ML25084A354
| ML25084A354 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 04/10/2025 |
| From: | Mahesh Chawla Plant Licensing Branch IV |
| To: | Diya F Union Electric Co |
| Chawla M | |
| References | |
| EPID L-2025-LLA-0016 | |
| Download: ML25084A354 (24) | |
Text
April 10, 2025 Fadi Diya Senior Vice President and Chief Nuclear Officer Ameren Missouri Callaway Energy Center 8315 County Road 459 Steedman, MO 65077
SUBJECT:
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)
Dear Fadi Diya:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 239 to Renewed Facility Operating License No. NPF-30 for Callaway Plant, Unit No. 1 (Callaway). The amendment consists of changes to the technical specifications (TSs) in response to your application dated January 29, 2025.
The amendment revises Callaway TS 4.2.2, Control Rod Assemblies, to add a note that for Operating Cycle 28, operation is permitted with 52 control rod assemblies with no control rod assembly installed in core location H-08.
A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Mahesh L. Chawla, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-483
Enclosures:
- 1. Amendment No. 239 to NPF-30
- 2. Safety Evaluation cc: Listserv
UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT NO. 1 DOCKET NO. 50-483 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 239 License No. NPF-30
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Union Electric Company (UE, the licensee),
dated January 29, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions 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 the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-30 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan*
The Technical Specifications contained in Appendix A, as revised through Amendment No. 239 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
This amendment is effective as of its date of issuance, and shall be implemented within 90 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Tony Nakanishi, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to Renewed Facility Operating License No. NPF-30 and the Technical Specifications Date of Issuance: April 10, 2025 TONY NAKANISHI Digitally signed by TONY NAKANISHI Date: 2025.04.10 17:36:11 -04'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 239 CALLAWAY PLANT, UNIT NO. 1 RENEWED FACILITY OPERATING LICENSE NO. NPF-30 DOCKET NO. 50-483 Replace the following pages of Renewed Facility Operating License No. NPF-30 and the Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Renewed Facility Operating License REMOVE INSERT Technical Specifications REMOVE INSERT 4.0-1 4.0-1
Renewed License No. NPF-30 Amendment No. 239 (3)
UE, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)
UE, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source of special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)
UE, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This renewed license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level UE is authorized to operate the facility at reactor core power levels not in excess of 3565 megawatts thermal (100% power) in accordance with the conditions specified herein.
(2)
Technical Specifications and Environmental Protection Plan*
The Technical Specifications contained in Appendix A, as revised through Amendment No. 239 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
Environmental Qualification (Section 3.11, SSER #3)**
Deleted per Amendment No. 169.
Amendments 133, 134, & 135 were effective as of April 30, 2000 however these amendments were implemented on April 1, 2000.
The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 239 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-30 UNION ELECTRIC COMPANY CALLAWAY PLANT UNIT NO. 1 DOCKET NO. 50-483
1.0 INTRODUCTION
By letter dated January 29, 2025 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML25029A188), Union Electric Company, doing business as, Ameren Missouri (the licensee) submitted a license amendment request (LAR) to the U.S.
Nuclear Regulatory Commission (NRC, the Commission) requesting changes to the technical specifications (TSs) for Callaway Plant Unit No. 1 (Callaway).
The proposed amendment would revise Callaway TS 4.2.2, Control Rod Assemblies, to add a note that for Operating Cycle 28, operation is permitted with 52 control rod assemblies with no control rod assembly installed in core location H-08.
In section 1.0, Summary Description, of the enclosure to the LAR, the licensee stated that
[t]he results of control rod drop time testing performed during recent refueling outages indicate a slowing rod drop time for the H-08 control rod. While the drop times are still within the TS allowable limits (per TS Surveillance Requirement 3.1.4.3), licensee troubleshooting is planned for the upcoming refueling outage, Refuel 27. The licensee stated that troubleshooting activity could indicate the need for a repair that would not be able to be performed during the outage. If so, the H-08 control rod would be removed with the intent of leaving it removed from the H-08 core location throughout Cycle 28 such that the H-08 control rod drive mechanism would be repaired or replaced during the next refueling outage, Refuel 28.
1.1
System Description
The licensee stated, in part, in section 3.1 of the enclosure to the LAR that, The Callaway reactor normally contains 53 control rod assemblies divided into four control banks (Control Banks A, B, C, D) and five shutdown banks (Shutdown Banks SA, SB, SC, SD, and SE). Of the nine banks, Control Bank D is used for reactivity control during normal at-power operation. The remaining control banks are normally used for reactor startup and shutdown. The shutdown banks provide additional negative reactivity to meet shutdown margin (SDM) requirements. During MODES 1 [power operation] and 2 [startup], the shutdown
banks are fully withdrawn from the core in accordance with TS 3.1.5 [Shutdown Bank Insertion Limits,] and as specified in the Core Operating Limits Report (COLR).
The H-08 control rod is part of Control Bank D and is located directly in the center of the core. With the removal of the H-08 control rod, the core during Cycle 28 will contain 52 control rod assemblies.
Note that the terms control rod and rod cluster control assemblies (RCCAs) are used synonymously by the licensee.
1.2 Proposed Change In Callaway TS section 4.0, Design Features, TS 4.2.2, Control Rod Assemblies, states:
The reactor core shall contain 53 control rod assemblies. The control rod material shall be silver indium cadmium, hafnium metal, or a mixture of both types, as approved by the NRC.
The licensee proposes to revise TS 4.2.2, to add a note reflecting that for Operating Cycle 28, operation is permitted with 52 control rod assemblies. Specifically, the licensee proposes to add the following:
NOTE-----------------------------------------------
Operation with 52 control rod assemblies (i.e., with no control rod assembly installed in core location H-08) is permitted during Cycle 28.
2.0 REGULATORY EVALUATION
Under Title 10 of the Code of Federal Regulations (10 CFR) Section 50.90, whenever a holder of a license wishes to amend the license, including technical specifications in the license, an application for amendment must be filed, fully describing the changes desired, and following as far as applicable, the form prescribed for original applications. 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 to the extent applicable and appropriate. Both the common standards in 10 CFR 50.40(a), and those specifically for issuance of operating licenses in 10 CFR 50.57(a)(3), provide that there must be reasonable assurance that the activities at issue will not endanger the health and safety of the public.
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.
The regulation in 10 CFR 50.36(c)(4) requires TS to include items in the Design features category, stating:
Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in
[10 CFR 50.36(c)(1), Safety limits, limiting safety system settings, and limiting control settings; 10 CFR 50.36(c)(2), Limiting conditions for operation; and 10 CFR 50.36(c)(3), Surveillance requirements.]
Section 3.1, Conformance with NRC General Design Criteria, of the Callaway Final Safety Analysis Report (FSAR) (ML23062A644) addresses the plants conformance with the general design criteria (GDC) specified in Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50. The GDC in Appendix A to 10 CFR Part 50 applicable to this LAR are listed below and discussed in detail in section 3.6 of this SE:
GDC 10 - Reactor design GDC 11 - Reactor inherent protection GDC 12 - Suppression of reactor power oscillations GDC 23 - Protection system failure modes GDC 25 - Protection system requirements for reactivity control malfunctions GDC 26 - Reactivity control system redundancy and capability GDC 27 - Combined reactivity control systems capability GDC 28 - Reactivity limits GDC 29 - Protection against anticipated operational occurrences The following guidance documents were 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 licensee's application to determine if the proposed changes are consistent with the regulations, guidance, and licensing and design-basis information discussed in section 2.0 of this SE. The NRC staff reviewed the licensees statements in the LAR, the licensees use of NRC-approved methodologies, and the relevant sections of the Callaway FSAR and TS. The NRC staff compared the impacted parameters listed in the LAR with the corresponding key safety parameters to verify that the list of parameters was complete and that they were bounded by the safety analysis in Chapter 15, Accident Analysis, of the FSAR (ML23062A635).
As stated in the enclosure to the LAR, the licensees reload safety analysis applies NRC-approved codes and methods to design the reload core, where the specific codes and methods are provided in TS 5.6.5, Core Operating Limits Report (COLR), and the cycle-specific COLR.
The licensee stated that the reload safety analysis methods are not invalidated by the removal of control rod H-08 from the Cycle 28 core design because these methods are not dependent on a particular RCCA configuration. The licensee stated that cycle-specific reload evaluations of TS limits, safety analysis limits, and operating limits without control rod H-08 for Cycle 28 were performed to ensure core protective and operating limits remain satisfied and safety analysis limits remain bounded. The NRC staff finds that the reload safety analysis methods and supporting computer codes used by the licensee remain applicable to model and evaluate the as-designed/operated configuration of the plant, as the reload methodology is not dependent upon control bank configuration.
The licensee stated that there were no changes in analytical methods or safety analysis limits used to perform the core reload safety evaluation for Cycle 28 with control rod H-08 removed.
The analysis supporting the evaluation of these impacted parameters was performed using the NRC-approved methodology described in Callaway TS 5.6.5.
3.1 Impact on Safety Analysis Parameters The licensee, as part of its reload process used for each new fuel cycle, determined the nuclear design changes and impact to core and fuel performance, as well as impact to the accident analyses described in Chapter 15 of the FSAR for removal of control rod H-08 for Cycle 28. The licensee stated that NRC-approved reload safety analysis codes and methods were used to determine if the change in core design parameters adversely impacted the bounding key safety parameters assumed in the safety analysis. Specific parameters are discussed in the sections below.
The NRC staff examined each parameter presented by the licensee to: (1) confirm the selected parameters (shutdown margin, trip reactivity, etc.) were comprehensive and consistent with the removal of control rod H-08, (2) that the updated values were reasonable given the proposed change, and (3) that the safety analysis parameters as defined in the COLR and FSAR remain bounding.
3.1.1 Shutdown Margin Removal of a control rod has a direct impact on the available SDM. SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences. SDM is a requirement in Callaway TS 3.1.1, SHUTDOWN MARGIN (SDM), and is referenced in action statements in TSs 3.1.4, Rod Group Alignment Limits; 3.1.5, Shutdown Bank Insertion Limits; 3.1.6, Control Bank Insertion Limits; and 3.1.8, PHYSICS TESTS Exceptions - MODE 2. These TSs state that the SDM shall be within the limits specified in the COLR. The current Callaway COLR section 2.1.1 states that the Shutdown Margin in MODES 1 through 4 shall be greater than or equal to 1.3 %k/k (alternatively denoted as % in the current LAR). Using NRC-approved methodology, the licensee calculated the SDM both before and after removal of control rod H-08 and attested that the SDM is reduced from 2.289 % to 2.154 % at the beginning-of-cycle (BOC) and from 1.861 % to 1.479 % at the end-of-cycle (EOC). The details are shown in the table below (reproduced as-is from table 2 of the enclosure to the LAR).
While it is noted that such a reduction in % (or %k/k) is due to the proposed removal of control rod H-08, these reduced values remain above the 1.3 % limit specified in the COLR.
As a result, the safety analysis described in Chapter 15 of the Callaway FSAR remains bounding with regards to SDM for accidents initiated in MODES 1 (Power Operation) and 2 (Startup).
In addition, the COLR also requires that SDM be greater than or equal to 1.3 % in subcritical Modes 3 (Hot Standby) and 4 (Hot Shutdown) and be at least 1.0 % in Mode 5 (Cold Shutdown). A means to ensure this would be to increase the boron concentration in the reactor coolant system (RCS) to offset the absence of control rod H-08 needed for safe shutdown. The table provided by the licensee (reproduced as-is from table 3 of the enclosure to the LAR) illustrates the minimum boron concentration required to maintain the safety margin above 1.3 % in Modes 3 and 4 and 1.0 % in Mode 5 for scenarios with and without control rod H-08 available. The licensee states that the critical boron concentration protocols as stated in FSAR section 15.4.6.2, Analysis of Effects and Consequences, for Modes 1 through 5 will be applicable to the proposed Cycle 28 and thus the analysis continues to remain bounding for the without control rod H-08 available.
Required Boron with H-08 (ppm)*
Required Boron without H-08 (ppm)*
1.3% SDM 350°F**
557°F**
350°F**
557°F**
BOC 1215 1072 1245 1116 MOC***
1038 860 1083 923 EOC 399 106 448 171 1.0% SDM 68°F**
200°F**
68°F**
200°F**
BOC 1230 1220 1253 1244 MOC***
1063 1050 1094 1088 EOC 504 462 524 500
- parts per million (ppm)
- degrees Fahrenheit (°F)
- middle of cycle (MOC)
After review of the information presented in section 3.3.1, Shutdown Margin, of the enclosure to the LAR, the NRC staff finds that the SDM limits are maintained (as per TSs), and the SDM assumptions used in the safety analyses listed in Chapter 15 of the FSAR remain valid after removal of control rod H-08.
3.1.2 Boron Concentration and Worth The licensee stated in its LAR that the removal of control rod H-08 increases the required boron concentration and reduces boron worth under conditions when control rods are inserted into the core. This results in a change in boron concentration requirements in the RCS for Modes 1 through 5 and impacts the uncontrolled boron dilution accident. The licensee attests that while the boron concentration requirements are revised, the uncontrolled boron dilution accident for Modes 1 through 5 continues to remain bounding for the removal of control rod H-08. The LAR does not indicate any changes in TS SDM requirement, instrumentation, system alignment, primary water flow rate, operator alarms, and operator actions and timing due to the removal of control rod H-08.
Therefore, the NRC staff finds that with the removal of control rod H-08, the boron concentration change based on meeting SDM requirements is still within the assumed values for the FSAR events and does not impact the results presented in FSAR section 15.4.6, Chemical and Volume Control System Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant, which describes the uncontrolled boron dilution accident in Modes 1 through 6.
3.1.3 Trip Reactivity The removal of control rod H-08 reduces the trip reactivity (%k, total negative reactivity inserted in the core during a trip event) as a function of rod insertion position. This reduces the trip reactivity as a function of time after the RCCAs begin to fall. The normalized trip reactivity as a function of RCCA insertion position and the normalized trip reactivity as a function of time after the RCCAs begin to fall are presented in Chapter 15 of the Callaway FSAR. The licensees evaluations provided in the LAR demonstrate that despite the removal of control rod H-08, trip reactivity as a function of RCCA insertion position and trip reactivity as a function of time after the RCCAs begin to fall used in the safety analyses remain bounded by the curves called out in the FSAR. Further, the licensee provided a comparison of the trip reactivity (%k) as a function of the rod position for with and without control rod H-08 inserted scenarios for Cycle 28. The results provided in table 4 in the enclosure to the LAR demonstrate that the trip reactivity is maintained above the required limits at all rod positions for both scenarios. The licensee also evaluated the minimum trip worth for the proposed cycle for scenarios with and without control rod H-08 inserted. The results indicate that the minimum trip reactivities are met at all rated thermal power conditions in both scenarios.
Therefore, the NRC staff finds that the removal of control rod H-08 does not impact the trip reactivity assumed in the Callaway FSAR Chapter 15 events.
3.1.4 Moderator Temperature Coefficient The moderator temperature coefficient (MTC) is one of the controlling parameters for core reactivity in both overheating and overcooling accidents. Both the most positive and most negative values of the MTC are important to safety, and both values must be bounded. Values used in the safety analyses consider worst case conditions to ensure that the accident results are bounding. The consequences of accidents that cause core overheating must be evaluated with the most positive MTC. Such accidents include the uncontrolled bank withdrawal transient from any power level, loss of electrical load, and loss of forced reactor coolant flow. The consequences of accidents that cause core overcooling must be evaluated with the most
negative MTC. Such accidents include sudden increase in feedwater flow and steam line breaks.
To ensure a bounding accident analysis, the MTC is assumed to be its most limiting value for the analysis conditions appropriate to each accident. The bounding value is determined by considering rodded and unrodded conditions, whether the reactor is at full or zero power, and whether it is the BOC or EOC. The most conservative combination appropriate to the accident is then used for the analysis.
The removal of control rod H-08 impacts the MTC calculated at the conservative bounding conditions determined for the FSAR accident analyses. The licensee evaluated the MTC for Cycle 28 with and without control rod H-08 with the results shown in the table below (reproduced as-is from Table 6 of the enclosure to the LAR).
Limit Description Limit Reload Values with H-08 Reload Values without H-08 Most positive HFP* MTC
< 0 pcm/F***
-3.784
-3.784 HFP error-adjusted rod insertion limit (Bank D at 149 steps withdrawn) EOC
> -40.4 pcm/F***
-34,684
-34,684 Near-EOC MTC at 300 ppmB****
> -40.4 pcm/F***
-28.482
-28.479 Near-EOC MTC at 60 ppmB****
> -45.5 pcm/F***
-33.259
-33.259 Most positive HZP** MTC
< 5 pcm/F***
3.936 3.936
- Hot full power (HFP)
- Hot zero power (HZP)
- Percent mille per degree F (pcm/F)
- Parts per million Boron (ppmB)
The licensee attests that the values in the table confirm the limits assumed in the safety analysis remain bounding. Based on the evaluation, the licensee states that the removal of control rod H-08 does not impact the results presented in FSAR sections 15.1.2, 15.1.5, 15.2.2, 15.3.2, 15.4.1, and 15.4.2.
The NRC staff reviewed the information and values provided in the LAR against those in the COLR, and those used in the FSAR Chapter 15 analysis along with the justifications and found that they remain conservative.
3.1.5 Miscellaneous Safety Analysis Neutronic Parameters Miscellaneous safety analysis neutronic parameters such as delayed neutron data (beta and prompt neutron lifetime), Doppler temperature coefficients, and fuel temperatures are not significantly impacted by the change in core configuration. These parameters are driven more directly by the core design.
The NRC staff finds acceptable, the licensees attestation, that the licensees cycle-specific parameter evaluations of these safety analysis values show negligible changes and confirm that the values assumed in the safety analysis remain bounding.
3.2 Chapter 15 Accident Analysis Impacts As stated by the licensee, removal of control rod H-08 has an impact on most comparisons to FSAR Chapter 15 accident analysis parameters routinely evaluated as part of the reload safety
evaluation. In addition to the items discussed above in section 3.1 (SDM, MTC, trip reactivity, boron concentration, etc.), the impact of removal of control rod H-08 on control rod worth, margin to peaking limits (departure from nucleate boiling), and centerline fuel melt), and other accident analysis parameters were also evaluated by the licensee.
The bounding key safety parameters are developed in the Callaway FSAR Chapter 15 accident analysis of record (AOR) 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. The licensee performed cycle-specific evaluations to determine if the change in control rod configuration adversely impacts bounding key safety parameters assumed in the FSAR Chapter 15 safety analysis and any impacts on departure from nucleate boiling and fuel thermal margins due to the change in power distribution.
Table 16, Impact o FSAR Chapter 15 Accident Analyses, of the enclosure to the LAR provides the impact on the FSAR Chapter 15 accident analyses due to the removal of control rod H-08.
The licensees table provides the FSAR section, description of the event, and the licensees comment on the events impact. For most cases, the licensee stated that the cycle specific reload evaluations verify that the AOR remains bounding with removal of control rod H-08. For the few remaining cases, removal of control rod H-08 has no impact. These include events such as inadvertent loading and operation of a fuel assembly in improper position, steam generator tube rupture, radioactive waste gas decay tank failure, and fuel handling accidents. In some of these cases, there is no impact on the analysis results as control rod insertion is not credited. In other cases, while the actual trip reactivity is affected by the removal of control rod H-08, the trip reactivity used in the analysis is conservative and remains bounding. Overall, the licensee found that the removal of control rod H-08 for Cycle 28 does not impact the results presented in FSAR Chapter 15.
The NRC staff reviewed the licensees results presented in the enclosure to the LAR for key parameters. The results of the cycle specific evaluations confirm that the limits assumed in the safety analysis remain bounding. Therefore, NRC staff finds the removal of control rod H-08 during Cycle 28 does not impact the results presented in the FSAR Chapter 15 accident analyses.
3.3 Thermal-Hydraulic Design Impacts With control rod H-08 and its associated driveshaft removed from service, the flow in the upper head region of the vessel will increase. The licensee bypass flow analysis confirmed that there is a negligible increase in core bypass flow and that the design core bypass limits are not exceeded. The licensee analysis estimated a total core bypass flow increase of approximately 0.08 percent at maximum. In addition, the changes in RCS water volume and metal mass are not appreciably impacted by removal of control rod H-08 and its associated driveshaft.
Therefore, NRC staff finds that the removal of control rod H-08 during Cycle 28 is acceptable.
3.4 Structural Evaluation In section 3.5.1, Thermal-Hydraulic Impacts, of the enclosure to the LAR, the licensee states that the removal of the H-08 drive rod results in a 0.239 percent increase in the flow in the upper reactor vessel head region. In section 3.5.2, Seismic and Structural Impacts, of the enclosure
to the LAR, the licensee states that [t]here is no impact on the functionality or structural integrity of the reactor vessel upper internals with the removal of the control rod drive shaft and RCCA at core location H-08.The licensee also states that [t}here is no impact on the current reactor vessel internals analyses.
FSAR section 3.7(N).3.14 discusses the control rod drive mechanism (CRDM) housing dynamic analysis (seismic and LOCA). The licensee states that the removal of the control rod drive shaft has negligible impact on the structural integrity of the CRDM housing dynamic analysis, so the current analysis would remain bounding with removal of the H-08 control rod.
The NRC staff has evaluated the licensees proposal to operate Callaway with the H-08 control rod removed throughout Cycle 28 with its repair or replacement during the next refueling outage, Refuel 28. The staff concurs that the absence of the H-08 control rod has negligible effect on the flow in the upper reactor vessel head region and the overall weight of the reactor vessel internals. The staff finds that the removal of H-08 control rod has an insignificant effect on the reactor vessel internals analyses, and will not cause an impact on the functionality or structural integrity of the reactor vessel upper internals. The staff has not identified any concerns with the structural integrity of the CRDM housing with the removal of the control rod drive shaft. Therefore, the NRC staff has determined that the removal of the H-08 control rod and its control rod drive shaft during plant operation during Cycle 28 is acceptable until its repair or replacement in Refuel 28.
3.5 Other Considerations 3.5.1 Impact on Operating Analysis Support In the enclosure to the LAR, the licensee stated that the reload safety evaluation methodology and computer code package (PARAGON/ANC9) currently used are applicable to model and evaluate the as-designed/operated configuration of the plant. The PARAGON/ANC9 models that calculate reactivity parameters and power distribution performance are not impacted nor invalidated due to removal of the control rod H-08 from Cycle 28 core, as the methodology is not dependent upon control bank configuration. Cycle-specific reload evaluations of TS limits and core operating limits without control rod H-08 for Cycle 28 are performed to ensure applicable safety analysis limits remain satisfied.
The licensee stated that there will not be any differences in the COLR limits for TS 3.2.1, Heat Flux Hot Channel Factor (FQ(Z)) (FQ Methodology), and 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor (FNDH), for core monitoring due to removal of control rod H-08 from the Cycle 28 core. Explicit modeling of the new core configuration is used in the generation of the cycle-specific peaking factor limits.
Based on its review of the above information, the NRC staff finds that the reload safety analysis methods and supporting computer codes used by the licensee remain applicable to model and evaluate the as-designed/operated configuration of the plant, as the reload methodology is not dependent upon control bank configuration.
3.5.2 Field Work Required to Remove the H-08 Control Rod from Service In the LAR, the licensee states that before commencement of Cycle 28, the licensee may remove control rod H-08 for repairs and operate Cycle 28 without control rod H-08. In such a scenario, the licensee submits that the impacts of any physical/electrical/software changes will
be governed by the current plant design change procedures. If necessary, this would include removal of the RCCA from the fuel assembly in core location H-08, removal of inputs to the digital rod position indication (DRPI) software, modifications to the DRPIs and alarm, and control system fuses for control power to the H-08 CRDM.
This calls to question if any of the above actions will adversely impact the DRPIs functioning.
The DRPI is a system that provides indication of actual and demanded control rod position, and alarms for any misaligned rods or excessive rod insertion to ensure SDM. The licensee states that the DRPI is a nonsafety-related system independent of the rod control and reactor protection systems.
3.5.3 Adequate Level of Safety The licensee addressed the level of safety based on the proposed removal of control rod H-08.
Although reduced, the SDM continues to be maintained above the limits specified in the COLR and TS 3.1.1 with additional margin. Further, the licensee states that the evaluations of the impact on the safety analyses have demonstrated that requirements for reactivity control provided by control rods continue to be met, even with removal of control rod H-08 during Cycle 28.
The NRC staff evaluated the information provided in the LAR to assess the impact of removal (or absence) of control rod H-08 on the SDM, boron concentration/worth, trip reactivity and the impact on DRPI functionality (nonsafety-related). Therefore, NRC staff finds that the licensees assumption that insertion of the remaining 52 control rods will provide sufficient negative reactivity to shut down the reactor remains valid and is compliant with the TS, hence providing reasonable assurance that the proposed change does not endanger the health and safety of the public.
3.5.4 Impact on Operator Actions As discussed above in sections 3.1 and 3.2 of this SE, the licensees evaluations performed for Cycle 28 with control rod H-08 removed validate that the impacts to the nuclear design parameters are within the bounds of those already assumed in Chapter 15 of the FSAR. The licensee stated that no new or revised operator actions are required to meet the safety analyses acceptance criteria, and as a result, there are no changes required to the emergency operating procedures or the operator actions.
3.6 Compliance with Applicable GDC The NRC staff's review of the applicable GDC listed in section 2.0 of this SE is summarized below.
GDC 10 - Reactor design The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.
The licensee performed a redesign reload analysis in accordance with the methods described in TS 5.6.5 and confirmed that the fuel design limits are not exceeded during any condition of
normal operation, including the effects of anticipated operational occurrences with control rod H-08 removed. The NRC staff finds that this meets the requirements of GDC 10, since acceptable fuel design limits will not be exceeded during normal operation, including the effects of anticipated operational occurrences.
GDC 11 - Reactor inherent protection The reactor core and associated coolant systems shall be designed so that in the power operating range the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity.
The fuel temperature coefficient is negative, and the moderator temperature coefficient of reactivity is non-positive for power operating conditions, thereby providing negative reactivity feedback characteristics. The NRC staff finds that this criterion remains satisfied because removal of control rod H-08 does not impact the ability to detect or control core power distribution.
GDC 12 - Suppression of reactor power oscillations The reactor core and associated coolant, control, and protection systems shall be designed to assure that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.
The licensee states that power oscillations of the fundamental mode are inherently eliminated by the negative Doppler and non-positive moderator temperature coefficient of reactivity.
Oscillations due to xenon spatial effects in the radial, diametral, and azimuthal overtone modes are heavily damped due to the inherent design and due to the negative Doppler and nonpositive moderator temperature coefficients of reactivity. Oscillations due to xenon spatial effects in the axial first overtone mode may occur. Assurance that fuel design limits are not exceeded by xenon axial oscillations is provided as a result of reactor trip functions using the measured axial power imbalance as an input. Oscillations due to xenon spatial effects in axial modes higher than the first overtone are heavily damped due to the inherent design and due to the negative Doppler coefficient of reactivity.
During operation, axial imbalance is maintained within the limits of the Callaway Technical Specifications (i.e., imbalances which are alarmed to the operator and are within the imbalance trip setpoints). As such, the operator can suppress xenon axial oscillations by control rod motions and/or temporary power reductions.
The NRC staff finds that this criterion remains satisfied, as the safety analysis with control rod H-08 removed demonstrates that it will not result in power oscillations, which would result in conditions exceeding specified acceptable fuel design limits.
GDC 23 - Protection system failure modes The protection system shall be designed to fail into a safe state or into a state demonstrated to be acceptable on some other defined basis if conditions such as disconnection of the system, loss of energy (e.g., electric power, instrument air),
or postulated adverse environments (e.g., extreme heat or cold, fire, pressure, steam, water, and radiation) are experienced.
As stated by the licensee, the overall design of the protection system was not changed. The removal of control rod H-08 from the reactor vessel does not impact the fail-safe function of the remaining 52 control rods, which will still reliably maintain an adequate reactor shutdown capability. The physical removal of the control rod drive shaft does not have any mechanical impact on the function of the remaining 52 control rods. The electrical removal from service of control rod H-08 involves pulling fuses to remove control power to the respective stationary, lift, and movable coils. The remaining control rods are not impacted by this electrical change and will continue to meet their design function. The licensees modification design change process ensures that the associated plant modifications involve only control rod H-08 and do not affect other control rods.
The NRC staff finds that this criterion remains satisfied by maintaining the control rod insertion capability with the remaining 52 control rods.
GDC 25 - Protection system requirements for reactivity control malfunctions The protection system shall be designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods.
The protection system is designed to limit reactivity transients so that fuel design limits are not exceeded. The licensees Cycle 28 redesign reload analysis confirmed that FSAR Chapter 15 events were bounding for analyzed malfunctions of the reactivity control systems.
The NRC staff finds that this criterion remains satisfied as the reactor trip function remains fully capable of performing its function with 52 control rods, and fuel design limits were not exceeded for analyzed malfunctions of the reactivity control systems with the removal of control rod H-08.
GDC 26 - Reactivity control system redundancy and capability Two independent reactivity control systems of different design principles shall be provided. One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions.
Two reactivity control systems are provided, including the RCCAs and chemical shim (boric acid). The RCCAs are inserted into the core by the force of gravity. The boron chemical shim is unaffected and will maintain the reactor in the cold shutdown state, independent of the position of the control rods, and can compensate for all xenon burnout transients.
The NRC staff finds that this criterion remains satisfied, as the licensees analysis has demonstrated that removal of control rod H-08 does not impact the ability of the reactivity
control system to perform its function. Under normal operating conditions, including anticipated operational occurrences, acceptable fuel design limits were demonstrated to not be exceeded.
GDC 27 - Combined reactivity control systems capability The reactivity control systems shall be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system, of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained.
Callaway is provided with means of making and holding the core subcritical under any anticipated conditions and with appropriate margin for contingencies. Combined use of the rod cluster control system and the chemical shim control system permits the necessary shutdown margin to be maintained during long-term xenon decay and plant cooldown. The single highest worth control cluster is assumed to be stuck full out upon trip for this determination.
The NRC staff finds that this criterion remains satisfied with the removal of control rod H-08, as the licensees analysis has demonstrated that the ability of the reactivity control systems to reliably control reactivity changes and that adequate SDM is maintained when considering the highest stuck rod worth. The licensees evaluations of the removal of control rod H-08 during Cycle 28 demonstrate that SDM and safety analysis limits are met throughout the fuel cycle.
GDC 28 - Reactivity limits The reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition.
The appropriate reactivity insertion rate for the withdrawal of the RCCAs and the dilution of the boric acid are controlled by the TSs. The specification includes or references appropriate graphs that show the permissible mutual withdrawal limits and overlap of functions of the several RCCA banks as a function of power.
The NRC staff finds that this criterion remains satisfied, as the licensees analysis with removal of control rod H-08 demonstrates trip reactivity insertion rate, SDM, and the safety analysis limits remain met for the FSAR Chapter 15 accidents for the entire fuel cycle (Cycle 28).
GDC 29 -Protection against anticipated operational occurrences The protection and reactivity control systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences.
The licensee stated that the protection and reactivity control systems are designed to ensure an extremely high probability of fulfilling their intended functions. The design principles of diversity and redundancy, coupled with a rigorous quality assurance program and analyses, support this probability, as does operating experience in plants using the same basic design.
The NRC staff finds that this criterion remains satisfied, as the removal of control rod H-08 does not impact the ability of the reactivity control systems to perform their safety functions. The mechanical removal of the control rod drive shaft and RCCA do not have any mechanical impact on the function of the remaining 52 control rods. The remaining 52 control rods are also not impacted by the related electrical changes when control rod H-08 is removed. Therefore, the NRC staff finds that a high probability control rod insertion continues to exist under anticipated operational occurrences, even with the removal of control rod H-08 during Cycle 28.
3.7 Conclusion The licensee proposed to modify TS 4.2.2 to add a note stating that operation with 52 control rod assemblies (with no control rod assembly installed in core location H-08) is permitted during Cycle 28. The NRC staff has reviewed the results of the reload safety evaluation process for use in justifying the specific circumstances addressed in this LAR (e.g., Cycle 28 of Callaway, the location of the rod in the core, the rod worth, the moderator density coefficient and the impact on the accident analyses, etc.) The licensee does not intend for the removal of the control rod H-08 to be a permanent plant configuration. Additional plant design changes under different circumstances for different cycles may require additional NRC staff review.
The NRC staff concludes that the licensee's proposed use of 52 control rod assemblies in Callaway for Cycle 28, is acceptable because the design change is consistent with the current design basis and does not challenge the safety analyses detailed in Chapter 15 of the Callaway FSAR. The NRC staff concludes that the licensee used methods consistent with regulatory requirements and guidance identified in section 2.0 of this SE. Based on the above, the NRC staff finds the proposed use of 52 control rod assemblies, with no control rod in location H-08, continues to meet the requirements of 10 CFR 50.36 (c)(4), GDC 10, 11, 12, 23, 25, 26, 27, 28, and 29.
4.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION
The NRC proposed to find that the requested amendment involves no significant hazards consideration (NSHC) in its Federal Register (FR) notice of February 27, 2025 (90 FR 108320).
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 NSHC 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.
An 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.
The proposed license amendment would add a note to Callaway Unit 1 Technical Specification (TS) 4.2.2, Control Rod Assemblies, to permit Cycle 28 to contain 52 control rods, i.e., with no control rod in core location H-08. Currently, TS 4.2.2 requires 53 control rod assemblies in each reactor core.
This proposed license amendment would allow for the temporary removal of the control rod in core location H-08 during Cycle 28. Operation of Callaway Cycle 28 with the H-08 control rod removed will not involve a significant increase in the probability or consequences of an accident previously evaluated. Shutdown margin (SDM) is reduced by the absence of the H-08 control rod but remains bounded by the limits specified by the Core Operating Limits Report (COLR). Because the impacts on the cycle-specific nuclear design parameters are bounded by the conservative input values used in the Final Safety Analysis Report (FSAR) accident analyses, the current accident analyses remain bounding. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed changes will not alter or prevent the ability of the remaining 52 control rods from performing their intended functions to mitigate the consequences of an initiating event within the assumed acceptance limits.
The mechanical removal of the control rod drive shaft and Rod Cluster Control Assembly (RCCA) does not have any mechanical impact on the function of the remaining 52 control rods. The electrical removal from service of the H-08 control rod involves pulling fuses to remove control power to the respective stationary, lift, and movable coils. The remaining control rods are not impacted by this electrical change and will continue to meet their design function. The modification design change process ensures that the associated plant modifications involve only the H-08 control rod and do not affect other control rods.
The proposed changes do not alter any accident analysis assumptions discussed in the FSAR. Shutdown Margin (SDM) is reduced by the absence of the H-08 control rod but remains bounded by the limits specified by the COLR.
Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2.
Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
Operation of Callaway Cycle 28 with the H-08 control rod removed will not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed change involves no physical change beyond the removal of the H-08 control rod. The safety evaluations performed for Cycle 28 with the H-08 control rod removed validated that the impacts to the nuclear design parameters were within the bounds of those already assumed in the FSAR Chapter 15 accident analyses. The change in core bypass flow through the upper head region has been evaluated, and it has been determined that the increase is negligible and there is no impact to the safety analysis due to this negligible increase. The current accident analyses remain bounding. All plant equipment will continue to meet applicable design and safety requirements.
Therefore, it is concluded that this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3.
Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated.
The proposed change does not alter any FSAR design basis or safety limit and does not change any setpoint at which automatic actuations are initiated. The proposed change has been evaluated for effects on available shutdown margin, boron worth, trip reactivity as a function of time, and moderator temperature coefficient. The results of these evaluations show that adequate margin is maintained such that the proposed change would not cause a design basis or safety limit to be altered or exceeded.
Therefore, it is concluded that the proposed change does not involve a significant reduction in a margin of safety.
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.
5.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Missouri State official was notified of the proposed issuance of the amendment on March 14, 2025. The State official had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to the 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. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration published in the Federal Register on February 27, 2025 (90 FR 10832), and there has been no public comment on such finding. Further, the Commission has made a final determination that no significant hazards consideration is involved for the proposed amendment as discussed in section 4.0 of this SE. 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.
7.0 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: Robert Beaton, NRR Gokul Vasudevamurthy, NRR Yuken Wong, NRR Thomas Scarbrough, NRR Clint Ashley, NRR Date: April 10, 2025
ML25084A354 Concurrence via email OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA* NRR/DEX/EMIB/BC*
NRR/DSS/SNSB*
NAME MChawla PBlechman SBailey DMurdock DATE 3/14/2025 3/27/2025 3/26/2025 3/11/2025 OFFICE NRR/DSS/STSB*
OGC*
NRR/DORL/LPL4/BC* NRR/DORL/LPL4/PM*
NAME SMehta DRoth TNakanishi MChawla DATE 3/25/2025 4/3/2025 4/10/2025 4/10/2025