36 to regulating group 1 and CEA 60 to regulating group 5 for rest of the CC2C26, is acceptable because the design change does not challenge the safety analyses detailed in CCNPP, Units 1 and 2, UFSAR and Chapter 14 UFSAR, when the operating restrictions proposed by the licensee are implemented. The NRC staff finds that the licensee used NRC-approved methods consistent with regulatory requirements and guidance identified in Section 2.0 of this SE.

Hence, the NRC staff finds the proposed LAR to be acceptable based on the technical evaluation provided in the LAR and operating restrictions proposed by the licensee.

The NRC staff notes that the acceptability of the present LAR is only on a temporary basis until the end of the current operating cycle, CC2C26.

4.0 EXIGENT CIRCUMSTANCE In its October 17, 2025, submittal, the licensee stated that in this instance, an emergency exists because the amendment is required to prevent an unnecessary shutdown of CCNPP, Unit 2.

This is because CEA 36 is currently unable to be moved with its assembly regulating group (though CEA 36 is still OPERABLE if a plant trip is experienced). During an upcoming plant downpower to address other plant equipment issues, movement of CEA 36, along with other CEAs in its current assembly group will be required. The inability to move CEA 36 will result in an inability to meet TS alignment requirements for the CEAs which would then require a plant shutdown.

The licensee stated that 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 automatic voltage regulator communication control measuring card. Further, the licensee stated that NRC approval was needed sooner than can be provided under exigent circumstances, and this LAR is timely considering the unplanned nature of the issues and time required to develop the technical justification supporting the proposed change. Additionally, CCNPP did not knowingly cause the emergent situation and CEG affirmed a best effort has been made for a timely license amendment application.

NRC Staff Conclusion regarding Exigent Circumstance The NRC staff reviewed the licensees basis for processing the proposed LAR as an emergency amendment and concluded that there was sufficient time to process the LAR using exigent procedures instead of emergency procedures. Accordingly, the NRC staff used the exigent provisions in 10 CFR 50.91(a)(6)(i)(B) to provide reasonable notice of the LAR and an opportunity to comment on NRCs proposed determination that no significant hazards consideration is involved in the requested amendment.

Accordingly, the NRC used local media and published a public notice in the Calvert Recorder located in Prince Frederick, Maryland (https://www.somdnews.com); a news organization local to the licensees facility on October 31, 2025. The notice included the NRC staffs proposed no significant hazards consideration determination. The notice also provided an opportunity for public comment until November 2, 2025, regarding the NRC staffs proposed determination. No public comments were received regarding the proposed amendment.

5.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

DETERMINATION The NRCs regulation in 10 CFR 50.92(c) states that the NRC may make a final determination, under the procedures in 10 CFR 50.91, that a license amendment involves no significant hazards consideration if operation of the facility, in accordance with the amendment, would not:

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

The NRC staffs evaluation of the issue of no significant hazards consideration is presented below:

1.

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

Response: No.

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 along with the proposed operating restriction will be 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, along with the proposed operating restriction and the updated COLR Figure 3.2.1-2, does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

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

Response: No.

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 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.

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. Although the proposed change has an impact on several accident analysis parameters routinely evaluated as part of the reload design process, the cycle-specific evaluations performed to determine if the change in core design adversely impacts bounding key safety parameters assumed in the safety analyses did not identify any significant reductions in margins of safety. The proposed change, including operational restrictions for restoring the original CEA group assignments, does not impact on previously evaluated accident analyses. The proposed change does not result in exceeding design basis limits. All licensed safety margins are maintained with the implementation of the operating restrictions.

Therefore, operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.

For the reasons noted above, there is no significant reduction in a margin of safety, after the incorporation of operating restrictions and update to COLR Figure 3.2.1-2. Based on the above evaluation, the NRC staff concludes that the three standards of 10 CFR 50.92(c) are satisfied.

Therefore, the NRC staff has made a final determination that no significant hazards consideration is involved for the proposed amendment and that the amendment should be issued as allowed by the criteria contained in 10 CFR 50.91.

6.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Maryland State official was notified of the proposed issuance of the amendment on October 21, 2025. There were no comments received from the Maryland State official.

7.0 ENVIRONMENTAL CONSIDERATION

The amendment changes requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. Further, the NRC staff has found that the amendment involves no significant hazards consideration. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

8.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that 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.

Principal Contributors: S. Bhatt, NRR M. Li, NRR Date: November 3, 2025

ML25304A004 NRR-058 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL1/LA NRR/DSS/SNSB/BC NRR/DEX/EICB/BC(A)

NAME GEMiller KEntz (SLent for)

NDiFrancesco (RPatton for)

SDarbali DATE 10/27/2025 10/23/2025 10/27/2025 10/27/2025 OFFICE OGC - NLO NRR/DORL/LPL2-2/BC NAME DRoth DWrona DATE 10/29/2025 11/03/2025