Text

License Amendment Request to Change Technical Specification 3.6.5, Containment Air Temperature Actions
	ML24201A108
Person / Time
Site:	Farley
Issue date:	07/18/2024
From:	Coleman J; Southern Nuclear Operating Co
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
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ML24201A106	List: NL-24-0281, License Amendment Request to Change Technical Specification 3.6.5, Containment Air Temperature Actions;
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	NL-24-0281
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._ Southern Nuclear Regulatory Affairs 3535 Colonnade Parkway Birmingham, AL 35243 205 992 5000 WITHHOLD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390 (DECONTROLLED UPON REMOVAL OF ATTACHMENT 6 TO THE ENCLOSURE)

July 18, 2024 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant Units 1 and 2 Docket Nos. 50-348 and 50-364 NL-24-0281 10 CFR 50.90

Subject:

License Amendment Request to Change Technical Specification 3.6.5, "Containment Air Temperature" Actions Pursuant to the provisions of 10 CFR 50.90, Southern Nuclear Operating Company (SNC) hereby requests a license amendment to the Technical Specifications (TS) for Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2 renewed facility operating licenses NPF-2 and NPF-8, respectively. The requested amendment would revise the operating license, Appendix A, Technical Specification (TS) 3.6.5, Containment Air Temperature, Actions upon exceeding the containment average air temperature limit and remove an expired Limiting Condition for Operation Note.

The change was previously discussed with the NRC Staff on July 11, 2024 [ML24193A028].

SNC requests expedited review and approval of the proposed license amendment by August 24, 2024. This license amendment will be implemented promptly upon issuance.

The enclosure to this letter provides the description, technical evaluation, regulatory evaluation (including the Significant Hazards Consideration Determination) and environmental considerations for the proposed changes.

Attachments 1 and 2 provide the marked-up TS pages and revised TS pages, respectively, depicting the requested changes. Attachment 3 provides a "for information" markup of the TS Bases that would be implemented along with the proposed TS change. An affidavit and request for withholding proprietary information is provided as Attachment 4. A non-proprietary version of the large break loss-of-coolant accident (LB LOCA) evaluation is provided as and a proprietary version of the LB LOCA evaluation is provided as Attachment 6.

U. S. Nuclear Regulatory Commission NL-24-0281 Page 2 contains information proprietary to Westinghouse Electric Company LLC

("Westinghouse"), and it is supported by an Affidavit signed by Westinghouse, the owner of the information (Attachment 4 ). The Affidavit sets forth the basis on which the information may be withheld from public disclosure by the Nuclear Regulatory Commission ("Commission") and addresses with specificity the considerations listed in paragraph (b )(4) of Section 2.390 of the Commission's regulations. Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CFR Section 2.390 of the Commission's regulations.

Correspondence with respect to the copyright, proprietary aspects, or the supporting Westinghouse Affidavit should reference CAW-24-037 and should be addressed to Zachary S.

Harper, Senior Manager, Licensing Engineering.

This letter contains no regulatory commitments. This letter has been reviewed and determined not to contain security-related information.

In accordance with 10 CFR 50.91, SNC is notifying the State of Alabama of this license amendment request by transmitting a copy of this letter, enclosure, and attachments to the designated State Official.

If you have any questions, please contact Ryan Joyce at 205-992-6468.

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

Executed on the 18th day of July 2024.

Respectfu I ly submitted,

~~

Jamie M. Coleman Director, Regulatory Affairs Southern Nuclear Operating Company JMC/was/cbg

Enclosure:

Evaluation of the Proposed Change cc:

NRC Regional Administrator, Region II NRR Project Manager - Farley 1 & 2 Senior Resident Inspector - Farley 1 & 2 Alabama - State Health Officer for the Department of Public Health RType: CFA04.054

Enclosure to NL-24-0281 Evaluation of the Proposed Change ENCLOSURE Evaluation of the Proposed Change

Subject:

License Amendment Request to Change Technical Specification 3.6.5, "Containment Air Temperature" Actions

1.

SUMMARY

DESCRIPTION

2. DETAILED DESCRIPTION 2.1 System Design and Operation 2.2 Current Technical Specifications Requirements 2.3 Description of the Proposed Change 2.4 Reason for the Proposed Change

3. TECHNICAL EVALUATION 3.1 Deterministic Evaluation 3.2 Defense-in-Depth 3.3 Safety Margins 3.4 Probabilistic Risk Assessment (PRA) 3.5 Implementation and Monitoring Plan 3.6 Conclusions and Integrated Decision Regarding Safety Impact

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusions

5. ENVIRONMENTAL CONSIDERATION

6. REFERENCES ATTACHMENTS:

1. Technical Specification Page Markups

2. Revised Technical Specification Pages

3. Technical Specification Bases Markup (For Information Only)

4. Westinghouse Electric Company Affidavit and Request for Withholding Proprietary Information

5. Large Break Loss of Coolant Evaluation (Non-Proprietary)

6. Large Break Loss of Coolant Evaluation (Proprietary) [WITHHOLD FROM PUBLIC DISCOSURE]

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Enclosure to NL-24-0281 Evaluation of the Proposed Change

1.

SUMMARY

DESCRIPTION Pursuant to the provisions of 10 CFR 50.90, Southern Nuclear Operating Company (SNC) hereby requests a license amendment to the Technical Specifications (TS) for Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2. The requested amendment would revise the operating license, Appendix A, Technical Specification (TS) 3.6.5, Containment Air Temperature, Actions upon exceeding the containment average air temperature limit of 120°F, and remove an expired Limiting Condition for Operation (LCO) Note.

2.

DETAILED DESCRIPTION 2.1 System Design and Operation The containment is a prestressed, reinforced concrete cylindrical structure with a shallow domed roof and a reinforced concrete foundation slab. A 1/4-in.-thick welded steel liner is attached to the inside face of the concrete. The floor liner is installed on top of the foundation slab and is then covered with concrete. The containment completely encloses the reactor, the reactor coolant systems, the steam generators, and portions of the auxiliary and engineered safeguards systems. It ensures that an acceptable upper limit for leakage of radioactive materials to the environment will not be exceeded even if gross failure of the reactor coolant system occurs. The structure is designed to contain radioactive material that may be released from the reactor core following a Design Basis Accident (OBA). Additionally, this structure provides shielding from the fission products that may be present in the containment atmosphere following accident conditions.

As described in Final Safety Analysis Report (FSAR) subsection 6.2.2, three systems are provided to reduce containment atmosphere temperature and pressure and/or to remove heat from the containment under post-accident conditions. These are the low-head safety injection/residual heat removal system, the containment spray system, and the containment cooling system. The two redundant trains of the low-head safety injection/residual heat removal system initially provide injection operation; following the injection operation, water collected in the containment sump is cooled and returned to the reactor coolant system by the low-head safety injection/ residual heat removal system recirculation flow paths. The two redundant trains of the containment spray system have been designed to provide sufficient heat removal capacity to prevent exceeding containment design pressure for all piping breaks. The containment cooling system has been designed to remove heat which will be released to the containment atmosphere during any Main Steam Line Break (MSLB) or Loss Of Coolant Accident (LOCA) up to and including the double-ended rupture of the largest system pipe. This is accomplished by one of four containment air coolers.

As described in FSAR subsection 6.2.1.3.3, Containment Pressure Transient Analysis, and shown in Table 6.2-3, Initial Conditions for Pressure Analysis, the analyses for containment pressure assumed an initial containment temperature of 127°F while the accumulator is assumed at 120°F and the refueling water storage tank (RWST) is assumed at 110°F.

2.2 Current Technical Specifications Requirements TS 3.6.5 requires that the containment average air temperature be limited to::; 120°F. Once this limit is reached, the plant has eight hours to restore the temperature within limits. If this action is not met, the plant must be in Mode 3 within six hours and Mode 5 within thirty-six hours. The LCO also contains an expired NOTE allowing a temporary increase in the containment average air temperature until September 9 of 2023.

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Enclosure to NL-24-0281 Evaluation of the Proposed Change 2.3 Description of the Proposed Change The TS 3.6.5 LCO NOTE allowing a one-time increase in the containment average air temperature limit until 0600 hours0.00694 days 0.167 hours 9.920635e-4 weeks 2.283e-4 months on September 9, 2023, has expired and is proposed to be removed.

The TS 3.6.5 Actions, upon exceeding the containment average air temperature limit, are proposed to be revised to allow continued operation provided:

the containment average air temperature remains less than or equal to 122°F (verified within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter);

the refueling water storage tank temperature remains less than or equal to 100°F (verified within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter); and the containment average air temperature has not exceeded the 120°F limit for more than 7 days cumulative during the calendar year.

The proposed actions and the associated limitations in these actions align with the supporting supplemental evaluations discussed in the Technical Evaluation section below.

Markups showing the TS change are provided in Attachment 1.

2.4 Reason for the Proposed Change In August of 2023, elevated site ambient temperatures were experienced which were projected to result in containment temperature exceeding the limit of 120°F. An emergency LAR was prepared and submitted by SNC and approved by the NRC which added the temporary use LCO Note [ML23235A296].

The high containment average ambient temperature conditions are projected to continue during future summer months potentially resulting in the potential for one or both FNP Unit 1 and Unit 2 containment average air temperatures exceeding 120°F. Previously unanticipated high summer temperatures are expected to continue (and possibly worsen); thus, it is prudent to request a change to the actions associated with exceeding the existing limit in order to avoid further expenditures of SNC and NRC resources on emergency License Amendment Requests (LAR) and Notices Of Enforcement Discretion (NOED).

The containment average air temperature is verified to be within the daily limit in accordance with Surveillance Requirement (SR) 3.6.5.1. During 2023, Unit 1 had 8 days during which the containment average temperature exceeded 118°F for at least a portion of the day with no days over 119°F; however, Unit 2 had 28 days during which the containment average temperature exceeded 118°F for at least a portion of the day and 6 days over 119°F.

The temperature data shows a general trend upward for periods of high temperatures. Such a projection is consistent with the discussion of climate change considerations provided in the NRC's Generic Environmental Impact Statement for License Renewal of Nuclear Plants (NUREG-1437, Revision 2 [ML23202A179]). Section 4.12.2 of this report indicates that while "the projections of possible climate change effects entail substantial uncertainty," the "Climate models indicate that over the next few decades, temperature increases will continue due to current GHG emission concentrations in the atmosphere (USGCRP 2014)."

Furthermore, given that the loss of generation during high ambient temperature conditions could cause large deviations in normal power flows in southern Alabama and Georgia and northern Florida, and will likely cause load curtailments in the Southern Balancing Area and create a higher probability of rotating load shed that could produce safety and wellness issues for E-3

Enclosure to NL-24-0281 Evaluation of the Proposed Change customers during sustained periods of extremely high temperatures, the request reflects a significant reduction in risk to public health and safety.

SNC is requesting an extension of completion time allowed to restore the containment average air temperature to within the limit based on additional remedial actions and compensatory measures as discussed below. Approval of this request is expected to alleviate the need for emergency LARs and requests for NOEDs related to containment temperature for the foreseeable future.

3. TECHNICAL EVALUATION SNC anticipates that the high containment average temperature experienced during past summers will occur during future summers and that those challenges will be exacerbated by continued increases in projected ambient temperatures, such that SNC can expect to experience challenges to containment temperature exceedances. In order to alleviate the need for unnecessary plant shutdowns during peak electrical demand or emergency amendments and requests for enforcement discretion, SNC requests a license amendment to allow for containment temperature exceedance while taking appropriate remedial actions and compensatory measures. These remedial actions include limitations on the amount of exceedance allowed (based on both the risk evaluation and the deterministic evaluation), on the total time per calendar year that the temperature can be exceeded (based on risk evaluation and a reasonable expectation of duration), and more restrictive limitations on the refueling water storage tank temperature (based on the deterministic evaluation).

The technical basis for these remedial actions is discussed in the LAR. These remedial actions, including verifying actual RWST temperature is at least 10°F less than what's assumed in the accident analysis and verifying containment temperature is less than the initial containment air temperature assumed in the limiting design basis accident analysis, provide a safety basis that's equivalent to or bounded by what's assumed in the accident analysis. In addition, the compensatory measures discussed in Section 3.5 will help limit the probability the LCO limit is exceeded, or if it is, will limit the duration the LCO limit is exceeded.

Pursuant to the content guidance provided by Regulatory Guide (RG) 1.177, "Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," a deterministic evaluation is provided in Section 3.1, the defense-in-depth (DID) is addressed in Section 3.2, the safety margins are evaluated in Section 3.3, a risk impact evaluation is provided in Section 3.4, an implementation and monitoring plan is provided in Section 3.5, and the summary conclusions and integrated decision of acceptability of the change is provided in Section 3.6.

3.1 Deterministic Evaluation As identified in FSAR Section 6.2, the containment, in conjunction with engineered safety features (ESFs), is designed to withstand the internal pressure and coincident temperature resulting from the energy release of the LOCA or MSLB associated with 2831 MWt and to limit the site boundary radiation dose to within the guidelines set forth in regulations.

The limiting postulated accidents considered are as follows:

A.

Double-ended pump suction guillotine (DEPSG), minimum ESF.

B.

Double-ended hot leg guillotine (DEHLG), blowdown phase only.

C.

Spectrum of main steam line breaks.

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Enclosure to NL-24-0281 Evaluation of the Proposed Change The DEPSG with minimum ESF (discussed below) bounds the response of the DEHLG. For the main steamline break accidents, the containment response is already bounded for a LCO 3.6.5 temperature up to 124 °F and do not require evaluation.

The proposed change would allow the internal containment average air temperature to exceed the Technical Specification limit for the temperature at the initiation of an accident up to 2°F for up to 7 days cumulative during a calendar year. The proposed Technical Specification actions during the increased containment temperature include a requirement to verify the RWST water temperature remains less than 100°F within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter.

Existing containment maximum temperature analyses currently assume an initial containment average air temperature of 127°F. The analysis also assumes an initial accumulator water temperature of 120°F (consistent with the current TS limit on containment average temperature) and an RWST initial water temperature of 110°F. Operational data shows the RWST to typically be below 95°F.

SNC has evaluated the impact of allowing a containment average air temperature 2°F higher than the TS limit on the maximum calculated containment pressure in FSAR Chapter 6 and on the dose analyses in FSAR Chapter 15. The margins in these analyses are sufficient to bound the impacts of a containment average air temperature 2°F higher than the TS limit (which could result in the accumulator temperature increasing to 122°F) in conjunction with a reduced RWST water temperature.

Details of the large break LOCA evaluation are provided in Attachment 5 (non-proprietary) and (proprietary). As noted therein, the results of the estimate of effect for the requested change will be tracked pursuant to 10 CFR 50.46(a)(3) and included in the FNP FSAR Subsection 15.4.1.5.3.

The requested removal of an expired Limiting Condition for Operation (LCO) Note is an administrative change and has no impact on safety. The Note was only applicable through September of 2023 and has no further bearing on operation of the plant.

3.2 Defense-in-Depth Defense-in-depth (DID) is often characterized by varying layers of defense, each of which may represent conceptual attributes of nuclear power plant design and operation or tangible objects such as the physical barriers between fission products and the environment.

The defense-in-depth provided for the units is maintained. The criteria identified in Regulatory Guide 1.17 4, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis", section 2.1.1.2, are addressed below.

1. Preserve a reasonable balance among the layers of defense.

The defense-in-depth (DID) design features continuing to provide the layers of defense applicable to containment heat removal include:

Containment temperature is maintained by using four safety related forced air heat exchangers. The containment cooling system consists of four containment air coolers, each with a one-third cooling capacity during normal operation, with up to four units operating. Each air cooler consists of a fan and finned tube coil supplied by water from the service water system. As the post-accident containment atmosphere, which consists of a steam-air mixture, is circulated through the bank of cooling coils, it is cooled and a portion of the steam is condensed. The capacity of one cooler in conjunction with one E-5

Enclosure to NL-24-0281 Evaluation of the Proposed Change containment spray train is adequate to maintain pressure and temperature within containment structural design limits.

The containment cooling and ventilating functions are augmented by the containment recirculation fans, which take suction from the containment dome and discharge downward to help provide mixing of the containment atmosphere during normal operation to augment heat removal and maintain uniform temperature distributions throughout the containment volume.

The control rod drive mechanism (CROM) cooling system consists of fans and ducting to draw air through the CROM shroud and eject it to the main containment atmosphere.

One hundred-percent redundancy is provided by a standby fan.

The reactor vessel support cooling system, consisting of two 100% capacity fans and ducting, is arranged to cool the reactor vessel supports by drawing air through the supports. One hundred percent redundancy of the active components is provided.

The containment spray system has been designed to spray water into the containment atmosphere, when appropriate, in the event of a MSLB or LOCA, to ensure the containment peak pressure is below its design value. This function can be accomplished by one of the two trains of containment spray.

Post-OBA, after the injection operation, water collected in the containment sump is cooled and returned to the RCS by the low-head/high-head recirculation flow paths.

The containment heat removal systems are designed such that the failure of any single active component, assuming the availability of either onsite or offsite power exclusively, does not prevent the systems from accomplishing their design safety functions.

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

The proposed licensing basis change includes remedial actions discussed below accompanied by the compensatory measures discussed in Section 3.5 of this evaluation.

The proposed licensing basis change remedial actions include additional initial and periodic confirmations along with a change to the completion time for the "restore" action included in the proposed TS Required Actions. These include confirmation that:

the containment average air temperature remains less than or equal to 122°F (verified within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter);

the refueling water storage tank temperature remains less than or equal to 100°F (verified within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter); and the containment average air temperature has not exceeded the 120°F limit for more than 7 days cumulative during the calendar year.

These are typical TS Required Actions related to a parameter temporarily not within the limits of the TS Limiting Condition for Operation, and thus, the reliance on programmatic activities as compensatory measures, is not excessive (i.e., not overly reliant). Since these actions are intended to confirm the plant remains within the considerations of the evaluation for the TS change on the capability of the design features, the use of such measures does not significantly reduce the capability of the design features (e.g., hardware).

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Enclosure to NL-24-0281 Evaluation of the Proposed Change Maintaining the containment average air temperature less than or equal to 122°F, and the RWST temperature less than or equal to 100°F, is consistent with the alternative deterministic analyses provided in this section of the Enclosure. Maintaining the containment average air temperature less than or equal to 122°F also maintains consistency with the supporting risk analysis.

The 8-hour Completion Times to verify these parameters are reasonable, based on operating experience, to verify the containment average air temperature and RWST temperature are less than or equal to the identified limit within the Required Action.

Limiting the time the containment average air temperature has exceeded the 120°F limit to 7 days cumulative during the current calendar year is conservative with respect to the risk analysis provided in Section 3.4 of this Enclosure. Presuming a single prolonged period of high temperatures, this would effectively limit the exceedance of the 120°F limit to 7 days.

If any of these required verifications are found to exceed the identified limits, Action B would lead to a shutdown of the unit.

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

System redundancy, independence, and diversity are addressed by the multiple functions identified in response to item 1 of this DID section (above) which show that system functions are not reliant on any single feature of the design.

4. Preserve adequate defense against potential common cause failures (CCFs).

While initial containment temperature is a consideration in equipment qualification and reliability for components located inside containment, engineering judgment is that minor, short-term excursions of temperatures do not impact the capabilities of qualified equipment, and thus, a 2°F for up to 7 cumulative days per calendar year is not expected to simultaneously have an adverse effect on multiple components important to safety.

5. Maintain multiple fission product barriers.

Fission product barriers considered consist of the fuel cladding, the reactor coolant system (RCS) pressure boundary, and the containment. These barriers are not challenged by the proposed change.

Fuel Integrity: The evaluation of the impact of allowing a potential increase in the containment average air temperature and accumulator temperature by 2°F for up to 7 cumulative days in the calendar year indicates a maximum of 2°F increase in peak clad temperature (PCT) as a result of a large break LOCA. The latest rack-up of Farley PCT is 2034 °F, so the PCT rack-up would increase to 2036°F under this containment temperature assumption. The increased PCT would still be less than the regulatory acceptance criterion of 2200°F.

The evaluation of the impacts of a small break LOCA show a 0°F impact.

Based on the relatively small impact of a 2°F increase in the accumulator fluid associated with a 122°F initial containment temperature (the increase in accumulator temperature corresponds to an enthalpy increase of ~2 Btu/lbm [~2.23%]), accumulator initial injection timing and characteristics remaining unaffected, and the stored energy associated with a small break transient, it is concluded that temporarily increasing the maximum containment temperature from 120°F to 122°F will have a negligible impact on the small break LOCA E-7

Enclosure to NL-24-0281 Evaluation of the Proposed Change analysis of record, leading to an estimated peak cladding temperature impact of 0°F and a negligible impact on the maximum local oxidation reaction on the cladding surfaces.

Details of the large break LOCA PCT evaluation are provided in Attachment 5 (non-proprietary) and Attachment 6 (proprietary).

RCS Pressure Boundary: The proposed increase in containment temperature is within the margins for the design of the RCS pressure boundary and thus, the RCS pressure boundary would not be impacted by the allowance for containment average air temperature to exceed its current limit of 120°F up to a maximum of 122°F for up to 7 cumulative days in the calendar year.

Containment: The proposed allowance for containment average air temperature to exceed its current limit of 120°F up to a maximum of 122°F for up to 7 cumulative days in the calendar year is within the margins for the containment structural design. The pressures and temperatures post-LOCA and post-MSLB have been evaluated at a higher temperature (127°F). The evaluation of conservatisms in the containment response analysis shows that existing margin in the analysis bound the impact of the potential for the accumulator temperature to increase from 120°F to 122°F. Thus, the containment fission product barrier would not be impacted by the proposed change.

6. Preserve sufficient defense against human errors.

Advance consideration of the actions necessary to respond to off-normal conditions and accidents are generally offset by the use of procedures and training. The discussions in response to considerations 1 through 5 above identify the continued effectiveness of the systems and the fission product barriers. The programmatic activities, i.e., remedial actions and compensatory measures, discussed above and in Section 3.5 are to be implemented through TS, procedures, and training on those revised TS and procedures. The layers of defense within the plant design and operation of the plant in accordance with accepted practices (of predetermined, proceduralized actions) preserve sufficient defense against the occurrence of human errors.

7. Continue to meet the intent of the plant's design criteria.

There are no changes to the design or performance of any plant system. There are no changes to the redundancy inherent in the containment heat removal design. Existing conservatisms in analysis methodology and limiting the RWST temperature will offset the additional stored energy in the accumulators due to increased temperature (refer to discussion in the following section), therefore, there are no significant changes to the mass and energy released into containment during an event. Therefore, the plant design provides reasonable assurance of the continued availability of the containment heat removal systems to perform their intended function after an anticipated operational occurrence or a postulated design-basis accident. Thus, the proposed licensing basis change continues to meet the intent of the plant's design criteria.

Considering items 1 through 7 above, allowing containment average air temperature increase above 120°F to 122°F for up to 7 cumulative days per calendar year does not impact the layers of defense-in-depth inherent to the containment heat removal systems. Thus, the proposed change maintains the defense-in-depth capability.

3.3 Safety Margins The safety margins provided for the units are maintained. The criteria identified in Regulatory Guide 1.17 4, section 2.1.2, are addressed below.

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Enclosure to NL-24-0281 Evaluation of the Proposed Change As noted in the Regulatory Guide, "With sufficient safety margins, (1) the codes and standards or their alternatives approved for use by the NRC are met and (2) safety analysis acceptance criteria in the licensing basis are met or proposed revisions provide sufficient margin to account for uncertainty in the analysis and data."

The codes and standards, and any alternatives approved for use by the NRC, continue to be met. No changes to the codes and standards are proposed.

SNC has evaluated the impact of the proposed increase in the containment average air temperature on the maximum calculated containment pressure and temperature in FSAR Chapter 6 and on the dose analyses in FSAR Chapter 15. SNC evaluation of the temperature instrumentation accuracy shows a +/-2.5°F uncertainty in the instrumentation.

The containment response analysis for LOCA is evaluated at a starting temperature of 127°F.

SNC evaluation of the temperature instrumentation accuracy shows a +/-2.5°F uncertainty in the instrumentation. Given the containment temperature accuracy, the assumed temperature continues to provide appropriate margin to the proposed Required Action limit of 122°F. The evaluation of conservatisms in the containment response analysis shows that existing margin in the analysis bound the impact of increasing the accumulator temperature from 120°F to 122°F (see detailed explanation below). Therefore, the margins in these analyses are sufficient to bound the impacts of the proposed 2°F increase in containment average air temperature. In addition, the current containment pressure evaluation post-LOCA (that develops Pa) is unaffected; thus, there is no impact on the post-LOCA doses with an initial containment average air temperature of 122°F.

The Containment response analysis for MSLB currently also assumes an initial containment average air temperature of 127°F. Given the containment temperature instrumentation accuracy, the proposed temperature continues to provide appropriate margin. There are no other MSLB analysis inputs that are affected by an increase in containment average air temperature. For a MSLB, the current analyses of record are bounding for the proposed containment initial temperature allowance. The limiting radiological consequences associated with an MSLB involve a break outside containment and are therefore not impacted by an initial containment average air temperature of 122°F.

The LOCA analysis assumes an accumulator liquid temperature of 120°F, which could be exceeded if containment average air temperature increased above 120°F. However, the LOCA analysis assumes a Refueling Water Storage Tank (RWST) initial temperature of 110°F. To address the uncertainty in the accumulator temperature when exceeding the TS 3.6.5 limit of 120°F, the proposed change will impose a verification that the RWST does not exceed 100°F.

Operational data shows the RWST to typically be below 95°F, and therefore this limit does not impose a challenge to plant operations.

Two limiting LOCA break cases are of interest, the hot leg break and the double ended pump suction guillotine (DEPSG) break.

The hot leg break case is bounded by the DEPSG case. As a result, evaluation of the impact of this change was performed against the bounding DEPSG case.

For the DEPSG case, increasing accumulator initial liquid water temperature by 2°F corresponds to an energy increase of 382,000 Btu. The increase in accumulator energy is more than offset by assuming an RWST initial temperature decrease from 110°F to 100°F, resulting in a decrease in the integrated break energy at 3600 seconds by 10.59E6 Btu. This is a net total decrease in energy into the containment of 10,208,000 Btu. The analyzed mass and energy releases at an accumulator initial temperature of 120°F and RWST initial temperature of 110°F E-9

Enclosure to NL-24-0281 Evaluation of the Proposed Change would remain bounding for a set of initial conditions where the accumulator temperature has increased from 120°F to 122°F and the RWST temperature has decreased from 110°F to 100°F. Therefore, the impact on containment pressure from a potential increase in accumulator temperature to 122°F is bounded by the conservative margin between RWST operating temperature and its analysis assumed initial temperature. As a result, there would be no appreciable increase in post-LOCA containment pressure from what's analyzed.

Based on these evaluations, there is no anticipated increase in the containment vapor temperature, sump temperature, or containment peak pressure following a MSLB or LOCA.

Therefore, containment releases will remain within the assumptions of the calculated offsite doses.

Thus, the proposed change maintains the safety margins.

Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions SNC response to Generic Letter 96-06, "Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions," continues to be based on the analysis of record and remains unaffected. The analyses determine that the water hammer pressure spikes at the containment air coolers are not significant enough to cause damage, demonstrating that the plant is not susceptible to the Generic Letter 96-06 concerns. An allowance for containment average air temperature to exceed its current limit of 120°F up to a maximum of 122°F for up to 7 cumulative days in the calendar year has a negligible impact on the current analysis. The waterhammer analysis identified that the region of the service water piping that is most susceptible to water hammer is the containment cooler return piping with lowest possible system backpressure. However, this analysis showed that no waterhammer will occur within this piping. Following a LOCA coincident with a LOSP, both the containment cooler fans and service water pumps are de-energized resulting in reduced air and service water flows through the coolers. Therefore, the heat transfer from the containment atmosphere will be less than the full capacity of the containment coolers. In the time interval of interest (25 seconds or less) following initiation of this event, the service water downstream of the containment coolers has been calculated to reach a maximum of 119°F. Even assuming that the service water downstream of the containment coolers picks up an additional 2°F based on the increased initial containment average air temperature, the maximum temperature will still be less than the 164 °F temperature required to form a vapor cavity.

Service water temperature, along with its associated vapor pressure, will rise rapidly following containment cooler fan restart. However, the increase in the service water's vapor pressure, caused by the service water's temperature rise, will not reach the increased pressure in the containment cooler discharge piping. Therefore, two phase flow conditions still will not occur.

Post-LOCA Subcriticality Assessment Post-LOCA subcriticality analyses minimize liquid mass inventories for boration sources such as accumulators. A potential increase of the maximum accumulator temperature to 122°F will change the accumulator mass slightly due to the density change. Due to the level of precision used for the density in those calculations and the small change in temperature (2°F), there would be no measurable impact to the resulting accumulator mass used in the subcriticality calculations.

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Enclosure to NL-24-0281 Evaluation of the Proposed Change Post-LOCA Sump Dilution and Hot Leg Switchover Assessment The accumulator mass is based on a higher density that is not associated with the maximum accumulator temperature. As a result, there is no impact to the post-LOCA sump dilution calculation due to the increase in maximum accumulator temperature. The hot leg switchover analysis used the same accumulator mass as that discussed above for post-LOCA sump dilution. As noted above, this accumulator mass is not impacted by an increase in accumulator temperature and as a result there is no impact to the hot leg switchover analysis.

Post-LOCA Decay Heat Removal Assessment Any minor changes in core voiding and core boil-off rates resulting from the 2°F accumulator temperature increase are relatively short term effects that do not persist into the long-term cooling phase of the emergency core cooling system (ECCS) performance evaluations.

Net Positive Suction Head (NPSH) Evaluation The NPSH calculations "NPSH Calculation from Containment Sump to the Residual Heat Removal (RHR) Pumps - Recirculation Mode" and "NPSH Calculation from CTMT Sump to the CTMT Spray Pumps - Recirculation Mode" identify the available NPSH margin at sump temperatures between 120°F and 291 °F. The NPSH margins up to 180°F are greater than 14 feet of head. The strainer head losses are also shown to decrease as the sump temperature increases above 140°F. Based on the competing effects between vapor pressure of the sump inventory and strainer head losses, the pump NPSH margin would be expected to increase or stay the same as the sump temperature increases above 212°F.

The impact of the temporary allowance for containment temperature increase on the LOCA sump temperature response is minimal, and therefore the effect of the increase in containment air temperature on the available NPSH for the pumps that draw water from the sump during the LOCA recirculation phase is also minimal. A bounding GOTHIC evaluation determined a LOCA sump temperature increase of no more than 0.34 °F during this time, leading to an available NPSH margin of more than 1 foot for the Residual Heat Removal and Core Spray pumps in recirculation mode at a temperature of 292.4 °F. The bounding high-end temperature of 292.4 °F is conservative and is the saturation temperature corresponding to the maximum containment peak pressure that occurs during a LOCA of 45 psig (59.7 psia). The temperature of 292.4°F utilizes the guidance in Regulatory Guide 1.82, "Water Sources for Long-Term Recirculation Cooling Following a Loss-Of-Coolant-Accident," that the containment pressure is equal to the vapor pressure of the sump inventory. This ensures that credit is not taken for containment pressurization during the transient.

Regarding biased inputs for a conservative analysis, the GOTHIC evaluation assumed a containment air temperature of 127°F, and an accumulator water temperature of 124 °F. As discussed in the LAR, the containment average air instrument uncertainty calculation demonstrates sufficient margin to the assumed safety analyses initial condition of 127°F to account for an increase from 120°F to 122°F. The assumed accumulator water temperature of 124 °F (4 °F increase from the FSAR analysis) bounds the expected response from the 2°F increase in containment air temperature. The GOTHIC evaluation did not account for the restriction on RWST temperature (i.e., it assumed 110°F versus 100°F). As discussed in the preceding paragraph, various conservative assumptions/biases were used to verify adequate NPSH margin based on the calculated change in containment sump temperature.

The GOTHIC code version used for the NPSH analysis (Version 8.1) was previously accepted by the NRC (ML22263A225).

E-11

Enclosure to NL-24-0281 Evaluation of the Proposed Change The GOTHIC code analysis of record utilizes an initial containment average air temperature of 127°F, an initial accumulator water temperature of 120°F, and an initial RWST water temperature of 110°F.

Updated LOCA mass and energy releases for accumulator temperature of 124 °F and an RWST temperature of 110°F were developed in accordance with the approved methodology for Plant Farley (WCAP-10325-P-A) and as approved by NRC (ML22263A225). These mass and energy releases were used as input in the GOTHIC analysis for evaluating the impact of this change.

The GOTHIC case used to evaluate how the proposed change in the Containment Average Air Temperature impacts LOCA sump temperature response did not apply additional uncertainties to the containment average air temperature and the accumulator water temperature. However, the chosen inputs provide additional margin (2 degrees for accumulator temperature, i.e., 124 °F input, and 5 degrees for containment temperature, i.e., 127°F input) above the 122°F air and water temperatures of concern. The initial accumulator water temperature was the only change (via updated mass and energy releases) from the analysis of record (ML22263A225) used to evaluate the impact of the containment temperature increase.

NPSH margin is unaffected for sump temperatures above the point at which containment (CTMT) has reached saturation conditions and the RHR pumps and the CTMT spray pumps are initiated. Table 3.3-1 identifies the available NPSH, required NPSH, and the NPSH margin.

This table was previously provided to NRC (ML22171A010) with prior amendment request.

Table 3.3-1 Pump NPSHA {ft)

NPSHR {ft)

NPSH Margin {ft)

U1 A Train RHR 20.7 18 2.7 U1 B Train RHR 19.6 18 1.6 U 1 A Train CTMT Spray 20.8 18 2.8 U 1 B Train CTMT Spray 21.2 18 3.2 U2 A Train RH R 19.2 18 1.2 U2 B Train RHR 19.3 18 1.3 U2 A Train CTMT Spray 20.7 18 2.7 U2 B Train CTMT Spray 19.1 18

1.1 Notes

NPSHA = Minimum net positive suction head (NPSH) available for the pumps drawing water from the sump during the LOCA recirculation phase.

NPSHR = Maximum NPSH required for the pumps at their respective applicable flow rates.

Minimum NPSH margin = NPSHA-NPSHR The minimum NPSH margin occurs in a sump water temperature range of 212-291 °F where the minimum values are the same.

NPSHR is 18 ft for all pumps at flow rates of 4500 gpm for RHR and 3400 gpm for CTMT sorav.

No impact is identified to the loss-of-coolant-accident (LOCA) sump temperature response E-12

Enclosure to NL-24-0281 Evaluation of the Proposed Change available net positive suction head (NPSH) during LOCA recirculation phase. Thus, the containment accident pressure does not need to be credited during the transient to maintain positive NPSH margin. No saturation pressure impact is identified. Sufficient margin exists to accommodate the minimal temperature increase.

Therefore, an increase in the initial containment average air temperature from 120°F to 122°F would be expected to have no adverse impacts on NPSH margin.

Service Water and Ultimate Heat Sink Evaluation The ultimate heat sink (UHS) analysis for the service water pond considers various combinations of units in shutdown and accident conditions. The service water discharge from both units is aligned back to the pond, resulting in the pond absorbing decay heat from both units for the potential 7-day period. The small increase in containment temperature at the start of the event represents an insignificant effect compared to the magnitude of heat transferred to the service water from both units.

While the UHS (i.e., the source for SW) temperature might be expected to rise in conjunction with the projected increased ambient temperatures, SNC does not anticipate the ultimate heat sink to exceed its Technical Specification limit of 95°F (Surveillance Requirement 3.7.9.2),

which is consistent with analysis assumptions. This LAR does not request a change the TS limits or required actions associated with the UHS temperature. As such, there will be no impact as a result of an increase in SW temperatures within the TS limits on the evaluations provided in the requested amendment.

No impact is identified on the environment or the ultimate heat sink caused by the additional heat load. Sufficient margin exists to accommodate the minimal temperature increase.

Based on the service water temperature evaluation, the containment fan cooler performance used in the containment response analysis is also not impacted.

Environmental Qualification (EQ) Evaluation A potential 2°F increase in the containment average air temperature is bounded by the existing environmental qualification (EQ) analyses, due to conservatisms and margins in the existing test programs and calculations. The temperature used within the calculation is >125°F which bounds the current Technical Specification temperature of 120°F and the potential increase to 122°F.

Thus, an increase from 120°F to 122°F in the containment average temperature for up to 7 cumulative days in the calendar year will have no impact on the qualification status or qualified lives of existing equipment located in containment in the EQ Program scope.

In accordance with EQ Program, the program scope includes electrical equipment that is important to safety. Equipment that is important to safety involves safety related and non-safety related electrical equipment whose failure can prevent satisfactory accomplishment of safety functions as described in 10 CFR 50.49 (b)(1) and (b)(2) and certain post-accident monitoring equipment as described in 10 CFR 50.49(b)(3).

In addition, no impact has been identified for normally operating equipment within containment resulting from the potential increase in temperature. SNC has reviewed the specifications of the non-EQ electric equipment (i.e., electric equipment not subject to the requirements in 10 CFR 50.49) within containment that is expected to perform a design function under normal operation (e.g., electrical equipment that is either relied upon by the plant operators to inform operational decisions or provides a signal input to other plant systems or processes) whose failure could mislead a plant operator or cause a plant transient and determined that this electrical equipment will not be adversely impacted by the proposed temporary increase in plant temperature.

E-13

Enclosure to NL-24-0281 Evaluation of the Proposed Change Instrumentation Uncertainties Evaluation The containment average air instrument uncertainty calculation demonstrates sufficient margin to the assumed safety analyses initial condition of 127°F to account for an increase from 120°F to 122°F.

The total Channel Statistical Allowance (CSA) or the total channel uncertainty for the containment average air temperature instrumentation is calculated to be +/-2.5°F when statistically combining 4 channels. The potential containment average air temperature increase to 122°F will continue to allow for adequate protection of the Safety Analysis Limit (SAL) of 127°F. Given the CSA of+/- 2.5°F, measuring 122°F still provides an additional margin of 2.5°F to 127°F.

The methodology used is the square root of the sum of the squares (SRSS) of independent components which is widely utilized in the industry. The use of probabilistic and statistical techniques to determine safety related and non-safety related set points has been endorsed by various industry standards. In particular, the methodology used in channel uncertainty for the containment average air temperature instrumentation calculation is consistent with the methodology in WCAP-13751, Westinghouse Setpoint Methodology For Protection Systems Farley Nuclear Plant Units 1 and 2 (which the NRC found acceptable for use in deriving Reactor Trip and Engineered Safety Feature Actuation System setpoints [ML013130715]).

Impact on Personnel and Operator Actions Time Critical Operator Actions (TCOAs) and Time Sensitive Operator Actions (TSOAs) were also evaluated, and it was determined that the change in Containment Temperature has no effect on the failure rate of the Human Error Probability (HEP) of concern.

The proposed amendment to extend the TS Actions completion time upon exceeding the LCO limit for the containment average air temperature does not affect the operation of the assumed mitigation systems or the containment fission product barrier assumptions. The potential for containment average air temperature to exceed its current limit of 120°F up to a maximum of 122°F for up to 7 cumulative days in the calendar year is within the existing margins in the safety analyses.

Other General Considerations Farley has installed Generation Ill RCP seals, adhering to PWROG-14001-P, Revision 1, including exceptions for Limitations and Conditions, as approved in amendments 217 and 214 (ML17261A087), respectively, for the containment leakage rate testing program.

No additional human actions or extended response times are identified to complete specified actions during a design basis accident.

No impact is identified on the capacity and capability of the emergency diesel generators.

Sufficient margin exists to accommodate the minimal temperature increase.

The proposed amendment does not alter any plant equipment or operating practices with respect to such initiators or precursors in a manner that the frequency of an accident would be increased. The internal containment average air temperature is not associated with an accident initiator or with an initiating sequence of events and no impact has been identified for normally operating equipment within containment resulting from the proposed increase in temperature.

E-14

Enclosure to NL-24-0281 Evaluation of the Proposed Change 3.4 Probabilistic Risk Assessment (PRA) 3.4.1 PRA Model Impacts The PRA Model is evaluated to identify the change in risk associated with Containment Temperature >120°F and ::;122°F. The Containment Temperature is postulated to be 122°F during the entire action statement window of 7 days. Containment Temperature would, however, be expected to slowly increase from the entry condition of 120°F toward the limiting temperature of 122°F. This assumption provides a conservative bias to the risk evaluation since it is difficult to predict how quickly Containment Temperature could increase.

Internal events human failure events (HFEs) and success criteria were screened and evaluated by rerunning associated MAAP cases at the higher initial Containment Temperature condition.

The potential impact to HFEs by increase Containment Temperature would impact all hazards uniformly since the HFEs are credited to mitigate initiating events, which are consistent among the different hazards.

Personnel are not required to enter containment to mitigate a transient, accident, or natural hazard. Personnel's capability is evaluated within Farley's Internal Events HRA Post-Initiators and Dependency Analysis Notebook. The capability of operators is analyzed within the Timing Analysis and is evaluated based on the timing required to complete the mitigating action. The time required is broken into the cognition time and execution time. The cognition time consists of detection, diagnosis, and decision making. The execution time includes the travel time, collection of tools, donning of personnel protective equipment, and manipulation of components.

The initial Containment Temperature affects none of the previously mentioned framework for the capability of operators. The operator actions that had the potential of being affected by the change in Containment Temperature and not eliminated by the PRA screening method (discussed in section 3.4.3) are shown in Table 3.4-1.

Table 3.4-1 HFE Description Initiator MAAP Input of Interest OAC-AF-Operator Fails to T sw is 191 min based on timing ISOLAFW-Isolate SG Feedwater SSB of peak Containment Pressure FAULTSG to Faulted Generator (T REc = 168 min)

T sw is 52.25 min based on timing Operators fail to align of Core Exit Temperature OAC-LH-RECI RC-ECCS low head LLOCA reaching 700°F and starting one LLOCA recirculation - L LOCA train of RHR injection in recirculation mode (T REC = 19.05 min)

Operators fail to align T sw is 438 min based on timing OAC-LH-RECIRC-ECCS low head MLOCA of core damage without transfer MLOCA recirculation - Medium SLOCA to recirculation mode LOCA (T REc = 138.3 min)

Operator Fails to T sw is 191 min based on timing OAL-AS-Locally Close TDAFW CLOSEV017ABC to SG Isolation Valve SSB of peak Containment Pressure Q 1 N23V0 17 A/B/C (T REc = 157 min)

E-15

Enclosure to NL-24-0281 Evaluation of the Proposed Change The Farley maximum acceptable containment temperature is used as a design input for the Internal Events PRA. This temperature input is used as an initial condition for MAAP analysis that supports and provides input to key elements of the PRA.

MAAP analysis is used as the tool to determine several aspects of the Farley PRA. This includes the Success Criteria (determination of the minimum equipment available to prevent core damage), accident sequence (determine of the plant response and verification of the beyond design basis accident mitigation strategies), human reliability analysis available timing, and support in determining the timing and release size to the environment following an accident.

The credited cases were reviewed comparing the base cases with the updated containment temperature case. For cases that may have had an impact on the PRA, or a part of the PRA, analyses were completed to determine what the values should be in the updated PRA model.

The change in core damage frequency (CDF) and large early release frequency (LERF) was then calculated to determine the overall effect of the various changes updated in the PRA model.

3.4.2 PRA Model Acceptability FNP Units 1 and 2 has previously implemented an NRC-approved risk-informed license amendment to permit the use of Risk-Informed Completion Times (RICTs) in accordance with Nuclear Energy Institute (NEI) 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines," Revision 0-A [ADAMS Accession No. ML19175A243].

FNP Units 1 and 2 has also implemented an NRC-approved amendment [ADAMS Accession No. ML21137A247] to adopt the 10 CFR 50.69 categorization process (including an integrated decision-making panel or IDP) as described in the SNC license amendment request [ADAMS Accession No. ML201708114].

Acceptability of the Farley PRA model was most recently assessed by NRC staff in the approval of the use of 10 CFR 50.69 for both units. The safety evaluation for use of 10 CFR 50.69 states that the peer review findings are considered fully resolved. A Focused Scope Peer Review (FSPR) was performed in January 2023 to review an upgrade to the Farley PRA. This included revisions to the Fire PRA to incorporate the updated methods provided in NUREG-2230 and NUREG-2178.

An approved process for performing a facts and observations (F&O) closure review is provided by Appendix X to NEI 05-04/07-12/12-06 [ML16158A035]. The process allows several options.

The review documented in this report is based on "Closeout F&Os by Independent Assessment" option. This process is similar to a Peer Review following NEI 05-04 but with a scope limited to evaluating the closure of F&Os. In addition, the NRC has provided some specific expectations when using this process that are included in the conduct of this assessment. It should be recognized that this process does not permit the closure of F&Os when the resolutions are assessed as a PRA upgrade as defined by the ASME standard. Therefore, this review only closes F&Os when the resolution is determined to be PRA maintenance. It is noted that a newer version of the peer review process has been developed as NEI 17-07 and has been endorsed for use by RG 1.200. SNC elected to use the process as documented in Appendix X. The differences do not impact the validity of the review.

Southern Nuclear Company (SNC) addressed and closed all open F&Os resulting from the FSPR performed in January 2023.

Some notable model maintenance items which have been implemented since the approval of 10 CFR 50.69 include:

E-16

Enclosure to NL-24-0281 Evaluation of the Proposed Change Updates to FLEX modeling including adoption of PWROG-18042 FLEX equipment data and addition of credit for FLEX pumps to mitigate certain reactor coolant pump (RCP) seal leakage sequences if the low leakage RCP seal fails. The updates were performed consistent with the 2022 NRC staff memorandum for crediting FLEX strategies in PRA.

Minor logic edits to close out Model Maintenance Log Items SNC is not proposing to use any newly developed PRA methods for this application. The approved PRA methods are documented according to the requirements for each specific PRA hazard.

3.4.3 PRA Modeling of the Proposed Change The scope of the Farley PRA model for the assessment of the increase in Containment Temperature consists of Internal Events, Internal Flooding, and Fire as approved in the safety evaluation for 10 CFR 50.69. The Farley PRA Seismic model, though Farley being a tier 1 plant, was also evaluated for Seismic insights from the impact of the increase in Containment temperature. The credited cases were reviewed comparing the base cases with the updated containment temperature case. For cases that may have had an impact on the PRA, or a part of the PRA, analyses were completed to determine what the values should be in the updated PRA model. The change in core damage frequency (CDF) and large early release frequency (LERF) was then calculated to determine the overall effect of the various changes updated in the PRA model.

A review of the current Human Reliability Analysis (HRA) Calculator file and the Success Criteria PRA Notebook was performed to determine which Human Failure Events (HFEs) and Success Criteria may be affected by an increase in initial Containment Temperature. For example, an increase in initial Containment Temperature was postulated to affect the timing of Containment Spray actuation which could affect the timing of emergency core cooling system (ECCS) recirculation mode related HFEs.

The HRA events were screened as follows:

Any HFEs calculated with the screening method are screened from further analysis.

Any HFEs without a supporting MAAP analysis used for timing are screened due to Containment Air Temperature only being an input to MAAP cases.

Any HFEs related to steam generator overfill are screened due to the possibility of a steam line break being eliminated. Steam generator overfill would cause water to be in the place of steam preventing a steam line break from occurring.

Any HFEs with long time windows and recovery timings (>1 hr) are screened due to those HFEs not being time sensitive enough to impact the success rate of completing those actions.

Using the above screening criteria, the remaining HFEs with MAAP parameters of interest were identified and subsequently evaluated by rerunning the associated MAAP case. Three MAAP cases were used to support the evaluations affected by the change in Containment Temperature:

MSLBX - Main Steam Line Break inside containment with one of the three SGs feeding the break. In addition, one train of ECCS and containment fan coolers are successful, containment E-17

Enclosure to NL-24-0281 Evaluation of the Proposed Change sprays fail, and AFW to the faulted steam generator is isolated at 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months . Transfer to recirculation is not modeled and all ECCS stops at the lo-lo RWST level.

LLOCA-nofcs-cs-norec - Double-ended cold leg break with two trains of ECCS and 3 accumulators available. In addition, containment fan coolers are assumed to fail, but both containment spray trains are successful and begin to inject automatically on high containment pressure. No AFW is available and transfer to recirculation fails and all pumps take suction from the RWST until the RWST is empty.

MLOCA inch diameter cold leg break with two trains of ECCS available. In addition, two accumulators and two containment spray pumps are available along with one train of containment fan coolers. ECCS pumps are assumed to fail when the RWST reaches the Lo-Lo RWST level while containment spray pumps successfully transfer suction to the containment sump.

MAAP cases require user defined variables that Containment Temperature plays a role in and are shown in Table 3.4-2.

Table 3.4-2 Figure of Merit User-Defined Variable Time of steam generator dry out TI_SG_DRY1 Time of core uncover Tl_ CR_ UNCO _L Time to onset of core damage TI_CR_ 1800f Time to reach Lo-Lo RWST water level Tl RWLL Time to RWST empty for cases without transfer to recirculation Event code 187 TRUE Maximum containment pressure PR_CN_MAX2_PSI Maximum core temperature for cases without core damage TE_CR_MAX_F 3.4.4 Risk Results Results show that there is negligible impact on the MAAP conclusions (timing, success criteria) currently used in support of the development of the PRA model for both CDF and LERF.

Similarly, the post-initiator operator actions currently credited in the PRA were reviewed for different parameters (timing, crew composition, cues, procedures, pathways, and training) and are also considered to have a negligible impact to HEPs and therefore to CDF and LERF.

Containment Spray would likely occur earlier in a LOCA scenario, but RWST cues are still available to inform operators of low level. Since there is minimal impact to peak containment pressures and containment sump water temperatures, no changes to RHR success criteria are noted. The quantitative sensitivity analysis performed by placing more restrictive success criteria for containment fan coolers from 2/4 to 3/4 showed that there is a negligible impact to the CDF. They are not credited for LERF mitigation. In each of these cases it was determined that the proposed increase in containment temperature is expected to have negligible impact on E-18

Enclosure to NL-24-0281 Evaluation of the Proposed Change the PRA risk metrics. The MAAP analysis key results are shown in Table 3.4-3.

Table 3.4-3 MAAP Case Name MSLBX LLOCA-nofcs-MLOCA cs-norec Associated HPEs OAC-AF-ISOLAFW-FAULTSG OAC-LH-OAC-LH-RECI RC-RECIRC-LLOCA MLOCA OAL-AS-CLOSEV017ABC Containment temperature, 0P 120 122 120 122 120 122 Time of SG 1 dryout, min 214.1 214.1 N/A N/A N/A N/A Time of core uncover, min N/A N/A 0.1 0.1 8.1 7.8 Time to onset of core damage, min.

N/A N/A 63.0 62.9 438.3 432.5 Time of Lo-Lo RWST level, min N/A N/A 27.2 27.2 370.9 370.9 Time of RWST empty, min N/A N/A 30.6 30.6 N/A N/A Max containment pressure, psia 74.0 74.2 71.8 70.4 39.5 39.3 Max core temperature for cases not 559.7 559.8 N/A N/A N/A N/A going to core damage, 0P In addition, the large and medium LOCA cases considered in the reports indicate no impact to the time to transfer to recirculation. Therefore, the RHR success criteria mission times are also unchanged as a result of a change in the containment initial average air temperature from 120°F to 122°F.

The results of the MAAP runs showed that none of the HFEs were substantially impacted by the change in initial Containment Temperature. The largest change occurred for the medium loss of coolant accident (LOCA) recirculation human failure event. Although the change results in an 8.6% decrease in the available time to perform the action, the human error probability result has no change due to the ample time (7.2 hrs) available for operators to transition to the recirculation mode to prevent core damage. Therefore, the HFEs are not impacted by this change.

E-19

Enclosure to NL-24-0281 Evaluation of the Proposed Change The Farley PRA Internal Events, Internal Flooding, Fire and Seismic models were quantified.

The quantification results show that the total ICCDP/.b.CDF and ICLERP/.b.LERF values for this proposed change are well within the Regulatory Guide (RG) 1.174 and RG 1.177 guidance criteria as shown in Table 3.4-4 and Table 3.4-5.

Table 3.4 Summa!}' of ICCDP/.b.CDFAVE Contribution IE IF Fire Seismic Total ICCDP iJ.CDFAvE 0.00 0.00 0.00 0.00 0.00 Unit 1 Unit 2 0.00 0.00 0.00 0.00 0.00 a e T bl 3 4 5 S ummary o f ICLERP/.b.LERFAVE C t "b f on n u 10n IE Flood Fire Seismic Total ICLERP iJ.LERFAvE 0.00 0.00 0.00 0.00 0.00 Unit 1 Unit 2 0.00 0.00 0.00 0.00 0.00 The ICCDP and ICLERP values for the proposed change are within the acceptance guidelines from RG 1.177 and represent no quantitative impact on plant risk. With no quantitative impact, the proposed change adds no additional cumulative risk to the currently implemented Risk Applications. Therefore, the.b.CDFAVE and.b.LERFAVE result in being in Region Ill for Figures 4 and 5 of RG 1.174 and support objectives of the risk-informed application, i.e., revising the TS LCO 3.6.5 action statements to allow containment average air temperature from ::;120°F to

122°F for a period not to exceed 7 cumulative days during a calendar year.

SNC has evaluated the cumulative risk impact of a 2°F increase in containment average air temperature consistent with the principles of RG 1.17 4 and RG 1.177. The issue was determined to contribute less than 1 x1 o-7/year to CDF and contribute less than 1 x1 o-8/year to LERF. The cumulative risk baseline remains less than 1 x1Q-4/year for CDF and less than 1 x1Q-5/year for LERF, with the impact of the proposed change is incorporated into baseline risk as shown in Table 3.4-6.

E-20

Enclosure to NL-24-0281 Evaluation of the Proposed Change Table 3.4 Summa!}' of Total Plant Baseline Risk including All Hazards Unit 1 Unit 1 Unit 2 Unit 2 CDF LERF CDF LERF Internal 2.69E-06 1.79E-08 3.18E-06 1.94E-08 Events Flood 6.22E-06 2.41 E-08 7.85E-06 3.99E-08 Fire 3.63E-05 3.49E-06 3.92E-05 2.93E-06 Seismic 2.94E-07 4.72E-08 2.97E-07 4.59E-08 Total 4.55E-05 3.58E-06 5.0SE-05 3.04E-06 3.4.5 Uncertainty Evaluations The quantification results show that the total ICCDP/.b.CDF and ICLERP/.b.LERF values for this proposed change are well within the RG 1.174 and RG 1.177 guidance. No compensatory measures were credited in the risk evaluation. The performance monitoring proposed in Section 3.5, coupled with the required action to maintain containment average air temperature to ::;122°F, reduces uncertainty associated with the risk analysis by ensuring an upper bound temperature consistent with the risk analysis.

RG 1.177, "Plant-Specific, Risk-Informed Decision making: Technical Specifications" Section 2.3.5, states that sensitivity analyses may be necessary to address the important assumptions in the submittal. As stated in RG 1.177, most sources of uncertainty for the PRA models have similar effects on the base case and the proposed completion time changed case.

The uncertainty analysis used to support the emergency TS change was performed by placing more restrictive success criteria for containment fan coolers from 2/4 to 3/4 to show that there is a negligible impact to CDF and LERF. This uncertainty analysis remains applicable, and the results are consistent with the determination of no impact to the deterministic analyses.

3.5 Implementation and Monitoring Plan During period(s) of operation approaching or exceeding 120°F, the following compensatory measures are proposed to be implemented as defense-in-depth efforts to prevent exceeding 122F.

Operate available containment coolers on high speed with service water aligned to the service water wet pit, i.e., to the emergency mode.

Operate available containment mini-purge continuously; Operate available containment recirculation fans in high speed; E-21

Enclosure to NL-24-0281 Evaluation of the Proposed Change Put in place work controls to prevent removal of containment cooling system components and supporting systems from service; and Put in place work controls to protect the containment cooling systems.

These actions may be initiated prior to exceeding 120°F consistent with best efforts to avoid exceeding the TS limiting containment average air temperature.

Operating the containment coolers, the mini-purge, and recirculation fans in this manner maximizes containment cooling. Aligning the service water system for the emergency mode to the containment coolers provides higher flow rates and also maximizes containment cooling.

The work controls maintain the systems in an operating condition until the high temperature situation is resolved.

The performance of the containment cooling systems is monitored in accordance with the requirements of the 10 CFR 50.65 Maintenance Rule and the compensatory measures identified above keep the systems in service during any periods of exceedance of the TS containment air temperature.

Should an emergency event occur, these systems will revert to their emergency alignment and function as assumed in the safety analysis.

3.6 Conclusions and Integrated Decision Regarding Safety Impact The deterministic evaluations provided in Section 3.1 demonstrate that with appropriate remedial actions of increased monitoring and a reduced maximum temperature for the refueling water storage tank, an increase in the containment average air temperature from 120°F to 122°F would not result in exceeding the impacts identified by the existing safety analyses. The risk evaluation provided in Section 3.4 demonstrate that the cumulative risk impact of the proposed exceedance of the containment average air temperature limit contribute less than 1 x10-7/year to CDF and contribute less than 1 x10-8/year to LERF.

The qualitative risk insights, integrated with considerations of defense-in-depth, safety margins, and other deterministic considerations provide reasonable assurance that the health and safety of the public will not be endangered by an exceedance of the containment average air temperature limit contained within the constraints of the proposed TS Required Actions.

Thus, with no significant impacts identified, SNC requests approval of the proposed change.

4.

REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria This activity involves changes to the operating license Appendix A, Technical Specifications; therefore, in accordance with 10 CFR 50.90, this activity requires an amendment. As such, NRC approval is required prior to making the proposed changes in this license amendment request.

10 CFR 50, Appendix A, General Design Criterion (GDC) 4, "Environmental and dynamic effects design bases," states, in part, that structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. SNC has evaluated that the potential to exceed the containment average air temperature above the current limit of 120°F up to a maximum of 122°F for up to 7 cumulative days in the calendar year will have no impact on the qualification E-22

Enclosure to NL-24-0281 Evaluation of the Proposed Change status or qualified lives of existing equipment important to safety located in containment in the Environmental Qualification Program scope.

10 CFR 50, Appendix A, GDC 38, Containment Heat Removal, requires a system to remove heat from the reactor containment. The change does not impact any containment heat removal functions, and therefore adequately satisfies the requirements of GDC 38.

10 CFR 50.36(c)(2) requires that TSs include Limiting Conditions for Operation (LCOs). Per 10 CFR 50.36(c)(2)(i), LCOs "are the lowest functional capability or performance levels of equipment required for safe operation of the facility." The regulation also requires that when an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TS until the condition can be met. The proposed change to the containment average air temperature limit continues to reflect the lowest functional capability required for safe operation and continues to provide appropriate remedial actions including a required plant shutdown if they are not met. Required Actions are proposed for this change rather than an LCO change since the "lowest functional capability or performance levels of equipment required for safe operation of the facility" has not changed, and the analysis of record has not changed. The requested exceedance of the LCO limit is temporary and remedial actions are appropriate during the exceedance. Per 10 CFR 50.36, the proposed remedial actions are appropriately included as Required Actions rather than as an LCO. These remedial actions are based on risk assessment, including considerations for defense in depth and safety margin. The assessment confirms the change in risk is low. The technical basis for these remedial actions is discussed in the LAR. These remedial actions, including verifying actual RWST temperature is at least 10°F less than what's assumed in the accident analysis and verifying containment temperature is less than the initial containment air temperature assumed in the limiting design basis accident analysis, provide a safety basis that's equivalent to or bounded by what's assumed in the accident analysis. Since the LCO is not being changed, and the remedial actions are supported as discussed in this request, the changes do not result in non-conservative TS as described in Regulatory Guide 1.239, Revision 0, and NEI, 15-03, Revision 3, "Licensee Actions to Address Nonconservative Technical Specifications."

10 CFR Part 50.49, "Environmental qualification of electric equipment important to safety for nuclear power plants," requires, in part, licensees to establish a program for qualifying the electric equipment important to safety. In accordance with FNP Environmental Qualification Program, the Environmental Program scope includes electrical equipment that is important to safety and evaluation of any temperature exceedance.

Regulatory Guide (RG) 1.155, Station Blackout, describes a means acceptable to the NRC staff for meeting the requirements of 10 CFR 50.63. NUMARC-87-00 also provides guidance acceptable to the staff for meeting these requirements. NUMARC-87-00, "Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors,"

Section 2.7.1, identifies the assumptions associated with loss of ventilation and indicates "Equipment Operability Inside Containment Temperatures resulting from the loss of ventilation are enveloped by the loss of coolant accident (LOCA) and high energy line break environmental profiles." Thus, the LAR discussions concluding that there is no anticipated increase in the containment vapor temperature provide continued compliance with the guidance of RG 1.155.

4.2 Precedent On August 19, 2010, the NRC issued Notice of Enforcement Discretion (NOED) for Southern Nuclear Operating Company (SNC) regarding Joseph M. Farley Nuclear Plant (FNP) Unit 1 (NOED No. 10-2-004) [ADAMS Accession No. ML102310595] for a temporary exceedance of the containment average air temperature limit from 120°F. This action included in part an SNC E-23

Enclosure to NL-24-0281 Evaluation of the Proposed Change commitment that the FNP Unit 1 would be shutdown if containment air temperature exceeded 122°F.

On August 24, 2023, the NRC issued Amendment Nos. 247 and 244 for Southern Nuclear Operating Company (SNC) regarding Joseph M. Farley Nuclear Plant (FNP) Units 1 and 2, respectively [ADAMS Accession No. ML23235A296] allowing a temporary increase in the containment average air temperature limit from 120°F to 122°F.

4.3 Significant Hazards Consideration Pursuant to 10 CFR 50.90, Southern Nuclear Operating Company (SNC) requests an amendment to Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2 renewed facility operating licenses NPF-2 and NPF-8, respectively. The requested amendment would revise the operating license, Appendix A, Technical Specification (TS) 3.6.5, Containment Air Temperature, actions upon exceeding the containment average air temperature limit of 120°F and remove an expired Note provision.

SNC has evaluated whether a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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

Response: No.

The proposed changes do not adversely affect the operation of any structures, systems, or components (SSCs) associated with an accident initiator or initiating sequence of events. The proposed changes do not affect the design of the containment heat removal systems.

The proposed amendment does not affect accident initiators or precursors nor adversely alter the design assumptions, conditions, and configuration of the facility. The proposed amendment does not alter any plant equipment or operating practices with respect to such initiators or precursors in a manner that the probability of an accident is increased. The proposed amendment to allow exceeding the containment average air temperature above the current limit of 120°F up to a maximum of 122°F for up to 7 cumulative days in the calendar year and remove an expired Note does not adversely affect the operation of the assumed mitigation systems or the containment fission product barrier assumptions. As demonstrated in the SNC request, the potential increase in containment average air temperature coupled with limiting the refueling water storage tank (RWST) temperature to ::;100°F is more than offset by existing margins in the safety analyses. As such, the proposed change will not alter assumptions relative to the mitigation of an accident or transient event. The proposed amendment does not increase the likelihood of the malfunction of an SSC or adversely impact analyzed accidents.

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

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

Response: No.

The proposed amendment does not introduce any new or unanalyzed modes of operation. The proposed changes do not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation.

The changes do not alter the limiting assumptions made in the safety analysis.

E-24

Enclosure to NL-24-0281 Evaluation of the Proposed Change Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3) Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The margin of safety is related to the ability of the fission product barriers to perform their design functions during and following an accident. These barriers include the fuel cladding, the reactor coolant system, and the containment. The fission product barriers continue to be able to perform their required functions; based on the pre-existing margins and conservatisms currently assumed in the safety analyses. Therefore, the margins to the onsite and offsite radiological dose limits are not significantly reduced.

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

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the 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.

ENVIRONMENTAL CONSIDERATION A review has determined that the proposed changes require an amendment to the operating license. A review of the anticipated effects of the requested amendment has determined that the requested amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9), in that:

(i) There is no significant hazards consideration.

As documented in Section 4.3, Significant Hazards Consideration, of this license amendment request, an evaluation was completed to determine whether or not a significant hazards consideration is involved by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment." The Significant Hazards Consideration evaluation determined that (1) the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated; (2) the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated; and (3) the proposed amendment does not involve a significant reduction in a margin of safety. Therefore, it is concluded that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of "no significant hazards consideration" is justified.

(ii) There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

The proposed changes are unrelated to any aspect of plant operation that would introduce any change to effluent types (e.g., effluents containing chemicals or biocides, sanitary system effluents, and other effluents) or affect any plant radiological or non-radiological effluent release E-25

Enclosure to NL-24-0281 Evaluation of the Proposed Change quantities. Furthermore, the proposed changes do not affect any effluent release path or diminish the functionality of any design or operational features that are credited with controlling the release of effluents during plant operation. Therefore, it is concluded that the proposed amendment does not involve a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite.

(iii) There is no significant increase in individual or cumulative occupational radiation exposure.

The proposed change in the requested amendment does not affect the shielding capability of, or alter any walls, floors, or other structures that provide shielding. Plant radiation zones and controls under 10 CFR 20 preclude a significant increase in occupational radiation exposure.

Therefore, the proposed amendment does not involve a significant increase in individual or cumulative occupational radiation exposure.

Based on the above review of the proposed amendment, it has been determined that anticipated effects of the proposed amendment do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criteria 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 amendment.

6.

REFERENCES None.

E-26 to the Enclosure of NL-24-0281 Technical Specification Page Markups Insertions Denoted by Blue Underline text and Deletions by Red Strikethrough (This Attachment consists of 3 pages, including this cover page)

3.6 CONTAINMENT SYSTEMS 3.6.5 Containment Air Temperature Containment Air Temperature 3.6.5 LCO 3.6.5 Containment average air temperature shall be ::::; 120° F.

Containment average air temperature shall be< 122°F until 0600 hours0.00694 days 0.167 hours 9.920635e-4 weeks 2.283e-4 months on September 9, 2023 APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Containment average air A.1 temperature not within limit.

AND A.2 AND A.34 B.

Required Action and B.1 associated Completion Time not met.

AND B.2 Farley Units 1 and 2 Verify containment 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months average air tem12erature

< 122°F.

AND Once 12er 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter Verify refueling water 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months storage tank tem12erature

< 100°F.

AND Once 12er 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter Restore containment 7 days cumulative average air temperature in calendar year to within limit.

B hours Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months 3.6.5-1 Amendment No.-247-_

(Unit 1)

Amendment No. 244_

(Unit 2)

SURVEILLANCE REQUIREMENTS SURVEILLANCE Containment Air Temperature 3.6.5 FREQUENCY SR 3.6.5.1 Verify containment average air temperature is within limit.

In accordance with the Surveillance Frequency Control Program Farley Units 1 and 2 3.6.5-2 Amendment No.~

(Unit 1)

Amendment No. ~

(Unit 2) to the Enclosure of NL-24-0281 Revised Technical Specification Pages (This Attachment consists of 3 pages, including this cover page)

3.6 CONTAINMENT SYSTEMS 3.6.5 Containment Air Temperature Containment Air Temperature 3.6.5 LCO 3.6.5 Containment average air temperature shall be ::::; 120°F.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Containment average air A.1 temperature not within limit.

AND A.2 AND A.3 B.

Required Action and B.1 associated Completion Time not met.

AND B.2 Farley Units 1 and 2 Verify containment 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months average air temperature

122°F.

AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter Verify refueling water 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months storage tank temperature

100°F.

AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter Restore containment 7 days cumulative average air temperature in calendar year to within limit.

Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months 3.6.5-1 Amendment No._ (Unit 1)

Amendment No. _

(Unit 2)

SURVEILLANCE REQUIREMENTS SURVEILLANCE Containment Air Temperature 3.6.5 FREQUENCY SR 3.6.5.1 Verify containment average air temperature is within limit.

In accordance with the Surveillance Frequency Control Program Farley Units 1 and 2 3.6.5-2 Amendment No._ (Unit 1)

Amendment No. _

(Unit 2) to the Enclosure of NL-24-0281 Technical Specification Bases Markups (For Information Only)

(This Attachment consists of 3 pages, including this cover page)

BASES LCO APPLICABILITY ACTIONS Farley Units 1 and 2 Containment Air Temperature B 3.6.5 During a OBA, with an initial containment average air temperature less than or equal to the LCO temperature limit, the resultant containment structure peak accident temperature is maintained below the containment design temperature. As a result, the ability of containment to perform its design function is ensured.

In MODES 1, 2, 3, and 4, a OBA could cause a release of radioactive material to containment. In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, maintaining containment average air temperature within the limit is not required in MODE 5 or 6.

A.1, A.2, and A.3 When containment average air temperature is not within the limit of the LCO, it must be restored to 'Nithin limit 'Nithin 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months the containment average air temperature must be verified to be less

122°F within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and once every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter (Required Action A.1 ). Required Action A.2 requires verification that the refueling water storage tank (RWST) temperature is: ::; 100°F to provide additional margin for the containment post-accident mass and energy release. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is reasonable to verify the containment average air temperature and the RWST temperature.

The once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter is adequate to confirm the temperatures remain within the Required Action limits.

+ms-Required Action A.3 is necessary to return operation to within the bounds of the containment analysis. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ? day cumulative in the calendar year Completion Time is acceptable considering the sensitivity of the analysis to variations in this parameter and provides sufficient time to correct minor problems. The cumulative time is tracked as actual time operating in Condition A and initially begins from the initial Action A entry in the calendar year. Each entry and exit time for Condition A are tracked and added to the prior cumulative time(s). The Required Action A.3 Completion Time expires when the cumulative time reaches 7 days in the calendar year.

Short-term exceedance of the containment average air temperature limit has been evaluated and determined to be of minimal impact to safety (Ref. 3).

B 3.6.5-3 Revision a_

BASES ACTIONS (continued)

SURVEILLANCE REQUIREMENTS REFERENCES Farley Units 1 and 2 B.1 and B.2 Containment Air Temperature B 3.6.5 If the containment average air temperature cannot be restored to within its limit within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SR 3.6.5.1 Verifying that containment average air temperature is within the LCO limit ensures that containment operation remains within the limit assumed for the containment analyses. In order to determine the containment average air temperature, an arithmetic average is calculated using measurements taken at four of the following sensor locations with at least two being containment air cooler intake sensors:

Instrument Number TE3187 E, F, G, & H TE3188 H & I TE3188 J Sensor Location Containment Air Cooler Intake Lower Compartment Reactor (lower)

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

1. FSAR, Section 6.2.

2. 10 CFR 50.49.

3. Amendment Nos. ###- and ##-# for Farley, Units 1 and 2, respectively, dated Month day, year.

B 3.6.5-4 Revision ~ _

to the Enclosure of NL-24-0281 Affidavit and Request for Withholding of Proprietary Information (This Attachment consists of# pages, including this cover page)

Commonwealth of Pennsylvania:

County of Butler:

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-24-037 (1)

I, Zachary Harper, Senior Manager, Licensing, have been specifically delegated and authorized to apply for withholding and execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse).

(2)

I am requesting the proprietary portions of Attachment 6 to the Enclosure ofNL-24-0281, Revision 0 be withheld from public disclosure under 10 CFR 2.390.

(3)

I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged, or as confidential commercial or financial information.

Page 1 of3 (4)

Pursuant to 10 CFR 2.390, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i)

The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse and is not customarily disclosed to the public.

(ii)

The information sought to be withheld is being transmitted to the Commission in confidence and, to Westinghouse's knowledge, is not available in public sources.

(iii)

Westinghouse notes that a showing of substantial harm is no longer an applicable criterion for analyzing whether a document should be withheld from public disclosure. Nevertheless, public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-24-037 (5)

Westinghouse has policies in place to identify proprietary information. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a)

The information reveals the distinguishing aspects of a process ( or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

Page 2 of3 (b)

It consists of supporting data, including test data, relative to a process ( or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage (e.g., by optimization or improved marketability).

( c)

Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

( d)

It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

( e)

It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f)

It contains patentable ideas, for which patent protection may be desirable.

(6)

The attached documents are bracketed and marked to indicate the bases for withholding. The justification for withholding is indicated in both versions by means of lower-case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower-case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (5)(a) through (f) of this Affidavit.

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-24-037 I declare that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief. I declare under penalty of perjury that the foregoing is true and correct.

Executed on: 7/18/2024 Signed electronically by Zachary Harper Page 3 of3 to the Enclosure of NL-24-0281 Large Break Loss of Coolant Accident Evaluation Non-Proprietary (This Attachment consists of 6 pages, including this cover page) to NL-24-0281 Large Break Loss-of-Coolant Accident Evaluation (Non-Proprietary)

Large-Break Loss-of-Coolant Accident (LOCA) Transients Southern Nuclear Company received approval to implement the Automated Statistical Treatment of Uncertainty Method (ASTRUM) Evaluation Model (EM) (Reference 1) by Farley Nuclear Plant (FNP)

Amendment No. 174 dated July 11, 2006 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML061810306, Reference 2). Changes to the FNP LOCA licensing basis analysis are tracked pursuant to 10 CFR 50.46(a)(3) and included in Section 15.4.1.5.3 of the FNP Final Safety Analysis Report (FSAR). The requested change to the actions upon exceeding the containment average air temperature limit of 120°F is likewise considered pursuant to 10 CFR 50.46(a)(3) by evaluating the effect of the specified 122°F air temperature to demonstrate that FNP continues to adhere to the regulatory acceptance criterion of 10 CFR 50.46(b )(1) under postulated large break LOCA (LBLOCA) conditions. The NRC staffs position on what constitutes a change or error to an evaluation model or the application thereof was recently clarified in NRC Regulatory Issue Summary (RIS) 2016-04 (ADAMS Accession No. ML15324A296, Reference 3). As stated on page 4 of 10 (emphasis added),

As discussed above, JO CFR 50.46(c)(2) defines an "evaluation model" as the "calculational framework for evaluating the behavior of the reactor coolant system during a [LOCA]. "In addition, JO CFR 50.46(c)(2) states that the evaluation model includes "all other information necessary for application of the calculational framework to a specific LOCA. " Therefore, the NRC considers changes to, or errors in, an evaluation model, or in the application thereof, to include not only changes to, or errors in, the physical models and model parameters that partially comprise the "calculational framework" (e.g., heat transfer correlations, etc.) but also changes to, or errors in, plant-specific inputs and design parameters (e.g., setpoints, initial conditions, etc.). All of these parameters and inputs are necessary for the analysis of a plant-specific LOCA, as described in JO CFR 50.46(c)(2).

The maximum containment air temperature is a plant-specific input. The results of the estimate of effect for the requested change will be likewise tracked pursuant to 10 CFR 50.46(a)(3) and included in Section 15.4.1.5.3 of the FNP FSAR.

The only change to the parameters for the LBLOCA evaluation for this amendment request is the maximum accumulator temperature, which is increased from 120°F to 122°F. An initial containment temperature of 90°F was assumed to determine a conservatively low containment backpressure in the application of the ASTRUM evaluation model to FNP; therefore, this input is not affected by the increase in maximum initial containment temperature.

The ASTRUM evaluation model relies on a statistical sampling technique to demonstrate that there is a high level of probability that the 10 CFR 50.46 acceptance criteria would not be exceeded under postulated LBLOCA conditions. The statistical sampling of the uncertainty contributors occurs simultaneously in the uncertainty analysis, leading to scatter in the analysis results when plotted as a function of a single uncertainty contributor. Furthermore, the change under consideration is small relative to the range of accumulator temperatures analyzed in the FNP ASTRUM analysis (90 to 120°F versus 90 to 122°F). As such, execution of the ASTRUM evaluation model framework is not considered to be the best option to estimate the effect of the maximum accumulator temperature increase from 120°F to 122°F.

Instead, existing accumulator temperature sensitivities performed with the same thermal-hydraulic code used in the FNP ASTRUM analysis from similar pressurized water reactor (PWR) plant designs with similar fuel assembly design, power level, and predicted cladding temperature response are utilized to determine an estimated effect to support the requested change.

A5-2 of 6 to NL-24-0281 Large Break Loss-of-Coolant Accident Evaluation (Non-Proprietary)

The use of existing sensitivity studies from similar PWR designs is a well-established practice to estimate the effect of changes or errors to acceptable evaluation models or the applications thereof. Indeed, FNP past precedent of applying existing sensitivity studies from similar PWR designs exists for the evaluation of the fuel thermal conductivity degradation (ADAMS Accession No. ML12293A093, Reference 4). The FNP 1 fuel thermal conductivity degradation (TCD) evaluation approach utilized a [

r,c from plants already evaluated for fuel TCD included a subset of the PWR fleet fueled by Westinghouse for which plant-specific code simulations were performed to estimate the effect of fuel TCD. The plant-specific calculations for the first 'wave' of plants evaluated were considered necessary due to the expected large impact of fuel TCD on the analysis results, the plants' margin to the licensing limit, and to establish an understanding of the effects of fuel TCD on the postulated LOCA transients in the context of Westinghouse Best Estimate LOCA evaluation models. The [

r,c considered plant-specific core design and fuel rod design data, relied on the most influential parameters related to TCD, considered the plant-specific analysis ofrecord, and provided estimates of effect that are reasonable and conservative, but not overly so.

Recent LARs for plant updates to the acceptable LOCA evaluation model included an evaluation for the error correction of the gamma energy redistribution multiplier in the FULL SPECTRUM' LOCA (FSLOCA ') evaluation model (Reference 6). Examples of the NRC staff safety evaluations on the LARs affected by this error can be found in ADAMS Accession No. ML20302Al 79 (Reference 7) and ML23 l 98A359 (Reference 8). As noted in the NRC staff safety evaluations, the PCT impact from the error correction was found by the licensee to be different for the transient phases based on parametric PWR sensitivity studies. The results from the parametric PWR sensitivity studies were then applied to estimate the effect of the error on multiple PWR analyses (as evident by the same PCT penalty being applied in both of the cited LARs). These examples provide recent precedents, where the NRC staff reviewed and accepted the use of sensitivity studies from similar PWR designs to estimate the effect of errors in the application of acceptable evaluation models.

In this case, accumulator temperature sensitivities from similar PWR plant designs with similar fuel assembly design, power level, and predicted cladding temperature response as FNP were used to estimate the effect of the maximum accumulator temperature increase, pursuant to 10 CFR 50.46(a)(3), to demonstrate continued adherence to the regulatory acceptance criteria of l O CFR 50.46(b) under postulated LBLOCA conditions. Table l provides plant parameters of interest used to establish that the existing sensitivity studies provide a realistic and appropriate estimate of effect for FNP. Further, it is noted that the predicted temperature response for the l O cases with the highest PCT from the FNP application of the ASTRUM evaluation model were characterized by a blowdown PCT on the order of l 350-to-l 550°F, a refill PCT slightly higher than the blowdown PCT, and a reflood PCT on the order of 300-400°F higher occurring roughly 100 seconds into the transient. The temperature transients from the Plant A and B Code Qualification Document (CQD) evaluation model (WCAP-12945-P-A (Reference 9))

analyses were reviewed and determined to be similar. As such, the Plant A and B accumulator temperature parametric sensitivities are considered realistic and appropriate to estimate an effect for FNP.

1 FNP was part of a Pressurized Water Reactor Owner's Group (PWROG) program to estimate the effect of fuel TCD for plants as part ofa second 'wave' of fuel TCD evaluations. The U.S. NRC audit of the PWROG evaluation approach was held on June 27, 2012, per ADAMS Accession No. ML12228A328, Reference 5.

AS-3 of 6 to NL-24-0281 Large Break Loss-of-Coolant Accident Evaluation (Non-Proprietary)

The estimated effect is a PCT increase of- 0.5°P per 1 °P increase in accumulator temperature. As such, with the conservative multiplier, the 2°P increase in the maximum accumulator temperature is estimated to have a 2°P effect on the analysis PCT. The latest racked up PCT is 2034°P. With the additional 2°P effect, the racked up PCT is estimated to be 2036°P, which maintains margin to the regulatory acceptance criterion of 2200°P.

Westinghouse has applied numerous evaluation techniques to estimate the effect of changes or errors to acceptable evaluation models or the applications thereof pursuant to 10 CPR 50.46(a)(3) or during the licensing process over the years. The applied evaluation techniques are established based on fuel vendor technical expertise and broad experience base, dependent on the given circumstances, and tend to have varying degrees of conservatism. In this circumstance, the applied approach is considered by Westinghouse to be a best practice, supported by past precedent (as cited above), and viewed as an improvement to previously performed accumulator temperature estimates of effect that were overly conservative (e.g., ADAMS Accession No. ML13067A328 (Reference 10)).

As a confirmation of the magnitude of impact, 18 accumulator temperature sensitivity studies based on the WCOBRA/TRAC code for Westinghouse-fueled nuclear plants were reviewed. The sensitivity studies span accumulator temperatures from approximately 40°P to 130°P, so the change proposed for PNP is within the range of these studies. It is acknowledged that the plants surveyed include variations in the number of RCS loops, fuel assembly design, power level, and peaking factors and that the survey was conducted only as a confirmation of the magnitude of the impact. The average impact on PCT for each 1 °P change in the accumulator temperature from these studies varied from -5°P to 5°P, with an average impact of 0.2°P. The PNP evaluation results are in good agreement with the existing database of accumulator temperature sensitivity studies, confirming the magnitude of the estimated effect for PNP.

AS-4 of 6 to NL-24-0281 Large Break Loss-of-Coolant Accident Evaluation (Non-Proprietary)

Table 1 Plant Parameters of FNP and Similar PWRs used to Determine Estimate of Effect Parameter FNP Plant A Plant B Number of Reactor 3

3 3

Coolant System Loops Fuel Assembly Design 17xl7 17xl7 17xl7 Power 2775 MWt 2900 MWt 2900 MWt System Thermal-WCOBRA/TRAC WCOBRA/TRAC WCOBRA/TRAC Hydraulic Code for Analysis' Total Core Peaking 2.5 2.5 2.52 Factor(-)

Analysis PCT2 1836 1988 1976 Notes:

1. ASTRUM and CQD evaluation models are based on the WCOBRA/TRAC code.

New code version changes are released on an as-needed basis.

2. Does not include PCT rackups and includes differences in statistical methods (ASTRUM vs. CQD)

References

1.

Westinghouse Report WCAP-16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment Of Uncertainty Method (ASTRUM),"

January 2005. (Westinghouse Proprietary Class 2)

2.

Letter from Robert E. Martin (NRC) to L. M. Stinson (SNC), "JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 -

ISSUANCE OF AMENDMENTS FOR BEST ESTIMATE LOSS-OF-COOLANT ACCIDENT (LOCA)ANALYSES USING ASTRUM (TAC NOS. MC8588 AND MC8589)," dated July 11, 2006. (Available in NRC ADAMS under Accession Number ML061810306.)

3.

NRC Regulatory Issue Summary (RIS) 2016-04, "Clarification of 10 CFR 50.46 Reporting Requirements and Recent Issues with Related Guidance Not Approved for Use," dated April 19, 2016. (Available in NRC ADAMS under Accession Number ML15324A296.)

4.

Letter from M. J. Ajluni (SNC) to the U.S. NRC, "Joseph M. Farley Nuclear Plant, 10 CFR 50.46 ECCS Evaluation Model, Annual Report for 2011 and Significant Change/Error Report," dated October 18, 2012. (Available in NRC ADAMS under Accession Number ML12293A093.)

A5-5 of 6 to NL-24-0281 Large Break Loss-of-Coolant Accident Evaluation (Non-Proprietary)

5.

Letter from Jack Stringfellow (PWROG) to the U.S. NRC, "NRC/PWROG Meeting to Discuss the PWROG Program for Providing an Estimate of the Impact of TCD on the LBLOCA PCT for Plants with Westinghouse Fuel - Transmittal of Presentation Slides (PA-ASC-1073)," dated August 7, 2012. (Available in NRC ADAMS under Accession Number ML12228A328.)

6.

Westinghouse Report WCAP-16996-P-A, Revision l, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology),"

November 2016. (Westinghouse Proprietary Class 2)

7.

Letter from G. Edward Miller (U.S. NRC) to Daniel G. Stoddard (Dominion Energy), "NORTH ANNA POWER STATION, UNIT NOS. l AND 2 - ISSUANCE OF AMENDMENT NOS. 286 AND 269 TO REVISE TECHNICAL SPECIFICATIONS TO ALLOW USAGE OF A FULL SPECTRUM LOSS-OF-COOLANT-ACCIDENT (LOCA) METHODOLOGY (EPID L-2019-LLA-0236)," dated October 29, 2020. (Available in NRC ADAMS under Accession Number ML20302Al 79.)

8.

Letter from Sujata Goetz (U.S. NRC) to Barry N. Blair (Energy Harbor), "BEAVER VALLEY POWER STATION, UNIT NOS. l AND 2 - ISSUANCE OF AMENDMENT NOS. 322 AND 212 RE: ANALYTICAL METHODOLOGY TO THE CORE OPERATING LIMITS REPORT FORA FULL SPECTRUM LOSS OF COOLANT ACCIDENT (EPID L-2022-LLA-0129),"

dated October 2, 2023. (Available in NRC ADAMS under Accession Number ML23 l 98A359.)

9.

Westinghouse Report WCAP-12945-P-A, Volume l, Revision 2 and Volumes 2 through 5, Revision l, "Code Qualification Document for Best Estimate LOCAAnalysis," March 1998.

(Westinghouse Proprietary Class 2)

10.

Letter from Joseph E. Pacher (Constellation Energy) to U.S. NRC, "License Amendment Request, Revise Section 3.6.5 of the Technical Specifications, "Containment Air Temperature,"

dated February 28, 2013. (Available in NRC ADAMS under Accession Number ML13067A328.)

A5-6 of 6

v t e Letter Sequence Request
EPID:L-2024-LLA-0098, License Amendment Request to Change Technical Specification 3.6.5, Containment Air Temperature Actions - Revised Completion Time (Approved, Closed)
Initiation Request, Request Acceptance Supplement Administration Withholding Request Acceptance, Withholding Request Acceptance Questions Answers Results Approval, Approval, Approval Other: ML24226B211
MONTHYEAR2024-08-01 [Table View]

NL-24-0281, License Amendment Request to Change Technical Specification 3.6.5, Containment Air Temperature Actions

Text

Subject:

Enclosure:

Subject:

SUMMARY

SUMMARY

1.1 Notes

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