NL-19-1475, Response to Request for Additional Information Regarding the License Amendment Request to Transition to 10 CFR 50.48(c)- NFPA-805 Performance Based Standard for Fire Protection for Light Water Reactor Generating Plants

Response to Request for Additional Information Regarding the License Amendment Request to Transition to 10 CFR 50.48(c)- NFPA-805 Performance Based Standard for Fire Protection for Light Water Reactor Generating Plants

ML19351D130
Person / Time
Site:	Hatch
Issue date:	12/13/2019
From:	Gayheart C Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-19-1475
Download: ML19351D130 (57)

Text

Attachments C, G, S1, S2, and W contain security-related lnfonnatlon and should. be withheld under 1 O CPR 2.390. Upon removal of these Attachments, this,correspondence Is de-controlled.

A., Southern Nuclear December 13, 2019 Docket Nos.: 50-321 50-366 l). S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Cheryl A. Gayheart Regulatory Affairs Dlractor Edwin I. Hatch Nuclear Plant - Units 1 and 2 3535 Colonnade Parkway Brrmmgham, AL 35243 205 992 5316 cagay~com NL-19-1475 Response to Request for Additional infonnation Regarding the License Amendment Request to Transition to 10 CFR 50.48(c)- NFPA-805 Perfonnance Based standard for Fire Protection for Light Water Reactor Generating Plants Ladies and Gentlemen:

!3y letter dated April 4, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18096A936), Southern Nuclear Operating Company (SNC) submitted a license amendment request (LAR) for the Edwin I. Hatch Nuclear Plant (HNP),

Un~ 1 and 2, to adopt National Fire Protection Association Standard 805 (NFPA 805),

"Perfonnance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition (ADAMS Accession No. ML010800360), as incorporated into Trtle 10 of the Code of Federal Regulations, Part 50, Section 50.48(c). On March 29, 2019, and August 8, 2019, the U.S. Nuclear Regulatory Comr11lsslon (NRC) staff issued requests for additional infonnation (RAls) to SNC. SNC responded to the RAls by letters dated May 28, 2019, August 9, 2019, an.d October 7, 2019, with the exception of the March 29, 2019 PRA RAI 03 and 18. to this letter provides the SNC response to PRA RAI 03 *and 18. As part of the RAI response process, necessary modifications were identified to various Attach.ments and to Section 4.1.2 of the transition report provided in the April 4, 2018 LAR. Enclosure 2 provides a markup to the LAR transition report. Additional markups to the LAR Attachments are provided as stated in the Attachment list below. Additionally, a "clean" Attachment S Is also provided (Attachment S2). This "clean" Attachment S incorporates all changes to the previously provided Attachm~nt S, and supersedes the previously provided Attachment Sin Its entirety.

Based on revisions to Table S-2, "Plant Modifications Committed" and Table S-3, "Implementation Items" of Attachment S, an update to the Facility Opera,:ing License (FOL)

Condition is necessary to reflect the latest Attachment S requirements. Enclosures 3 and 4 provide the revised marked-up and clean FOL Conditions.

U.S. Nuclear Regulatory Commission,

NL-19-1475 Page 2 The conclusions of the No Significant Hazards Consideration and Environmental Consideration contained in the original License Amendment Request (LAR) have been reviewed and are unaffected by this RAI response.

This letter contains no NRC commitments. If you have any questions, please contact Jamie Coleman at 205.992.6611.

I declare under penalty of perjury that the foregoing Is true and correct. Executed on the 13th day of December 2019.

Respectfully submitted, y

art Director, Regulatory Affairs Southern Nuclear Operating Company CAG/RMJ

Enclosures:

1.

Response to PRA RAI 03 and 18

2.

Markup to Transition Report Section 4.1.2

3.

Revised Marked-Up FOL Condition

4.

Revised Clean FOL Condition Attachments: A.

Markup to LAR Attachment A C.

Markup to LAR Attachment C (Security-related information)

G.

Markup to LAR Attachment G (Security-related information)

H.

Markup to LAR Attachment H J.

Markup to LAR Attachment J L.

Markup to LAR Attachment L S1. Markup to LAR Attachment S (Security-related information)

S2. Revised LAR Attachment S (Security-related information)

W. Markup to LAR Attachment W (Security-related Information) cc:

Regional Administrator, Region II NRR Project Manager - Hatch Senior Resident Inspector - Hatch Director, Environmental Protection Division - State of Georgia RType: CHA02.004

Edwin I. Hatch Nuclear Plant - Units 1 and 2 Response to Request for Additional lnfonnatlon Regarding the License Amendment Request to Transition to 10 CFR 60.48(c)- NFPA -806 Perfonnance Based Standard for Fire Protection for Light Water Recictor Generating Plants Response to PRA RAI 03 and 18 to NL-19-1475

, Response to PRA RAI 03 and 18 PRARAl3:

Section 2.4.4.1 of NFPA 805 states that the change in public health risk arising from transition from the current fire protection program (FPP) to an NFPA 805 based program, and all Mure plant changes to the program, shall be acceptable to the NRC. RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk.,lnformed Decisions on Plant-Specific Changes to the Licensing Basis," Revision i, (ADAMS Accession No. ML100910006), provides quantitative guidelines on core damage frequency (CDF), and large early release frequency (LERF), and identifies acceptable changes to these frequencies th'at result from proposed changes to the plant'-s licensing basis and describes a general framework to determine the acceptability of risk-informed changes.

Based on other NRC staff RAls, the PRA methods discussed in the following RAls may need to be revised to be acceptable to the NRC:

PRA RAI 02.b regarding update of the FPRA for IE F&O resolutions' PRA RAI 04.b regarding replacement of methods that deviate from NRC guidance

• _ PRA RAI 07.b regarding treatment of sensitive electronics PRA RAI 08 regarding treatment of the impact of violations on influence factors PRA RAI 09.c.il regarding minimum joint Human Error Probabilities (HEPs)

PRA RAI 1 O regarding obstructed plume modelling PRA RAI 11.e regarding fire modeling of the MCB enclosure PRA RAI 11.h regarding credit for MCB pa~ons

• PRA RAI 12.e regarding MCR abandonment due to loss of habitability (LOH)

PRA RAI 13.f regarding MCR abandonment due to loss of control (LOC)

• . PRA RAI 13.i regarding inclusion of the decision-to-abandon the MCR in LOC scenarios PRA RAI 16.c regard-ing the impact of untraced cables on change-in-risk PRA RAI 16.d regarding the impact of other modeling conservatisms on change-in-risk

/

This list may be revised following the NRC review of the licensee's response to all the RAls (not just those listed here).

a) Provide the results of an aggregate analysis that provides the integrated impact on the fire risk (i.e., the total tr:ansition CDF and LERF, and the change (Ll) in CDF (LlCDF), and LlLERF), of replacing specific methods identified above with alternative methods which are_ acceptab_le to the NRC. In this pggregate analysis, for those cases where the individual issues have a synergistic impact on the results, a simultaneous analysis must be performed. For those cases where no synergy exists, a on&:-at-a-time analysis may be done. For those cases that have a negligible impact, a qualitative evaluation may be done.

b) For each method above, explain how the Issue will be addressed in (1) the final aggregate analysis results provided In support of the LAR, and (2) the PRA that will be used at the beginning of the self-approval of post-transition changes. In addition, provide a method to ensure that all changes will be made, that a focused-scope peer review will be performed on changes that are PRA upgrades as defined in the PRA standard, and that any findings wilt be resolved before self-approval of post-transition changes.

E1-1 to NL-19-1475 Response to PRA RAI 03 and 18 c) Explain how the RG 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," Revision 1, (ADAMS Accession No. ML092730314.) risk acceptance guidelines are satisfied for the aggregate analysis. If applicable, Include a description of any new modifications or operator actions being credited to reduce delta risk as well as a discussion of the associ.ated impacts to the FPP.

d) If any unacceptable methods or weaknesses will be retained in the PRA that will be used to estimate the change-in-risk of post-transition changes to support self-approval, explain how the quantification r~s1,11ts for each future change will account for the use of these unacceptable methods or weaknesses.

e) Identify and summarize the changes to the FPRA model beyond those associated with the RAls cited above that may need to be revised and confirm that the changes do not introduce approaches unacceptable to NRC.

PRA RAI 3 RESPONSE:

a) The PRA was changed to address any issues identified in response to each of the RAls listed. ~e response to b). Given the PRA was changed, only a single aggregate_

analysis is required. Furthermore, a qualitative evaluation is not required. The results of the aggregate analysis are provided In the following table.

CDF (/yr.)

ACDF LERF (/yr.)

ALERF Unit 1 5.24E-05 8.12E-06 1.89E-06 4.88E-07 Unit 2 5.38E-05 8.31E-06 3.39E-06

' 5.74E-07 b) The PRA was changed to address any issues identified in response to each of the RAls listed. The PRA with the changes was used for the final aggregate analysis results and will be used at. the beginning of the self-approval of post.:.transition changes. The changes were performed in accordance with the SNC PRA maintenance procedures.

These procedures are used to determine if changes are defined as upgrades per the PRA ~tandard and if a focused-scope peer review is required. The changes to the PRA are summarized in the following table. Refer to the RAI responses for details of the PRA treatment for each RAI listed. No changes to the PRA satisfied the definition of a PRA-upgrade and a focused-scope peer review was not required.

PRARAI PRA Resolution PRA.Uoarade Comment 02.b No chanae reauired NIA NIA 04.b No change required NIA NIA 07.b No change required NIA N/A 08 The FPRAwas No No new or change in revised to incorporate method.

the changes to the transient weighting factors.

E1-2 to NL-19-1475 Response to PRA RAI 03 and 18 PRA RAI PRA Resolution 09.c The FPRA was changed to use a minimum JHEP floor value of 1 E-5 10 No change reouired 11.e No chanoe reauired 11.h No change required 12.e No change reouired 13.f No chanoe reauired 13.i No change reouired 16.c No change reouired 16.d No change.reauired PRA Uoorade Comment No The minimum JHEP floor value approach was used in the peer reviewed model; however, exceptions were justified. These exceptions have been replaced with the recommended floor value.

N/A NIA N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A NIA N/A NIA c) See response to a. The results provided show that the RG 1.205 risk acceptance guidelines are satisfied. The results do not credit any new modifications or operator actions to reduce delta risk.

d) See response to b. The PRA did not retain unacceptable methods or identified weaknesses.

e) The PRA model was revised in response to PRA RAI 18 to remove the assumed conditional probability of the loss of NPSH for low pressure emergency core cooling system pumps following containment venting. Changes to the model were based on existing PRA approaches.

PRA RAJ 18:

Section 2.4.3.3 of NFPA 805 states that the PRA 1approach, methods, and data shall be acceptable to the NRC. Section 2.4.4.1 of NFPA 805 further states that the change in public health risk arising from transition from the current FPP to an NFPA 805 based program, and all future plant changes to the program, shall be acceptable to the NRC. RG 1.174 provides quantitative guidelines on CDF, LERF, and identifies acceptable changes to these frequencies that result from proposed changes to the plant's licensing basis and describes a general framework to detennine the acceptability of RI changes. Wrth regard to model uncertainty, Section 2.5.3 of RG 1.174 states that Min many cases, the industry's state of knowledge is incomplete, and there may be different opinions on how the models should be fonnulated." It I

also states that understanding the impact of key assumptions may be addressed by "perfonning the appropriate sensitivity studies."

E1-3 to NL-19-1475 Response to PRA RAI 03 and 18, *01sposition of Key Assumptions/Sources of Uncertainty,* of your application to adopt 1 O CFR 50.69, provides dispositions for candidate key assumptions and sources of uncertainty for RI categorization. One uncertainty identified that may impact the NFPA 805 application. concerns the assumed conditional probability of 1 E-02 used account for the loss of net positive suction head (NPSH) following emergency containment venting which leads to failure of the of low pressure emergency core cooling system pumps. The 50.69 LAR did not explain the basis for the 1 E-02 value or indicate how much uncertainty may exist in this assumption. NFPA 805 LAR Attachment C, Table C-1 Identifies VFDRs associated with spurious opening of the Safety Relief Valves (SRVs) which suggests that assumptions made regarding loss of NPSH following containmE;!nt venting might have an impact on tne estimated change-in-risk. Accordingly, the NRC staff observes that the sensitivity of the fire change-in-risk results for the NFPA 805 application may be sensitive to the same modeling uncertainty as the 1 O CFR 50.69 application. In light of these observations:

a) Describe the basis for the assumed conditional probability of 1 E-02 for loss of NPSH given containment venting and indicate the degree of uncertainty that exists.

b) Justify why uncertainty in the assumed probability for loss of NPSH following.

containment venting has a minimal impact on fire risk estimates (i.e., CDF, LERF, LlCDF, and LlLERF).

c) If it cannot be qualitatively Justified that the impact from the assumed probability for loss of NPSH following emergency venting has a minimal impact the fire risk estimates, then perfonn a sensitivity study on the Integrated analysis provided in response to PRA RAI 03 demonstrating that the uncertainty associated with the assumed conditional probability of 1 E-02 does not impact the. N FPA 805 applica_tion.

RAI 18 RAJ RESPONSE:

a) The PRA model has been revised to remove the uncertainty associated with the assumed conditional probability for loss of NPSH for low pressure emergency core cooling system pumps following containment venting (model basic event NPSHLOSSPROB). The basic event NPSHLOSSPROB identified in the RAI was removed from the PRA model. Thermal hydraulic analysis and operator response analysis were used to remove the uncertainty and revise the model to include the dependencies of the pump NPSH.on torus conditions and the operator response to throttle pump flow to maintain NPSH following containment venting. Thus, no new key sources of uncertainty were introduced with the model changes. The updated an~lysis results provided within PRA RAI 3 use the revised PRA model that removed the assumed conditional probability for loss of NPSH following containment venting. This model change also requires a procedure change to provide guiqance to the operators to swap suction for core spray to the CST when the required NPSH for the core spray pumps is lost. See Attachment S, Table 8-3, Implementation Item IMP-23.

b) See response to a.

c) See response to a.

E1-4

Edwin I. Hatch Nuclear Plant - Units 1 and 2 Response to Request for Additional Information Regarding the License Amendment Request to Transition to 10 CFR 50.48(c}- NFPA -805 Performance Based Standard for Fire Protection for Light Water Reactor Generating Plants Markup to Transition Report Section 4.1.2

Southern Nuclear Operating Company 4.0 Compliance with NFPA 805 Requirements 4.1.2 Results of the Evaluation Process 4.1.2.1 NFPA 805 Chapter 3 Requirements Met or Previously Approved by the NRC Attachment A contains the NEI 04-02 Table 8-1, Transition of Fundamental Fire Protection Program and Design Elements. This table provides the compliance basis for the requirements in NFPA 805 Chapter 3. Except as identified in Section 4.1.2.3, Attachment A demonstrates that the fire protection program at HNP either:

Complies directly with the requirements of NFPA 805 Chapter 3, Complies with clarification with the requirements of NFPA 805 Chapter 3, Complies through the use of existing engineering equivalency evaluations which are valid and of appropriate quality, or Complies with a previously NRC approved alternative to NFPA 805 Chapter 3 and therefore the specific requirement of NFPA 805 Chapter 3 is supplanted.

Complies, with Required Action, with applicable implementation items identified in the compliance basis, and also appear in Attachment S.

4.1.2.2 NFPA 805 Chapter 3 Requirements Requiring Clarification of Prior NRC Approval NFPA 805 Section 3.1 states in part, "Previously approved alternatives from the fundamental protection program attributes of this chapter by the AHJ take precedence over the requirements contained herein." In some cases, prior NRC approval of an NFPA 805 Chapter 3 program attribute may be unclear. SNC requests that the NRC concur with their finding of prior approval for the following sections of NFPA 805 Chapter 3:

None.

4.1.2.3 NFPA 805 Chapter 3 Requirements Not Met and Not Previously Approved by NRC The following sections of NFPA 805 Chapter 3 are not specifically met nor do previous NRC approvals of alternatives exist:

3.2.3(1) - Approval is requested for the use of EPRI performance-based fire protection inspection, testing, and maintenance frequencies 3.3.4 - Approval is requested for thermal insulation materials 3.3.5.1 - Approval is requested for wiring above suspended ceilings 3.3.5.2 - Approval is requested for 1) the use of nonmetallic conduit in embedded applications, 2) existing installations of flexible metallic and PVC coated flexible metallic conduits in lengths greater than 3-feet, and 3) the future use of flexible metallic and PVC coated flexible metallic conduits in lengths up to 6-feet.

3.5.2 and 3.5.10 - Approval is requested for the lack of check valves in the fire water tanks discharge piping 3.5.3-Approval is requested for fire pump controller NFPA 20 compliance HNP Page 16

Edwin I. Hatch Nuclear Plant - Units 1 and 2 Response to Request for Additional Information Regarding the License Amendment Request to Transition to 10 CFR 50.48(c)- NFPA -805 Performance Based Standard for Fire Protection for Light Water Reactor Generating Plants Revised Marked-Up FOL Condition for sample analysis or instrument calibration, or associated with radioactive apparatus or components (6)

Southern Nuclear, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

This renewed license shall be deemed to contain, and is subject to, the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 of Part 50, and Section 70.32 of Part 70; all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and the additional conditions specified or incorporated below:

(1)

Maximum Power Level Southern Nuclear is authorized to operate the facility at steady-state reactor core power levels not in excess of 2,804 megawatts thermal.

(2)

Technical Specifications The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix B), as revised through Amendment No. 298, are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

The Surveillance Requirement (SR) contained in the Technical Specifications and listed below, is not required to be performed immediately upon implementation of Amendment No. 195. The SR listed below shall be successfully demonstrated before the time and condition specified:

SR 3.8.1.18 shall be successfully demonstrated at its next regularly scheduled performance.

..---I ---,1.(

3

)

Unit 1

_Insert

~

Fire Protection Soutl=lern Nuolear sl=lall iFApleFAent and FAaintain in effeot all pro*;isions of tl=le fire proteotion prograFA, wl=liol=l is referenoed in tl=le Updated Final Safety Analysis Report for tl=le ffloili*)i, as oontained in tl=le updated Fire 1-lazards Analysis and Fire Proteotion PrograFA for tl=le Edwin I. l-latol=l

~Juolear Plant, Units 1 and 2, wl=liol=l was originally subFAitted by letter dated cluly 22, 198e. Soutl=lern Nuolear FAay FAake ol=langes to tl=le fire proteotion prograFR witl=lout prior GoFAFRission appro11al only if tl=le ol=langes Renewed License No. DPR-57 Amendment No. 298 woule Rot aevoFsoly affoot tho ability to aohio\\lo aAe FAaiAtaiA safe shuteo1.\\'A iA tho OVOAt of a fiFO.

(4.a)

Physical Protection Southern Nuclear shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans, including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 1 O CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plan is entitled: "Southern Nuclear Operating Company Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan," with revisions submitted through May 15, 2006.

Southern Nuclear shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Southern Nuclear CSP was approved by License Amendment No. 265, as supplemented by a change approved by License Amendment No. 274.

(4.b)

Mitigation Strategy License Condition Develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(a)

Fire fighting response strategy with the following elements:

1. Pre-defined coordinated fire response strategy and guidance

2. Assessment of mutual aid fire fighting assets

3. Designated staging areas for equipment and materials

4. Command and control

5. Training of response personnel (b)

Operations to mitigate fuel damage considering the following:

1. Protection and use of personnel assets

2. Communications

3. Minimizing fire spread

4. Procedures for implementing integrated fire response strategy

5. Identification of readily-available pre-staged equipment

6. Training on integrated fire response strategy

7. Spent fuel pool mitigation measures (c)

Actions to minimize release to include consideration of:

1. Water spray scrubbing

2. Dose to onsite responders (4.c)

The licensee shall implement and maintain all Actions required by to NRC Order EA-06-137, issued June 20, 2006, except the last action that requires incorporation of the strategies into the site security plan, contingency plan, emergency plan and/or guard training and qualification plan, as appropriate.

Renewed License No. DPR-57 Amendment No. 274 (6)

Southern Nuclear, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

This renewed license shall be deemed to contain, and is subject to, the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 of Part 50, and Section 70.32 of Part 70; all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and the additional conditions2 specified or incorporated below:

(1)

(2)

(3)

Maximum Power Level Southern Nuclear is authorized to operate the facility at steady-state reactor core power levels not in excess of 2,804 megawatts thermal, in accordance with the conditions specified herein.

Technical Specifications The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix 8), as revised through Amendment No. 243, are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Additional Conditions The matters specified in the following conditions shall be completed to the satisfaction of the Commission within the stated time periods following the issuance of the renewed license or within the operational restrictions indicated. The removal of these conditions shall be made by an amendment to the license supported by a favorable evaluation by the Commission.

(a)

Fire Protection

~

Seutl:lerR Nuolear sl:lall iR"lpleR"leRt aRel R"laiRtaiR iA ef.feet all pr011isi0As

~

of tl:le fire preteetieR pregraffl, wl:liol=t is refereReea iR tl=te Upaateel FiRal Safety /t,Ralysis Report fer tl=te faoility, as oeRtaiReel 2

The original licensee authorized to possess, use, and operate the facility was Georgia Power Company (GPC). Consequently, certain historical references to GPC remain in certain license conditions.

Renewed License No. NPF-5 Amendment No. 243 iR tRe updated FiFO l=lazaFds ARalysis aRd FiFe PFotootioR PFOJFaA=I fOF tRO e:dwiR I. l=latoR ~JuoloaF PlaRt URits 1 aRd 2, WRiOR was OFijiRally sueA=titted ey letter fFOR=t GPG to tRe GoA=tmissioR dated duly 22, 1 Q8e.

SoutReFA ~JuoleaF may R=take ORaRJeS to tRe fire proteotioR pFOJFaA=I

• .vitROUt prior GoA=tFRissioR appF01Jal ORiy if tRO GRanJOS would Rot adversely a#eot tRo aeility to aoRiove aRd A=taiRtaiA safe sRutdovm iA tRo evoRt of a fire.

(b.1)

Physical Protection Southern Nuclear shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans, including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 1 O CFR 73.55 (51 FR 27817 and 27822) and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plan is entitled: "Southern Nuclear Operating Company Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan," with revisions submitted through May 15, 2006.

Southern Nuclear shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP),

including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p ). The Southern Nuclear CSP was approved by License Amendment No. 209, as supplemented by a change approved by License Amendment No. 219.

(b.2)

Mitigation Strategy License Condition Develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(a)

Fire fighting response strategy with the following elements:

1. Pre-defined coordinated fire response strategy and guidance

2. Assessment of mutual aid fire fighting assets

3. Designated staging areas for equipment and materials

4. Command and control

5. Training of response personnel (b)

Operations to mitigate fuel damage considering the following:

1. Protection and use of personnel assets

2. Communications

3. Minimizing fire spread

4. Procedures for implementing integrated fire response strategy

5. Identification of readily-available pre-staged equipment

6. Training on integrated fire response strategy

7. Spent fuel pool mitigation measures Renewed License No. NPF-5 Amendment No. 219

!Unit 1 Insert Page 1 of 2 Southern Nuclear Operating Company shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c},

as specified in the licensee amendment request dated April 4, 2018, supplemented by letters dated May 28, 2019, August 9, 2019, October 7, 2019, and December 13, 2019, and as approved in the NRC safety evaluation (SE) dated

. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c}, and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 1 O CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.

(a) Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact.

(1) Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.

(2) Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1 x 1 o-7 /year (yr) for CDF and less than 1 x 1 o-s/yr for LERF.

The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.

(b) Other Changes that May Be Made Without Prior NRC Approval (1) Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3, fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3, element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the component, system, procedure, or physical arrangement functionality using a relevant technical requirement or standard.

I Unit 1 Insert

_Page 2 of 2 The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3, elements are acceptable because the alternative is "adequate for the hazard." Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805, Chapter 3, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the component, system, procedure, or physical arrangement functionality using a relevant technical requirement or standard. The four specific sections of NFPA 805, Chapter 3, are:

Fire Alarm and Detection Systems (Section 3.8);

Automatic and Manual Water-Based Fire Suppression Systems (Section 3.9);

Gaseous Fire Suppression Systems (Section 3.1 O); and, Passive Fire Protection Features (Section 3.11).

This License Condition does not apply to any demonstration of equivalency under Section 1.7 of NFPA 805.

(2) Fire Protection Program Changes that Have No More than Minimal Risk Impact Prior NRC review and approval are not required for changes to the licensee's fire protection program that have been demonstrated to have no more than a minimal risk impact. The licensee may use its screening process as approved in NRC SE dated to determine that certain fire protection program changes meet the minimal criterion. The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program.

(c) Transition License Conditions (1) Before achieving full compliance with 1 O CFR 50.48(c), as specified by (2) and (3) below, risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in (b)(2) above.

(2) The licensee shall implement the modifications described in Attachment S2, Table S-2, "Plant Modifications Committed," of SNC letter NL-19-1475, dated December 13, 2019, to its facility to complete transition to full compliance with 10 CFR 50.48(c) by the startup of the second refueling outage (for each unit) after the issuance of the NRC SE. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.

(3) The licensee shall implement the items as listed in Attachment S2, Table S-3, "Implementation Items," of SNC letter NL-19-1475, dated December 13, 2019, within 365 days after the issuance of the NRC SE. An exception to this statement is for the completion date for Implementation Item IMP-19. This item will be completed for each unit at a time not to exceed 180 days after all modifications for the respective unit (as listed in Attachment S2, Table S-2) are operable.

!Unit 2 Insert Page 1 of 2 Southern Nuclear Operating Company shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a} and 10 CFR 50.48(c),

as specified in the licensee amendment request dated April 4, 2018, supplemented by letters dated May 28, 2019, August 9, 2019, October 7, 2019, and December 13, 2019, and as approved in the NRC safety evaluation (SE} dated

. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c}, and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 1 O CFR 50.48(a} and 1 O CFR 50.48(c}, the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.

(1 ) Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact.

a} Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.

b} Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1x10*7 /year (yr} for CDF and less than 1 x 1 o-a/yr for LERF.

The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.

(2) Other Changes that May Be Made Without Prior NRC Approval a) Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3, fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3, element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the component, system, procedure, or physical arrangement functionality using a relevant technical requirement or standard.

!Unit 2 Insert Page 2 of 2 The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3, elements are acceptable because the alternative is "adequate for the hazard." Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805, Chapter 3, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the component, system, procedure, or physical arrangement functionality using a relevant technical requirement or standard. The four specific sections of NFPA 805, Chapter 3, are:

Fire Alarm and Detection Systems (Section 3.8);

Automatic and Manual Water-Based Fire Suppression Systems (Section 3.9};

Gaseous Fire Suppression Systems (Section 3.1 O); and, Passive Fire Protection Features (Section 3.11).

This License Condition does not apply to any demonstration of equivalency under Section 1.7 of NFPA 805.

b) Fire Protection Program Changes that Have No More than Minimal Risk Impact Prior NRC review and approval are not required for changes to the licensee's fire protection program that have been demonstrated to have no more than a minimal risk impact. The licensee may use its screening process as approved in NRC SE dated to determine that certain fire protection program changes meet the minimal criterion. The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program.

(3) Transition License Conditions a) Before achieving full compliance with 10 CFR 50.48(c), as specified by (b) and {c) below, risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in (2)(b}

above.

b) The licensee shall implement the modifications described in Attachment S2, Table S-2, "Plant Modifications Committed," of SNC letter NL-19-1475, dated December 13, 2019, to its facility to complete transition to full compliance with 10 CFR 50.48(c) by the startup of the second refueling outage (for each unit) after the issuance of the NRC SE. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.

c) The licensee shall implement the items as listed in Attachment S2, Table S-3, "Implementation Items," of SNC letter NL-19-1475, dated December 13, 2019, within 365 days after the issuance of the NRC SE. An exception to this statement is for the completion date for Implementation Item IMP-19. This item will be completed for each unit at a time not to exceed 180 days after all modifications for the respective unit (as listed in Attachment S2, Table S-2) are operable.

Edwin I. Hatch Nuclear Plant - Units 1 and 2 Response to Request for Additional lnfonnatlon Regarding the License Am~ndment Request to Transition to 10 CFR 50.48(c)- NFPA.;.805 Performance Based Standard for Fire Protection for Light Water Reactor Generating Plants Revised Clean FOL Condition for sample analysis or instrument calibration, or associated with radioactive apparatus or components (6)

Southern Nuclear, pursuarJt to the Act and 1 O CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

This renewed license shall be deemed to contain, and is subject to, the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34. of Part *30, Section 40.41 of Part 40, Section 50.54 of Part 50, and Section, 70.~2 of Part 70; all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and the additional conditions specified or incorporated below: *

(1)

Maximum Power Level Southern Nuclear is at,rthorized to operate the facility at steady-state reactor core power levels not in excess of 2,804 megawatts them,al.

(2)

Technical Specifications (3)

The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix B), as revised through Amendment No._,

are hereby Incorporated in the renewed license. Southern Nuclear shall operate the facility In accordance with the Technical Specifications and the Enviror:,mental Protection Plan.

The Surveillance Requirement (SR) contained in the Technical Specifications and listed below, is not required to be performed immediately upon implementation of Amendment No. 195. The SR listed below shall be successfully demonstrated before the tim~ and condition specified:

SR 3.8.1.18 shall be successfully demonstrated at its next regularly scheduled perfom,ance.

Fire Protection Southern Nuclear Operating Company shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated April 4, 2018, supplemented by letters dated May 28, 2019, August 9, 2019, October 7, 2019, and December 13, 2019, and as approved in the NRC safety evaluation (SE) dated ____ _

Except where NRC approval for changes or deviations Is required by 1 O CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission If those changes satisfy the provisions set forth In 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not Renewed License No. DPR-57 Amendment No.

require a change to a technical specification or a license condition, and the criteria listed below are satisfied.

(a) Risk-lntonned Changes that May Be Made Without Pdor NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met The risk asse5SIT1ent approach, methods, a_nd data shall be accep~ble to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; arid reflect the operating experience at the plant Acceptable methods to assess the risk of the change may Include methods that have been used in the peer-reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact (1) Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.

(2) Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1 x10-7/year (yr) for CDF and less than 1x10-S/yr for LERF. The proposed change must also be consistent with the def~nse-in-depth philosophy_

and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.

(b) other Changes that May Be Made Without Prior NRC Approval (1) Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3, fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equlvalem or adequate for the hazard.

The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3, element is functionally equivalent to the corresponding technical requirement A qualified fire protection engineer shall perfonn the engineering evaluation and conclude that the change has not affected the component, system, procedure, or physical Renewed License No. DPR-57 Amendment No.

-, arrangement functionality using a relevant technical requirement or standard.

The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3, elements are acceptable because the alternative is Nadequate for the hazard."

Prior NRG review and approval would not be required fo_r alternatives to four specific sections of NFPA 805, Chapter 3, for which an ~nglneering evaluation demonstrate$ ttu~t the alternative to the Chapter 3 element ls adequate for the hazard.

A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the component, system, procedure, or physical arrangement functionality using a relevant technical requirement or standard.

The four specific sections of NFPA 805; Chapter 3, are:

Fire Alarm and Detection Systems (Section 3.8);

Automatic and Manual Water-Based Fire Suppression Systems (Section 3.9);

Gaseous Fire Suppression Systems (Section 3.1 O); and, Passive Fire Protection Features (Section 3.11 ).

This License Condition does not apply to any demonstration of equivalency und~r Section 1.7 of NFPA 805.

(2) Fire Protection Program Changes th~t Have No More than Minimal Risk Impact Prior NRG review and approval are not required for changes to the licensee's fire protection program that have been demonstrated to have no more than a minimal risk impact. The licensee may use its screening process as approved in NRG SE dated to determine that certain fire protection.

program changes meet the minimal criterion. The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program.

(c) Transition License Conditions (1) Before achieving full compliance with 10 CFR 50.48(c), as specified by (2) and (3) below, risk-informed changes to the licensee's fire protection program may not be made without prior NRG review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in (b)(2) above.

(2) The licensee shall implement the modifications described In Attachment S2, Table S-2, "Plant Modifications Committed," of SNC letter NL-19-1475, dated December 13, 2019, to its facility Renewed License No. DPR-57 Amendment No.

to complete transition to full compliance with 10 CFR 50.48(c) by the startup of the second refueling outage (for each unit) after the issuance of the NRC SE. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.

(3) The licensee shall implement the iter:ns as listed In Attachment S2, Table S-3, Mlmplementatlon ltems,D of SNC letter NL 1475, dated December 13, 2019, within 365 days after the issuance of the NRC SE. An exception to this statement is for the completion date for Implementation Item IMP-19. This item will be completed for each unit at a time not to exceed 180 days after all modifications for the respective unit (as listed in Attachment S2, Table S-2) are operable.

(4.a)

Physical Protection Southern Nuclear shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans, including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 1 O CFR 50.54(p). The plan is entitled: "Southern

• Nuclear Operating Company Security Ptan, Training and Qualification Plan, and_ Safeguards Contingency Plan, D with revisions submitted through May 15, 2006.

Southern Nuclear shall fully impiement and maintain in effect all pro~isions of the Commission-approved cyber security plan (CSP), Including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Southern Nuclear CSP was approved by License Amendment No. 265, as supplemented by a change approved by License Amendment No. 274.

(4.b)

Mitigation Strategy License Condition Develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(a)

Fire fighting response strategy with the following elements:

1. Pre-defined coordinated fire response strategy and guidance

2. Assessment of mutual aid fire fighting as~ts

3. Designated staging areas for equipment and materials

4. Command and control

5. Training of response personnel (b)

Operations to mitigate fuel damage considering the following:

1. Protection and use of personnel assets

2. Communications

3. Minimizing fire spread

4. Procedures for implementing integrated_ fire response strategy Renewed License No. DPR-57 Amendment No.

5. Identification of readily-available pre-staged equipment

6. Training on Integrated fire response strategy

7. Spent fuel pool mitigation measures (c)

Actlorls to minimize release to include consider:ation of:

1. Water spray scrubbing

2. Dose to onsite responders (4.c)

The licensee shall Implement and maintain all Actions requ_lred by to NRC Order EA-06-137, issued June 20, 2006, except the last action that requires incorporation of the strategies into the site security plan, contingency plan, emergency plan and/or guard training and qualification plan, ~ appropriate.

(5)

FSAR Supplement (6)

Toe licensee'~ Final Safety Analysis Report Supplement, dated September 5, 2001, shall be Included in the next Updated Final Safety Evaluation Analysis Report*update, required by 10 CFR 50.71(e).

Safety Analysis Report The licensee's Final Safety Analysis Report Supplement, dated September 5,*2001, submitted pursuant to 10 CFR 54.21(d), describes certain future Inspection activities to be completed before the period of extended operations begins. The licensee shall complete those activities no later than August 6, 2014.

(7) _

Integrated Surveiilance Program The licensee shall in:iplement a staff-approved reactor v~el integrated surveillance program for the extended period of operation which satisfies the requirements of 10 CFR Part 54. Such a program will be implemented through a staff-approved Boiling Water Reactor Vessel and Internals Project program or through a staff-approved plant-specific program. The plant-specific program, if needed, will be developed in a manner that is consistent with other aging management programs, will include consideration of the 1 O program attributes utilized for other aging management programs, and will providE.3 a technical justification for any Renewed License No. DPR-57 Amendment No.

program attribute not covered by the plant-specific surveillance material testing program. The plant-specific program, if needed, will Include the folloWing ~ctions:

(a)

Capsules will periodically be removed to determine the rate of embrittlemenl (b)

Capsules will be removed at neutron fluence levels that provide relevant data for assessing the integrity of the Plant Hatch, Unit 1 reactor pressure vessel (in particular, for the qetermination of reactor pressure vessel pressure-temperature limits through the period of extended operation).

(c)

Capsules VJill contain material to monitor the impact of irradiation on the Plant Hatch Unit 1 reactor pressure vessel and will contain dosimetry to monitor neutron fluence.

Before the renewal t~rm begins, the licensee will notify the NRC of its decision to implement the integrated surveillance program or a plant-specific program, and provide the appropriate revisions to the Updated Final Safety Analysis Report Supplement summary desc;riptions of the vessel surveillance material testing program.

(8)

'Design Bases Accident Radiological Consequences Analyses

• Southern Nuclear Is authorized to credit administering potassium iodide to reduce the 30 day post-accident thyroid radiological dose to the operators in the main control room until May 31, 2012. Should design basis changes be completed rendering the crediting of potassium iodide no longer necessary prior to May 31, 2012, Southern Nuclear will remove the crediting of potassium iodide from the design basis accident radiological consequences analyses (reference Unit 2 FSAR paragraph 15.3.3.4.2.2) in the next Updated Final Safety Analysis Report as required by 10 CFR 50.71(e).

(9)

Alternative Source JE?rm

1)

Southern Nuclear Operating Company, Inc (SNC, the licensee) shall complete actions by April 30, 2010, as*pesciil;>ed in SNC'l?

letters dated October 18, 2007, and March 13, 2008, fo complete the design modifications to the HNP turbine building ventilation exhaust systems. Specifically, the HNP Units 1 and 2 turbine building exhaus~ fans shall be capable of being manually switched over from normally operating power supplies, to a Class 1 E circuit that will be isolated by an appropriately rated safety related, environmentally and seismically qualified circuit breaker. For further protection and isolation, the licensee shall also use fuses Renewed License No. DPR-57 Amendment No.

2)

3)

4) that will be located in a seismically qualified manual transfer switch housing. The aforementioned circuit breaker and fuses shall be adequately coordinated with the upstream load center breaker over the entire range. These devices shall be adequately rated to prevent adverse effects of a fault to the rest of the distribution system.

SNC shall implement modifications by May 31, 2010, as described In, se_ction 2. 7.3.2, of the LAR and section 5. 7 of SN C's letter dated February 25, 2008 (NL 08-0175) to modify the design for the air supply to the turbine building exhaust ventilation dampers, such that operating air to the dampers will be supplied from a non-interruptible instrument air source to eliminate single failure point vulnerability to loss of system/instrument air.

SNC shall complete actions by May 31, 2010, as.described in SNC's letter dated February 25, 2008 (NL-08-0175) to Install and Implement the capability for standby liquid Control System hand switch jumpers for HNP Units 1 and 2.

SNC shall complete actions by May 31, 2012 for HNP Unit 1, as described in SNC's letters dated February 25, 2008 (NL-08-0175) and July 2, 2008 (NL-08-1022), to modify the following Main steam Isolation Valve alternate leakage treatment boundary valves, such that they can be closed In the event of a loss of offsite power without requiring local operation:

1N38-F101A, 1N38-F1018, 1N33-F012, 1N33-F013

5)

SNC shall implement actions by May 31, 2010, as described In SNC's letter dated February 27, 2008, to assure that temperature swit~es which monitor charcoal bed temperature meet the environmental qualification requirements of 10 CFR 50.49.

(10) TSTF-448, Control Room Habitability Upon Implementation of the Amendments adopting TSTF-448,- Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by S~ 3.7.4.4, in accordance with TS 5.5.14.c.(i), the assessment of CRE habitability as required by Specification 5.5.14.c.(ii), and the measurement of CRE pressure as required by Specification 5.5.14.d, shall be considered mel Following implementation:

a.

The first performance of SR 3.7.4.4, in accordance with Specification 5.5.14.c.(i), shall be within the next 18 months.

b.

The first performance of the periodic assessment of CRE habitability, Specification 5.5.14.c.(ii), shall be within 3 years, plus the 9-month allowance of SR 3.0.2, of the next successful tracer gas test.

Renewed License No. DPR-57 Amendment No.

c.

The first performance of the periodic measurement of CRE pressure, Specification 5.5.14.d, shall be within 24 months, plus the 6 months allowed by SR 3.0.2, from the date of the most recent successful pressure measurement test D.

Southern Nuclear shall not market or broker power or energy from Edwin I. Hatch Nuclear Plant, Unit 1.

3.

This renewed license is effective as of the date of issuance and shall expire at midnight, August 6, 2034.

FOR THE U.S. NUCLEAR REGULATORY COMMISSION

~irector Office of Nuclear Reactor Regulation Attachments:

Appendix A - Technical Specifications Appendix B - Environmental Protection Plan Date of Issuance: January 15, 2002 Renewed License No. DPR-57 Amendment No.

2 (6)

Southern Nuclear, pursuant to the Act and 1 O CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

This renewed license shall be deemed to contain, and is subject to, the conditions specified in the following Commission regulations In 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 of Part 50, and Section 70.32 of Part 70; all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and the additional condition2 specified or incorporated below:*

(1)

(2)

Maximum Power Level Southern Nuclear is authorized to operate the facility at steady-state reactor core power levels not in excess of 2,804 megawatts thermal, in accordance with the conditions specified herein.

Technical Specifications The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix 8), as revised through Amendment No. _, are hereby

, incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Additional Conditions The matters specified in the following conditions shall be completed to the satisfaction of the Commission within the stated time periods following the issuance of the renewed license or within the operational restrictions indicated. The removal of these conditions shall be made by an amendment to the license supported by a favorable evaluation by the Commission.

(a)

Fire Protection Southern Nuclear Operating Company shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c),

as specified in the licensee amendment request dated April 4, 2018, supplemented by letters dated May 28, 2019, August 9, 2019, October 7, 2019, and December 13, 2019, and as approved in the NRC safety evaluation (SE) dated

. Except where NRC approval for changes or deviations is required by 1 O CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, The original licensee authorized to possess, use, and* operate the facility was Georgia Power Company (GPC). Consequently, certain historical references to GPC remain in certain license conditions.

Renewed License No. NPF-5 Amendment No.

the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions setforth In 1 O CFR 50.48(a) and 1 O CFR 50.48(c),

the change does not require a change to a technical specification, or a license condition, and the ctiteria listed below are satisfied.

(1) Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use In NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk Impact.

a) Prior NRC review and approval is not required for changes that clearly result In a decrease in risk. The proposed change must also be consistent with the defense in-depth philosophy and must maintain sufficient safety margins. The change may be Implemented following completion of the planf change evaluation.

b) Prior NRC review and approval is not required for indMdual changes that result In a risk increase less than 1x10-7 /year (yr) for CDF and less than 1 x 1 o-6/yr for LERF. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be Implemented following completion of the plant change evaluation.

(2) Other Changes that May Be Made Without Prior NRC Approval a) Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3, fundamental fire protection program elements and design requirements for which an engineering evaluation demon~trates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3, element is functionally equivalent to the Renewed License No. NPF-5 Amendment No.

corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the component, system, procedure, or physical arrangement functionality using a relev.ant technical requirement or standard.

The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3, elements are acceptable because the alternative is Madequate for the hazard. D Prjor NRC review and approval would not be required for alternatives to four specific sections of NFPA 805, Chapter 3, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the ~mponent, system, procedure, or physical arrangement functionality using a relevant technical requirement or standard. The four specific sections of NFPA 805, Chapter 3, are:

Fire Alarm and Detection Systems (Section 3.8);

Automatic and Manual Water-Based Fire ~uppression Systems (Section 3.9);

Gaseous Fire Suppression Sy~tems (Section 3.10); and, Passive Fire Protection Features (Section 3.'11).

This License Condition does not apply to any demonstration of equivalency under Section 1. 7 of NFPA 805.

b) Fire Protection Program Changes that Have No More than Minimal Risk Impact Prior NRC review and ~pproval are not required for changes to the licensee's fire protection program that have been demonstrated to have no more than a minimal risk impact.

The licensee may use its screening process as approveid in NRC SE dated to determine that certain fire protection program changes meet ttie minimal criterion. The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when ch~nges are made to the fire protection program.

(3) Transition License Conditions a) Before achieving full compliance with 10 CFR 50.48(c), as specified by (b) and (c) below, risk-informed changes to the Renewed License No. NPF-5 Amendment No.

licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated t9 have no more than a minimal risk impact, as descrjbed-in (2)(b) above, b) The licensee shall implement the modifications described in Attachment S2, Table S-2, "Plant Modifications Committed,"

of SNC letter NL-19-1475, dated December 13, 2019, to its facility to complete transition to full compliance with 10 CFR 50.48(c) by the startup of the second ~eling outage (for each unit) ~fter ttie issuance of the NRC SE. The licensee shall maintain appropriate compensatory measures in place.

until completion of these modification$.

c) The licensee shall implement the items as listed in Attachment S2, Table S-3, "Implementation Items," of SNC letter NL-19-1475, dated December 13, 2019, within 365 days after the issuance of the NRC SE. An exceptio-n to this statement is for the completion date for Implementation Item IMP-19. This item will be completed for eaeh unit at a time not to exceed 180 days after all modifications for the respective unit (as listed in Attachment S2, Table S-2) are operable.

(b.1)

Physical Protection Southern Nucl_ear shall fully Implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans, including amendments made pursuant to provisions of the Miscellaneous Amendments arid Search Requi_rements revisions to 1 O CFR 73.55 (51 FR 27817 and 27822) and the authority of 10 CFR 50.90 and.10 CFR 50.54(p). The plan is entitled: asouthern Nuclear Operating Company Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, 0 with revisions submitted through M~y 15, 2006.

J Southern Nuclear shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP),

including changes made pursuant to the authority of 10 CFR 50.90 and 1 O CFR 50.54(p}. The Southern Nuclear CSP was approved by License Amendment No. 209, as supplemente;d by a change approved by License Amendment No. 219.

(b.2)

Mitigation Strategy License Condition Develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(a)

Fire fighting response strategy with the following elements:

Renewed License No. NPF-5 Amendment No.

I

1. Pre-defined coordinated fire response strategy and guidance

2. Assessment of mutual aid fire fighting assets

3. Designated staging areas for equ.ipment and materials

4. Command and control 1

5. Training of response personnel (b)

Operations to mitigate fuel damage (?Onsidering the following:

1. Protection and use of personnel assets

2. Communications

3. 'Minimizing fire spread

4. Procedures for 'implementing integrated fire response strategy

5. lde.ntification of readily-available pre-staged equipment

6. Training on Integrated fire response strategy

7. Spent fuel pool mitigation measures (c)

Actions to minimize release to Include consideration of:

1. Water spray scrubbing

2. Dose to onsite responders (b.3)

The licensee. shall implement and maintain all Actions required by to NRG Order EA-06-1_37, issued June 20, 2006, except the last action that requires incorporation *of the strategies into the site security plan, contingency plan, emergency plan and/or guard training and qualification plan, as appropriate.

(c)

FSAR Supplement (d)

(e)

The licensee's Final Safety Analysis Report Supplement dated September 5, 2001, shall be included in the next Updated Final Safety Analysis Report update, required by 10 CFR 50.71(e).

Safety Analysis Report The licensee's Final Safety Analysis R~port Supplement dated September S, 2001, submitted pursuant to 10 CFR54.21(d),.

describes certain future inspection activities to be completed before the period of extended operations begins. The licensee shall complete those activities no later than June 13, 2018.

• Integrated Surveillance Program The licensee shall implement a staff-approved reactor vessel integrated surveillance program for the extended period of operation which satisfies the requirements of 1 O CFR Part 54.

Such a program will be implemented through a staff-approved Renewed License No. NPF-5 Amendment No.

(f)

(g) Bolling Water Reactor Vessel Internals Project program or through a staff-approved plant-specifie program. The plant-specific program, if needed, will be developed in a manner consistent with other agihg management programs, will iriclude consideration of the 1 O program attributes utilized for other aging management programs, and will provide a technical justification for any program attri_bute not covered by the plant-specific surveillance material testing program: The plant-specific program, if needed, will *include the following actions:

i.

Capsules will periodically be removed to determine the rate of embrittlernent.

ii.

Capsules will be removed at neutron fluence levels that provide relevant data for assessing the integrity of the Plant Hatch Unit 2 reactor pressure vessel (in particular, for the det~rmination of reactor pressure vessel pressure-temperature limits through the period of extended operation).

iii.

Capsules will contain material to monitor the impact of irradiation on the Plant Hatch Unit 2 reactor pressure vessel and will contain dosimetry to monitor neutron fluence.

Before the renewal tenn begins, the licensee will notify the NRC of its decision to implement the integrated surveillance program or a plant-specific program, and provide the appropriate revisions to the Updated Final Safety Analysis Repqrt.Supplement summary descriptions of the vessel surveillance material testing program.

Design Bases Accjdent Radiological Consequences Analyses Southern Nuclear is authorized to credit administering potassium iodide to reduce the 30 day post-accident thyroid radiological dose to the operators in the main control room until ¥ay 31, 2011. Should design basis changes be completed rendering the crediting pf potassium iodide no'longer necessary prior to May 31, 2011, Southern Nuclear will remove the crediting of potassium iodide from the design basis accident radiological consequences analyses (reference Unit 2 FSAR paragraph 15.3.3.4.2.2) in the next Updated

- Final Safety Analysis Report update as required by 10 CFR 50.71(e).

Alternative Source Tenn i)

Southern Nuclear Operating Company, Inc ($NC, the licensee) shall complete actions by April 30, 2010, as described in SNC's letters dated October 18, 2007, and March 13, 2008, to complete the design modifications to the

___ HNP turbine buildlng ventilation exhaust systems. Specifically, the HNP Units 1 and 2 turbine building exhaust fans shall be capable of being

• Renewed License No. NPF-5 Amendment No.

manually switched over fro111 nom,ally operating power supplies, to a Class -1 E circuit that will be isolated by an appropriately rated safety related, environmentally and seismically qualified circuit breaker. For further protection and Isolation, the licensee shall also use fuses that will be located in a seismically qualified manual transfer switch housing. The aforementioned circuit breaker and fuses shall be adequately coordinated with the upstream load center breaker over the entire range. These devices shall be adequately rated to prevent adverse effects of a fault to the rest of the distribution system.

iQ

, SNC shall implement modifications by May 31, 2010, as described in Enclosure 1, section 2.7.3.2, of the LAR and section 5.7 of SNC's letter dated February 25, 2008, (NL 08-0175) to modify the design for the air supply to the turbine building exhaust ventilation dampers, such that operating air to the dampers will be supplied from a 11on-interruptible instrument air source to eliminate single failure point

,-..

• vulnerability to loss of system/instrument air.

iiQ SNC shall complete actions by May 31, 2010, as described in SNC's letter dated February 25, 2008 (NL-08-0175) to install and Implement the capability for Standby Liquid

• Control System hand switch jumpers for HNP Units 1 and 2.

. iv)

SNC shall complete actions by May 31, 2011, for HNP Unit 2, as described in SNC's letters dated February 25, 2008 (NL 0175) and July 2, 2008 (NL-08-1022), to modify.the following Main Steam.Isolation Valve alternate leakage treatment boundary valves, such that they can be closed in the event of a loss of offsite power without requiring local operation:

2N11-FQ,04A, 2N11-F004B, 2N33-F003,.2N33-F004 v)

_SNC shall implement actions by May 31, 2010, as described in SNC's letter dated February 27, 2008, to assure that temperature switches which monitor charcoal bed temperature meet the environmental qualification requirements of 1 O CFR 50.49.

Renewed License No. NPf-5 Amendment No.

( (h)

TSTF-448 Control Room Habitability Upon Implementation of the Amend merits adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air lnleakage as required by SR 3.7.4.4, in accordance with TS 5.5.14.c.(i), the assessment of CRE habitability as required by Specification 5.5.14.c.(ii), and the measurement of CRE pressure as req1,1i(ed by Specification 5.5.14.d, shall be considered met fo_llowing implementation:

i)

The first performance of SR 3.7.4.4, in accordance With Specification 5.5.14.c.(Q, shall be within tl)e r;iext 18 months.

ii)

The first perfom,ance of the periodic assessment of CRE habltabillty, Specification 5.5.14.c.(ii), shall be within 3 years, plus-the 9-month allowance of SR 3.0.2, of the next successful tracer gas test.

ii~

The first perfom,ance of the periodic meas_urement of CRE pressure, Specification 5.5.14.d, shall be within 24 months, plus the 6 months allowed by SR 3.0.2, from the date of the most recent successful pressure measurement test.

D.

This renewed license is subject to the following antitrust conditions:

(1)

As used herein:

(a)

"Entity" means any financially responsible person, private or public corporation, municipality, county, cooperative, association, joint stock association or business trust, owning, operating or proposing to own or operate equipment_ or facilities within the state of Georgia (other than Chath.am, Effingham, Fannin, Towns and Union Counties) for Renewed License No. NPF-5 Amendment No.

the generation, transmission, or distribution of electricity, provided that, except for municipalities, counties, or rural electric cooperatives, Mentity" is restricted to those which are or will be public utilitles under the laws of the State of Georgia or uhder the laws of the Unit09 States, and are or will be providing retail electric service under a contract or rate schedule on file with and subject to the regulation of the Public Service Commission of the State of Georgia or any regulatory agency of the United States, and proyided further:, that as to municipalities, counties, or rural electric cooperatives, "entity" is restricted to those which provide electricity to the public at retail within the State of Georgia (other than Chatham, Effingham, Fannin, Towns and Union Counties) or to responslble and legally qualified organizations of such municipalities, counties, and/or cooperatives In the State of Georgia (other than Chatham, Effingham, Fannin, Towns and Union Counties) to the extent they may bind their members.

(b)

"Power Company" means Georgia Power Company, any successor, assignee of this license, or assignee of all or substantially ali of Georgia Power Company's assets, and any affiliate or subsidiary of Georgia Power Company to the extent It engages In the ownership of any bulk power supply generation or transmission resource in the State of Georgia (but specifically not including (1) flood rights and other land rights acquired in the State of Georgia incidental to hydroelectric generation facilities located in another state and (2).facilities located west of the thread of the stream on that part

• of the Chattahoochee River serving as th~ boundary between the states of Georgia and Alabama).

(2)

Power Company recognizes that it is often in the public Interest for those engaging in bulk power supply and purchases to Interconnect, coordinate for reliability and economy, and engage in bulk power supply transactions in order to increase.interconnected system reliability and reduce the costs of electric power. Such arrangements must provide for Power Company's costs

• (including a reasonable return) in connection therewith and allow other participating entities full access to the benefits available from interconnected bulk power supply operations and must provide net benefits to Power Company.' In entering into such arrangements neither Power Company nor any other participant should be required to violat(;3 the principles of sound engineering practice or forego a reasonably contemporaneous alternative arrangement with another, developed in good faith in arms length negotiations (but not including arrangements between Power Company and its affiliates or subsidiaries which impair entities' rights hereunder more than they would be impaired were such arrangements made in good faith between Power Company a non-affiliate or non-subsidiary) which affords it greater benefits. Any such arrangement must provide for adequate notice and Joint planning procedures consistent with sound engineering practice, and must relieve Power Company from obligations undertaken by it in the event such procedures are not followed by any participating entity.

Renewed License No. NPF-5 Amendment No.

Power Company recognizes that each entity may acquire some or all of its bulk power supply from sources other than Power Company_.

In the implementation of the obligations stated in tt,e succeeding paragraphs, Power Company and entities shall act in accordance with the foregoing principles, and these principles are conditions to *each of Power Company's obligations herein undertaken.

(3)

Power Company shall interconnect with any entity which provides, or which has undertaken firm contractual obligations to provide, some or all of its bulk power supply from source other than Power Company on terms to be included in an interconnection agreement which shall provide for appropriate allocation of the costs of interconnection facilities; provided, however, that If an entity undertakes to negotiate such a firm contractual obligation, the Power Company shall, in good faith, negotiate with such entity concerning any proposed interconnection. Such interconnection agreement shall provide, without undue preference or discrimination, for the following, among other things, insofar as consistent with the operating necessities of Power Company's and any participating entity's systems:

(a) maintenance and coordination of reserves, including, where appropriate, the purGhase and sale thereof, (b) emergency support, (c) maintenance support, (d) economy energy exct,anges, (e) purchase and sale of firm and non-firm capacity and energy, (f) economic dispatch of power resources within the State of Georgia, provided, however, that in no event shall such arrangements impose a higher percentage of reserve requirements on the participating entity than that maintained by Power Company for similar resources.

(4)

Power Company shall sell full requirements power to any entity. Power Company shall sell partial requirements power to any entity. Such sales shall be made pursuant to rates on file with the Federal Power Commission, or any successor regulatory agency, and subject to reasonable terms and conditions.

Renewed License No. NPF-5 Amendment No.

(5)

(a) Power Company shall transmit ("transmission servicej bulk power over its system to 'any entity or entitles with which it is interconnected, pursuant to rate schedules on file with the Federal Power Commission which will fully compensate Power Company for the use of its system, to the extent that such arrangements can be accommodated from ~ functional engineering standpoint and to the extent that Power Company has surplus llne capacity or reasonably available funds to finance new construction for this purpose. To the extent the entity or entities are able, they shall reciprocally provide transmission service to Power Company.

Transmission service will be provided under this subparagraph for the delivery of power to an entity for its or its members' consumption and retail distribution or for casual ~le to another entity for (1) its consumption or (2) its retail distribution. Nothing contained herein shall require the Power Company to transmit bulk power so as to have the effect of making the Tennessee Valley Authority ("T\\/Aj or its di$butors, directly or indirectly, a source of power supply outside the area determined by the TVA Board of Directors by resolution of May 16, 1966 to be the area for which the TVA or its distributors were the primary source of power supply on July 1, 1957, the date specified in the Revenue Bond Act of 1959, 16 USC 831 n,-4.

(b)

Power Company shall transmit over its system from any entity or entities with which it is interconnected, pursuant to rate schedules on ijle with the Federal Power Commission which will fully compensate Power Company for the use of its system, bulk power which results from any such entity having excess capacity available froni self-owned generating resources in the State of Georgia, to the extent such excess necessarily results from economic unit sizing or from failure to forecast load accurately or from such generating resources becoming operational earlier than the planned in-service date, to the extent that such arrangements cah be accommodated from a functional engineering standpoint, and to the extent Power Company has surplus line capacity available.

(6)

Upon request, Power Company shall provide service to any entity purchasing partial requirements service, full requirements service or transmission service from Power Company at a delivery voltage appropriate for loads served by such entity, commensurate with Power Company's available transmission facilities. Sales of such service shall be made pursuant to rates on file with the Federal Power Commission or any successor regulatory agency, and subject to reasonable terms and conditions.

Renewed *License No. NPF-5 Amendment No.

(7)

Upon reasonable notice, Power Company shall grant any entity the opportunity to purchase an appropriate share in the ownership of, or, at the option of the entity, to purchase an appropriate share of unit power from each (>'f the following nuclear generating unl,ts at Power Company's costs, to the extent the same are constructed and operated: Hatch 2, Vogtle 1, Vogtle 2, and any other nuclear generating unit constructed by Power Company in the State of Georgia which, in the application filed with USAEC or its successor agency, is scheduled for commercial operation prior to January 1, 1989.

An entity's request for a share must have regard for the economic size of such nuclear unit(s), for the entity's load size, growth and characteristics, and for demands upon Power Company's system from other entities and Power Company's retail customers, all in accordance with sound engineering practice. Executory agreements to accomplish the foregoing shall contain provisions reasonably specified by Power Company requiring the entity to consummate and pay for such purchase by an early date or dates certain. For purposes of this provision, "unit power" shall mean capacity and associated energy from a specified generating unit.

(8)

Southern Nuclear shall not market or broker power or energy from Edwin I. Hatch Nuclear Plant, Unit 2. Georgia Power Company shall continue to be responsible for compliance with the obligations imposed on it in its antitrust license conditions. Georgia Power Company is responsible and aCC9untable for the actions of Southern Nuclear, to the extent that Southern Nuclear's actions may, in any way, contravene the existing antitrust license conditions.

(9)

To effect the foregoing conditions, the following steps shall be taken:

(a)

Power Company shall file with the appropriate regulatory authorities and thereafter maintain in force as needed an appropriate transmission tariff available to any entity; (b)

Power Compar:iy shall file with the appropriate regulatory authorities and thereafter maintain in force as needed an appropriate partial requirements tariff available to any entity; Power Company shall have its liability limited to the partial requi.rements service actually contracted for and the entity shall be made responsible for the security of the bulk power supply resources acquired by the entity from sources other than the Power Company; Renewed License No. NPF-5 Amendment No.

r (c)

(d)

(e)

(f)

(g) Power Company shall amend the general terms and conditions of its current Federal Power Commission tariff and thereafter maintain in force as needed provisions to enable any entity to receive bulk power at transmission voltage at appropriate rates; Power Company shall not have the unilateral right to defeat the intended access by each entity to alternative sources of bulk power supply provided by the cond!fions to this license; but Power Company shall retain the right to seek regulatory approval of changes in its tariffs to the end that it be adequately compensated for services it provides, specifically including, but not limited to, the provisions of Section 205 of the Federal Power Act; Power Company shall use its best efforts to amend any outstanding contract to which it is a party that contains provisions which are Inconsistent with the conditions of this license; Power Company affirms that no consents are or will become necessary from Power Company's parent, affiliates or subsidiaries to enable Power Company to carry out Its obligations hereunder or to enab,le the entities to enjoy their rights hereunder; All provisions of these conditions shall be subject to and implemented in accordance with the laws of the United States and of the State of Georgia, as applicable, and with rules, regulations, and orders of agencies of both, as applicable.

3.

This renewed license is effective as of the date of issuance and shall expire at midnight, June 13, 2038.

FOR THE U.S. NUCLEAR REGULA TORY COMMISSION u

. o ms, irector Office of Nuclear Reactor Regulation Attachments:

Appendix A-.Technical Specifications Appendix B - Environmental Protection Plan Date of l~uance: January 15, 2002 Renewed License No. NPF-5 Amendment No.

/

I Edwin I. Hatch Nuclear Plant - Units 1 and 2 Response to Request for Additional Information Regarding the *License Amendment Request to Transition to 10 CFR 60.48(c)-NFPA -806 Performance Based Standard for Fire Protection for Light Water Reactor Generating Plants Attachment A Markup to LAR Attachment A

/

Southern Nuclear Operating Company Attachment A - NEI 04-02 Table B Transition of Fundamental FP Program and Design Elements NFPA 805 Ch. 3 Ref.

Requirements/Guidance Compliance Statement 3.3.5.2 Only metal tray and metal Complies conduits shall be used for electrical raceways. Thin wall metallic tubing shall not be used for power, instrumentation, or control cables. Flexible metallic conduits shall only be used in short lengths to connect components.

Replace with:

NRC approval for: 1) the use of nonmetallic conduit in embedded applications, 2) existing installations of flexible metallic and PVC coated flexible metallic conduits in lengths greater than 3-feet, and 3) the future use of flexible metallic and PVC coated flexible metallic conduits in lengths up to 6-feet is being requested in Attachment L, Approval Request 4.

Insert:

Submit for NRC Approval Complies, with Required Action IMPLEMENTATION ITEMS (See Attachment S, Tabla S-3):

Compliance Basis Except as identified below, HNP complies with no additional clarification.

FAQ 06-0021 defines "short lengths" as approximately three feet of flexible metallic conduit.

NRC approval of tl:le 1.1se of PVC coated flexible c:ond1.1it in lengtl:ls 1,1p to 6 feet and embedded non metallic: c:0Ad1.1it is being r:eq1.1ested in Attac:l:lment L, Approval Req1.1est 4, Implementation items are identified below.

Reference Document Drawing A29500, Conduit and Conduit Support, Ver. 3.0 I Section 3.1 Drawing A29501, General Design Document and Details for the Installation of Nonsafety-Related Electrical Work, Ver. 2.0 I Section 6.0 Drawing 613000, Conduit &

Grounding Installation Notes, Ver.

5.0 E-1-03, SNC Raceway Design Standard, Rev. 6 Specification SS-2123-009, Technical Specification for Cable Trays and Cable Tray Accessories for the Edwin I. Hatch Nuclear Plant - Unit 2, Rev. A/ All FAQ 06-0021, Cable Air Drops, Rev. 0/ All None FPE RAI02 IMP-20 Current cable/raceway installation procedures allow for flexible metallic conduit installations up to 6-feet, or greater if approved by the Architectural Engineer. Cable/raceway procedures will be revised to ensure that installation guidance is limited to a maximum of 6 feet for future flexible metallic conduit installations.

HNP Page A-20

Southern Nuclear Operating Company Attachment A - NEI 04-02 Table 8 Transition of Fundamental FP Program and Design Elements NFPA 805 Ch. 3 Ref.

Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.4.1(e)

Each industrial fire brigade Complies member shall pass an annual No Additional Clarification Procedure MS-MED-009, Fire Brigade Physical Examination Procedure, Ver. 1.1 / All physical examination to determine that he or she can perform the strenuous activity required during manual firefighting operations. The physical examination shall determine the ability of each member to use respiratory protection equipment.

Procedure NMP-ES-035-010, Fire Brigade, Ver. 5.0 I Section 3.0 Procedure NMP-TR-426, Fire Training Program, Ver. 5.1 /

Sections 4.2.1.k and 4.3 3.4.2 Pre-Fire Plans.

Current and detailed pre-fire Complies, with Required Current and detailed pre-fire plans for fighting fires in all areas have been instituted at HNP and are readily available. The pre-fire plans describe actions to be taken by firefighting personnel during the fire, including instructions on use of firefighting equipment.

Drawing Series A-43965, Pre-Fire Plan for Powerblock Areas HNP plans shall be available to the Action industrial fire brigade for all areas in which a fire could jeopardize the ability to meet the performance criteria described in Section 1.5.

Implementation items are identified below.

IMPLEMENTATION ITEMS (See Attachments, Table S-3):

Drawing Series A-43966, Pre-Fire Plan for Non-Powerblock Areas IMP-4 Fii:e Brigade training materials and pi:e fii:e plans will be i:e1.*ised to address U1e radioaGtive i:elease req1a1irements of NFPA 8Q5 and to ens1a1re that details i:egarding n1a1clear saf.et:y components that ai:e pi:esent in the area are provided.

Replace with:

Fire Brigade training materials and pre-fire plans will be revised to address the radioactive release requirements of NFPA 805. Pre-fire plans will also be updated to include components necessary to achieve the nuclear safety performance criteria where entry to the affected fire area is required.

PageA-33

Southern Nuclear Operating Company Attachment A - NEI 04-02 Table B Transition of Fundamental FP Program and Design Elements NFPA 805 Ch. 3 Ref.

Requirements/Guidance 3.4.2.1 The plans shall detail the fire area configuration and fire hazards to be encountered in the fire area, along with any nuclear safety components and fire protection systems and features that are present.

Compliance Statement Complies, with Required Action Compliance Basis The pre-fire plans at Plant Hatch detail the fire area configuration and fire hazards to be encountered for the fire area. The pre-fire plans also list any fire protection systems and features that are present in the fire area.

Implementation items are identified below.

IMPLEMENTATION ITEMS {See Attachment S, Table S-3):

Reference Document Drawing Series A-43965, Pre-Fire Plan for Powerblock Areas / All Drawing Series A-43966, Pre-Fire Plan for Non-Powerblock Areas I All Procedure NMP-ES-035-GL01,

Fire Protection Program Guideline, Ver. 3.0 / Section 3.2.4.f IMP-4 Fir:e Brigade training materials and pre fire plans will be r:evised tg addr:ess the radigactive r:elease req1a1ir:ements Qf NFPA B05 and tg ens1a1re that details r:egarding n1a1clear safety cgmpgnents that ar:e pr:esent in the ar:ea ar:e prgvided.

3.4.2.2 HNP Pre-fire plans shall be reviewed Complies No Additional Clarification and updated as necessary.

Replace with:

Fire Brigade training materials and pre-fire plans will be revised to address the radioactive release requirements of NFPA 805. Pre-fire plans will also be updated to include components necessary to achieve the nuclear safety performance criteria where entry to the affected fire area is required.

PM NCFIREPLAN1, Review Pre-Fire Plans / All PM NCFIREPLAN2, Review Pre-Fire Plans / All Procedure NMP-ES-035-GL01,

Fire Protection Program Guideline, Ver. 3.0 I Sections 3.2.4.e(1) and 3.2.4.e(3)

Procedure NMP-ES-084-001,

Plant Modification and Configuration Change Processes, Ver. 7.0/ All PageA-34

Southern Nuclear Operating Company Attachment A - NEI 04-02 Table B Transition of Fundamental FP Program and Design Elements NFPA 805 Ch. 3 Ref.

Requirements/Guidance Compliance Statement 3.5.7 Individual fire pump connections Complies to the yard fire main loop shall be provided and separated with sectionalizing valves between connections.

3.5.8 A method of automatic pressure Complies maintenance of the fire protection water system shall be provided independent of the fire pumps.

3.5.9 Means shall be provided to Complies immediately notify the control room, or other suitable constantly attended location, of operation of fire pumps.

3.5.10 An underground yard fire main Complies with Use of loop, designed and installed in EEEE's accordance with NFPA 24, Standard for the Installation of Private Fire Service Mains and Their Appurtenances, shall be installed to furnish anticipated water requirements.

Compliance Basis No Additional Clarification No Additional Clarification No Additional Clarification The underground yard fire main loop is designed in accordance with NFPA 24 as identified in Calculation SMNH-16-031, NFPA 24 Code Compliance Review.

Insert:

Complies, with Required Implementation items are identified below.

Action IMPLEMENTATION ITEMS (See Attachment S, Table S-3):

Reference Document Drawing H-11033 Sheet 1, Fire Protection-P&ID Pumphouse Layout, Ver. 51.0 / All Drawing H-11033 Sheet 1, Fire Protection-P&ID Pumphouse Layout, Ver. 51.0 / All A-42162, Unit No. 1 / 2 Fire Protection Detection/Annunciation Multiplex Database, Rev. 10 / All Procedure 34SV-X43-001-1, Fire Pump Test, Ver. 3.5 / Section 7.0 Calculation SMNH-16-031, NFPA 24 Code Compliance Review, Ver.

1 / All Drawing H-11033 Sheet 1, Fire Protection-P&ID Pumphouse Layout, Ver. 51.0 / All NFPA 24, Standard for Outside Protection, 1973 Edition I All None FPE

RAI 04

IMP-21 Bollards or other appropriate protection will be provided for yard post indicator valves 1Y43-F308P and 1Y43-F316P in accordance with approved plant design.

HNP Page A-46

Southern Nuclear Operating Company Attachment A - NEI 04-02 Table B Transition of Fundamental FP Program and Design Elements NFPA 805 Ch. 3 Ref.

Requirements/Guidance Compliance Statement Compliance Basis Reference Document HNP Complies, with Required Bypass lines or pressure relief trim kits will be Action installed to prevent pressure buildup in wet pipe sprinkler systems subject to excessive static pressures. (See Attachment S, Table 5-2, Item 3.)

Implementation items are identified below.

IMPLEMENTATION ITEMS (See Attachments, Table S-3):

Calculation SMNH-16-026, NFPA 13 Code Compliance Review, Ver.

1 / All NFPA 13, Standard for Installation of Sprinkler Systems, 1983 Edition

/ All IMP-7 Sprinkler system testing procedures will be revised to ensure inspectors' test connections are appropriately sized during system testing.

IMP-8 Update plant documentation to perform periodic internal sprinkler piping obstruction testing and monitoring.

Insert:

of preaction and dry pipe sprinkler systems Page A-66

Southern Nuclear Operating Company NFPA 805 Ch. 3 Ref.

Requirements/Guidance 3.9.1 (2)

NFPA 15, Standard for Water Spray Fixed Systems for Fire Protection HNP Attachment A - NEI 04-02 Table 8 Transition of Fundamental FP Program and Design Elements Compliance Statement Compliance Basis Complies with Use of Water spray systems are installed in accordance EEEE's with NFPA 15 as identified in Calculation SMNH 028, NFPA 15 Code Compliance Review.

Calculation SMNH-16-062 evaluates the separation distances between the transformers and the Turt>ine building and documents that the current configuration is adequate for the hazards.

Calculation SMNH-16-067 documents the functionality of hose station supply and the pressure and application density of the suppression system in the Intake Structure.

Eval1a1atign SMNH 16 107 dgc1,1ments the ade~1,1acy Reference Document Calculation SMNH-16-028, NFPA 15 Code Compliance Review, Ver.

1 / All Calculation SMNH-16-062, Engineering Evaluation of NFPA 805 Power Block Building Separation, Ver. 1 / All Calculation SMNH-16-067, Hydraulic Calculation and Evaluation of Hose Stations and Water Spray / Sprinkler System -

Intake Structure, Ver. 1 / All gf the applicatign density fQr the water spray system Calc1a1latign SMNH 16 107, (2X43129W-03) protecting the SQQ kVA1,1tg Hydra1a1lic Calc1a1latign and Transformers (Fire.Area QS01). ~

Eval1a1atign gf\\11.'ater Spi:ay System SQQ kV A1,1tg TransfQrmers, Ver. 1 / All Delete NFPA 15, Standard for Water Spray Fixed Systems for Fire Protection, 1982 Edition / All NFPA 15, Standard for Water Spray Fixed Systems for Fire Protection, 2001 Edition / All NFPA 15, Standard for Water Spray Fixed Systems for Fire Protection, 2007 Edition / All NFPA 15, Standard for Water Spray Fixed Systems for Fire Protection, 2012 Edition / All Page A-67

Edwin I. Hatch Nuclear Plant - Units 1 and 2 Response to Request for Additional Information Regarding the License Amendment Request to Transition to 10 CFR 50.48(c)- NFPA -805 Performance Based Standard for Fire Protection for Light Water Reactor Generating Plants Attachment H Markup to LAR Attachment H

Southern Nuclear Operating Company Attachment H - NEI 04-02 FAQs Summary Table This table includes the approved FAQs that have not been incorporated into the current endorsed revision of NEI 04-02 and utilized in this submittal:

Table H NEI 04-02 FAQs Utilized in LAR Submittal No.

Rev.

Title FAQ Ref.

Closure Memo 06-0008 9

NFPA 805 Fire Protection Engineering ML090560170 ML073380976 Evaluations 06-0022 3

Acceptable Electrical Cable M L090830220 ML091240278 Construction Tests 07-0030 5

Establishing Recovery Actions ML103090602 ML110070485 07-0032 2

Clarification of 10 CFR 50.48(c),

ML081300697 ML081400292 10 CFR 50.48(a) and GDC 3 clarification 07-0035 2

Bus Duct Counting Guidance for High ML091610189 ML091620572 Energy Arcing Faults 07-0038 3

Lessons learned on Multiple Spurious ML103090608 ML110140242 Operations 07-0039 2

Lessons Learned - NEI B-2 Table M L091420138 ML091320068 07-0040 5

Non-Power Operations Clarification ML173318109 ML173318108 08-0042 0

Fire Propagation from Electrical ML080230438 ML092110537 Cabinets ML091460350 08-0043 Electrical Cabinet Fire Location ML083540152 ML092120448 ML091470266 08-0044 0

Large Oil Fires ML081200099 ML092110516 ML091540179 08-0049 0

Cable Fires ML081200309 ML092100274 ML091470242 08-0052 0

Transient Fire Growth Rate and ML081500500 ML092120501 Control Room Non-Suppression ML091590505 08-0054 Demonstrating Compliance with ML103510379 ML15016A280 Chapter 4 of NFPA 805 09-0056 2

Radioactive Release Transition ML102810600 ML102920405 09-0057 3

New Shutdown Strategy ML100330863 ML100960568 10-0059 e

~IFFl,A. 8Qe MeRiteFiR!I Ml::~2Q4rne8Q Ml::~2Q7eQrn8 12-0062 UFSAR Content ML121430035 ML121980557 12-0063 Fire Brigade Make-Up ML121670141 ML121980572 Replace with:

6 NFPA 805 Monitoring ML18208A450 ML18208A409 HNP Page H-2

Edwin I. Hatch Nuclear Plant - Units 1 and 2 Response to Request for Additional Information Regarding the License Amendment Request to Transition to 10 CFR 50.48(c)- NFPA -805 Performance Based Standard for Fire Protection for Light Water Reactor Generating Plants Attachment J Markup to LAR Attachment J

Southern Nuclear Operating Company Attachment J - Fire Modeling V&V Table J-1 V & V Basis for Fire Models / Model Correlations Used Calculation Replace with:

(Smoke optical density method)

Heat Detection Actuation Correlation HNP Application A111er:t CeiliR!I.Jet eeFFelatieR ie 1:1688 le ael8FFRiR8 l8FRJl8Fall:IF8 aRa tile Heskestaa aRa Qeliettatsies leFRJl8Fal1:1Fe le SFReke aeRsity eeFFelalieR feF sFReke aeteelieR liFRiR!I esliFRales Replace with:

An optical density method, developed from Alpert's ceiling jet temperature correlation, is used to determine the lime to automatic smoke detection.

Estimates heat detector timing based on the Alpert ceiling jet temperature, velocity, and thermal response of sprinkler.

Replace with:

V& V Basis Technical

Reference:

NUREG-1805, Chapter 11, 2004 NUREG-1824, Volume 4, 2007 NUREG-1934, Chapter 2, 2012 SFPE Handbook of Fire Protection Engineering, 5th Edition, Chapter 40, Custer R., Meacham B., and Schifiliti, R., 2016 SFPE Handbook of Fire Protection Engineering, 5th Edition, Chapter 14, Alpert, R., 2016 Technical

Reference:

NUREG-1805, Chapter 11, 2004 NFPA Fire Protection Handbook, 19th Edition, Chapter 3-9, Budnick, E., Evans, 0., and Nelson, H.,

2003 NUREG-1824, Volume 4, 2007 NUREG-1934, Chapter 2, 2012 The Alpert's ceiling jet correlation temperature is transformed to a soot optical density by application of basic laws of thermodynamics and fluid mechanics.

Discussion The smoke detection correlation is contained in NUREG-1805.

Alpert's ceiling jet correlation V&V is documented in NUREG-1824.

The correlation has been applied within its limits of applicability and the validated range reported in NUREG-1824 or, if applied outside the validated range, the model has been justified as acceptable, either by qualitative analysis, or by quantitative sensitivity analysis. The methodology for justifying application of the fire model outside the range is in accordance with methods documented in NUREG-1934.

. ~

.......,....,..=

• • =

~

Th~

l1~

t1onm~

H FPRA is described in the following HNP Calculations:

H-RIE-FIREPRA-U00-008A, "Hatch Fire PRA Task 8/11 a, Fire Scenario Development and Detailed Fire Modeling", Version 2, Attachment A, Fire Modeling Workbook Methodology for HNP The heat detection correlation is contained in NUREG-1805.

The correlation is documented in an authoritative publication of the NFPA Fire Protection Handbook.

Alpert's ceiling jet correlation V&V is documented in NUREG-1824.

The correlation has been applied within its limits of applicability and the validated range reported in NUREG-1824 or, if applied outside the validated range, the model has been justified as acceptable, either by qualitative analysis, or by quantitative sensitivity analysis. The methodology for justifying application of the lire model outside the range is in accordance with methods documented in NUREG-1934.

The applicability of the V&V basis to the model implementation in the HNP FPRA is described in the following HNP Calculations:

H-RIE-FIREPRA-U00-008A, "Hatch Fire PRA Task 8/11 a, Fire Scenario Development and Detailed Fire Modeling", Version 2, Attachment A, Fire Modeling Workbook Methodology for HNP Page J-6 FMRAI 03(c)

I FMRAI 03(c)

Edwin I. Hatch Nuclear Plant - Units 1 and 2 Response to Request for Additional Information Regarding the License Amendment Request to Transition to 10 CFR 50.48(c)- NFPA -805 Performance Based Standard for Fire Protection for Light Water Reactor Generating Plants Attachment L Markup to LAR Attachment L

Southern Nuclear Operating CompanyAttachment L - NFPA 805 Chapter 3 Requirements for Approval Approval Request 4 NFPA 805 Section 3.3.5.2 states:

"Only metal tray and metal conduits shall be used for electrical raceways. Thin wall metallic tubing shall not be used for power, instrumentation, or control cables. Flexible metallic conduits shall only be used in short lengths to connect components."

Plant Hatch uses embedded PVC conduits. In addition, flexible metallic and PVC coated flexible metallic conduits in lengU1s of 1:1p to 6 feet lengths greater than 3-feet are used to route cables between equipment and rigid conduits. This exceeds the d-foet maximum allowable "short length" as clarified in FAQ 06-0021.

SNC requests NRC approval forJ,1 the use of nonmetallic conduit in embedded applications, 2) existing installations and for the 1:1se of flexible metallic and PVC coated flexible metallic conduits in lengths greater than 3-feet, and 3) the future use of flexible metallic and PVC coated flexible metallic conduits in lengths up to 6-feet.. as acceptable variances from the requirements of NFPA 805, Chapter 3.

Basis for Request:

The basis for the approval request of the deviation for the use of flexible metallic and PVC coated flexible metallic conduits in lengths 1:1p to 6 greater than 3-feet in existing installations is:

Current cable/raceway installation procedures allow for flexible metallic conduit installations up to 6-feet. Flexible metallic conduit installations that exceed 6-feet require Architect/Engineer (AIE) approval. An implementation item will ensure revision to cable/raceway installation guidance that limits future flexible metallic conduit installations to a maximum of 6 feet (See Attachment S, Table S-3, Implementation Item IMP-20).

PVC coated flexible metallic conduit provides equivalent physical and electrical protection to uncoated flexible metallic conduit, because the characteristics of the metallic body of the conduit are not affected by the coating.

According to vendor specifications, the PVC coating on the metallic conduit is very thin and is not expected to provide any credible influence on fire propagation behavior and the amount of PVC introduced to a given fire area is considered negligible.

If a fire were to occur in a fire area containing these conduits, existing controls such as fire-rated barriers, electrical raceway fire barrier systems, spatial separation, etc. would ensure redundant cabling and circuitry would not be affected by the fire.

HNP Page L-16 FPE RA102 FPE RAI02 FPE

RAI 02

Southern Nuclear Operating CompanyAttachment L - NFPA 805 Chapter 3 Requirements for Approval Flexible metallic and PVC coated flexible metallic conduits exceeding the 3-foot length clarified in FAQ 06-0021 are installed such that the conduits are not in danger of being damaged by equipment or personnel.

The basis for the approval request of the deviation for the future use of flexible metallic and PVC coated flexible metallic conduits in lengths up to 6-feet is similar to that for flexible metallic and PVC coated flexible metallic conduits in lengths greater than 3-feet above, with the exception of the assurance that future installation of flexible conduit will be no greater than 6-feet in length. As committed above, an implementation item will ensure revision to cable/raceway installation guidance that limits future flexible metallic*

conduit installations (including PVC coated flexible metallic conduit) to a maximum of 6-feet (See Attachment S, Table S-3, Implementation Item IMP-20).

The basis for the approval request of the deviation for the use of nonmetallic conduit in embedded applications is:

For instances where nonmetallic conduit is used in concrete embedded applications, the concrete provides physical protection and separation for the conduit.

The embedded PVC conduits, while combustible material, are not subject to flame or heat impingement from an external source which would result in structural failure, contribution to the fire load, and/or damage to circuits contained within where the conduit is embedded in concrete and exposure is minimal.

NFPA 70 (National Electric Code (NEC)), Article 352, allows the use of rigid nonmetallic conduit for underground and embedded applications.

Failure of circuits within embedded non-metallic conduits resulting in a fire would not result in damage to external targets (i.e., other circuits would not be exposed to the effects of a circuit failure in the embedded conduit).

The non-metallic conduits are installed such that the conduits are not in danger of being damaged by equipment or personnel.

Failure of circuits 'l.'itt:liR ROA metallio coRduits resultiRg iR a fire 'Nould Rot result iR aamage to eM-terRal targets (i.e., other circuits woula Rot be e><posea to tt:le effects of a circuit failure iR tt:le coRauit).

Acceptance Criteria Evaluation:

Nuclear Safety and Radiological Release Performance Criteria:

The use of nonmetallic conduit in embedded applications aRa tt:le use of fle><ible PVC coated metallic coRduits iR leRgtt:ls up to 6 feet does not affect nuclear safety as the material in which conduits are run are located such that they are not subject to failure mechanisms that potentially result in circuit damage or damage to external targets.

HNP Page L-17 FPE IRAI02 FPE RAI02 FPE IRAI02 FPE RAI02 I

FPE

RAI 02

Southern Nuclear Operating CompanyAttachment L - NFPA 805 Chapter 3 Requirements for Approval Additionally, NFPA 70 allows for the use of rigid nonmetallic conduit for underground and embedded applications.

Also. the use of flexible conduit (both metallic and PVC coated metallic) in lengths greater than 3-feet does not affect nuclear safety. If a fire were to occur in a fire area containing these conduits, existing controls such as fire-rated barriers, electrical raceway fire barrier systems, spatial separation, etc. would ensure redundant cabling and circuitry would not be affected by the fire. Therefore, there is no impact on the nuclear safety performance criteria.

The use of nonmetallic conduit in embedded applications and the use of flexible conduit (both metallic and PVC coated metallic} conduits in lengths greater than 3-feet up toe feet have no impact on the radiological release performance criteria. The radiological release review was performed based on the manual fire suppression activities in areas containing or potentially containing radioactive materials and is not dependent on the type of conduit material. The conduit material or length of conduit does not change the radiological release evaluation, which concluded that potentially contaminated water is contained and smoke is monitored. The conduits for which NRC approval is requested do not add additional radiological materials to the area or challenge system boundaries.

Safety Margin and Defense-in-Depth:

The areas with nonmetallic conduit in embedded applications and flexible PVC coated metallic conduits (both metallic and PVC coated metallic) in lengths up to e greater than J-feet have been analyzed in their current configuration. The precautions and limitations of the use of these materials do not impact the analysis of the fire event. PVC coated flexible metallic conduit introduces a negligible amount of combustibles to a fire area due to the thickness of the PVC coating. Although, the PVC coating introduces a potential smoke toxicity issue due to its corrosive nature to electrical circuits and sensitive electronics in the event of a fire, the PVC coating is of minimal thickness and would not result in smoke production that would impact electrical circuits or sensitive electronics. This conclusion also applies to any future installations of PVC coated flexible metallic conduit in lengths not to exceed 6-feet. Embedded nonmetallic conduit is protected from an exposure fire and possible mechanical damage. PVC conduit that is not embedded The PVC coating on flexible metallic conduit introduces a negligible amount of combustibles to an area. Therefore, the inherent safety margin and conservatisms in these methods remain unchanged.

The three echelons of defense-in-depth are:

(1) To prevent fires from starting (combustible/hot work controls)

(2) Rapidly detect, control and extinguish fires that do occur, thereby limiting damage (fire detection systems, automatic fire suppression, manual fire suppression, pre-fire plans)

(3) Provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed (fire barriers, fire rated cable, success path remains free of fire damage, recovery actions)

HNP Page L-18 I

FPE

RAI 02

FPE

RAI 02

FPE RAI02 FPE

RAI 02

Southern Nuclear Operating CompanyAttachment L - NFPA 805 Chapter 3 Requirements for Approval Per NFPA 805 Section 1.2, defense-in-depth is achieved when an adequate balance of each of these elements is provided.

The current configuration of the conduit at Plant Hatch, and the future use of flexible conduit (metallic and PVC coated metallic) in lengths up to 6-feet, does not impact fire protection defense-in-depth.

PVC coated flexible metallic conduit used within the plant is constructed of a metallic core coated with a thin layer of PVC. The metal core is expected to withstand any potential exposure fire or flame impingement and the PVC coating is thin enough so that it is not expected to provide any credible influence on fire propagation behavior, therefore not affecting the three echelons of defense-in-depth. When installed in configurations exceeding 3 feet in length, the conduit is not expected to negatively affect the three echelons of defense-in-depth as the additional combustibles added by exceeding 3 feet in length is negligible.

Nonmetallic conduit in embedded applications does not affect the three echelons of defense-in-depth. The use of nonmetallic conduits in embedded applications has no effect on the ability for the plant to rapidly detect, control and extinguish any fires that may occur. Additionally, embedded conduit will be shielded from an exposure fire.

Lastly, failure of circuits within embedded non-metallic conduits resulting in a fire would not result in damage to redundant circuits and would not prevent essential safety functions from being performed. iR e\\!ery area of tl=te plaRt wl=tere reEhrndaRt patl=tways or required safe sl=tutdov.'R related cables are located, oRe patl=t*Nay is protected 'Nitl=t a fire protectioR barrier allowiRg for esseRtial safety fuRctioRs to be completed.

The use of these conduits does not directly result in compromising automatic fire suppression functions, manual fire suppression functions, or post-fire safe shutdown capability, and will not prevent essential functions from being performed.

==

Conclusion:==

NRC approval is requested for~

the use of nonmetallic conduit in embedded applications, 2) existing installations a Rd for tl=te use of flexible metallic and PVC coated flexible metallic conduits in lengths greater than 3-feet. and 3) the future use of flexible metallic and PVC coated flexible metallic conduits in lengths up to 6-feet. The engineering analysis performed determined that the performance-based approach utilized to evaluate a variance from the requirements of NFPA 805 Chapter 3:

HNP (A) Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; (B) Maintains safety margins; and (C) Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire shutdown capability).

Page L-19 I

FPE RAI02 FPE IRAI02 FPE RAI02 FPE RAI02

Southern Nuclear Operating CompanyAttachment L - NFPA 805 Chapter 3 Requirements for Approval elevation 130' and the bottom of the fire zone at elevation 87'. Class Ill hose stations are available in the fire zones above, 1203F and 2203F respectively.

The Class Ill hose stations are located approximately 75 feet from the top of the enclosed stairway leading down to Fire Zones 1203C and 2203C. These hose stations therefore provide some, but not full, redundant manual suppression coverage in Fire Zone 1205C. Fire Zones 1203C and 2203C are also protected by automatic wet-pipe sprinkler systems. Water flow alarms will initiate response by the plant fire brigade.

Fire Zones 1205Z and 2205Z have upper level metal grate platform accessed by a metal grate open stairwell. Additional 1 %-inch fire hose stations are located in the adjacent Fire Zones 1205B and 2205B near the doorways leading to Fire Zones 1205Z and 2205Z. These additional hose stations provide redundant coverage to the entire floor level of Fire Zones 1205Z and 2205Z, which contains the majority of the hazards. Fire Zones 1205Z and 2205Z are also protected by automatic wet-pipe sprinkler systems. These fire zones also have fire extinguishers on the 97' elevation and the 118' elevation. Tl=te Rortl=t wall eo1:JRetapt to Fire ZoRe 22058 is proteotee ey etireotioRal spray Rozzles. The wet-pipe sprinkler system is expected to control the fire and initiate a water flow alarm both locally and in the MCR to initiate response by the plant fire brigade.

Unit 1 Hose Stations HS-ROS, HS-R07, HS-ROB and HS-R09 are installed to protect the contiguous Torus Fire Zones 1205A and 1203A. Similarly, Unit 2 Hose Stations 2HS-R06, 2HS-R07, 2HS-R08 and 2HS-R09 are installed to protect the contiguous Torus Fire Zones 2205A and 2203A. The spacing and hose lengths are sufficient to reach all portions of these fire zones with almost fully-overlapping redundant coverage. There are also wall-mounted fire extinguishers located next to two of the four hose stations in each unit. These fire zones are partially protected by a common wet-pipe sprinkler system and linear heat detection system at the open boundary between Fire Area 1203 (Fire Zone 1203A) and Fire Area 1205 (Fire Zone 1205A) in Unit 1 and between Fire Area 2203 (Fire Zone 2203A) and Fire Area 2205 (Fire Zone 2205A) in Unit 2. The suppression systems are designed to prevent fire propagation across the open fire area boundaries between 1203A and 1205A in Unit 1 (2203A and 2205A in Unit 2). Actuation of the installed suppression or detection system provides both local and MCR alarms to initiate response by the plant fire brigade.

The fire brigade members are trained in the use standpipe and hose systems, as well as the use backup capabilities located in adjacent fire areas/fire zones when needed. NMP-TR-426, Fire Training Program, indicates that backup lines (i.e.,

safety lines) from independent water supplies are used to reinforce and protect personnel in case the initial attack line proves inadequate. NMP-TR-426 also describes the responsibility of the Site Lead Fire Instructor to ensure adequate protection for personnel on training attack lines by always providing backup lines.

The fire brigade will properly use the hose stations in adjacent fire zones identified herein which provide additional/redundant hose coverage.

HNP Page L-38