ML22153A176

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Enclosure 1 - Description and Assessment of the Proposed Change
ML22153A176
Person / Time
Site: Callaway Ameren icon.png
Issue date: 06/02/2022
From:
Ameren Missouri, Union Electric Co
To:
Office of Nuclear Reactor Regulation
Shared Package
ML22153A174 List:
References
ULNRC-06729, LDCN 22-0002
Download: ML22153A176 (38)
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Text

Enclosure 1 to ULNRC-06729 Page 1 of 38

ENCLOSURE 1

DESCRIPTION AND ASSESSMENT OF THE PROPOSED CHANGE

to ULNRC-06729 Page 2 of 38

ENCLOSURE 1

DESCRIPTION AND ASSESSMENT OF THE PROPOSED CHANGE

1.

SUMMARY

DESCRIPTION Page 2

2. DETAILED DESCRIPTION Page 5
3. TECHNICAL EVALUATION Page 10
4. REGULATORY EVALUATION Page 28

4.1 Applicable Regulatory Requirements/Criteria Page 28 4.2 Precedent Page 32 4.3 No Significant Hazards Consideration Determination Page 33 4.4 Conclusions Page 36

5. ENVIRONMENTAL EVALUATION Page 36
6. REFERENCES Page 36
7. ATTACHMENTS

1 Marked-up Existing Technical Specifications 2 Revised Clean Technical Specifications 3 Marked-up Technical Specification Bases 4 Framatome Affidavits for Withholding 5 Non-Proprietary Version LAR Inputs Document 6 Non-Proprietary Version RLBLOCA Summary Report

7. Non-Proprietary Version SBLOCA Summary Report
8. Non-Proprietary Version Non-LOCA Summary Report
9. Proprietary Version LAR Inputs Document 10 Proprietary Version RLBLOCA Summary Report 11 Proprietary Version SBLOCA Summary Report 12 Proprietary Version Non-LOCA Summary Report to ULNRC-06729 Page 3 of 38

DESCRIPTION AND ASSESSMENT OF THE PROPOSED CHANGE

1.0

SUMMARY

DESCRIPTION

Ameren Missouri (Union Electric Company) is proposing to amend Renewed Operating License NPF-30 for Callaway Plant Unit 1 (Callaway). The proposed amendment would revise the Technical Specifications to support the use of Framatome GAIA fuel with M51 as a fuel cladding material. Since the Framatome GAIA fuel will use M5 fuel rod cladding, a 10 CFR 50.46 and 10 CFR 50 Appendix K exemption request is included as part of this license amendment request (LAR). The exemption request is provided in Enclosure 2.

Ameren Missouri is implementing a plan that provides for the option of transitioning from the use of fuel manufactured by Westinghouse Electric Company (Westinghouse), as currently used in the Callaway reactor core, to the use of fuel manufactured by Framatome, Inc. Implementation of the fuel transition plan has already begun, as four Framatome lead fuel assemblies were situated in the Callaway core (in non-limiting locations) during operating cycle 25. Due to core design considerations, these four assemblies were removed during the current operating cycle (cycle 26). The transition will proceed to and through the point in time (mid-to late 2023) when a management decision will be made on whom to award the fuel contract for subsequent Callaway operation.

Ameren Missouri has contracted with Framatome to establish a Vendor Qualification Program (VQP) for the use of GAIA fuel at Callaway as well as Framatomes evaluation methodologies for application at Callaway. The GAIA fuel assembly mechanical design is generically approved by the NRC per Reference 36. A description of the Framatome GAIA fuel, including its design attributes, thermohydraulic and neutronic characteristics, and mechanical and structural compatibility, along with information related to transient and accident analysis performance, is provided in Attachment 9. The attachment also provides a description of the analysis methodologies employed to ensure an acceptable transition from the current Westinghouse supplied Performance+ fuel to Framatome's GAIA fuel during transition and full-core operating cycles.

The requested license amendment would modify the Technical Specifications, including the approved Callaway reload analysis methodologies listed therein, to temporarily accommodate both Westinghouse and Framatome fuel, and to support the implementation of Framatome methodologies, parameters and correlations. The ability to use either vendor's analysis methodologies is necessitated by two separate but related/sequential conditions. First, during operating cycle 27, a total of eight (8)

Framatome GAIA fuel assemblies will be loaded in the core. These eight assemblies will consist of four (4) Framatome GAIA lead fuel assemblies (LFAs) previously installed in the core during operating cycle 25 and four (4) additional Framatome GAIA assemblies acquired under the VQP.

These assemblies will be analyzed to operate in neutronic, thermohydraulic, mechanical and structural locations that are unrestricted. Due to core reload analysis considerations and limitations, no Framatome GAIA fuel will be loaded in the core during operating cycle 26.

1 M5 is a registered trademark of Framatome, Inc. to ULNRC-06729 Page 4 of 38

Under the terms of the current fuel supplier contract with Westinghouse, and because the vast majority of the co-resident fuel is supplied by Westinghouse, for operating cycle 27 Westinghouse will provide core reload analysis support for Callaway, compatible with the Technical Specification (TS)

Section 3.2, "Power Distribution Limits" (i.e., Limiting Condition for Operation (LCO) 3.2.1 through LCO 3.2.4), using the Westinghouse methodologies for Core Operating Limits Report (COLR) development currently in place. To accommodate the eight Framatome GAIA assemblies during operating cycle 27, changes to TS Section 2, "Safety Limits"; Section 4, "Design Features"; and Section 5, "Administrative Controls," are proposed along with a regulatory exemption to accommodate the Framatome GAIA M5 fuel clad material. To accommodate the core design for operating cycle 27, the proposed TS changes require approval by September 30, 2023.

The second condition necessitating TS changes and the regulatory exemption request involves the selection of the fuel vendor for future operating cycles, starting with operating cycle 28. Since initial plant start up in 1984, Westinghouse has provided the reactor fuel and supporting core reload analyses for Callaway. The current fuel supplier contract expires at the conclusion of operating cycle 27, and the selection of a new fuel and reload analysis vendor starting with operating cycle 28 is underway.

Therefore, per this license amendment request, changes are also proposed for TS Section 3.2, "Power Distribution Limits," (specifically, LCO 3.2.1 through LCO 3.2.4) and TS Section 5.6.5, "COLR," in order to accommodate either Westinghouse or Framatome as the selected reactor fuel and reload analysis vendor. (This requires proposing two alternative sets of Technical Specifications, as further explained in Section 2.0 of this enclosure.) To accommodate the potential acquisition of Framatome GAIA fuel and adoption of reload analysis methodologies, the TS changes described in this LAR are desired well in advance of operating cycle 28.

It is Ameren Missouri's intent to select only one fuel vendor, and Ameren Missouri does not intend to alternate between vendors for future fresh fuel batches or to use multiple vendors resulting in mixed fresh fuel batches. To address possible concerns with alternating between suppliers or mixing fresh fuel batches, Ameren Missouri proposes a regulatory commitment described in Enclosure 3, "Regulatory Commitments," to preclude this practice. Further, Ameren Missouri proposes an additional regulatory commitment, also described in Enclosure 3, in which after selection of the future fuel supplier and reload analysis vendor, Ameren Missouri will submit a license amendment request to remove the TS LCOs associated with the non-selected fuel vendor after that vendor's fuel is no longer present in the core. These regulatory commitments will be managed in accordance with NEI 99-04, "Guidelines for Managing NRC Commitment Changes," Revision 0.

The subject of this license amendment request was presented to the NRC staff during pre-submittal meetings held on May 20, 2021 (References 9, 18); October 14, 2021 (References 1, 2, 19); and February 28, 2022 (References 20, 21). Feedback, recommendations and discussion points from those meetings are reflected in this LAR.

This enclosure (Enclosure 1) contains twelve attachments. Regarding the first three attachments, provides the marked-up version of the current Technical Specifications which depicts the proposed changes. Attachment 2 provides the clean, re-typed version of the Technical Specifications assuming approval of the proposed changes. Attachment 3 provides marked-up TS to ULNRC-06729 Page 5 of 38

Bases pages to show the conforming changes for information only. TS Bases changes will be managed in accordance with TS 5.5.14, "Technical Specification (TS) Bases Control Program."

In support of Ameren Missouri's intent to license the Framatome GAIA fuel design and Framatome's evaluation methodologies, Framatome developed a series of technical reports that provide details supporting this license amendment request, including summaries of the evaluations performed for transients and accidents described in the Callaway Final Safety Analysis Report (FSAR). These evaluations demonstrate continued compliance with license basis and regulatory acceptance criteria.

The technical reports are contained in Attachments 9 through 12.

It should be noted that the technical reports provided as Attachments 9 through 12 contain information that is proprietary to Framatome. Attachment 4 contains four (4) affidavits signed by Framatome that set forth the basis on which the proprietary information in Attachments 9 through 12 may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in 10 CFR 2.390(b)(4). Non-proprietary (public) versions of Attachments 9 through 12 are provided as Attachments 5 through 8, respectively, and are described below.

provides a non-proprietary version of technical report ANP-3947P, "Callaway Unit 1 License Amendment Request Inputs for Use of Framatome Fuel," Revision 0. Attachment 9 provides the proprietary version this report. As noted above, the proprietary version must be withheld from public disclosure.

provides a non-proprietary version of technical report ANP-3944P, "Callaway Realistic Large Break LOCA Analysis with GAIA Fuel Design," Revision 0. Attachment 10 provides the proprietary version this report. As noted above, the proprietary version must be withheld from public disclosure.

provides a non-proprietary version of technical report ANP-3943P, "Callaway Small Break LOCA Analysis with GAIA Fuel Design," Revision 0. Attachment 11 provides the proprietary version this report. As noted above, the proprietary version must be withheld from public disclosure.

provides a non-proprietary version of technical report ANP-3969P, "Callaway Non-LOCA Summary Report," Revision 0. Attachment 12 provides a proprietary version of this report. As noted above, the proprietary version must be withheld from public disclosure.

2.0 DETAILED DESCRIPTION

2.1 Technical Specification Changes

Changes to the Callaway Plant Technical Specifications are required to allow the partial use of the Framatome GAIA fuel in operating cycle 27 and transition to batch loading of Framatome GAIA fuel beginning in operating cycle 28. As noted previously, Attachment 1 to this Enclosure provides the marked-up version of the current Technical Specifications which depicts the proposed TS changes. to ULNRC-06729 Page 6 of 38 provides the clean re-typed version of the TS assuming approval of the proposed changes.

Specifically, the proposed TS changes are driven by the need to accommodate a limited number (8) of Framatome GAIA fuel assemblies during operating cycle 27, which affects TS 2.1.1, TS 4.2.1 and TS 5.6.5, during which time the plant will continue to operate in accordance with the TS Section 3.2, "Power Distribution Limits," (i.e., TS 3.2.1 through TS 3.2.4) as supported by the current Westinghouse methods for COLR development. A second portion of TS changes is proposed to accommodate the future fuel supplier, yet to be chosen, by providing two/alternate sets of "Power Distribution Limit" Technical Specifications based on either Westinghouse or Framatome-supported COLR limits. As discussed in the Section 3.0, Technical Evaluation, only one vendor's methods may be used to establish the COLR limits for a given operating cycle. A summary of the changes is provided below:

TS 2.1.1 - Reactor Core Safety Limits

Revise TS 2.1.1.1 and TS 2.1.1.2 - Clarify that the existing Technical Specifications specify a Westinghouse-supplied fuel limiting Departure from Nucleate Boiling Ratio (DNBR), DNB correlation, and peak fuel centerline temperature. Other than the designation as Westinghouse fuel limits, the current Safety Limits and their supporting methods are unchanged.

Add new TS 2.1.1.3 and TS 2.1.1.4 - Add corresponding Framatome GAIA fuel Safety Limits for DNBR and peak fuel centerline temperature, along with the Framatome-specific DNB correlation.

TS 3.2.1 - Heat Flux Hot Channel Factor (FQ(Z)) (FQ Methodology)

Revise TS 3.2.1 - Add clarifying annotation that the current LCO 3.2.1 applies only when Westinghouse core design analysis methods are used to derive the COLR limits. Designate the Westinghouse-applicable TS as LCO 3.2.1A. Correspondingly, designate the associated Surveillance Requirements (SRs) as SR 3.2.1A.1 and SR 3.2.1A.2. The requirements of the LCO and SRs are unchanged other than the adoption of the 'A' designation.

Add new TS 3.2.1B - Add corresponding Framatome fuel Heat Flux Hot Channel Factor (FQ(Z)) LCO 3.2.1B that applies only when the Framatome core design analysis methods are used to derive the COLR limits. The adoption of this LCO includes the LCO requirement, Actions and SRs specific to the Framatome methodologies.

TS 3.2.2 - Nuclear Enthalpy Rise Hot Channel Factor (F NH)

Revise TS 3.2.2 - Add clarifying annotation that the current TS LCO 3.2.2 applies only when Westinghouse core design analysis methods are used to derive the COLR limits. Designate the Westinghouse-applicable TS as LCO 3.2.2A. Correspondingly, designate the associated SR as SR 3.2.2.A.1. The requirements of the LCO and SR are unchanged other than the adoption of the 'A' designation. to ULNRC-06729 Page 7 of 38

Add new TS 3.2.2B - Add corresponding Framatome fuel Nuclear Enthalpy Rise Hot Channel Factor (FNH) LCO that applies only when the Framatome core design analysis methods are used to derive the COLR limits. The adoption of this LCO includes the LCO requirement, Actions and SRs specific to the Framatome methodologies.

TS 3.2.3 - Axial Flux Difference (AFD) (Relaxed Axial Offset Control Methodology, RAOC)

Revise TS 3.2.3 - Add clarifying annotation that the current TS LCO 3.2.3 applies only when Westinghouse core design analysis methods are used to derive the COLR limits. Designate the Westinghouse-applicable TS as LCO 3.2.3A. Correspondingly, designate the associated SR as SR 3.2.3A.1. The requirements of the LCO and SR are unchanged other than the adoption of the 'A' designation.

Add new TS 3.2.3B - Add corresponding Framatome fuel Axial Flux Difference (AFD) LCO that applies only when the Framatome core design analysis methods are used to derive the COLR limits. The adoption of this LCO includes the LCO requirement, Actions and SRs specific to the Framatome methodologies.

TS 3.2.4 - Quadrant Power Tilt Ratio (QPTR)

Revise TS 3.2.4 - Revise Required Actions A.3 and A.6 to specify that the required SRs align with those of the fuel vendor's methodologies used to establish the core design and associated COLR limits.

TS 4.2.1 - Fuel Assemblies

Revise TS 4.2.1 to include reference to the M5 fuel cladding material (and correct the spelling of Zircalloy2). This change is necessary to support the GAIA fuel assembly use of M5 fuel cladding material. The corresponding regulatory exemption request is provided in Enclosure 2.

TS 5.6.5 - Core Operating Limits Report (COLR)

Revise the listing of supported LCOs in TS 5.6.5.a.4, a.5 and a.6 to provide reference to both the Westinghouse and Framatome applicable Power Distribution Limit LCO and revise the listing of analytical methods in TS 5.6.5.b to include the applicable analytical methods that support Framatome-developed core operating limits.

These changes are supported by the evaluations presented in Attachments 9 through 12. Attachment 9 provides a detailed description of the GAIA fuel, supporting bases for these proposed TS changes, and a summary of the analyses performed to support the acceptability of the use of GAIA fuel at Callaway.

Attachments 10 through 12 provide the summary results of the analyses performed to support the

2 Callaway TS 4.2.1 misspells zircaloy (as given in 10 CFR 50.46) as Zircalloy. to ULNRC-06729 Page 8 of 38

FSAR Chapter 15 Accident Analyses and continued compliance with the NRC regulatory requirements.

2.2 Additional Considerations

Spent Fuel Pool Criticality Analysis

To accommodate the Framatome GAIA fuel in the spent fuel pool and during fuel handling, the spent fuel criticality safety analyses are being reperformed and updated. Therefore, independent of this license amendment request, a separate LAR will be developed and submitted for the spent fuel pool criticality reanalysis. Depending on the outcome of that reanalysis, changes to TS 3.7.17, "Spent Fuel Assembly Storage," and TS 4.3, "Fuel Storage," may be needed. The analysis activities for that LAR are underway and the LAR is expected to be submitted during the summer of 2022.

GL 2004-02 / GSI-191

Regarding Ameren Missouri's resolution of Generic Letter (GL) 2004-02 (i.e., Generic Safety Issue (GSI) 191 involving the potential for containment sump screen clogging and downstream effects as a result of a loss of coolant accident debris generation and transport), Ameren Missouri is in the process of closing out GL 2004-02 for the resident (Westinghouse) fuel (References 3, 4, 5, 6, 7, and 8) using the NRC staff guidance provided in Reference 30. Ameren Missouri determined that Callaway is a Category 4 plant that required a plant specific analysis.

Regarding operation with eight GAIA fuel assemblies in operating cycle 27, a qualitative assessment confirms the acceptability of the core against the NRC staff guidance provided in Reference 30. A restriction will be applied that precludes the GAIA assemblies from residing in a face-to-face configuration in operating cycle 27.

Regarding batch quantity loading of GAIA fuel starting in operating cycle 28, the use of GAIA fuel only affects the assessment of the core fiber limit. That is, a change in fuel design will have no effect on the sump switchover time, rated thermal power, or alternate flow path resistance. Therefore, it need only be shown that GAIA fuel can meet the core fiber limit requirements outlined in Reference 30.

The Callaway-specific analysis described in the GL 2004-02 response demonstrated that the total amount of fiber reaching the reactor vessel was 89 grams per fuel assembly (g/FA). While this mass of fiber exceeds the allowed mass of fiber at the core inlet, conservatisms are available that can be credited as discussed in Reference 30 and Reference 32, Section 4.5.1.1. Specifically, the fibrous debris that reaches the core inlet will not build up in a uniform fashion due to variations in the flow velocities expected at the core inlet. This non-uniform buildup of debris will allow more debris to be tolerated at the core inlet than the amount for any specific fuel type. At some point, sufficient debris may build up at the core inlet to divert flow through the alternate flow path. Therefore, as long as the total amount of fiber reaching the Reactor Cooling System (RCS) is below the maximum value allowed by WCAP-17788P (Reference 31, Section 9.1.1), the requirements of GL 2004-02 are met.

This is true regardless of fuel design. Since the total amount of fiber that reaches the RCS is not to ULNRC-06729 Page 9 of 38

dependent on use of GAIA fuel, Callaway will continue to meet the requirements of GL 2004-02 following transition to Framatome's GAIA fuel.

Control Rod Ejection Event Analysis

With regard to the ejected rod analysis described in FSAR 15.4.8, "Spectrum of Rod Cluster Control Assembly Ejection Accidents," this postulated event is being evaluated using the NRC-approved AREA methodology using the GALILEO fuel rod thermal mechanical methodology as described in Section 6.1.1 of Attachment 9 against the acceptance criteria given in Reference 29. The results of the analysis will be available for NRC review.

Alternative Source Term (AST)

Ameren Missouri currently has a license amendment request under review by the NRC staff to adopt the provisions described in 10 CFR 50.67, "Alternative Source Term." (References 33 and 34) The proposed amendment would revise the technical specifications and authorize changes to the final safety analysis report to support application of the regulations in 10 CFR 50.67, as described in Regulatory Guide 1.183. (Reference 35) Specifically, the amendment would revise TS 3.7.10, "Control Room Emergency Ventilation System (CREVS)," TS 5.5.11, "Ventilation Filter Testing Program," and TS 5.5.17, "Control Room Envelope Habitability Program." Following approval of the AST LAR, an evaluation, including a 10 CFR 50.59 review, will be performed using the GAIA fuel characteristics against the approved analysis methods to determine if an additional license amendment request is required. Framatome has prepared an analysis of the postulated accident consequences using the AST analysis methods proposed in the Ameren Missouri submittal. The results of that analysis will be made available for review. (Reference 13)

Dose Consequence Evaluation Against the Current Analysis of Record

For operating cycle 27, in which eight GAIA fuel assemblies will reside in the core, the core reload analysis and the core design process will ensure no unacceptable change in radiological consequences results from a postulated anticipated operational occurrence or accident. For subsequent operating cycles with batch loaded quantities of GAIA fuel, assuming that Framatome is selected as the fuel vendor, an evaluation consistent with the previously mentioned AST LAR, which is currently under review with approval expected well before transition to batch loading of GAIA fuel, will be performed consistent with the approved methods of analysis and associated acceptance criteria to determine the acceptability of the core design. This approach is based on fuel transition precedent described in Reference 37.

2.3 Technical Specification Bases Changes

Changes to the TS Bases will be needed to ensure consistency with the proposed TS changes. The marked-up pages for the TS Bases changes described within this LAR are provided in Attachment 3 for information only. Final TS Bases changes will be processed under the program for updates per TS 5.5.14, "Technical Specifications Bases Control Program," at the time this amendment is to ULNRC-06729 Page 10 of 38

implemented. The TS Bases changes will describe the method used to designate the TS 3.2, "Power Distribution Limits," LCOs associated with the respective vendor's methods used to derive the COLR limits. In addition, the TS Bases changes will include the addition of the Bases associated with the proposed "B" series of LCOs associated with the Framatome methods of analysis.

2.4 Core Operating Limits Report (COLR) Changes

The COLR will be maintained in accordance with the administrative controls governing core reload design control and coordination, as specified in TS 5.6.5, "Core Operating Limits Report (COLR)."

For operating cycle 27, the Westinghouse methods will continue to be utilized, with confirmatory analyses by Framatome, consistent with the methods specified in TS 5.6.5. For this reason, no significant changes are necessary for the COLR to support operating cycle 27 and no mark-up is provided.

In support of operating cycle 28 and assuming both the approval of this amendment and that Ameren Missouri selects Framatome as the fuel supplier, the COLR will be updated using the normal reload administrative controls and will reflect limits derived using the Framatome methods being added to TS 5.6.5.b. Section 4, "Power Distribution Control," in Attachment 9 provides a description of a proposed modification of the previously approved power distribution analysis methodology and includes examples of the tables and figures that would be included in the COLR when developed using the Framatome methods.

3.0 TECHNICAL EVALUATION

The following paragraphs provide the proposed TS changes and their justification. Attachment 1 to this Enclosure provides the marked-up version of the current TS which depicts the proposed changes discussed below. Attachment 2 provides the clean re-typed version of the TS following approval of these proposed changes.

TS 2.1.1 - Reactor Core Safety Limits

As discussed in the TS Bases, General Design Criteria (GDC) 10 requires that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs). This is accomplished by having a departure from nucleate boiling (DNB) design basis, which requires that the minimum departure from nucleate boiling ratio (DNBR) of the limiting rod during Condition I and II events is greater than or equal to the DNBR correlation limits. The restrictions of the Safety Limits (SLs) prevent overheating of the fuel and cladding, as well as possible cladding perforation, which would result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate (LHR) below the level at which fuel centerline melting occurs. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime, where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of to ULNRC-06729 Page 11 of 38

the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant. The combination of operation within a defined set of power distribution limits, proper operation of the Reactor Trip System, and proper operation of safety valves prevents violation of the reactor core SLs.

The current TS Bases and the Final Safety Analysis Report (FSAR), as updated, provide the description regarding how these limits are derived and met during operation.

The current Callaway SLs are supported by Westinghouse analysis methods as described in the TS Bases, FSAR and documentation supporting the current COLR methodologies. The current SLs are:

TS 2.1.1 Reactor Core SLs

In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the COLR, and the following SLs shall not be exceeded:

TS 2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained 1.17 for the WRB-2 DNB correlation.

TS 2.1.1.2 The peak fuel centerline temperature shall be maintained < 5080°F, decreasing by 58°F per 10,000 MWd/MTU of burnup.

This LAR will revise TS 2.1.1.1 and TS 2.1.1.2 to clarify that the existing TS specify a Westinghouse-supplied fuel limiting Departure from Nucleate Boiling Ratio (DNBR) and peak fuel centerline temperature. Other than the designation as Westinghouse fuel limits, the current Safety Limits are unchanged. This LAR will add a new TS 2.1.1.3 and TS 2.1.1.4 which provide the corresponding Framatome GAIA fuel Safety Limits for DNBR and peak fuel centerline temperature, along with the applicable DNB correlation. The revised TS are given below. Additions are shown with underlined text.

TS 2.1.1.1 For Westinghouse fuel, the departure from nucleate boiling ratio (DNBR) shall be maintained 1.17 for the WRB-2 DNB correlation.

TS 2.1.1.2 For Westinghouse fuel, the peak fuel centerline temperature shall be maintained < 5080°F, decreasing by 58°F per 10,000 MWd/MTU of burnup.

TS 2.1.1.3 For Framatome fuel, the DNBR shall be maintained 1.12 for the ORFEO-GAIA DNB correlation.

TS 2.1.1.4 For Framatome fuel, the peak fuel centerline temperature shall be maintained < 4901°F, decreasing linearly by 13.7°F per 10,000 MWd/MTU of burnup.

to ULNRC-06729 Page 12 of 38

Section 3 and 4 of Attachment 9 provides the neutronics design characteristics of the reactor and Section 5 of Attachment 9 provides the thermal-hydraulic design of the reactor that ensures the core can meet steady-state and transient performance requirements without violating the core safety limits.

Supporting technical analyses are provided in Attachments 10 through 12. The addition of these Framatome fuel-specific safety limits ensures continued compliance with the requirements of 10 CFR 50.36.

The Callaway TS Bases will be updated to reflect the addition of the DNBR and the peak fuel centerline temperature safety limits for the GAIA fuel.

Power Distribution Limit (PDL) TS Changes

The following sections describe the changes made to the PDL TS 3.2.1 through TS 3.2.4. These changes are necessary to preserve the use of Westinghouse methodologies for COLR development during operating cycle 27 and subsequent operating cycles if Westinghouse is selected as the fuel vendor, and to accommodate transition to Framatome fuel beginning in operating cycle 28 if selected as the fuel vendor. The regulatory requirements for the PDL TS are: 10 CFR 50.36(c)(2)(ii),

10 CFR 50.36(c)(2)(ii)B Criterion 2, and 10 CFR 50.36(c)(3) Surveillance requirements (Reference 14).

10 CFR 50.36(c)(2)(ii) states, A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria.

10 CFR 50.36(c)(2)(ii)(B) Criterion 2 states, A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

10 CFR 50.36(c)(3) Surveillance requirements states, Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits and that limiting conditions for operation will be met."

TS 3.2.1 - Heat Flux Hot Channel Factor (FQ(Z)) (FQ Methodology)

As described in the TS Bases, the purpose of the limits on the values of FQ(Z) is to limit the local (i.e.,

pellet) peak power density. The value of FQ(Z) varies along the axial height (Z) of the core. FQ(Z) is defined as the maximum local fuel rod linear power density divided by the average fuel rod linear power density, assuming nominal fuel pellet and fuel rod dimensions. Therefore, FQ(Z) is a measure of the peak fuel pellet power within the reactor core. The limits on FQ(Z) work in combination with other power distribution limits and with limitations on control rod alignment and insertion. FQ(Z) varies with fuel loading patterns, bank insertion, fuel burnup, and changes in axial power distribution.

FQ(Z) is not directly measurable but is inferred from a power distribution measurement obtained with either the movable incore detector system or from an operable power distribution monitoring system. to ULNRC-06729 Page 13 of 38

The results of the three-dimensional power distribution measurement are analyzed to derive a measured value for FQ(Z). These measurements are generally taken with the core at or near equilibrium conditions.

As described further below, and as proposed per this LAR, alternative versions of TS 3.2.1, i.e.,

TS 3.2.1A and TS 3.2.1B, would be specified in the Technical Specifications such that either TS 3.2.1A or TS 3.2.1B would apply, depending on which fuel vendor is selected to supply the fuel starting with operating cycle 28 and, therefore, which vendor's analysis methods govern development of the COLR limits for the operating cycle.

TS 3.2.1A

The requested license amendment will revise TS 3.2.1 to add an "A" designation and annotate that the current LCO 3.2.1 applies only when Westinghouse core design analysis methods are used to derive the COLR limits. The changes include:

  • The APPLICABILITY is modified to include the following Note:

NOTE-----------------------------------

This LCO is not applicable when Framatome COLR methods govern COLR development

  • Required Action A.4 is revised to specify that SR 3.2.1A.1 is to be performed
  • Item b. in the NOTE modifying SR 3.2.1.2 is revised to specify the performance of SR 3.2.1A.2

These changes do not alter the LCO requirement, ACTIONS, or SRs. The APPLICABILITY is modified by a Note that would allow transition to the Framatome methods for COLR development beginning in operating cycle 28 should Framatome be selected as the fuel vendor. The NOTE modifying the APPLICABILITY, in combination with the NOTE modifying the corresponding Framatome specification (e.g., TS 3.2.1B), create a mutually exclusive arrangement where only one fuel vendor's methods may be applicable at a time. The addition of the "A" designation is effectively an editorial change that preserves compliance with 10 CFR 50.36 requirements to establish limiting conditions for operation, establishment of actions should the limits not be met, and establishment of surveillance requirements that demonstrate compliance with the limits.

The Callaway TS Bases for LCO 3.2.1 will be updated to reflect the addition of the "A" designation as described above. An explanation for the Note modifying the APPLICABILITY statement will also be to ULNRC-06729 Page 14 of 38

added to the TS Bases to ensure that it is understood that this LCO applies when the Westinghouse-methods are being used to derive the COLR limits.

TS 3.2.1B

The requested license amendment will add a new specification TS 3.2.1B to denote the Heat Flux Hot Channel Factor (FQ(Z)) limits that apply when Framatome core design analysis methods are used to derive the COLR limits. The adoption of TS 3.2.1B includes the LCO requirement, APPLICABILITY, ACTIONS and SRs specific to the Framatome methodologies. Refer to Attachments 1 and 2 for the new specification. The specification is drafted in a manner consistent with Callaway's Technical Specifications and TSTF-GG-05-01, "Writer's Guide for Plant Specific Improved Technical Specifications," along with NUREG-1431, "Standard Technical Specifications -

Westinghouse Plants," Volume 1, Revision 5, as a reference.

This new specification is necessary to accommodate the development of the COLR using Framatome methods of analysis, as applicable, beginning in operating cycle 28 should Framatome be selected as the fuel supplier. The Power Distribution Control (PDC) methodology is described in Section 4 of using the methodology to be added to TS 5.6.5.b. Section 4.2.6 of Attachment 9 provides a correlation between the proposed Technical Specifications and their supporting basis.

As described in the draft TS Bases (Attachment 3), TS 3.2.1B will establish an LCO requiring that the Heat Flux Hot Channel Factor (FQ(Z)), as approximated by FCQ(Z) and FVQ(Z), be within the limits specified in the COLR when the limits are derived using Framatome methodologies. FCQ(Z) is the measured value of FQ(Z) multiplied by factors that account for fuel manufacturing tolerance and measurement uncertainty. During equilibrium, steady-state conditions, FCQ(Z) closely correlates with FQ(Z). The values for fuel manufacturing and measurement uncertainty are provided in the COLR.

FVQ(Z) is a limit which accounts for power distribution transients that could potentially occur during normal operation. FVQ(Z) is the product of FCQ(Z) and V(Z) where V(Z) is a cycle-specific, axially dependent function that accounts for power distribution changes expected during normal operation.

These correlations are demonstrated in Section 4 of Attachment 9.

This LCO ensures that the peak power generation rate is limited to values shown to not violate the fuel design criteria during anticipated operational occurrences (AOOs) and accident conditions. This LCO is Applicable in MODE 1 which represents the conditions when the reactor's Thermal Power level is high enough to challenge the fuel integrity. The Applicability is modified by a Note that makes clear that this LCO is not applicable when the COLR limits are derived using the Westinghouse methods identified in TS 5.6.5.b.

TS 3.2.1B will also establish the Actions necessary when either FCQ(Z) or FVQ(Z) are identified as not being within their limits. When FCQ(Z) is not within limits, Required Actions A.1 through A.4 conservatively require a reduction in Thermal Power, a reduction of the Reactor Protection System (RPS) Power Range Neutron Flux-High and Overpower T trip setpoints, and performance of SRs, i.e., SR 3.2.1B.1 and SR 3.2.1B.2, that confirm core power distribution limits are met prior to increasing Thermal Power following the reduction in Thermal Power dictated by Required Action A.1.

The amount of Thermal Power reduction and Power Range Neutron Flux - High trip setpoint to ULNRC-06729 Page 15 of 38

reduction is a function of the amount by which FCQ(Z) exceeds its limit. The reduction in RPS trip setpoints provides an automatic protective action that ensures the capability of the RPS to continue to provide compliance with the fuel design criteria in the event of an AOO. The Completion Times are comparable to those established in the corresponding Westinghouse LCO (i.e., TS 3.2.1A).

When FVQ(Z) is not within limits, Required Actions B.1.1 and B.1.2 and B.2.1 through B.2.4 provide a set of alternative actions. Required Action B.1.1 requires the Axial Flux Difference (AFD) limits based on the current target AFD band and the V(Z) function be reduced by an amount greater than or equal to an amount specified in the COLR. The target AFD band represents the allowed AFD variation about the target AFD. V(Z) is a cycle-specific, axially dependent function that accounts for power distribution changes expected during normal operation. These actions effectively tighten the allowed core power distribution envelope because of the potential for FCQ(Z) to become excessively high if a normal operational transient were to occur. Refer to Section 4.2.4 in Attachment 9 for a complete description. Required Action B.1.2 requires performance of SR 3.2.1B.1 and SR 3.2.1B.2 to verify that FCQ(Z) and FVQ(Z) are within their limits prior to lessening the restrictions on the AFD band imposed by Required Action B.1.1. The Completion Time for Required Action B.1.1 is reasonable given the time necessary to perform and verify the calculations necessary to reduce the AFD limits.

Required Actions B.2.1 through B.2.4 conservatively require a reduction in Thermal Power, reduction of the Reactor Protection System (RPS) Power Range Neutron Flux-High and Overpower T trip setpoints, and performance of SRs, i.e., SR 3.2.1B.1 and SR 3.2.1B.2, that confirm core power distribution limits are met. The amount of Thermal Power reduction and Power Range Neutron Flux -

High trip setpoint reduction is a function of the amount by which FVQ(Z) exceeds its limit. The reduction in RPS trip setpoints provides an automatic protective action that ensures the capability of the RPS to continue to provide compliance with the fuel design criteria in the event of an AOO. The Completion Times are comparable to those established in the corresponding Westinghouse LCO (i.e.,

TS 3.2.1A).

Condition C addresses those circumstances where the Required Actions and associated Completion Times for Conditions A or B are not met. The 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Completion Time is consistent with the corresponding Westinghouse LCO and NUREG-1431 time-frame to be in Mode 2 when operating in Mode 1.

Two SRs, i.e., SR 3.2.1B.1 and SR 3.2.1B.2, are proposed that require verification of FCQ(Z) and FVQ(Z) once following each refueling outage prior to exceeding 75% Rated Thermal Power (RTP),

once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after achieving equilibrium conditions after exceeding, by 10% RTP, the Thermal Power at which FCQ(Z) or FVQ(Z) was last verified, and in accordance with the Surveillance Frequency Control Program (SFCP). Initially, the SFCP will stipulate a Frequency of 31 Effective Full Power Days (EFPD). This Frequency is reasonable given the slow changes in core flux profile during cycle operation (also referred to as core burnup or fuel exposure).

The adoption of the limits on the Heat Flux Hot Channel Factor (FQ(Z)) in TS 3.2.1B preserves compliance with 10 CFR 50.36 requirements to establish limiting conditions for operation, establishment of actions should the limits not be met, and establishment of surveillance requirements to ULNRC-06729 Page 16 of 38

that demonstrate compliance with the limits, when the Framatome methods of analysis are used to derive the COLR limits.

As shown in Attachment 3, the Callaway TS Bases will be updated to add the Framatome-specific description for LCO 3.2.1B consistent with TSTF-GG-05-01 and NUREG-1431. The Bases will make clear that this LCO applies when the Framatome-methods are being used to derive the COLR limits.

TS 3.2.2 - Nuclear Enthalpy Rise Hot Channel Factor (FNH)

As described in the TS Bases, the purpose of this LCO is to establish limits on the power density at any point in the core so that the fuel design criteria are not exceeded and the accident analysis assumptions remain valid. The design limits on local (pellet) and integrated fuel rod peak power density are expressed in terms of hot channel factors. Control of the core power distribution with respect to these factors ensures that local conditions in the fuel rods and coolant channels do not challenge core integrity at any location during either normal operation or a postulated accident analyzed in the safety analyses. FNH is defined as the ratio of the integral of the linear power along the fuel rod with the highest integrated power to the average integrated fuel rod power. Therefore, FNH is a measure of the maximum total power produced in a fuel rod. FNH is sensitive to fuel loading patterns, bank insertion, and fuel burnup. FNH is not directly measurable but is inferred from a power distribution measurement obtained with either the movable incore detector system or from an operable power distribution monitoring system. The limits on FNH work in combination with other power distribution limits and with limitations on control rod alignment and insertion.

As described further below, and as proposed per this LAR, alternative versions of TS 3.2.2, i.e.,

TS 3.2.2A and TS 3.2.2B, would be specified in the Technical Specifications such that either TS 3.2.2A or TS 3.2.2B would apply, depending on which fuel vendor is selected to supply the fuel starting with operating cycle 28 and, therefore, which vendor's analysis methods govern development of the COLR limits for the operating cycle.

TS 3.2.2A

The requested license amendment will revise TS 3.2.2 to add an "A" designation and annotate that the current LCO 3.2.2 applies only when Westinghouse core design analysis methods are used to derive the COLR limits. The changes include:

  • The APPLICABILITY is modified to include the following Note:

to ULNRC-06729 Page 17 of 38

NOTE-----------------------------------

This LCO is not applicable when Framatome COLR methods govern COLR development

  • Required Action A.2 is revised to specify that SR 3.2.2A.1 is to be performed
  • Required Action A.3 is revised to specify that SR 3.2.2A.1 is to be performed

These changes do not alter the requirements of the LCO, ACTIONS, or SRs. The APPLICABILITY is modified by a Note that would allow transition to the Framatome methods for COLR development beginning in operating cycle 28 should Framatome be selected as the fuel vendor. The NOTE modifying the APPLICABILITY, in combination with the NOTE modifying the corresponding Framatome specification (e.g., TS 3.2.2B), create a mutually exclusive arrangement where only one fuel vendor's methods may be applicable at a time. The addition of the "A" designation is effectively an editorial change that preserves compliance with 10 CFR 50.36 requirements to establish limiting conditions for operation, establishment of actions should the limits not be met, and establishment of surveillance requirements that demonstrate compliance with the limits.

The Callaway TS Bases for LCO 3.2.2 will be updated to reflect the addition of the "A" designator as described above. An explanation for the Note modifying the APPLICABILITY statement will also be added to the TS Bases to ensure that it is understood that this LCO applies when the Westinghouse-methods are being used to derive the COLR limits.

TS 3.2.2B

The requested license amendment will add a new specification TS 3.2.2B to denote the Nuclear Enthalpy Rise Hot Channel Factor (FNH) limits that apply when Framatome core design analysis methods are used to derive the COLR limits. The adoption of TS 3.2.2B includes the LCO requirement, APPLICABILITY, ACTIONS and SRs specific to the Framatome methodologies. Refer to Attachments 1 and 2 for the new specification. The specification is drafted in a manner consistent with Callaway's Technical Specifications and utilized TSTF-GG-05-01, along with NUREG-1431, Volume 1, Revision 5, as a reference.

This new specification is necessary to accommodate the development of the COLR using Framatome methods of analysis, as applicable, beginning in operating cycle 28 should Framatome be selected as the fuel supplier. The Power Distribution Control (PDC) methodology is described in Section 4 of using methodology to be added to TS 5.6.5.b.

As described in the draft TS Bases (Attachment 3), TS 3.2.2B will establish an LCO requiring that the Nuclear Enthalpy Rise Hot Channel Factor (FNH(Z)) be within the limits specified in the COLR when the limits are derived using Framatome methodologies. This LCO ensures that the power density at any point in the core is limited so that the fuel design criteria are not exceeded and the accident analysis assumptions remain valid. This LCO will be Applicable in MODE 1 which represents the conditions when the Thermal Power level is high enough to challenge the fuel integrity. The to ULNRC-06729 Page 18 of 38

Applicability is modified by a Note that makes clear that this LCO is not applicable when the COLR limits are derived using the Westinghouse methods identified in TS 5.6.5.b.

TS 3.2.2B will establish the Actions necessary when FNH(Z) is identified as not being within its limit.

Condition A addresses the situation where FNH(Z) is not being within its limit. Condition A is modified by a Note that stipulates that Required Actions A.2 and A.3 must be completed anytime that Condition A is entered. Required Action A.1.1 provides an opportunity to restore FNH to within its limit with a Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Alternatively, Required Actions A.1.2.1 and A.1.2.2 may be performed to reduce Thermal Power to less than or equal to 50% RTP and reduce the Power Range Neutron Flux-High trip setpoint to less than or equal to 55% RTP. Required Actions A.1.2.1 and A.1.2.2 have Completion Times of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, respectively, which experience has shown to be reasonable for completion of the power reduction and an acceptable time frame for reducing the Power Range Flux-High trip setpoints given the complexity of the evolution and the small probability of an operational event that relies on the high flux trip for protection. The Completion Times are comparable to those established in the corresponding Westinghouse LCO (i.e.,

TS 3.2.2A). The reduction in RPS trip setpoints provides an automatic protective action that ensures the capability of the RPS to continue to provide compliance with the fuel design criteria in the event of an AOO.

In addition to the completion of Required Actions A.1.1 or A.1.2.1 and A.1.2.2, Required Actions A.2 and A.3 require the performance of SR 3.2.2B.1 which verifies that FNH. is within the limits specified in the COLR. Required Action A.2 has a Completion Time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> which ensures that FNH is periodically verified when complying with the Required Actions or anytime that Condition A is entered. Required Action A.3 has three Completion Times that are associated with increasing Thermal Power levels when increasing reactor power to RTP. The first two Completion Times require SR 3.2.2B.1 be completed prior to exceeding 50% RTP and 75% RTP, respectively. These Completion Times provide confirmation that FNH is within limits as Thermal Power is increased following a situation that required entry into Condition A. Similarly, the third Completion Time requires confirmation that FNH is within its limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of Thermal Power exceeding 95% RTP. Required Action A.3 is modified by a Note that says that Thermal Power does not have to be reduced to comply with this Required Action. This Note ensures that situations where FNH was identified as not within limits, i.e., Condition A was entered, and FNH was subsequently restored to within its limit, i.e., Required Action A.1.1 was successful, and either a minimal, or no, Thermal Power reduction was performed, that FNH is verified to be within its limits.

Condition B addresses those circumstances where the Required Actions and associated Completion Times for Condition A are not met. The 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Completion Time is consistent with the corresponding Westinghouse LCO and NUREG-1431 time-frame to be in Mode 2 when operating in Mode 1.

One SR, i.e., SR 3.2.2B.1, is proposed that requires verification that FNH is within the limit specified in the COLR once following each refueling outage prior to exceeding 75% RTP, and in accordance with the SFCP. Initially, the SFCP will stipulate a Frequency of 31 EFPD. These Frequencies ensure that FNH is verified prior to attaining Thermal Power levels that are likely to challenge the limits following a refueling outage and are then periodically verified as the core power distribution gradually changes during power operation. This SR is modified by a Note that imposes two options for a to ULNRC-06729 Page 19 of 38

required action when an increasing trend in the measured value of FNH is identified although remaining within limits. First, FVQ(Z) is increased by a value specified in the COLR or, second, the Frequency for performance of SR 3.2.1B.2 and SR 3.2.3B.2 is increased to once per 7 EFPD until two successive measurements indicate that FNH is not increasing. The increased frequency of once per 7 EFPD allows detection of power distribution changes prior to FVQ(Z) exceeding its limit for any significant period of time without detection. Recall that FVQ(Z) is an approximation of the Heat Flux Hot Channel Factor, FQ(Z), when accounting for power distribution transients during normal power operation (V(Z)).

The adoption of the limits on the Nuclear Enthalpy Rise Hot Channel Factor (FNH) in TS 3.2.2B preserves compliance with 10 CFR 50.36 requirements to establish limiting conditions for operation, establishment of actions should the limits not be met, and establishment of surveillance requirements that demonstrate compliance with the limits, when the Framatome methods of analysis are used to derive the COLR limits.

As shown in Attachment 3, the Callaway TS Bases will be updated to add the Framatome-specific description for LCO 3.2.2B consistent with TSTF-GG-05-01 and NUREG-1431. The Bases will make clear that this LCO applies when the Framatome-methods are being used to derive the COLR limits.

TS 3.2.3 - Axial Flux Difference (AFD)

As described in the TS Bases, the purpose of this LCO is to establish limits on the values of the AFD in order to limit the amount of axial power distribution skewing to either the top or bottom of the core.

By limiting the amount of power distribution skewing, core peaking factors are consistent with the assumptions used in the safety analyses. Limiting power distribution skewing over time also minimizes the xenon distribution skewing, which is a significant factor in axial power distribution control. The Westinghouse-specific RAOC methodology is a calculational procedure that defines the allowed operational space of the AFD versus THERMAL POWER. The AFD limits are selected by considering a range of axial xenon distributions that may occur as a result of large variations of the AFD. Subsequently, power peaking factors and power distributions are examined to ensure that the loss of coolant accident (LOCA), loss of flow accident, and anticipated transient limits are met.

Violation of the AFD limits invalidates the conclusions of the accident and transient analyses with regard to fuel cladding integrity. The AFD is monitored on an automatic basis using the plant process computer.

As described further below, and as proposed per this LAR, alternative versions of TS 3.2.3, i.e.,

TS 3.2.3A and TS 3.2.3B, would be specified in the Technical Specifications such that either TS 3.2.3A or TS 3.2.3B would apply, depending on which fuel vendor is selected to supply the fuel starting with operating cycle 28 and, therefore, which vendor's analysis methods govern development of the COLR limits for the operating cycle.

to ULNRC-06729 Page 20 of 38

TS 3.2.3A

The requested license amendment will revise TS 3.2.3 to add an "A" designation and annotate that the current LCO 3.2.3 applies only when Westinghouse core design analysis methods are used to derive the COLR limits. The changes include:

  • The APPLICABILITY is modified to include the following Note:

NOTE-----------------------------------

This LCO is not applicable when Framatome COLR methods govern COLR development

These changes do not alter the requirements of the LCO, ACTIONS, or SRs. The APPLICABILITY is modified by a Note that would allow transition to the Framatome methods for COLR development beginning in operating cycle 28 should Framatome be selected as the fuel vendor. The NOTE modifying the APPLICABILITY, in combination with the NOTE modifying the corresponding Framatome specification (e.g., TS 3.2.3B), create a mutually exclusive arrangement where only one fuel vendor's methods may govern COLR development at a time. The addition of the "A" designation is effectively an editorial change that preserves compliance with 10 CFR 50.36 requirements to establish limiting conditions for operation, establishment of actions should the limits not be met, and establishment of surveillance requirements that demonstrate compliance with the limits.

As shown in Attachment 3, the Callaway TS Bases for LCO 3.2.3 will be updated to reflect the addition of the "A" designator as described above. An explanation for the Note modifying the Applicability statement will also be added to the TS Bases to ensure that it is understood that this LCO applies when the Westinghouse-methods are being used to derive the COLR limits.

TS 3.2.3B

The requested license amendment will add a new specification TS 3.2.3B to AXIAL FLUX DIFFERENCE (AFD) limits that apply when Framatome core design analysis methods are used to derive the COLR limits. The adoption of TS 3.2.3B includes the LCO requirement, APPLICABILITY, ACTIONS and SRs specific to the Framatome methodologies. Refer to Attachments 1 and 2 for the new specification. The specification is drafted in a manner consistent with Callaway's Technical Specifications and TSTF-GG-05-01, "Writer's Guide for Plant Specific Improved Technical Specifications," along with NUREG-1431, "Standard Technical Specifications -

Westinghouse Plants," Volume 1, Revision 5, as a reference.

to ULNRC-06729 Page 21 of 38

This new specification is necessary to accommodate the development of the COLR using Framatome methods of analysis, as applicable, beginning in operating cycle 28 should Framatome be selected as the fuel supplier. The Power Distribution Control (PDC) methodology is described in Section 4 of using the methodology to be added to TS 5.6.5b.

As described in the draft TS Bases (Attachment 3), TS 3.2.3B will establish an LCO requiring that the Axial Flux Difference (AFD) be controlled within the AFD limits defined by the sum of the target AFD and the target AFD band specified in the COLR when the limits are derived using Framatome methodologies. This LCO limits the amount of axial power distribution skewing to either the top or the bottom of the core, thereby, ensuring that the core peaking factors are consistent with the assumptions used in the safety analyses. The AFD limits are derived using the PDC-A calculational procedure that establishes an operating envelope. As described in Section 4.2.2 of Attachment 9, the Framatome PDC-A analysis methodology generates set of axial shapes for postulated load follow maneuvers at various times in core life about a target AFD value. This target AFD value defines the axial power shapes that are acceptable for LOCA and loss of flow accidents, and for the initial conditions of AOOs as a function of core life. Thus, the COLR defines an operating band based on the target AFD value at a given time in core life. This non-static AFD operating band is referred to as the floating barn.

An important element in recognizing the requirements of TS 3.2.3B is its relationship with TS 3.2.1B, "Heat Flux Hot Channel Factor (FQ(Z)) Framatome COLR Methods." An example is given TS 3.2.1B Required Action B.1.1, in which the AFD limits for the current target AFD band are reduced when FVQ(Z) is not within its limit. Similarly, the third Note modifying SR 3.2.3B.2 requires that the AFD limits be updated to reflect the new target AFD value determined when FQ(Z) is measured in accordance with SR 3.2.1B.2. This interrelationship preserves the limitations on the axial power profile necessary to ensure that peaking factors remain within their limits.

This LCO is Applicable in MODE 1 with Thermal Power greater than or equal to 50% RTP. The Applicability is modified by a Note that makes clear that this LCO is not applicable when the COLR limits are derived using the Westinghouse methods identified in TS 5.6.5.b. TS 3.2.3B also establishes the Actions necessary when AFD is identified as not being within its limits. Upon identification that AFD is not within limits, Condition A is entered. Required Action A.1 directs that Thermal Power be reduced to less than 50% RTP which conservatively increases operating margins and ultimately removes the unit from the Mode of Applicability. The Completion Time of 30 minutes is comparable to that established in the corresponding Westinghouse LCO (i.e., TS 3.2.3A) and is reasonable, based on operating experience, to reach 50% RTP without challenging plant systems. Condition A is modified by a Note that states that AFD shall be considered not within its limits when two or more Operable excore channels indicate AFD is not within the limits. This Note precludes a significant unit down power that could result from one anomalous excore channel indicating that AFD is not within the limits.

Two SRs, i.e., SR 3.2.3B.1 and SR 3.2.3B.2, are proposed. SR 3.2.3B.1 requires verification that AFD is within limits for each Operable excore nuclear instrumentation channel. The Frequency will be in accordance with the SFCP which will initially stipulate a Frequency of 7 days. This Proposed Frequency is consistent with the corresponding Westinghouse SR, i.e., SR 3.2.3A.1. SR 3.2.3B.2 to ULNRC-06729 Page 22 of 38

establishes the target AFD and its performance coincides with SR 3.2.1B.2, i.e., verification that FVQ(Z) is within the limit specified in the COLR. Tying performance of SR 3.2.3B.2 to the performance of SR 3.2.1B.2 ensures consistency between the AFD operating barn and the V(Z) penalty function used to derive FVQ(Z). This relationship is described in Sections 4.2.2 and 4.2.5 of. Optimally, SR 3.2.3B.2 is performed at RTP with steady states, equilibrium conditions which most closely approximate the analysis conditions. Thus, the SR clarifies that steady state and high power level means as near RTP as practical given the plant operating conditions, at equilibrium xenon conditions such that the core flux profile is not changing significantly, and with the control rods positioned at the normal bank positions (withdrawal and alignment) such that significant power peaking or localized power suppression is not present. SR 3.2.3B.2 has a Frequency of once after each refueling outage, and in accordance with the SFCP. Initially, the SFCP will stipulate a Frequency of 31 Effective Full Power Days (EFPD) based on the slow changes in the target AFD values with core burnup. The SR is modified by three Notes. The first Note allows the initial target AFD to be determined using design predictions. This allows the SR to be met above 50% RTP, i.e., the Applicability, during initial power ascension following a refueling outage. This provision is necessary to allow the unit to achieve the conditions optimum to perform the SR. The second Note allows the target AFD value to be updated in between performance of the SR, i.e., in between the 31 EFPD Frequency intervals following the first completion of the SR. This is accomplished by adding the most recently measured AFD value and the change in the predicted AFD values since the last measurement.

This provision allows updating the target AFD, and corresponding AFD limits, without having to perform flux mapping used during the normal performance of SR 3.2.3B.2. The third Note works in conjunction with the second Note and requires that the AFD limits based on the current target AFD band be updated with the new target AFD value.

The adoption of the limits on the AXIAL FLUX DIFFERENCE (AFD) in TS 3.2.3B preserves compliance with 10 CFR 50.36 requirements to establish limiting conditions for operation, establishment of actions should the limits not be met, and establishment of surveillance requirements that demonstrate compliance with the limits when the Framatome methods of analysis are used to derive the COLR limits.

As shown in Attachment 3, the Callaway TS Bases will be updated to add the Framatome-specific description for LCO 3.2.3B consistent with TSTF-GG-05-01 and NUREG-1431. The Bases will make clear that this LCO applies when the Framatome-methods are being used to derive the COLR limits.

TS 3.2.4 - Quadrant Power Tilt Ratio (QPTR)

As described in the TS Bases, the purpose of the limits on QPTR ensure that the gross radial power distribution remains consistent with the design values used in the safety analyses. Precise radial power distribution measurements are made during startup testing, after refueling, and periodically during power operation. The power density at any point in the core must be limited so that the fuel design criteria are maintained. The limits on QPTR work in combination with other power distribution limits and with limitations on control rod alignment and insertion.

Because the structure for TS 3.2.4 is similar for both the existing Westinghouse and the proposed Framatome Technical Specifications, it is unnecessary to adopt an "A" or "B" designation scheme. to ULNRC-06729 Page 23 of 38

The proposed changes to this TS are conforming changes to ensure that the Required Actions, i.e., A.3 and A.6, refer to the appropriate SRs associated with the governing Westinghouse or Framatome power distribution TS.

Required Actions A.3 and A.6 are modified to reference the corresponding SRs applicable when either Westinghouse or Framatome analysis methods are governing reload analyses and development of the COLR limits. A NOTE is added to stipulate that the SR to be performed is dictated by the vendor whose analysis methods were used to develop the COLR limits. The changes, shown in underlined text, include:

A.3 ------------NOTE-----------

Perform applicable SR based on governing COLR development method.

Perform SR 3.2.1A.1, SR 3.2.1A.2 and SR 3.2.2A.1.

OR

Perform SR 3.2.1B.1, SR 3.2.1B.2 and SR 3.2.2B.1.

to ULNRC-06729 Page 24 of 38

A.6 ------------NOTE-----------

1. Perform Required Action A.6 only after Required Action A.5 is completed.
2. Perform applicable SR based on governing COLR development method.

Perform SR 3.2.1A.1, SR 3.2.1A.2 and SR 3.2.2A.1.

OR

Perform SR 3.2.1B.1, SR 3.2.1B.2 and SR 3.2.2B.1.

The addition of Note 2 and specification of the vendor method-dependent SR in Required Actions A.3 and A.6 preserves compliance with 10 CFR 50.36 requirements to establish limiting conditions for operation, establishment of actions should the limits not be met, and establishment of surveillance requirements that demonstrate compliance with the limits.

As shown in Attachment 3, the Callaway TS Bases for LCO 3.2.4 will be updated to reflect the modification of Required Actions A.3 and A.6 as described above. This will ensure that the required SRs align with those of the fuel vendor's methodologies used to establish the core design and associated COLR limits.

TS 4.2.1 - Fuel Assemblies

The requested license amendment will revise TS 4.2.1 to add reference to the M5 cladding material.

This change is necessary to support the GAIA fuel assembly use of M5 fuel cladding material. In addition, Zircalloy is revised to read zircaloy consistent with 10 CFR 50.46 which corrects a spelling error.

The use of nuclear fuel cladding material M5 in PWR reactor fuel is approved by the NRC in topical report BAW-10227P-A, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel," (Reference 11) and its use in Framatome approved methods is approved in topical report BAW-10240(P)(A), "Incorporation of M5 Properties in Framatome ANP Approved Methods."

(Reference 12) The M5 cladding is a Framatome Proprietary material composed of zirconium and to ULNRC-06729 Page 25 of 38

niobium. This composition has demonstrated superior corrosion resistance and reduced irradiation induced growth relative to both standard and low-tin zircaloy. The resulting alloy microstructure is highly stable under irradiation and provides improved in-reactor thermal and mechanical performance over other zirconium alloys.

A 10 CFR 50.46 and 10 CFR 50 Appendix K exemption request for the implementation of M5 fuel rod cladding is provided in Enclosure 2 of this LAR. This request is necessary since both of these regulatory requirements state or assume that either zircaloy or ZIRLO3 is to be used as the fuel rod cladding material and the requirements are not applicable to M5 cladding.

TS 5.6.5 - Core Operating Limits Report (COLR)

The listing of supported LCOs in TS 5.6.5.a.4, a.5 and a.6 is revised as depicted below to provide references to both the Westinghouse and Framatome Power Distribution Limit LCOs. This is a conforming change to correctly reference the appropriate "A" or "B" TS LCO and to include reference to the LCO-specific parameters in TS 3.2.1A, TS 3.2.1B, TS 3.2.2A and TS 3.2.2B.

5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
4. Axial Flux Difference Limits for Specifications 3.2.3A and 3.2.3B.
5. Heat Flux Hot Channel Factor, FQ(Z), FQRTP, K(Z), W(Z) and FQ Penalty Factors for Specification 3.2.1A, and FCQ(Z), FVQ(Z), Measurement Uncertainty, Manufacturing Tolerance, and V(Z) for Specification 3.2.1B.
6. Nuclear Enthalpy Rise Hot Channel Factor FNH, FHRTP, and Power Factor Multiplier, PFH, limits for Specification 3.2.2A, and FNH, measured FH uncertainty, and UH,, for Specification 3.2.2.B.

TS 5.6.5.b will also be revised to include the Framatome-developed NRC approved analytical methods for neutronics, fuel mechanical, thermal-hydraulics, and safety analyses listed below:

16. EMF-2103(P)(A), Realistic Large Break LOCA Methodology for Pressurized Water Reactors. [Methodology for Specification 3.2.1, FQ]

3 Note that ZIRLO is a registered trademark of Westinghouse Electric Co. to ULNRC-06729 Page 26 of 38

17 EMF-2328(P)(A), PWR Small Break LOCA Evaluation Model, S-RELAP5 Based.

[Methodology for Specification 3.2.1, FQ]

18. EMF-2310(P)(A), SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors.
19. XN-NF-82-21(P)(A), Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations.
20. EMF-92-081(P)(A), Statistical Setpoint/Transient Methodology for Westinghouse Type Reactors.
21. ANP-10341(P)(A), The ORFEO-GAIA and ORFEO-NMGRID Critical Heat Flux Correlations.
22. XN-75-21(P)(A), XCOBRA-IIIC: A Computer Code to Determine the Distribution of Coolant During Steady State and Transient Core Operation.
23. ANP-10311P-A, COBRA-FLX: A Core Thermal-Hydraulic Analysis Code.
24. XN-NF-82-06(P)(A) Supplement 2, 4, and 5, Qualification of Exxon Nuclear Fuel for Extended Burnup.
25. XN-75-32(P)(A) Supplements 1, 2, 3, and 4, Computational Procedure for Evaluating Fuel Rod Bowing.
26. Insert reference to Callaway SER for fuel transition LAR. [For PDC-A methodology]
27. ANP-10297P-A, The ARCADIA Reactor Analysis System for PWRs Methodology Description and Benchmarking Results.
28. ANP-10297P-A, Supplement 1PA, The ARCADIA Reactor Analysis System for PWRs Methodology Description and Benchmarking Results.
29. ANP-10338P-A, AREA - ARCADIA Rod Ejection Accident.
30. ANP-10323P-A, GALILEO Fuel Rod Thermal-Mechanical Methodology for Pressurized Water Reactors.
31. BAW-10231P-A, COPERNIC Fuel Rod Design Computer Code.

Methodologies to be added to TS 5.6.5.b has been prepared by Framatome to augment this LAR discussion by providing a description of the analyses performed, a description of the analysis methods, a discussion of limitations and conditions associated with the methods, and the conclusions of the analyses. In addition, Attachments 10 through 12 contain summaries of the accident and transient analyses performed by Framatome in support of this LAR. They similarly include a description of the analyses performed, a description of the analysis methods, a discussion of limitations and conditions associated with the methods, and the conclusions of the analyses.

The following correlate the analysis area with the methods added to TS 5.6.5.b:

Methodologies proposed to be listed in TS 5.6.5.b as Items 16 and 17 are used in the realistic large break and small break LOCA analyses, respectively.

Methodologies proposed to be listed in TS 5.6.5.b as Items 18, 19, 20, 21, 22, 23, 24, 25 and 26 are used, in combination, in the thermal hydraulics and non-LOCA safety analyses. to ULNRC-06729 Page 27 of 38

Methodologies proposed to be listed in TS 5.6.5.b as Items 27, 28, 29, and 30 are used, in combination, as part of the neutronics analyses.

A methodology proposed to be listed in TS 5.6.5.b as Item 31 is used, in combination, in the thermal mechanical analyses.

As described in Section 5.5.6 of Attachment 9, the methodologies proposed to be listed in TS 5.6.5.b are used, in combination, in the setpoint verification analyses for the Overtemperature T and Overpower T parameters listed in TS 5.6.5.a.9.

The method proposed as TS 5.6.5.b Item 26 involves the NRC approval of the PDC-A methodology:

The PDC-A method will be used to establish the limits on power distribution peaking. PDC-A represents the Power Distribution Control (PDC) procedure that ensures that the TS on power peaking limits, dictated by the LOCA and the Loss-of-Coolant-Flow (LOCF) transient, are not exceeded during the interval between the flux maps. As discussed in Section 4 of Attachment 9, the PDC-A method is a modification of the Framatome PDC methodology referred to as PDC-3 that has been previously approved by the NRC. (References 22, 23, 24) The basic concept of the PDC-A methodology is to control the variation in the core power distribution during reactor operation by controlling the variation in core Axial Flux Difference (AFD) or, equivalently, core power Axial Offset (AO).

With respect to TS 5.6.5.b Item 29, Section 6.1.1 of Attachment 9 provides a description of the analyses performed and the acceptance criteria used to ensure that the FSAR 15.4.8, Spectrum of Rod Cluster Control Assembly Ejection Accidents, conclusions are addressed. These new analyses were performed using the "AREA - ARCADIA Rod Ejection Accident" model (Reference 28) against the acceptance criteria defined in Reference 29. At the time of LAR submittal, these analyses were still being performed; however, the results will be available for NRC during the LAR review process.

Assessment of Current Methodologies Retained in TS 5.6.5.b

Since Westinghouse fuel will remain in the core during operating cycle 27, during subsequent operating cycles should Westinghouse be selected as the fuel supplier, and during subsequent transition cycles, assuming that Framatome is selected as the fuel supplier, all of the Westinghouse Topical Reports, which are listed in TS 5.6.5.b, are retained since they will be applicable to the Westinghouse fuel analyses during the transition cycles or subsequent operation with Westinghouse fuel. Consistent with Ameren Missouri's commitment provided in Enclosure 3, unneeded methodologies will be deleted from TS 5.6.5.b after the completion of fuel transition to a full core of Framatome fuel or following removal of the last of the eight Framatome VQP and LFA assemblies.

Since the FSAR events not related to the fuel design change are not currently re-analyzed with Framatome methodology, some of the Westinghouse methodologies will continue to maintain the to ULNRC-06729 Page 28 of 38

analyses of record in the FSAR. As such, these methodologies will be retained in TS 5.6.5.b until such time as the corresponding events are re-analyzed.

Provisions are in place to have Westinghouse continue to provide supporting analyses demonstrating the acceptability of their fuel during transition cycles.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements / Criteria

The regulatory requirements and/or guidance documents associated with this amendment application include the following:

Technical Specifications

Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include Technical Specifications (TS) as part of the license. The TS ensure the operational capability of structures, systems, and components that are required to protect the health and safety of the public. The U.S. Nuclear Regulatory Commissions (NRCs) requirements related to the content of the TSs are contained in Section 50.36 of Title 10 of the Code of Federal Regulations (10 CFR 50.36) which requires that the TSs include items in the following specific categories: (1) safety limits, limiting safety systems settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements per 10 CFR 50.36(c)(3); (4) design features; and (5) administrative controls. This amendment request involves each of the five categories in 10 CFR 50.36(c)(1) through 10 CFR 50.36(c)(5).

This LAR describes the manner in which the 10 CFR 50.36 requirements continue to be met through the specification of appropriate safety limits, limiting conditions for operation, surveillances, design feature description, and establishment of administrative controls. The analyses described in Attachments 9 through 12 support the demonstration of the adequacy of the proposed TS.

10 CFR 50.46 requires, in part, that each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical zircaloy or ZIRLO cladding must be provided with an emergency core cooling system (ECCS) that must be designed so that its calculated cooling performance following postulated loss-of-coolant accidents conforms to the criteria set forth in 10 CFR 50.46(b). Appendix K to 10 CFR Part 50 establishes the regulations for conservative ECCS evaluation models. Enclosure 2 contains an exemption request from 10 CFR 50.46 and Appendix K to 10 CFR Part 50, in accordance with 10 CFR 50.12. In Reference 11, the NRC approved the use of M5 as an acceptable fuel rod cladding material for Framatome fuel designs. The analyses discussed in Attachments 10 and 11 demonstrate the continued acceptability of the Callaway ECCS to meet the performance criteria in 10 CFR 50.46.

to ULNRC-06729 Page 29 of 38

General Design Criteria

Callaway Final Safety Analysis Report-Standard Plant (FSAR-SP) Section 3.1 discusses the extent to which the design criteria for Westinghouse Standardized Nuclear Unit Power Plant System (SNUPPS) plant structures, systems, and components important to safety comply with Title 10, Code of Federal Regulations, Part 50 (10 CFR 50), Appendix A, "General Design Criteria for Nuclear Power Plants" (GDC). Compliance with pertinent GDC are discussed in the following paragraphs:

Criterion 10 - Reactor Design

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

As described in Attachment 9, the reactor core will be designed with appropriate margin to assure that SAFDLs are not exceeded during any condition of normal operation, including the effects of AOOs. Therefore, GDC 10 continues to be satisfied.

Criterion 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."

As described in Attachment 9, the reactor core will be designed so that the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity. Therefore, GDC 11 continues to be satisfied.

Criterion 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."

As described in Attachment 9, the reactor core will be designed to assure that power oscillations, which can result in conditions exceeding SAFDLs, are not possible or can be reliably and readily detected and suppressed. Therefore, GDC 12 continues to be satisfied.

Criterion 20 - Protection System Functions

"The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to to ULNRC-06729 Page 30 of 38

sense accident conditions and to initiate the operation of systems and components important to safety."

As described in Attachments 9 through 12, the reactor core will be designed such that automatic initiation of the reactivity control systems will continue to assure that acceptable fuel design limits are not exceeded as a result of AOOs and to automatically initiate operation of systems and components important to safety under accident conditions. Therefore, GDC 20 continues to be satisfied.

Criterion 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."

As described in Attachments 9 through 12, the protection system will continue to be designed to assure that the protection system assures that SAFDLs are not exceeded for any single malfunction of the reactivity control systems. Therefore, GDC 25 continues to be satisfied.

Criterion 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 that the acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions."

As described in Attachments 9 through 12, two independent reactivity control systems will continue to be provided, with both systems capable of reliably controlling the rate of reactivity changes from planned, normal power changes. Therefore, GDC 26 continues to be satisfied.

Criterion 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."

As described in Attachments 9 through 12, the reactivity control systems will continue to be designed to have a combined capability, in conjunction with poison addition by the ECCS, of reliably controlling reactivity changes under postulated accident conditions, with appropriate to ULNRC-06729 Page 31 of 38

margin for stuck rods, to assure the capability to cool the core is maintained. Therefore, GDC 27 continues to be satisfied.

Criterion 28 - Reactivity Limits

"The reactivity control system 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."

As described in Attachments 9 through 12, the reactivity control systems will continue to be designed to assure that the effects of postulated reactivity accidents can neither result in damage to the RCPB greater than limited local yielding, nor disturb the core, its support structures, or other reactor vessel internals so as to significantly impair the capability to cool the core. Therefore, GDC 28 continues to be satisfied.

Criterion 35 - Emergency Core Cooling

"A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts."

As described in Attachments 9 through 12, the fuel design installed in Callaway's core configuration has been evaluated and analyses demonstrate that fuel and cladding damage will not occur to the extent that it interferes with the continued effective core cooling following a loss of coolant accident. Further, analyses demonstrate that the clad metal-water reaction is limited to amounts that meet acceptance criteria. The use of the Framatome GAIA fuel has no effect on the design characteristics or functional capability of the emergency core cooling systems. Therefore, GDC 35 continues to be satisfied.

There are no changes being proposed in this amendment application such that conformance or commitments to the regulatory requirements and/or guidance documents above would come into question. The evaluations documented herein confirm that Callaway will continue to comply with all applicable regulatory requirements.

In conclusion, based on considerations discussed herein, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. to ULNRC-06729 Page 32 of 38

4.2 Precedent

The following fuel transition licensing activities have involved either NRC review of GAIA fuel or have involved the review of the Framatome analysis methods associated with this LAR:

Duke Energy submitted a license amendment request to transition from Framatome High Thermal Performance (HTP) fuel to GAIA fuel at the Shearon Harris Nuclear Power Plant, Unit 1 (ADAMS Accession No. ML19100A442) that was approved in 2019 by the NRC (ADAMS Accession No. ML20212L594). The NRC evaluated the analytical methods specific to GAIA fuel as part of the amendment application. Shearon Harris is a Westinghouse three loop pressurized water reactor sharing many comparable design elements and operating characteristics with Callaway.

Arizona Public Service submitted a license amendment request to transition to Framatome Advanced Combustion Engineering 16x16 HTP fuel with M5 as a fuel rod cladding material and gadolinia as a burnable absorber for all three units at the Palo Verde Nuclear Generating Station that was approved in 2020 by the NRC (ADAMS Accession No. ML20031C947). Palo Verde is a Combustion Engineering two steam generator and four cold leg pressurized water reactor.

NextEra Energy submitted a license amendment request to transition to Framatome (then AREVA) Combustion Engineering 16x16 HTP fuel with M5 as a fuel rod cladding material for its St. Lucie Plant, Unit 2, that was approved in 2016 by the NRC (ADAMS Accession No. ML16063A121). St. Lucie Unit 2 is a Combustion Engineering two steam generator and four cold leg pressurized water reactor.

Constellation Energy submitted a license amendment request to transition to Framatome (then AREVA) Advanced Combustion Engineering 14x14 HTP fuel with gadolinium oxide burnable poison with M5 as a fuel rod cladding material for its Calvert Cliffs Nuclear Power Plant, Units 1 and 2, that was approved in 2011 by the NRC (ADAMS Accession No. ML110390263). Calvert Cliffs is a Combustion Engineering two steam generator and four cold leg pressurized water reactor.

Exelon Generation submitted a license amendment request to allow use of up to two accident tolerant fuel lead test assemblies in its Calvert Cliffs Nuclear Power Plant, Units 1 and 2. The application requested the ability to load Framatome PROtect' fuel design (ADAMS Accession No. ML19347A779). The NRC's permission to load the lead test assemblies is documented in the amendment approval letter along with the safety evaluation dated January 26, 2021 (ADAMS Accession No. ML20363A242). Calvert Cliffs is a Combustion Engineering two steam generator and four cold leg pressurized water reactor.

Florida Power and Light submitted a license amendment request for an extended power uprate for a core using AREVA 14x14 high thermal performance fuel. The review of the amendment request included a review of Framatome (then AREVA) analysis methods that are also included in this amendment request. The license amendment request was approved as documented in the safety evaluation dated July 9, 2012 (ADAMS Accession No. ML12181A019). to ULNRC-06729 Page 33 of 38

Section 2.2 of Attachment 9 provides a synopsis of the operational experience of Framatome GAIA fuel assemblies in Westinghouse 3-loop and 4-loop designs.

4.3 No Significant Hazards Consideration Determination

The standards used to arrive at a determination that a request for amendment involves a No Significant Hazards Consideration are included in the Commission's regulation, 10 CFR 50.92, "Issuance of amendment," which states that No Significant Hazards Considerations are involved if the 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.

Ameren Missouri has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92 as discussed below:

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

Response: No

The proposed license amendment would modify the TS to include Framatome GAIA fuel-specific safety limits, adopt Framatome-method specific limiting conditions for operation, allow the use of M5 fuel cladding material, and add Framatome analysis methods to the Core Operating Limits Report (COLR) listing. In combination, these changes allow the use of eight GAIA fuel assemblies in operating cycle 27 and transition to full batch and full core use of Framatome GAIA fuel should Framatome be selected as the next fuel supplier for Callaway.

The proposed change incorporates into the TS a limit on the departure from nucleate boiling ratio (DNBR) safety limit and the peak fuel centerline temperature safety limit for the GAIA fuel design. These limits are based on NRC reviewed and approved correlations, and do not require a physical change to plant systems, structures or components. Plant operations and analysis will continue to be in accordance with the Callaway licensing basis. This change does not impact any of the accident initiators. The departure from nucleate boiling ratio and the peak fuel centerline temperature are the basis for protecting the fuel and are consistent with the safety analysis.

The proposed safety limits ensure that fuel integrity will be maintained during normal operations and anticipated operational transients. The operating limits specified in the COLR will be developed in accordance with approved methodologies. The proposed safety limit values and proposed analysis methods do not affect the performance of any equipment used to mitigate the consequences of an analyzed accident. There is no impact on the pathways for radionuclide to ULNRC-06729 Page 34 of 38

release assumed in accidents previously evaluated. The available radiological source term available for release is not significantly affected by the use of the GAIA design. An evaluation of the projected does consequences of a loss of accident while operating with GAIA fuel against the dose consequences in the current analysis of record demonstrates that the health and safety of the public continues to be assured.

The adoption of Framatome method-specific power distribution limits TS establishes limiting conditions for operation derived from approved methods of analysis added to the listing of approved methods for development of the COLR. These limiting conditions for operation will establish appropriate actions to take to place the plant in a safe condition should the limits not be met. The establishment of these limits does not affect the initiating conditions for any evaluated event.

This change also allows the use of Framatome M5 cladding material which has been previously evaluated and approved for use in pressurized water reactors. Analysis demonstrates the suitability of the cladding material and its ability to satisfy regulatory performance criteria during transients and anticipated operational occurrences. The presence of this material has no direct cause and effect relationship with the initiation of any evaluated accident.

The proposed change also adds reference to previously approved topical reports used to develop the COLR all of which have been reviewed and approved by the NRC. This addition is administrative in nature and has no impact on a plant configuration or system performance relied upon to mitigate the consequences of an accident. The list of topical reports in the TS used to develop the core operating limits does not impact either the initiation of an accident or the mitigation of its consequences.

For the reasons above, this proposed license amendment does not significantly increase the probability or consequences of an accident previously evaluated in the FSAR.

The proposed changes will allow use of Framatome GAIA fuel in the reactor core. The design characteristics have been analyzed to ensure its compatibility with co-resident fuel and that it complies with regulatory acceptance criteria. The presence of the GAIA fuel in the core does not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended functions to mitigate the consequences of an initiating event within the assumed acceptance limits.

The proposed use of Framatome GAIA fuel has been evaluated using NRC-approved methods to demonstrate that the consequences of a previously evaluated accidents are not significantly increased. This change will not alter the licensed thermal power level of the reactor, nor will it alter any assumption regarding assumed release pathways to the environment. For these reasons, the consequences of an accident are not significantly increased.

Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

to ULNRC-06729 Page 35 of 38

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

Response: No

No new or different accidents result from the use of the proposed Framatome GAIA fuel design.

Other than the fuel design change, the proposed license amendment does not involve a physical alteration of the plant or plant systems (i.e., no new or different type of equipment will be installed which would create a new or different kind of accident).

Physical changes associated with the proposed Framatome GAIA fuel design do not introduce any new accident initiators and do not adversely affect the performance of any structure, system, or component previously credited for accident mitigation. Use of Framatome fuel with M5 cladding in the Callaway reactor core is compatible with the plant design and does not introduce any new safety functions for plant structures, systems, or components. Analysis demonstrates that the fuel design performs within the fuel design limits.

The adoption of the Framatome specific safety limits and power distribution limiting conditions for operation, and the use of M5 cladding do not affect or create any new or different accident initiator. The proposed change also adds reference to topical reports for NRC reviewed and approved analytical methods to the list of topical reports in the TS, which is administrative in nature and has no impact on a plant configuration or system performance. The proposed change updates the list of NRC-approved topical reports used to develop the core operating limits. There is no change in the parameters within which the plant is normally operated and the changes do not impose any new or different operating 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 changes proposed in this license amendment request are related to the use of the GAIA fuel design with M5 cladding and adoption of the methodologies supporting the analysis of accidents impacted by the fuel design change. The analysis methods are previously approved by the NRC for similar applications. The analyses demonstrate continued compliance with the regulatory acceptance limits in a manner comparable to the existing safety analysis for the fuel supplied by the current vendor. The reactor will continue to be operated within its analyzed operating and design envelop. The overpressure limits for the reactor coolant system integrity and the containment integrity remain unchanged. The LOCA analyses meet all the applicable 10 CFR 50.46 acceptance criteria, and, thus, the proposed changes do not affect margin to safety for any accidents previously evaluated.

to ULNRC-06729 Page 36 of 38

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

In consideration of the above, Ameren Missouri concludes that the proposed changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and on that basis, a finding of "no significant hazards consideration" is justified.

4.4 Conclusions

Based on the considerations discussed above, 1) there is a reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner with the fuel design change, 2) such activities will be conducted in compliance with the Commission's regulations, and

3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL EVALUATION

The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

6.0 REFERENCES

1. Ameren Missouri letter ULNRC-06695, "Supplement to Presentation Information For Second Pre-Application Meeting Regarding Use of Framatome Fuel At Callaway Plant," dated October 6, 2021 (ADAMS Accession No. ML21279A286)
2. Ameren Missouri letter ULNRC-06694, Presentation Information for Second Pre-Application Meeting Regarding Use of Framatome Fuel at Callaway Plant, dated September 28, 2021 (ADAMS Accession No. ML21271A611)
3. Ameren Missouri letter ULNRC-06692, " Third Supplement to Request For License Amendment and Regulatory Exemptions for a Risk-Informed Approach To Address GSI-191 and Respond To GL 2004-02 (LDCN 19-0014) (EPID L-2021-LLA-0059 AND EPID L-2021-LLE-0021)," dated October 7, 2021 (ADAMS Accession No. ML21280A379)
4. Ameren Missouri letter ULNRC-06526, "Request for License Amendment and Regulatory Exemptions for a Risk-Informed Approach to Address GSI-191 and Respond to GL 2004-02 (LDCN 19-0014)," dated March 31, 2021 (ADAMS Accession No. ML21090A184) to ULNRC-06729 Page 37 of 38
5. Ameren Missouri Letter ULNRC-06664, "Supplement to Request for License Amendment and Regulatory Exemptions for a Risk-Informed Approach to Address GSI-191 and Respond to GL 2004-02 (LDCN 19-0014)," dated May 27, 2021 (ADAMS Accession No. ML21147A222)
6. Ameren Missouri letter ULNRC-06651, "Supplement to Request for License Amendment and Regulatory Exemptions for a Risk-Informed Approach to Address GSI-191 and Respond to GL 2004-02 (LDCN 19-0014)," dated July 22, 2021 (ADAMS Accession No. ML21203A192)
7. Ameren Missouri letter ULNRC-06690, " Fourth (Post-Audit) Supplement to Request for License Amendment and Regulatory Exemptions for a Risk-Informed Approach to Address GSI-191 and Respond to GL 2004-02 (LDCN 19-0014) (EPID L-2021 LLA 0059 AND EPID L-2021-LLE-0021)," dated January 27, 2022 (ADAMS Accession No. ML22027A805)
8. Ameren Missouri letter ULNRC-06683, "Transmittal of Documents Identified from NRC Audit of License Amendment Request Regarding Risk-Informed Approach to Closure of Generic Safety Issue 191 (EPID L-2021-LLA-0059)", dated August 23, 2021 (ADAMS Accession No. ML21237A136)
9. Ameren Missouri letter ULNRC-06654, "Presentation Information for Pre-Application Meeting Regarding Use of Framatome Fuel at Callaway Plant," dated May 3, 2021 (ADAMS Accession No. ML21123A259)
10. ANP-3947P, "Callaway Unit 1 License Amendment Request Inputs for Use of Framatome Fuel," Revision 0, dated April 2022.
11. BAW-10227P-A, Revision 1, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel," dated June 2003.
12. BAW-10240(P)(A) Revision 0, "Incorporation of M5 Properties in Framatome ANP Approved Methods," dated May 2004.
13. FS1-0061145, Revision 2.0, "Callaway Impact of GAIA on Dose Analyses of Record (Alternate Source Term)" dated April 2022.
14. NRC Regulations, Title 10, Code of Federal Regulations, Part 50, Domestic Licensing of Production and Utilization Facilities.
15. ANP-3969P, Revision 0, "Callaway Non-LOCA Summary Report," dated March 2022.
16. ANP-3943P, Revision 0, "Callaway Small Break LOCA Analysis with GAIA Fuel Design,"

dated March 2022.

17. ANP-3944P, Revision 0, "Callaway Realistic Large Break LOCA Analysis with GAIA Fuel Design," dated March 2022.

18 NRC letter dated July 1, 2021, Summary of May 20, 2021, Partially Closed Pre-Application Teleconference with Union Electric Company Regarding Approval to Insert Framatomes GAIA Fuel Assemblies in Their Core in Reload Quantities for Callaway Plant, Unit No. 1 (EPID L-2021-LRM-0043), (ADAMS Accession No. ML21168A002) 19 NRC letter dated November 17, 2021, Summary of October 14, 2021, Partially Closed Pre-Application Teleconference with Union Electric Company Regarding Approval to Insert Framatomes GAIA Fuel Assemblies in its Core in Reload Quantities for Callaway Plant, Unit No. 1 (EPID L-2021-LRM-0099). (ADAMS Accession Number ML21308A074) 20 Ameren Missouri letter ULNRC-06716, "Presentation Information for Third Pre-Application Meeting Regarding Use of Framatome Fuel at Callaway Plant," dated February 15, 2022.

(ADAMS Accession No. ML22047A093) 21 NRC letter dated April 6, 2022, Summary of February 28, 2022, Partially Closed Pre-Application Teleconference with Union Electric Company Regarding Approval to Insert to ULNRC-06729 Page 38 of 38

Framatomes GAIA Fuel Assemblies in its Core in Reload Quantities for Callaway Plant, Unit No. 1 (EPID L-2022-LRM-0013), (ADAMS Accession No. ML22089A229) 22 XN-NF-77-57(A) & Supplement 1(A), Exxon Nuclear Power Distribution Control for Pressurized Water Reactors - Phase II, March 1981.

23 XN-NF-77-57(P)(A) Supplement 2(P)(A) and Supplement 2 Addendum 1, Exxon Nuclear Power Distribution Control for Pressurized Water Reactors - Phase II, October 1982.

24 ANF-88-054-(P)(A), PDC-3: Advanced Nuclear Fuels Corporation Power Distribution Control for Pressurized Water Reactors and Application of PDC-3 to H.B. Robinson Unit 2, October 1990.

25 Deleted 26 BAW-10227 Revision 2, Q3P Revision 0, Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel, July 2021.

27 XN-75-32(P)(A) Supplements 1, 2, 3, and 4, Computational Procedure for Evaluating Fuel Rod Bowing, February 1983.

28 ANP-10338P-A, Revision 0, AREATM - ARCADIA Rod Ejection Accident, December 2017.

29 Regulatory Guide 1.236, Pressurized-Water Reactor Control Rod Ejection and Boiling-Water Reactor Control Rod Drop Accidents, dated June 2020. (ADAMS Accession No. ML20055F490) 30 U.S. Nuclear Regulatory Commission Staff Review guidance for In-Vessel Downstream Effects Supporting Review of Generic Letter 2004-02 Responses, September 4, 2019, (ML19228A011).

31 WCAP-17788P, Volume 1, "Comprehensive Analysis and Test Program for GSI-191 Closure (PA-SEE-1090)."

32 PWROG-16057-P, "TSTF-567 Implementation Guidance, Evaluation of In-Vessel Debris Effects, Submittal Template for Final Response to Generic Letter 2004-02 and FSAR Changes."

33 Ameren Missouri letter ULNRC-06636, "License Amendment Request for Adoption of Alternate Source Term and Revision of Technical Specifications," dated September 28, 2021.

(ADAMS Accession No. ML21272A167 (Package), ML21335A451 (Package)).

34 Ameren Missouri letter ULNRC-06696, "Supplement to License Amendment Request for Adoption of Alternate Source Term and Revision of Technical Specifications (EPID L-2021-LLA-0177)," dated December 1, 2021 (ADAMS Accession No. ML21335A451 (Package)).

35 Regulatory Guide 1.183, Revision 0, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors."

36 ANP-10342P-A, Revision 0, "GAIA Fuel Assembly Mechanical Design," dated September 2019.

37 NRC Safety Evaluation Report, "Transition to AREVA NP Fuel and Safety Analysis Methodology," Calvert Cliffs Nuclear Power Plant, dated February 18, 2011 (ADAMS Accession No. ML110390263).

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