Text

RIPE License Amendment Request to Change Containment Air Temperature Actions
	ML24110A126
Person / Time
Site:	Farley
Issue date:	04/19/2024
From:	Coleman J; Southern Nuclear Operating Co
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	NL-24-0137
	Download: ML24110A126 (1)
	v • d • e

Regulatory A ffair s 3535 Colonnade Parkway Birmingham, AL 35243 205 992 5000

April 19, 202 4 NL-2 4-0137 10 CFR 50.90

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

Subject:

RIPE License Amendment Request to Change 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 Conditions for Operation Note.

The change was previously discussed with the NRC Staff on April 3, 2024. In that discussion, SNC identified the intent to utilize the Risk -Informed Process for Evaluations (or RIPE) as described in Temporary Staff Guidance No. TSG -DORL-2021-01, Revision 3 (ADAMS Accession No. ML23122A014).

SNC requests approval of the proposed license amendment in accordance with the provisions of the RIPE review. A discussion of the use of the RIPE is provided in the enclosure to this letter. 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 Determ ination) 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 inform ation markup of the TS Bases that would be implemented along with the proposed TS change.

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

U. S. Nuclear Regulatory Commission NL-24-0137 Page 2

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 19 th day of April 2024.

Respectfully submitted,

Jamie M. Coleman Director, Regulatory Affairs Southern Nuclear Operating Com pany

JMC/was/cbg

Enclosure:

Evaluation of the Proposed Change

cc: NRC Regional Administrator, Region ll 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-24-0137 Evaluation of the Proposed Change

ENCLOSURE

Evaluation of the Proposed Change

Subject:

RIPE License Amendment Request to Change Containment Air Temperature Actions

1.

SUMMARY

DESCRIPTION

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

3. TECHNICAL EVALUATION 3.1 IDP Evaluation Results 3.2 PRA Model 3.3 PRA Use and Results 3.4 Risk Management Actions (RMAs) 3.5 RIPE Screening 3.6 Cumulative Risk 3.7 Performance Monitoring Strategies 3.8 Safety Impact

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration Determination Analysis 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)

E-1 Enclosure to NL-24-0137 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 120qF, and remove an expired Limiting Conditions for Operation (LCO) Note.

The NRC is requested to consider this license amendment request under the Risk -

Informed Process for Evaluations (RIPE) as described in Temporary Staff Guidance No. TSG-DORL-2021-01, Revision 3 (ADAMS Accession No. ML23122A014). The RIPE provides a risk-informed method to disposition issues of very low safety significance.

In order to characterize an issue as having a minimal safety impact, all of the following m ust apply:

The issue contributes less than 1x10 -7/year to core dam age frequency (CDF).

The issue contributes less than 1x10 -8/year to large early release frequency (LERF).

The issue screens to no impact or minimal impact (per NEI 21 -01).

Cumulative risk is acceptable using the guidelines in Section 5 of NEI 21 -01.

If any of the criteria above are not met, then the proposed change cannot be characterized as having minimal impact on safety.

The RIPE was used to evaluate the safety significance of this issue, and it was determined to meet each of the above criteria ; thus, the NRC is requested to apply a streamlined NRC review process consistent with TSG-DORL-2021-01, Revision 3.

A multi-discipline Integrated Decision-Making Panel (IDP) of plant-knowledgeable experts has reviewed the results of the initial categorization of SSCs/functions to confirm the appropriate considerations from plant design, operating practices and experience are reflected in the categorization input.

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 (DBA). Additionally, this structure provides shielding from the fission products that may be present in the containment atmosphere following accident conditions.

E-2 Enclosure to NL-24-0137 Evaluation of the Proposed Change

As described in 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 coole d 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, and Table 6.2-19, Containment Results for the Design Basis LOCA, the analyses for containment pressure assumed an initial containment temperature of 127 qF.

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 previously 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. ML20170B114].

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 during August and September of 2023.

2.3 Reason for the Proposed Change In August of 2023, elevated site ambient temperatures were experienced which drove the containment average air temperature very close to the existing 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]. A notice of enforcement discretion (NOED) request was also prepared but it was not submitted as the current temperature was not exceeded.

The high ambient temperature conditions are projected to continue during the summer months potentially resulting in both FNP Unit 1 and Unit 2 containment average air temperatures exceeding 120°F. Summer high temperatures are expected to continue (and possibly worsen); thus, it is prudent to request a change to

E-3 Enclosure to NL-24-0137 Evaluation of the Proposed Change

the actions associated with exceeding the existing limit in order to avoid further expenditures of SNC and NRC resources on emergency LARs and NOEDs.

The temperature for Surveillance Requirement (SR) 3.6.5.1 is taken daily at the approximate time when the containment internal temperature is expected to peak.

During 2023, Unit 1 had 8 days during which the containment internal 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 internal 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 the extreme temperatures and periods of high temperatures. Such a projection is consistent with the discussion of climate change considerations provided in the NRCs 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).

SNC is not requesting to change the LCO for containment temperature at this time, but SNC is requesting an extension of completion time along with prudent actions as discussed below. Approval of this request is expected to alleviate the need for emergency LARs and requests for enforcement discretion related to containment temperature for the foreseeable future.

2.4 Description of the Proposed Change The TS 3.6.5 LCO expired 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, is removed.

The TS 3.6.5 Actions, upon exceeding the containment average air temperature limit, are proposed to be revised to allow continued operation for up to 30 cumulative days provided:

x the containment average air temperature remains less than or equal to 122 qF (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) ;

x the containment average air temperature has not exceeded the 120 qF limit for more than 720 cumulative hours during the current calendar year (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 x the refueling water storage tank temperature remains less than or equal to 100qF (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 proposed actions and the associated limitations in these actions align with the supporting evaluation assumptions. The 8-hour Completion Times are reasonable, based on operating experience, to confirm the containment average air temperature and refueling water storage tank temperature and verify they are less than or equal to the identified limit within the Required Action. Additionally, since the ambient air temperature has led to the rising temperatures, verifying the containment average air temperature, the cumulative time above 120°F, and the refueling water storage tank temperature every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is appropriate to confirm the temperatures remain less than or equal to the identified limit within the Required Actions.

E-4 Enclosure to NL-24-0137 Evaluation of the Proposed Change

Markups showing the TS change are provided in Attachment 1.

3. TECHNICAL EVALUATION 3.1 IDP Evaluation Results The RIPE process includes evaluation by an Integrated Decision -Making Panel (IDP) in accordance with station procedures and in accordance with Enclosure 1 of NEI 21-01 [ML23306A074]. The procedure includes the following related to the IDP:

The IDP is composed of a group of experts of varying disciplines who have site-specific expertise. The minimum quorum requirements for RIPE IDP meetings will be composed of a group of at least five experts with collective expertise in the following fields: Operations, Design Engineering, Systems Engineering, Probabilistic Risk Assessment (PRA), and Safety Analysis.

The IDP members are trained in the requirements related to the RIPE process.

The training addresses, at a minimum, the purpose of the risk-informed categorization; the categorization process; the risk-informed defense-in-depth philosophy and criteria to maintain this philosophy; PRA fundamentals including details of the plant-specific PRA analyses used for the preliminary categorization (including the modeling scope and assumptions), interpretation of risk importance measures, and the role of sensitivity st udies and change in risk evaluations; and the IDP process, including roles and responsibilities.

The decision criteria for the IDP are documented. Decisions of the IDP are arrived at by consensus. Differing opinions are documented and resolved, if possible. However, a simple majority of the panel is enough for final decisions regarding the safety impact of the issues.

The IDP was held using the Farley 50.69 IDP. The quorum consisted of representatives from Operations (chair), System Engineering, Safety Analysis, Risk Informed Engineering, and Design Engineering. Licensing representatives and other engineering subject matter experts were also in attendance. Quorum members have been additionally qualified on the RIPE requirements, and a draft procedure was piloted during the process.

The IDP was divided into two separate sessions: preliminary and final. During the preliminary IDP, the preliminary screening question responses and risk assessment were presented. The IDP provided their comments, challenges on the responses, and requests for additional information to support the analysis and conclusions.

These comments were incorporated along with feedback from outside organizations (NEI) and additional information from topics noted during an NRC pre -submittal meeting. An additional IDP was held to review the additional input and changes; this IDP resulted in an additional round of comment incorporation before the analyses and conclusions were finalized and approved.

The IDP report is available to the NRC Staff for review.

3.2 PRA Model 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

E-5 Enclosure to NL-24-0137 Evaluation of the Proposed Change

Modular Accident Analysis Program (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 verif ication of the beyond design basis accident m itigation strategies),

human reliability analysis available timing, and support in determining the timing and release size to the environment following an accident.

MAAP calculations provide integrated analyses of plant response to postulated accident sequences including an assessment of available NPSH and pump NPSH requirements following the transfer to recirculation. Specifically, the risk evaluation shows MAAP results comparing the times to transfer to recirculation, changes to containment pressures, and impacts on sump water temperature and NPSH for several different accident sequences due to a change in the containment initial gas temperature from 120°F to 122°F.

Results of the analyses indicate no significant changes to sequence timing and minimal impact to peak containment pressures and containment sump water temperatures. Therefore, no changes to RHR success criteria are noted.

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 gas temperature from 120°F to 122°F.

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.

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

E-6 Enclosure to NL-24-0137 Evaluation of the Proposed Change

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 Com pany (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:

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 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.

Refueling outages and forced outages during peak summer temperatures are very infrequent. The plant response and initial conditions are bounded by the at -power results due to containment average air temperature dropping once the reactor is shutdown. Once the plant enters MODE 4, 5, 6, and fuel removed, the increase in average containment air temperature is bounded by meeting the defense in depth safety functions.

3.3 PRA Use and Results 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:

x Any HFEs calculated with the screening m ethod are screened from further analysis.

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

x Any HFEs related to steam generator overfill are screened due to the possibility of a steam line break being eliminated.

x 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.

For success criteria, those related to containment fan coolers were changed from two-out-of-four (2/4) to three-out-of-four (3/4) to evaluate for uncertainty.

E-7 Enclosure to NL-24-0137 Evaluation of the Proposed Change

Using the above screening criteria, the remaining HFEs with MAAP parameters of interest were identified and subsequently evaluated by rerunning the associated MAAP case. 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 available for operators to transition to the recirculation mode to prevent core damage. Therefore, the HFEs are not impacted by this change.

The Farley PRA Internal Events, Internal Flooding, Fire and Seismic models were quantified. The quantification results show that the total ICCDP/CDF and ICLERP/LERF values for this proposed change are well within the Regulatory Guide (RG) 1.174 and RG 1.177 guidance and the NEI 21 -01 criteria to use the RIPE process. No compensatory measures were credited in the risk evaluation.

RG 1.177, 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 (ML23234A151) 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 for this application and the results are consistent with the determination of no impact to the deterministic analyses.

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 CDFAVE and LERFAVE result in being in Region III for Figures 4 and 5 of RG 1.174 and meet the RIPE risk criteria, 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 30 cumulative days during a calendar year.

Results - The issue was determined to contribute less than 1x10 -7/year to CDF and contribute less than 1x10 -8/year to LERF. The cumulative risk baseline remains less than 1x10 -4/year for CDF and less than 1x10 -5/year for LERF, with the impact of the proposed change is incorporated into baseline risk.

3.4 Risk Management Actions (RMAs)

During period(s) of operation approaching 120 qF, the following compensatory measures are proposed to be implemented as defense-in-depth efforts to prevent exceeding the limit.

x Operate available containment coolers on high speed with service water aligned to the em ergency mode ;

x Operate available containment mini-purge continuously;

E-8 Enclosure to NL-24-0137 Evaluation of the Proposed Change

x Operate available containment recirculation fans in high speed ;

x Put in place work controls to prevent removal of containment cooling system components and supporting systems from service; and x Put in place work controls to protect the containment cooling systems.

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.

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

3.5 RIPE Screening 3.5.1 No Impact Screening SNC has considered the following screening questions and determined that the change has no impact for Question 1. The evaluation of Questions 2 through 5 determined that there may be an adverse impact and thus, the impact is further evaluated in Section 3.5.2.

3.5.1.1 Question 1: Does the issue result in an adverse impact on the frequency of occurrence of an accident initiator or result in a new accident initiator?

No, the issue does not result in an adverse impact on the frequency of occurrence of an accident initiator or result in a new accident initiator.

In accordance with NUREG/ CR-3862, Containment Pressure Problems result from hardware failure or operator error that causes containment pressure to exceed setpoint limits.

The proposed change affects the internal containment average air temperature and would allow the temperature to exceed the Technical Specification limit for the temperature at the initiation of an accident up to 2 qF for up to 30 cumulative days (720 cumulative hours) during a calendar year. The representative accident initiators identified in the guidance [ML22088A135] include exam ple categories that are either equipment failure or external hazards. The guidance also notes that external hazard frequencies cannot be reduced or increased by a plant -initiated or NRC-initiated change.

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. The proposed amendment does not increase the likelihood of the malfunction of an SSC or analyzed accidents.

Thus, the exceedance of the current internal containment average air temperature up to 2qF does not have an adverse impact on the frequency of occurrence of an accident initiator or result in a new accident initiator.

E-9 Enclosure to NL-24-0137 Evaluation of the Proposed Change

3.5.1.2 Question 2: Does the issue result in an adverse impact on the availability, reliability, or capability of structures, systems, or components (SSCs) or personnel relied upon to mitigate a transient, accident, or natural hazard?

Yes, the issue may result in an adverse impact on the availability, reliability, or capability of structures, systems, or components (SSCs) or personnel relied upon to m itigate a transient, accident, or natural hazard. See Section 3.5.2.2 for further discussion of the impact.

3.5.1.3 Question 3: Does the issue result in a n adverse impact on the consequences of an accident sequence?

Yes, the issue may result in an adverse impact on the consequences of an accident sequence. See Section 3.5.2.3 for further discussion of the impact.

3.5.1.4 Question 4: Does the issue result in a n adverse impact on the capability of a fission product barrier?

Yes, the issue may result in an adverse impact on the capability of a fission product barrier. See Section 3.5.2.4 for further discussion of the impact.

3.5.1.5 Question 5: Does the issue result in an adverse impact on defense-in-depth capability or impact in safety margin?

Yes, the issue may result in an adverse impact on defense-in-depth capability or in safety margin. See Section 3.5.2.5 for further discussion of the impact.

3.5.2 More Than Minimal Impact Screening SNC has considered the following more than minimal impact" screening questions and determined that the impact of the change is not more than a minimal adverse impact on safety based on the following screening questions.

3.5.2.1 Question 1: Does the issue result in more than a minimal increase in the frequency of occurrence of a risk-significant accident initiator or result in a new risk-significant accident initiator?

The issue was determined to have no adverse impact in response to Question 1 as addressed in Section 3.5.1.1 above.

Thus, the proposed internal containment average air temperature temporary increase does not result in a more than a minimal increase in the frequency of occurrence of a risk-significant accident initiator or result in a new risk-significant accident initiator.

3.5.2.2 Question 2: Does the issue result in more than a minimal decrease in the availability, reliability, or capability of structures, systems, or components (SSCs) or personnel relied upon to mitigate a risk -significant transient, accident, or natural hazard?

No, the issue does not result in more than minimal decrease in the availability, reliability, or capability of SSCs or personnel relied upon to mitigate a risk -significant transient, accident, or natural hazard.

E-10 Enclosure to NL-24-0137 Evaluation of the Proposed Change

The proposed change affects the internal containment average air temperature and would allow the temperature to exceed the current limit by up to 2 qF for up to 30 cumulative days (720 cumulative hours) during a calendar year.

SNC assessments of a 122 °F containment and accumulator temperature for the post-LOCA subcriticality, sump dilution, hot leg switchover, and decay heat removal analyses show no impact to the respective analyses.

No impact has been identified for normally operating equipment within containment resulting from the proposed temporary increase in temperature. Personnel are not required to enter containment to mitigate a transient, accident, or natural hazard.

Personnels capability is evaluated within Farleys 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 previous ly 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 are shown in Table 3.5.2.2-1.

Table 3.5.2.2-1

HFE Description Initiator MAAP Input of Interest OAC-AF-Operator Fails to TSW is 191 m in based ISOLAFW-Isolate SG SSB on timing of peak FAULTSG Feedwater to Containment Pressure Faulted Generator (TREC = 168 min)

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

E-11 Enclosure to NL-24-0137 Evaluation of the Proposed Change

TSW is 438 m in based Operators fail to align on timing of core OAC-LH-RECIRC-ECCS low head MLOCA damage without MLOCA recirculation - Medium SLOCA transfer to recirculation LOCA mode (TREC = 138.3 m in)

Operator Fails to TSW is 191 m in based OAL-AS-Locally Close TDAFW SSB on timing of peak CLOSEV017ABC to SG Isolation Valve Containment Pressure Q1N23V017A/B/C (TREC = 157 min)

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 HEPs of concern. OAC-AF-ISOLAFW-FAULTSG is the only TCOA potentially affected by the change in Containment Temperature.

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 temporary increase in allowed containment temperature is within the 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.

Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions SNC response to GL 96-06 continues to be based on the current calculation of record and remains unaffected. The current 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. A temporary containment air temperature increase from 120°F to 122°F 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 the maximum temperature will still be less than the 164°F temperature required to form a vapor cavity.

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

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 hea t from both units for the 30 day period. The small increase in containment temperature at the start of the event represents an insignificant effect compared to the magnitude of decay heat 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.

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.

Environmental Qualification (EQ) Evaluation A temporary 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 average monthly temperature for August used within the calculation is >125°F which bounds the current Technical Specification temperature of 120°F and the temporary increase to 122°F. Thus, a temporary increase from 120°F to 122°F for the containment average temperature limit will have n o 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).

Post-LOCA Subcriticality Assessment Post-LOCA subcriticality analyses minimize liquid mass inventories for boration sources such as accumulators. A temporary 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.

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

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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 A review of NPSH calculation NPSH Calculation from Containment Sump to the Residual Heat Removal (RHR) Pumps - Recirculation Mode documents available NPSH m argin at sum p temperatures between 120°F and 291°F. The NPSH m argins 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. Therefore, a temporary increase in containment temperature from 120°F to 122°F would be expected to have no adverse impacts on NPSH margin.

Thus, the proposed internal containment average air temperature temporary increase does not result in more than minimal decrease in the availability, reliability, or capability of SSCs or personnel relied upon to mitigate a risk-significant transient, accident, or natural hazard.

3.5.2.3 Question 3: Does the issue result in more than a minimal increase in the consequences of a risk-significant accident sequence?

No, the issue does not result in more than minimal increase in the consequences of a risk-significant accident sequence.

Containment analyses assume an internal containment average air temperature of 127°F. The proposed change affects the internal containment average air temperature and would allow the temperature to exceed the Technical Specification limit for the temperature at the initiation of an accident up to 2 qF for up to 30 cumulative days (720 cumulative hours) during a calendar year.

SNC has evaluated the impact of the proposed increase in the bulk containment average temperature 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 the proposed 2°F increase in containment average air temperature.

Containment maximum temperatures analyses for the LOCA and MSLB currently assume an initial bulk containment temperature of 127°F. For a MSLB, the current analyses of record are bounding for the proposed containment initial temperature allowance.

The LOCA analysis assumes an accumulator liquid temperature of 120°F. However, the LOCA analysis assumes a Refueling Water Storage Tank (RWST) initial temperature of 110°F. Operational data shows the RWST to typically be below 95°F.

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With the accumulator initial liquid water temperature increased by 2°F, the corresponding energy will increase by 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 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 response 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 or temperature from whats analyzed. Based on the service water temperature evaluation in Section 3.5.2.2, the containment fan cooler performance curves used in the containment response analysis are not impacted.

Operations logs RWST temperature once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at midnight each day. The proposed Technical Specification actions during the increased containment temperature include a requirement to verify RWST 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.

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 d oes not result in more than minimal increase in the consequences of a risk-significant accident sequence.

3.5.2.4 Question 4: Does the issue result in more than a minimal decrease in the capability of a fission product barrier?

No, the issue does not result in more than minimal decrease in the capability of a fission product barrier.

The fission product barriers consist of the fuel cladding, reactor coolant system (RCS) pressure boundary, and containment. The proposed change affects the internal containment average air temperature and would allow the temperature to exceed the Technical Specification limit for the temperature at the initiation of an accident up to 2 qF for up to 30 cumulative days (720 cumulative hours) during a calendar year.

Fuel Integrity: The evaluation of the impact of increasing the containment temperature and accumulator temperature by 2°F 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 reduction in total energy removal capability of the accumulator fluid associated with a 122°F initial containment temperature (the increase in containment temperature corresponds to an enthalpy increase of ~2 Btu/lbm

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[~2.23%]), accumulator initial injection timing and characteristics remaining unaffected, and the low core and vessel internals 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 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.

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 temporary temperature increase.

Containment: As discussed in Section 3.5.2.3 above, the proposed increase in containment temperature is within the margins for the containment structural design and temperature instrumentation accuracy (+/-2.5 °F). 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 increasing the accumulator temperature from 120 °F to 122 °F. Thus, the containment fission product barrier would not be impacted by the temporary temperature increase.

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 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 containm ent spray pum ps 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 3.5.2.4-1.

Table 3.5.2.4-1 User-Defined Variable from Figure of Merit Include File

Time of steam generator dry out TI_SG_DRY1

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Time of core uncover TI_CR_UNCO_L

Tim e to onset of core dam age TI_CR_1800f

Time to reach Lo-Lo RWST water level TI_RWLL

Time to RWST empty for cases without transfer to Event code 187 TRUE recirculation

Maximum containment pressure PR_CN_MAX2_PSI

Maximum core temperature for cases without core TE_CR_MAX_F damage

The MAAP analysis produces various results, however, the key figures of merit pertaining to Containment integrity and Core Damage are shown in Table 3. 5.2.4-2.

Table 3.5.2.4-2 MAAP Case Name MSLBX LLOCA-nofcs-cs-norec MLOCA

Associated HFEs OAC-AF-ISOLAFW-OAC-LH-OAC-LH-RECIRC-FAULTSG RECIRC-LLOCA MLOCA OAL-AS-CLOSEV017ABC

Containment temperature, 120 122 120 122 120 122

°F

Time of SG 1 dryout, min 214.1 214.1 N/A N/A N/A N/A

Time of core uncovery, min N/A N/A 0.1 0.1 8.1 7.8

Time to onset of core N/A N/A 63.0 62.9 438.3 432.5 damage, min

Time of Lo-Lo RWST level, N/A N/A 27.2 27.2 370.9 370.9 min

Time of RWST empty, min N/A N/A 30.6 30.6 N/A N/A

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Max containment pressure, 74.0 74.2 71.8 70.4 39.5 39.3 psia

Max core temperature for cases not going to core 559.7 559.8 N/A N/A N/A N/A damage, °F

Thus, the proposed change does not result in more than minimal decrease in the capability of a fission product barrier.

3.5.2.5 Question 5: Does the issue result in more than a minimal decrease in defense-in-depth capability or safety margin?

No, the issue does not result in more than a minimal decrease in defense -in-depth capability or safety margin.

Defense-in-depth 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 with no more than minimal impact to the defense-in-depth identified.

The proposed change affects the internal containment average air temperature and would allow the temperature to exceed the Technical Specification limit for the temperature at the initiation of an accident up to 2 qF for up to 30 cumulative days (720 cumulative hours) during a calendar year. Should an analyzed event occur during the higher temperatures, the results of the existing containment analyses are considered.

Defense-in-Depth The defense-in-depth provided for the units is maintained. The criteria identified in Regulatory Guide 1.174, 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:

x 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 containment spray train is adequate to maintain pressure and temperature within containment structural design limits.

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x 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.

x The control rod drive mechanism (CRDM) cooling system consists of fans and ducting to draw air through the CRDM shroud and eject it to the main containment atmosphere. One hundred -percent redundancy is provided by a standby fan.

x 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.

x 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.

x Post-DBA, 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 minimal compensatory measures.

Three simple periodic confirmations are proposed and each is included in the proposed TS Required Actions. These include confirmation that:

x the containment average air temperature remains less than or equal to 122 qF (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);

x the containment average air temperature has not exceeded the 120 qF limit for more than 720 cumulative hours during the current calendar year (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 x the refueling water storage tank temperature remains less than or equal to 100 qF (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).

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).

E-19 Enclosure to NL-24-0137 Evaluation of the Proposed Change

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 m ultiple functions identified in response to item 1 of this section (above) which show that system functions are not reliant on any single feature of the design.

With regard to uncertainly, the containment average air instrument uncertainty calculation demonstrates sufficient margin to the assumed safety analyses initial condition of 127qF to account for an increase from 120 qF 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 proposed temporary Technical Specification increase to 122°F protects the Safety Analysis Limit (SAL) of 127°F and given the CSA of +/- 2.5°F yields a margin of 2.5°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 Metho dology 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]).

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 qF for less than 30 cumulative days per calendar year is not expected to simultaneously affect several components important to risk.

5. Maintain multiple fission product barriers.

The fission product barriers consist of the fuel cladding, reactor coolant system (RCS) pressure boundary, and containment. The proposed change affects the internal containment average air temperature and would allow the temperature to exceed the Technical Specification limit for the temperature at the initiation of an accident up to 2 qF for up to 30 cumulative days (720 cumulative hours) during a calendar year.

Fuel Integrity: The impact of increasing the containment temperature and accumulator temperature by 2°F was evaluated and the evaluation 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.

E-20 Enclosure to NL-24-0137 Evaluation of the Proposed Change

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 temporary temperature increase.

Containment: The proposed increase in containment temperature is within the margins for the containment structural design and thus, the containment would not be impacted by the temporary temperature increase.

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., compensatory measures, discussed above 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 such human errors.

7. Continue to meet the intent of the plants design criteria.

There are no changes to the design or performance of these systems. There are no changes to the redundancy inherent in the containm ent heat rem oval design.

Existing conservatisms in analysis methodology 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 plants design criteria.

Considering items 1 through 7 above, temporarily allowing Containment Air Temperature increase above 120°F to 122°F for up to 30 cumulative days per calendar year does not impact the layers of DID inherent to the containment heat removal systems.

Safety Margins The safety margins provided for the units are maintained. The criteria identified in Regulatory Guide 1.174, section 2.1.2, are addressed below.

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 bulk containment average temperature on the maximum calculated containment pressure and temperature in FSAR Chapter 6 and on the dose analyses in FSAR Chapter 15.

E-21 Enclosure to NL-24-0137 Evaluation of the Proposed Change

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. Given the containment temperature accuracy, the assumed temperature provides appropriate margin. 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.

Since the current containment pressure evaluation post -LOCA (that develops P a) is bounding, no impact on the post -LOCA doses will occur with the higher containment bulk average temperature (122°F).

The Containment response analysis for MSLB currently assumes an initial bulk containment temperature of 127°F. Given the containment temperature instrumentation accuracy, the assumed temperature provides appropriate margin.

There are no other analysis inputs that are affected by an increase in containment bulk operating 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 the containment temperature change to 122°F.

The LOCA analysis assumes an accumulator liquid temperature of 120°F. However, the LOCA analysis assumes a Refueling Water Storage Tank (RWST) initial temperature of 110°F. Operational data shows the RWST to typically be below 95°F.

Two limiting LOCA break cases are of interest, the hot leg break and the double ended pump suction guillotine (DEPSG) break. For the hot leg break the transient is over in less than 30 seconds, and the ECCS injection phase (and therefore the RWST temperature and accumulator temperature) has no impact on the transient results.

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 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 whats analyzed. Based on the service water temperature evaluation in Section 3.5.2.2, the containment fan cooler performance curves used in the containment response analysis are not impacted.

Operations logs RWST temperature once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at midnight each day. The proposed Technical Specification actions during the increased containment

E-22 Enclosure to NL-24-0137 Evaluation of the Proposed Change

temperature include a requirement to verify RWST 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.

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 does not result in more than minimal decrease in defense-in-depth capability or safety margin.

3.6 Cumulative Risk SNC has evaluated the cumulative risk impact of a 2 qF increase in containment average air temperature for 30 cumulative days per calendar year consistent with the principles of RG 1.174 and, as applicable, RG 1.177. The cumulative risk baseline remains less than 1 x 10-4/year for CDF and less than 1 x 10-5/year for LERF, with the impact of the proposed change is incorporated into baseline risk.

3.7 Performance Monitoring Strategies With essentially no identified potential impacts from the change, the performance monitoring strategies are included in the proposed Technical Specification Actions to ensure the engineering evaluation that support s the proposed change remains valid.

3.8 Safety Impact SNC has developed risk insights related to the proposed change. A qualitative and quantitative risk analysis was performed to demonstrate with reasonable assurance that the increase in containment temperature will have negligible impact on the PRA.

In support of this evaluation, three specific areas were reviewed: 1) MAAP analyses,

2) Human Error Probability (HEP) Development, and 3) PRA model sensitivity evaluating the criteria for containment coolers.

Overall, it is judged 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 core damage frequency (CDF) and large early release frequency (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.

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 the PRA risk metrics.

Furthermore, given that the loss of generation during high ambient temperature conditions will 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 customers during sustained periods of extremely high temperatures, the request reflects a significant reduction in risk to public health and safety.

E-23 Enclosure to NL-24-0137 Evaluation of the Proposed Change

No additional hum an actions or extended response tim es are identified to com plete specified actions during a design basis accident.

No impact is identified to the peak calculated containment internal pressure for the design basis loss of coolant accident, P a. Sufficient margin exists to prevent exceeding the peak calculated containment internal pressure for the design basis loss of coolant accident.

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.

No impact is identified on the capacity and capability of the emergency diesel generators. Sufficient margin exists to accommodate the minimal temperature increase.

No design basis change is necessary for the proposed license am endment.

No impact is identified to the loss-of-coolant-accident (LOCA) sump temperature response, 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, nor during transients. No saturation pressure impact is identified. Sufficient margin exists to accommodate the minimal temperature increase.

Finally, the qualitative risk insights, integrated with considerations of defense -in-depth and safety margins, provide reasonable assurance that the health and safety of the public will not be endangered by temporary operation with an increased containment average air temperature of up to 122°F.

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 am endm ent 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 temporary increase in containment average air temperature by 2°F will have no impact on the qualification 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.

E-24 Enclosure to NL-24-0137 Evaluation of the Proposed Change

10 CFR 50.36(c)(2) requires that TSs include Limiting Condition s 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.

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. 10 004) [ADAMS Accession No. ML102310595] for a temporary exceedance of the containment average air temperature limit from 120qF. This action included in part an SNC commitment that the FNP Unit 1 would be shutdown if containment air temperature exceeded 122 qF.

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 qF to 122qF.

In March of 2022, the NRC issued an exemption to Palo Verde Nuclear Generating Station, Units 1, 2, and 3 - Exemption from the Requirements of 10CFR 50.62 to Delete Diverse Auxiliary Feedwater Actuation System Using Risk -Informed Process for Evaluations [ADAMS Accession No. ML22054A005].

4.3 No Significant Hazards Consideration Determination Analysis Pursuant to 10 CFR 50.90, Southern Nuclear Operating Com pany (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

E-25 Enclosure to NL-24-0137 Evaluation of the Proposed Change

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 qF.

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 o f the containment heat removal systems.

The proposed am endment 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 respe ct to such initiators or precursors in a manner that the probability of an accident is increased. The proposed amendment to allow temporary exceedance of the initial containment average air temperature 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 temporary increase in allowed containment temperature 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.

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

E-26 Enclosure to NL-24-0137 Evaluation of the Proposed Change

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 m a rg in o f sa fe ty.

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

5. ENVIRONMENTAL CONSIDERATION The proposed changes to the Technical Specifications (TS) are described in Section 2. 4 of this Enclosure.

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, No Significant Hazards Consideration Determination Analysis, of this license am endment 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

E-27 Enclosure to NL-24-0137 Evaluation of the Proposed Change

or non-radiological effluent release 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 CF R 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-28 Attachment 1 to the Enclosure of NL-24-0137

Technical Specification Page Markups

Insertions Denoted by Blue text.

(This Attachment consists of 3 pages, including this cover page) to NL-24-0137 Technical Specification Mark-Ups

3.6 CONTAINMENT SYSTEMS

3.6.5 Containment Air Temperature

LCO 3.6.5 Containment average air temperature shall be d 120°F.

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

Containment average air temperature shall be 122ºF 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 Verify containment 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months temperature not within average air temperature AND limit. d 122°F.

Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND A.2 Verify by administrative 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months means that containment AND average air temperature Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months has not exceeded 120°F for > 720 cumulative thereafter hours in the current calendar year.

AND 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months A.3 Verify refueling water AND storage tank temperature Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months d 100°F.

thereafter AND A.4 Restore containment 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months average air temperature 30 days to within limit.

(continued)

A1-2 to NL-24-0137 Technical Specification Mark-Ups

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND

B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

SR 3.6.5.1 Verify containment average air temperature is within In accordance with limit. the Surveillance Frequency Control Program

A1-3 Attachment 2 to the Enclosure of NL-24-0137

Revised Technical Specification Pages

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

Containment Air Temperature 3.6.5

3.6 CONTAINMENT SYSTEMS

3.6.5 Containment Air Temperature

LCO 3.6.5 Containment average air temperature shall be d 120°F.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

A. Containment average air A.1 Verify containment 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months temperature not within average air temperature AND limit. d 122°F.

Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND A.2 Verify by administrative 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months means that containment AND average air temperature Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months has not exceeded 120°F thereafter for > 720 cumulative hours in the current calendar year.

AND A.3 Verify refueling water 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months storage tank temperature AND d 100°F. Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND A.4 Restore containment 30 days average air temperature to within limit.

(continued)

Farley Units 1 and 2 3.6.5 -1 Amendm ent No. ___ (Unit 1)

Amendm ent No. ___ (Unit 2)

Containment Air Temperature 3.6.5

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND

B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

SR 3.6.5.1 Verify containment average air temperature is within In accordance with limit. the Surveillance Frequency Control Program

Farley Units 1 and 2 3.6.5 -2 Amendm ent No. ___ (Unit 1)

Amendm ent No. ___ (Unit 2)

Attachment 3 to the Enclosure of NL-24-0137

Technical Specification Bases Markups (For Information Only)

(This Attachment consists of 3 pages, including this cover page) to NL-24-0137 Technical Specification Bases Mark -Ups (For Information Only)

APPLICABLE SAFETY ANALYSES

The temperature limit is also used in the depressurization analyses to ensure that the minimum pressure limit is maintained following an inadvertent actuation of the Containment Spray System (Ref. 1).

The containment pressure transient is sensitive to the initial air mass in containment and, therefore, to the initial containment air temperature. The limiting DBA for establishing the maximum peak containment internal pressure is a SLB. The temperature limit is used in this analysis to ensure that in the event of an accident, the maximum containment internal pressure will not be exceeded.

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

Containment average air temperature satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).

ACTIONS A.1, A.2, A.3 and A.4

When containment average air temperature is not within the limit of the LCO, the containment average air temperature must be verified to be less than or equal to 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).

With the containment average air temperature less than or equal to 122°F, it must also be verified (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) that the containment average air temperature has not exceeded the limit of 120°F for more than 720 cumulative hours (30 cumulative days) within the calendar year (Required Action A.2).

An evaluation has determined that the small exceedance (less than or equal to 2°F) for this limited cumulative time per summer does not have a significant impact on the cumulative risk.

Required Action A.3 requires verification that the re fueling water storage tank temperature is less than or equal to 100°F to support the Containment Spray System and the containment atmosphere cooling function during post-accident conditions.

For Required Actions A.1, A.2, and A.3, the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is reasonable, based on operating experience, to confirm the containment average air temperature and refueling water storage tank temperature and verify they are less than or equal to the identified limit within the Required Action. Additionally, since the ambient air

Farley Units 1 and 2 B 3.6.5-Revision __ to NL-24-0137 Technical Specification Bases Mark -Ups (For Information Only)

temperature has led to the rising temperatures, verifying the containment average air temperature, the cumulative time above 120°F, and the refueling water storage tank temperature every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is appropriate to confirm the temperatures remain less than or equal to the identified limit within the Required Action.

Required Action A.4 requires that the containment average air temperature it must be restored to within limit within 30 days 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

This Required Action is necessary to return operation to within the bounds of the containment analysis. The 30 day 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is acceptable considering the evaluated risk and the sensitivity of the analysis to variations in this parameter and provides sufficient time to reduce the containment average air temperature to within the limit identified in the LCO, i.e., less than or equal to 12 0°F. correct m inor problem s.

REFERENCES 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.

Farley Units 1 and 2 B 3.6.5-Revision __

v t e Letter Sequence Request
EPID:L-2024-LLA-0052, Single Positive Test Form Collected on 01/03/2024 (Approved, Closed)
Initiation Request, Request, Request Acceptance Supplement Administration Questions Answers Results Approval Other:
MONTHYEAR2024-05-28 [Table View]

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