NL-17-1962, Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors

Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors

ML18158A583
Person / Time
Site:	Hatch
Issue date:	06/07/2018
From:	Gayheart C Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-17-1962
Download: ML18158A583 (56)

Text

A Southern Nuclear JUN 0 7 2018 Docket Nos.: 50-321 50-366 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Cheryl A. Gayheart Regulatory Affairs Director Edwin I. Hatch Nuclear Plant-Units 1 &2 3535 Colonnade Parkway Birmingham, AL 35243 205 992 5316 tel 205 992 7601 fax cagayhea@ southernco.com NL-17-1962 10 CFR 50.90 10 CFR 50.69 Application to Adopt 10 CFR 50.69... Risk-informed categorization and treatment of structures.

systems and components for nuclear power reactors..

In accordance with the provisions of Title 10 of the Code of Federal Regulations (1 0 CFR), Part 50.69 and 10 CFR 50.90, Southern Nuclear Operating Company (SNC) is requesting an amendment to the renewed licenses of Edwin I. Hatch, Units 1 (DPR-57) and 2 (NPF-5).

The proposed amendment would modify the licensing bases, by the addition of a License Condition, to allow for the implementation of the provisions of Part 10 CFR 50.69,.. Risk-informed categorization and treatment of structures, systems and components (SSCs) for nuclear power reactors... The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance, alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance, requirements will not be changed or will be enhanced. This allows improved focus on equipment that has safety significance resulting in improved plant safety.

The enclosure to this letter provides the basis for the proposed change to the Hatch, Units 1 and 2 Operating Licenses. The categorization process being implemented through this change is consistent with Nuclear Energy Institute (NEI) 00-04, "1 0 CFR 50.69 SSC Categorization Guideline,.. Revision 0, dated July 2005 which was endorsed by the Nuclear Regulatory Commission (NRC) in Regulatory Guide 1.201,.. Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance,.. Revision 1, dated May 2006. Attachment 1 of the enclosure provides a list of categorization prerequisites.

Use of the categorization process on a plant system will only occur after these prerequisites are met.

Though routine maintenance updates have been applied, the NRC has previously reviewed the technical adequacy of the Hatch Probabilistic Risk Assessment (PRA) internal event/internal flood model quality identified in this application for the Surveillance Frequency Control Program (NRC Safety Evaluation Report ML 111 OBA 129) dated January 3, 2012. SNC requests that the NRC utilize the review of the PRA technical adequacy for that application when performing the review for this application.

U. S. Nuclear Regulatory Commission NL-17-1962 Page 2 SNC submitted a separate license amendment request for.. Transition to 10 CFR 50.48(c) -

NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition),.. by letter dated April 4, 2018 using the same PRA models described in the enclosure. SNC requests that the NRC coordinate its review of the PRA technical adequacy description in Sections 3.2 and 3.3 of this enclosure for both applications.

This would reduce the number of SNC and NRC resources necessary to complete the review of the applications. This request should not be considered a linked requested licensing action, as the details of the PRA models in each LAR are complete which will allow the NRC staff to independently review and approve each LAR on their own merits without regard to the results from the review of the other.

SNC requests approval of the proposed license amendment by exactly one year of the date of this letter, with the amendment being implemented within ninety days of receipt of the NRC Safety Evaluation.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated Georgia Official.

This letter contains no regulatory commitments. This letter does contain License Conditions described in Attachment 1 to the enclosure.

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

_1_ day of June 2018.

Cheryl A}.

eart Director, Regulatory Affairs Southern Nuclear Operating Company CAG/PDB/SCM

Enclosure:

Evaluation of the Proposed Change 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 &2 Application to Adopt 10 CFR 50.69,.. Risk-Informed categorization and treatment of structures, systems, and components (SSCs) for nuclear power reactors..

Enclosure Evaluation of the Proposed Change

Enclosure to NL-17-1962 Evaluation of the Proposed Change Enclosure Evaluation of the Proposed Change TABLE OF CONTENTS 1

SUMMARY

DESCRIPTION............................................................................................ E-1 2

DETAILED DESCRIPTION............................................................................................. E-1 2.1 CURRENT REGULATORY REQUIREMENTS....................................................... E-1 2.2 REASON FOR PROPOSED CHANGE................................................................... E-1

2.3 DESCRIPTION

OF THE PROPOSED CHANGE.................................................... E-2 3

TECHNICAL EVALUATION............................................................................................ E-3 3.1 CATEGORIZATION PROCESS DESCRIPTION (10 CFR 50.69(b)(2)(i))............... E-4 3.1.1 Overall Categorization Process................................................................. E-4 3.1.2 Passive Categorization Process............................................................... E-8 3.2 TECHNICAL ADEQUACY EVALUATION (10 CFR 50.69(b)(2)(ii))......................... E-9 3.2.1 Internal Events and Internal Flooding........................................................ E-9 3.2.2 Fire Hazards............................................................................................. E-9 3.2.3 Seismic Hazards....................................................................................... E-9 3.2.4 Other External Hazards.......................................................................... E-1 0 3.2.5 Low Power & Shutdown.......................................................................... E-1 0 3.2.6 PRA Maintenance and Updates.............................................................. E-1 0 3.2.7 PRA Uncertainty Evaluations.................................................................. E-11 3.3 PRA REVIEW PROCESS RESULTS (10 CFR 50.69(b)(2)(iii))......... ;.................. E-12 3.4 RISK EVALUATIONS (10 CFR 50.69(b)(2)(iv))................................................... E-13 4

REGULATORY EVALUATION..................................................................................... E-14 4.1 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA............................... E-14 4.2 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS.............................. E-14

4.3 CONCLUSION

S................................................................................................... E-15 5

ENVIRONMENTAL CONSIDERATION........................................................................ E-16 6

REFERENCES............................................................................................................. E-17 LIST OF ATTACHMENTS : List of Categorization Prerequisites............................................................... E-19 : Description of PRA Models Used in Categorization........................................ E-29 : Disposition and Resolution of Open Peer Review Findings........................... E-31 : External Hazards Screening........................................................................... E-39 : Progressive Screening Approach for Addressing External Hazards............... E-47 : Disposition of Key Assumptions/Sources of Uncertainty................................ E-56 E-i

Enclosure to NL-17-1962 Evaluation of the Proposed Change 1

SUMMARY

DESCRIPTION The proposed amendment would modify the licensing basis to allow for the implementation of the provisions of Title 10 of the Code of Federal Regulations (1 0 CFR), Part 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Reactors." The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance, alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance, requirements will not be changed or will be enhanced. This allows improved focus on equipment that has safety significance resulting in improved plant safety.

2 DETAILED DESCRIPTION 2.1 CURRENT REGULATORY REQUIREMENTS The Nuclear Regulatory Commission (NRC) has established a set of regulatory requirements for commercial nuclear reactors to ensure that a reactor facility does not impose an undue risk to the health and safety of the public, thereby providing reasonable assurance of adequate protection to public health and safety. The current body of NRC regulations and their implementation are largely based on a.. deterministic.. approach.

This deterministic approach establishes requirements for engineering margin and quality assurance in design, manufacture, and construction. In addition, it assumes that adverse conditions can exist (e.g., equipment failures and human errors) and establishes a specific set of design basis events (DBEs). The deterministic approach then requires that the facility include safety systems capable of preventing or mitigating the consequences of those DBEs to protect public health and safety. Those structures, systems, and components (SSCs) necessary to defend against the DBEs are defined as.. safety-related,.. and these SSCs are the subject of many regulatory requirements, herein referred to as "special treatments," designed to ensure that they are of high quality and high reliability, and have the capability to perform during postulated design basis conditions. Treatment includes, but is not limited to, quality assurance, testing, inspection, condition monitoring, assessment, evaluation, and resolution of deviations. The distinction between.. treatmene and.. special treatment.. is the degree of NRC specification as to what must be implemented for particular SSCs or for particular conditions.

Typically, the regulations establish the scope of SSCs that receive special treatment using one of three different terms:.. safety-related,.... important to safety,.. or.. basic component... The terms

.. safety-related.. and.. basic component.. are defined in the regulations, while.. important to safety,.. used principally in the general design criteria (GDC) of Appendix A to 10 CFR Part 50, is not explicitly defined.

2.2 REASON FOR PROPOSED CHANGE A probabilistic approach to regulation enhances and extends the traditional deterministic approach by allowing consideration of a broader set of potential challenges to safety, providing a logical means for prioritizing these challenges based on safety significance, and allowing consideration of a broader set of resources to defend against these challenges. In contrast to the deterministic approach, Probabilistic Risk Assessments (PRAs) address credible initiating events by assessing the event frequency. Mitigating system reliability is then assessed, E-1

Enclosure to NL-17-1962 Evaluation of the Proposed Change including the potential for common cause failures. The probabilistic approach to regulation is an extension and enhancement of traditional regulation by considering risk in a comprehensive manner.

To take advantage of the safety enhancements available through the use of PRA, in 2004 the NRC published a new regulation, 10 CFR 50.69. The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance, alternative treatment requirements can be implemented in accordance with the regulation. For equipment determined to be of high safety significance, requirements will not be changed or will be enhanced. This allows improved focus on equipment that has safety significance resulting in improved plant safety.

The rule contains requirements on how a licensee categorizes SSCs using a risk-informed process, adjusts treatment requirements consistent with the relative significance of the sse, and manages the process over the lifetime of the plant. A risk-informed categorization process is employed to determine the safety significance of SSCs and place the SSCs into one of four risk-informed safety class (RISC) categories. The determination of safety significance is performed by an integrated decision-making process, as described by NEI 00-04, "1 0 CFR 50.69 SSC Categorization Guideline" (Reference 1 ), which uses both risk insights and traditional engineering insights. The safety functions include the design basis functions, as well as functions credited for severe accidents (including external events). Special or alternative treatment for the SSCs is applied as necessary to maintain functionality and reliability, and is a function of the SSC categorization results and associated bases. Finally, periodic assessment activities are conducted to make adjustments to the categorization and/or treatment processes as needed so that SSCs continue to meet all applicable requirements.

The rule does not allow for the elimination of SSC functional requirements or allow equipment that is* required by the deterministic design basis to be removed from the facility. Instead, the rule enables licensees to focus their resources on SSCs that make a significant contribution to plant safety. For SSCs that are categorized as high safety significant, existing treatment requirements are maintained or enhanced. Conversely, for SSCs that do not significantly contribute to plant safety on an individual basis, the rule allows an alternative risk-informed approach to treatment that provides reasonable, though reduced, level of confidence that these SSCs will satisfy functional requirements.

Implementation of 10 CFR 50.69 will allow Southern Nuclear Operating Company (SNC) to improve focus on equipment that has safety significance resulting in improved plant safety.

2.3 DESCRIPTION

OF THE PROPOSED CHANGE SNC proposes the addition of the following condition to the renewed operating licenses of Hatch, Units 1 and 2 to document the NRC's approval of the use of 10 CFR 50.69.

Southern Nuclear Operating Company is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) specified in the renewed license amendment dated DATE.

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Enclosure to NL-17-1962 Evaluation of the Proposed Change Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

The Southern Nuclear Operating Company shall complete all items listed in Attachment 1, List of Categorization Prerequisites, of Southern Nuclear Operating Company letter ML Number, dated DATE, prior to implementation.

3 TECHNICAL EVALUATION 10 CFR 50.69 specifies the information to be provided by a licensee requesting adoption of the regulation. This request conforms to the requirements of 10 CFR 50.69(b)(2), which states:

A licensee voluntarily choosing to implement this section shall submit an application for license amendment under § 50.90 that contains the following information:

(i) A description of the process for categorization of RISC-1, RISC-2, RISC-3 and RISC-4 SSCs.

(ii) A description of the measures taken to assure that the quality and level of detail of the systematic processes that evaluate the plant for internal and external events during normal operation, low power, and shutdown (including the plant-specific PRA, margins-type approaches, or other systematic evaluation techniques used to evaluate severe accident vulnerabilities) are adequate for the categorization of SSCs.

(iii) Results of the PRA review process conducted to meet§ 50.69(c)(1)(i).

(iv) A description of, and basis for acceptability of, the evaluations to be conducted to satisfy§ 50.69(c)(1 )(iv). The evaluations must include the effects of common cause interaction susceptibility, and the potential impacts from known degradation mechanisms for both active and passive functions, and address internally and externally initiated events and plant operating modes (e.g., full power and shutdown conditions).

Each of these submittal requirements are addressed in the proceeding sections.

Though routine maintenance updates have been applied, the NRC has previously reviewed the technical adequacy of the Hatch PRA internal event/internal flooding model quality identified in this application for the Surveillance Frequency Control Program (NRC Safety Evaluation Report (SER) ML 111 OBA 129) dated January 3, 2012. SNC requests that the NRC utilize the review of the PRA technical adequacy for that application when performing the review for this application.

SNC submitted a separate license amendment request (LAR) for.. Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition),.. using the same PRA models described in this enclosure. SNC requests that the NRC coordinate their review of the PRA technical adequacy description in Sections 3.2 and 3.3 of this enclosure for both applications. This would reduce the number of SNC and NRC resources necessary to complete the review of the applications.

This request should not be considered a linked requested licensing action (RLA), as the details of the PRA models in each LAR are complete which will allow the NRC staff to independently E-3

Enclosure to NL-17-1962 Evaluation of the Proposed Change review and approve each LAR on their own merits without regard to the results from the review of the other.

3.1 CATEGORIZATION PROCESS DESCRIPTION (1 0 CFR 50.69(b)(2)(i))

3.1.1 Overall Categorization Process SNC will implement the risk categorization process in accordance with the NEI 00-04, Revision 0, "1 0 CFR 50.69 SSC Categorization Guideline" (Reference 1) as endorsed by Regulatory Guide (RG) 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance," (Reference 2). NEI 00-04 Section 1.5 states "Due to the varying levels of uncertainty and degrees of conservatism in the spectrum of risk contributors, the risk significance of SSCs is assessed separately from each of five risk perspectives and used to identify SSCs that are potentially safety-significant."

Separate evaluation is appropriate to avoid reliance on a combined result that may mask the results of individual risk contributors.

The process to categorize each system will be consistent with the guidance in NEI 00-04, "1 0 CFR 50.69 SSC Categorization Guideline," as endorsed by RG 1.201. RG 1.201 states that "the implementation of all processes described in NEI 00-04 (i.e., Sections 2 through 12) is integral to providing reasonable confidence" and that "all aspects of NEI 00-04 must be followed to achieve reasonable confidence in the evaluations required by §50.69(c)(1 )(iv)."

However, neither RG 1.201 nor NEI 00-04 prescribe a particular sequence or order for each of the elements to be completed. Therefore, the order in which each of the elements of the categorization process (listed below) is completed is flexible, and as long as they are all completed they may even be performed in parallel. Note that NEI 00-04 only requires Item 3 to be completed for components/functions categorized as Low Safety Significance (LSS) by all other elements. Similarly, NEI 00-04 only requires Item 4 to be completed for safety related active components/functions categorized as LSS by all other elements.

1. PRA-based evaluations (e.g., the internal events, internal flooding, and fire PRAs) non-PRA approaches (e.g., fire safe shutdown equipment list (FSEL), seismic safe shutdown equipment list (SSEL), other external events screening, and shutdown assessment)

2. Seven qualitative criteria in Section 9.2 of NEI 00-04

3. The defense-in-depth assessment

4. The passive categorization methodology Categorization of SSCs will be completed per the NEI 00-04 process, as endorsed by RG 1.201, which includes the determination of safety significance through the various elements identified above. The results of these elements are used as inputs to arrive at a preliminary component categorization (i.e., High Safety Significant (HSS) or LSS) that is presented to the Integrated Decision-Making Panel (lOP). Note: the term "preliminary HSS or LSS" is synonymous with the NEI 00-04 term "candidate HSS or LSS."

A component or function is preliminarily categorized as HSS if any element of the process results in a preliminary HSS determination in accordance with Table 3-1 below. The safety significance determination of each element, identified above, is independent of each other and therefore the sequence of the elements does not impact the resulting preliminary categorization of each component or function. Consistent with NEI 00-04, the categorization of a component or function will only be "preliminary until it has been confirmed by the IDP. Once the IDP E-4

Enclosure to NL-17-1962 Evaluation of the Proposed Change confirms that the categorization process was followed appropriately, the final RISC category can be assigned.

The IDP may direct and approve detailed categorization of components in accordance with NEI 00-04 Section 1 0.2. The IDP may always elect to change a preliminary LSS component or function to HSS, however the ability to change component categorization from preliminary HSS to LSS is limited. This ability is only available to the IDP for select process steps as described in NEI 00-04 and endorsed by RG 1.201. Table 3-1 summarizes these IDP limitations in NEI 00-04. The steps of the process are performed at either the function level, component level, or both. This is also summarized in the Table 3-1. A component is assigned its final RISC category upon approval by the IDP.

T bl a e 3-1: c ategonzat1on E I

va uat1on s urn mary Categorization Step -

IDP Change Drives Element Evaluation Level Associated NEI 00-04 Section HSS to LSS Functions Internal Events Base Not Allowed Yes Case - Section 5.1 Fire, Seismic and Other External Allowable No Risk (PRA Events Base Case Component Modeled)

PRA Sensitivity Studies Allowable No Integral PRA Assessment -

Not Allowed Yes Section 5.6 Fire, Seismic and Risk (Non-Other External Component Not Allowed No modeled)

Hazards-Shutdown - Section 5.5 Function/Component Not Allowed No Core Damage -

Function/Component Not Allowed Yes Defense-in-Section 6.1 Depth Containment -

Component Not Allowed Yes Section 6.2 Qualitative Considerations-Function Allowable1 N/A Criteria Section 9.2 Passive Passive - Section 4 SeQment/Component Not Allowed No Notes:

1 The assessments of the qualitative considerations are agreed upon by the lOP in accordance with Section 9.2 (Reference 1). In some cases, a 50.69 categorization team may provide preliminary assessments of the seven considerations for the lOP's consideration, however the final assessments of the seven considerations are the direct responsibility of the lOP.

The seven considerations are addressed preliminarily by the 50.69 categorization team for at least the system functions that are not found to be HSS due to any other categorization step. Each of the seven considerations requires a supporting justification for confirming (true response) or not confirming (false response) that consideration. If the 50.69 categorization team determines that one or more of the seven considerations cannot be confirmed, then that function is presented to the lOP as preliminary HSS.

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Enclosure to NL-17-1962 Evaluation of the Proposed Change Conversely, if all the seven considerations are confirmed, then the function is presented to the IDP as preliminary LSS.

The System Categorization Document, including the justifications provided for the qualitative considerations, is reviewed by the IDP. The IDP is responsible for reviewing the preliminary assessment to the same level of detail as the 50.69 team (i.e. all considerations for all functions are reviewed). The IDP may confirm the preliminary function risk and associated justification or may direct that it be changed based upon their expert knowledge. Because the Qualitative Criteria are the direct responsibility of the IDP, changes may be made from preliminary HSS to LSS or from preliminary LSS to HSS at the discretion of the IDP. If the IDP determines any of the seven considerations cannot be confirmed (false response) for a function, then the final categorization of that function is HSS.

The mapping of components to system functions is used in some categorization process steps to facilitate preliminary categorization of components. Specifically, functions with mapped components that are determined to be HSS by the PRA-based assessment (i.e., Internal events PRA or Integral PRA assessment) or defense-in-depth evaluation will be initially treated as HSS. However, NEI 00-04 Section 10.2 allows detailed categorization which can result in some components mapped to HSS functions being treated as LSS; and NEI 00-04 Section 4 discusses additional functions that may be identified (e.g., fill and drain) to group and consider potentially LSS components that may have been initially associated with a HSS function but which do not support the critical attributes of that HSS function. Note that certain steps of the categorization process are performed at a component level (e.g. Passive, Non-PRA-modeled hazards-see Table 3-1). These components from the component level assessments will remain HSS (IDP cannot override) regardless of the significance of the functions to which they are mapped. Therefore, if a HSS component is mapped to a LSS function, that component will remain HSS. If an LSS component is mapped to an HSS function, that component may be driven HSS based on Table 3-1 above! or may remain LSS.

The following are clarifications to be applied to the NEI 00-04 categorization process:

The IDP will be composed of a group of at least five experts who collectively have expertise in plant operation, design (mechanical and electrical) engineering, system engineering, safety analysis, and probabilistic risk assessment. At least three members of the IDP will have a minimum of five years of experience at the plant, and there will be at least one member of the IDP who has a minimum of three years of experience in the modeling and updating of the plant-specific PRA.

The IDP will be trained in the specific technical aspects and requirements related to the categorization process. Training will address at a minimum the purpose of the categorization; present treatment requirements for SSCs including requirements for design basis events; PRA fundamentals; details of the plant specific PRA including the modeling, scope, and assumptions, the interpretation of risk importance measures, and the role of sensitivity studies and the change-in-risk evaluations; and the defense-in-depth philosophy and requirements to maintain this philosophy.

The decision criteria for the IDP for categorizing SSCs as safety-significant or low safety-significant pursuant to 10 CFR 50.69(f)(1) will be documented in SNC procedures. Decisions of the IDP will be arrived at by consensus. Differing opinions will be documented and resolved, if possible. If a resolution cannot be achieved concerning the safety significance of an sse, then the sse will be classified as safety-significant.

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Enclosure to NL-17-1962 Evaluation of the Proposed Change Passive characterization will be performed using the processes described in Section 3.1.2. Consistent with NEI 00-04, an HSS determination by the passive categorization process cannot be changed by the lOP.

An unreliability factor of 3 will be used for the sensitivity studies described in Section 8 of NEI 00-04. The factor of 3 was chosen as it is representative of the typical error factor of basic events used in the PRA model.

NEI 00-04 Section 7 requires assigning the safety significance of functions to be preliminary HSS if it is supported by an SSC determined to be HSS from the PRA-based assessment in NEI 00-04 Section 5, but does not require this for SSCs determined to be HSS from non-PRA-based, deterministic assessments in Section 5. This requirement is further clarified in the Vogtle SER (Reference 6) which states "...if any SSC is identified as HSS from either the integrated PRA component safety significance assessment (Section 5 of NEI 00-04) or the defense-in-depth assessment (Section 6), the associated system function(s) would be identified as HSS."

Once a system function is identified as HSS, then all the components that support that function are preliminary HSS. The lOP must intervene to assign any of these HSS Function components to LSS.

With regard to the criteria that considers whether the active function is called out or relied upon in the plant Emergency/ Abnormal Operating Procedures, SNC will not take credit for alternate means unless the alternate means are proceduralized and included in Licensed Operator training.

The risk analysis being implemented for each hazard is described below:

Internal Event Risks: Unit 1 internal events including internal flooding PRA model Revision 5, dated July 2017. Unit 2 internal events including internal flooding PRA model Revision 5.2, dated September 2017.

Fire Risks: Units 1 and 2 Fire PRA model Revision 1, April 2016.

Seismic Risks: Units 1 and 2 Seismic PRA model Revision 3, October 2017.

Other External Risks (e.g., tornados, external floods, etc.): An evaluation of external hazards was performed using Part 6 of the ASMEIANS PRA Standard (Reference 3).

These other external hazards were determ*ined to be insignificant contributors to plant risk.

Low Power and Shutdown Risks: Qualitative defense-in-depth (DID) shutdown model for shutdown configuration risk management (CAM) based on the framework for DID provided in NUMARC 91-06, "Guidance for Industry Actions to Assess Shutdown Management" (Reference 4), which provides guidance for assessing and enhancing safety during shutdown operations.

A change to the categorization process that is outside the bounds specified above (e.g.,

change from a seismic margins approach to a seismic probabilistic risk assessment approach) will not be used without prior NRC approval. The SSC categorization process documentation will include the following elements:

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Enclosure to NL-17-1962 Evaluation of the Proposed Change

1. Program procedures used in the categorization

2. System functions, identified and categorized with the associated bases

3. Mapping of components to support function(s)

4. PRA model results, including sensitivity studies

5. Hazards analyses, as applicable

6. Passive categorization results and bases

7. Categorization results including all associated bases and RISC classifications

8. Component critical attributes for HSS SSCs

9. Results of periodic reviews and SSC performance evaluations

10. lOP meeting minutes and qualification/training records for the lOP members 3.1.2 Passive Categorization Process For the purposes of 10 CFR 50.69 categorization, passive components are those components that have a pressure retaining function. Passive components and the passive function of active components will be evaluated using the Risk-Informed Repair/Replacement Activities (RI-RRA) methodology consistent with the SER by the Office of Nuclear Reactor Regulation.. Request for Alternative AN02-R&R-004, Revision 1, Request to Use Risk-informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 and 3 Moderate and High Energy Systems, Third and Fourth 1 0-Year In-service Inspection Intervals," dated April 22, 2009 (ML090930246} (Reference 5).

The RI-RRA methodology is a risk-informed safety classification and treatment program for repair/replacement activities for pressure retaining items and their associated supports. In this method, the component failure is assumed with a probability of 1.0 and only the consequence evaluation is performed. It additionally applies deterministic considerations (e.g., defense-in-depth, safety margins) in determining safety significance. Component supports are assigned the same safety significance as the highest passively ranked components within the bounds of the associated analytical pipe stress model. Consistent with NEI 00-04, an HSS determination by the passive categorization process cannot be changed by the I DP.

The use of this method was previously approved to be used for a 10 CFR 50.69 application by NRC in the final Safety Evaluation for Vogtle dated December 17, 2014 (Reference 6). The RI-RRA method as approved for use at Vogtle for 1 0 CFR 50.69 does not have any plant specific aspects and is generic. It relies on the conditional core damage and large early release probabilities associated with postulated ruptures. Safety significance is generally measured by the frequency and the consequence of the event. However, this RI-RRA process categorizes components solely based on consequence, which measures the safety significance of the passive component given that it ruptures. This approach is conservative compared to including the rupture frequency in the categorization as this approach will not allow the categorization of SSCs to be affected by any changes in frequency due to changes in treatment. The passive categorization process is intended to apply the same risk-informed process accepted by the E-8

Enclosure to NL-17-1962 Evaluation of the Proposed Change NRC in the AN02-R&R-004 for the passive categorization of Class 2, 3, and non-class components. This is the same passive SSG scope the NRC has conditionally endorsed in ASME Code Cases N-660 and N-662 as published in Regulatory Guide 1.147, Revision 15.

(Reference 21) Both code cases employ a similar risk-informed safety classification of SSCs in order to change the repair/ replacement requirements of the affected LSS components. All ASME Code Class 1 SSCs with a pressure retaining function, as well as supports, will be assigned HSS, for passive categorization, which will result in HSS for its risk-informed safety classification and cannot be changed by the IDP. Therefore, this methodology and scope for passive categorization is acceptable and appropriate for use at Hatch for 10 CFR 50.69 SSG categorization.

3.2 TECHNICAL ADEQUACY EVALUATION (1 0 CFR 50.69(b)(2)(ii))

The following sections demonstrate that the quality and level of detail of the processes used in categorization of SSCs are adequate. All of the PRA models described below have been peer reviewed and there are no PRA upgrades that have not been peer reviewed. The internal event/internal flooding PRA models credited in this request are the same PRA models credited in the TSTF-425, Revision 3 application dated October 29, 2010, as supplemented by letters dated February 21, 2011; May 27, 2011 (Unit 2 only); and October 13, 2011 (ADAMS Accession Number ML11108A129) (Reference 7) with routine maintenance updates applied.

3.2.1 Internal Events and Internal Flooding The Hatch categorization process for the internal events and internal flooding hazard will use the plant-specific PRA model. The SNC risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for each of the Hatch units. of this enclosure identifies the applicable internal events and internal flooding PRA models.

3.2.2 Fire Hazards The Hatch categorization process for fire hazards will use a peer reviewed plant-specific fire PRA model. The internal Fire PRA model was developed consistent with NUREG/CR-6850 and only utilizes methods previously accepted by the NRC. The SNC risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for each of the Hatch units. Attachment 2 at the end of this enclosure identifies the applicable Fire PRA model.

3.2.3 Seismic Hazards The Hatch categorization process for seismic hazards will use a peer reviewed plant-specific Seismic Probabilistic Risk Assessment (SPRA) model. The SNC risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for each of the Hatch units. Industry standard methods were utilized in the development of the seismic hazards for the SPRA. Updates to seismic hazard curves will be reflected in the PRA used for the categorization in accordance with the PRA model maintenance process. at the end of this enclosure identifies the applicable SPRA model. RG1.200 (Reference 8) endorses ASMEIANS PRA Standard Addendum A (Reference 3) but, as noted in an NRC letter to ASME (Reference 9), does not endorse PRA Standard Addendum 8 (Reference 1 0). The Plant Hatch SPRA peer review was performed using the SPRA E-9

Enclosure to NL-17-1962 Evaluation of the Proposed Change requirements in Addendum B. A generic discussion of the differences between* Addendum A and B is included in the Plant Vogtle.. License Amendment Request to Modify Approved 10 CFR 50.69 Categorization Process," dated June 22, 2017 (ADAMS Accession Number ML17173A875) (Reference 25). The generic discussion of differences is also applicable to the Plant Hatch SPRA peer review. Further, since the Plant Hatch SPRA was peer reviewed relative to Capability Category II as defined in the PRA Standard, and since Capability Category II establishes the appropriate technical capability for most risk-informed applications, including 10 CFR 50.69, the Plant Hatch SPRA meets the technical adequacy of Addendum A of the PRA Standard.

3.2.4 Other External Hazards All other external hazards were screened from applicability to Hatch Units 1 and 2 per a plant-specific evaluation in accordance with the criteria in Section 6 of ASME PRA Standard RA-Sb-2013. (Reference 4) RG 1.200 endorses the RA-Sa-2009 version of the standard. (Reference

8) The 2013 version of the standard contains the same technical requirements as the 2009 version, however editorial changes to the layouts of the tables and attachments were performed. Use of the 2013 version requirements meets all of the 2009 version requirements.

An internal self-assessment of the standard requirements was performed. Attachment 4 provides a summary of the other external hazards screening results. Attachment 5 provides a summary of the progressive screening approach for external hazards.

3.2.5 Low Power & Shutdown Consistent with NEI 00-04, the Hatch categorization process will use the shutdown safety management plan described in Nuclear Management and Resource Council (NUMARC) 91-06 (Reference 4) for evaluation of safety significance related to low power and shutdown conditions. The overall process for addressing shutdown risk is illustrated in Figure 5-7 of NEI 00-04.

NUMARC 91-06 specifies that a defense-in-depth approach should be used with respect to each defined shutdown key safety function. The key safety functions defined in NUMARC 91-06 are evaluated for categorization of SSCs.

SSCs that meet the two criteria (i.e., considered part of a "primary shutdown safety system" or a failure would initiate an event during shutdown conditions) described in Section 5.5 of NEI 00-04 will be considered preliminary HSS.

3.2.6 PRA Maintenance and Updates The SNC risk management process ensures that the applicable PRA models used in this application continue to reflect the as-built and as-operated plant for each of the Hatch units. The process delineates the responsibilities and guidelines for updating the PRA models, and includes criteria for both regularly scheduled and interim PRA model updates. The process includes provisions for monitoring potential areas affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, industry operational experience) for assessing the risk impact of unincorporated changes, and for controlling the model and associated computer files. The process will assess the impact of these changes on the plant PRA model in a timely manner but no longer than once every two refueling outages. If there is a significant impact on the PRA model, the SSC categorization will be re-evaluated.

E-10

Enclosure to NL-17-1962 Evaluation of the Proposed Change In addition, SNC will implement a process that addresses the requirements in NEI 00-04, Section 11, "Program Documentation and Change Control." The process will review the results of periodic and interim updates of the plant PRA that may affect the results of the categorization process. If the results are affected, adjustments will be made as necessary to the categorization or treatment processes to maintain the validity of the processes. In addition, any PRA model upgrades will be peer reviewed prior to implementing those changes in the PRA model used for categorization.

3.2.7 PRA Uncertainty Evaluations Uncertainty evaluations associated with any applicable baseline PRA models used in this application were evaluated during the assessment of PRA technical adequacy and confirmed through the self-assessment and peer review processes as discussed in Section 3.3 of this enclosure.

Uncertainty evaluations associated with the risk categorization process are addressed using the processes discussed in Section 8 of NEI 00-04 and in the prescribed sensitivity studies discussed in Section 5 of NEI 00-04.

In the overall risk sensitivity studies SNC will utilize a factor of 3 to increase the unavailability or unreliability of LSS components consistent with that approved for Vogtle in Reference 6.

Consistent with the NEI 00-04 guidance, SNC will perform both an initial sensitivity study and a cumulative sensitivity study. The initial sensitivity study applies to the system that is being categorized. In the cumulative sensitivity study, the failure probabilities (unreliability and unavailability, as appropriate) of all LSS components modeled in PRAs for all systems that have been categorized are increased by a factor of 3. This sensitivity study together with the periodic review process assures that the potential cumulative risk increase from the categorization is maintained acceptably low. The performance monitoring process monitors the component performance to ensure that potential increases in failure rates of categorized components are detected and addressed before reaching the rate assumed in the sensitivity study.

The detailed process of identifying, characterizing and qualitative screening of model uncertainties is found in Section 5.3 of NUREG-1855 (Reference 11) and Section 3.1.1 of EPRI TR-1 016737 (Reference 12). The process in these references was mostly developed to evaluate the uncertainties associated with the internal events PRA model; however, the approach can be applied to other types of hazard groups.

The list of assumptions and sources of uncertainty were reviewed to identify those which would be significant for the evaluation of this application. If the Hatch PRA model used a non-conservative treatment, or methods which are not commonly accepted, the underlying assumption or source of uncertainty was reviewed to determine its impact on this application.

Only those assumptions or sources of uncertainty that could significantly impact the risk calculations were considered key for this application. This review did not identify any key model specific assumptions or sources of uncertainty. Hatch PRA model specific assumptions and sources of uncertainty that may be key for this application are identified and dispositioned in. The conclusion of this review is that a few system specific sensitivity analyses may be required to address Hatch PRA model specific assumptions or sources of uncertainty.

E-11

Enclosure to NL-17-1962 Evaluation of the Proposed Change 3.3 PRA REVIEW PROCESS RESULTS (1 0 CFR 50.69(b)(2)(iii))

The PRA models described in Section 3.2 have been assessed against RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2 (Reference 8) consistent with NRC RIS 2007-06.

The internal events PRA model was subjected to a self-assessment and a full-scope peer review conducted in November 2009.

The fire PRA model was subjected to a self-assessment and a full-scope peer review conducted in May 2016.

The SPRA model was subjected to a self-assessment and a full-scope peer review conducted in October 2016.

Resolutions to the Findings and Observations (F&Os) resulting from the above peer reviews were reviewed and closed using the process documented in Appendix X to NEI 05-04 (Reference 22), NEI 07-12 (Reference 23) and NEI 12-13, "Close-out of Facts and Observations (F&Os)" (Reference 13) as accepted by the NRC in the staff memorandum dated May 3, 2017 (ML17079A427) (Reference 14). The results of these reviews have been documented and are available for NRC audit. Appendix X closure was performed for the Internal Events model in April2017, performed for the Fire model in October of 2017 and for the Seismic PRA model in June 2017. provides a summary of the remaining findings and open items. All peer review F&Os from the Fire and Seismic PRA peer reviews have been closed. Four findings from the Internal Events/Internal Flooding model remain open. Attachment 3 lists these findings, the proposed resolution, and the potential impact on use of the model for 50.69 evaluations. The attachments identified above demonstrate that the PRA is of sufficient quality and level of detail to support the categorization process, and has been subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC as required 10 CFR 50.69(c)(1 )(i).

E-12

Enclosure to NL-17-1962 Evaluation of the Proposed Change 3.4 RISK EVALUATIONS (1 0 CFR 50.69(b)(2)(iv))

The Hatch 1 0 CFR 50.69 categorization process will implement the guidance in NEI 00-04. The overall risk evaluation process described in the NEI guidance addresses both known degradation mechanisms and common cause interactions, and meets the requirements of

§50.69(b)(2)(iv). Sensitivity studies described in NEI 00-04 Section 8 will be used to confirm that the categorization process results in acceptably small increases to core damage frequency (CDF) and large early release frequency (LERF). The failure rates for equipment and initiating event frequencies used in the PRA include the quantifiable impacts from known degradation mechanisms, as well as other mechanisms (e.g., design errors, manufacturing deficiencies, human errors, etc.). Subsequent performance monitoring and PRA updates required by the rule will continue to capture this data, and provide timely insights into the need to account for any important new degradation mechanisms.

E-13

Enclosure to NL-17-1962 Evaluation of the Proposed Change 4

REGULATORY EVALUATION 4.1 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA The following NRC requirements and guidance documents are applicable to the proposed change.

The regulations in Title 1 0 of the Code of Federal Regulations (1 0 CFR) Part 50.69,

.. Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors...

• NRC Regulatory Guide 1.201,.. Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance," Revision 1, May 2006.

Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2, April 2015.

Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, US Nuclear Regulatory Commission, March 2009.

The proposed change is consistent with the applicable regulations and regulatory guidance.

4.2 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS Southern Nuclear Operating Company (SNC) proposes to modify the licensing basis to allow for the voluntary implementation of the provisions of Title 10 of the Code of Federal Regulations (1 0 CFR), Part 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Reactors." The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g.,

quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance, alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance, requirements will not be changed or will be enhanced. This allows improved focus on equipment that has safety significance resulting in improved plant safety.

SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendments by focusing on the three standards set forth in 10 CFR 50.92,

.. Issuance of amendment,.. as discussed below:

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

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The process used to evaluate SSCs for changes to NRC special treatment requirements and the use of alternative requirements ensures the ability of the SSCs to perform their design function. The potential change to special treatment requirements does not change the design and E-14

Enclosure to NL-17-1962 Evaluation of the Proposed Change operation of the SSCs. As a result, the proposed change does not significantly affect any initiators to accidents previously evaluated or the ability to mitigate any accidents previously evaluated. The consequences of the accidents previously evaluated are not affected because the mitigation functions performed by the SSCs assumed in the safety analysis are not being modified. The SSCs required to safely shut down the reactor and maintain it in a safe shutdown condition following an accident will continue to perform their design functions.

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

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

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not change the functional requirements, configuration, or method of operation of any SSC.

Under the proposed change, no additional plant equipment will be installed.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not affect any Safety Limits or operating parameters used to establish the safety margin.

The safety margins included in analyses of accidents are not affected by the proposed change. The regulation requires that there be no significant effect on plant risk due to any change to the special treatment requirements for SSCs and that the SSCs continue to be capable of performing their design basis functions, as well as to perform any beyond design basis functions consistent with the categorization process and results.

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

Based on the above, SNC concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of 11nO significant hazards consideration.. is justified.

4.3 CONCLUSION

S In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the E-15

Enclosure to NL-17-1962 Evaluation of the Proposed Change Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

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

E-16

Enclosure to N L-17 -1962 Evaluation of the Proposed Change 5

REFERENCES

1.

NEI 00-04, "1 0 CFR 50.69 SSC Categorization Guideline,.. Revision 0, Nuclear Energy Institute, July 2005.

2.

Regulatory Guide 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance," Revision 1, May 2006.

3.

ASME/ANS RA-Sa-2009, "Standard for Leveii/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum A to RA-S-2008," ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, February 2009.

4.

NUMARC 91-06, "Guidelines for Industry Actions to Assess Shutdown Management,"

December 1991.

5.

ANO SEA, Arkansas Nuclear One, Unit 2 - Approval of Request for Alternative AN02-R&R-004, Revision 1, "Request to Use Risk-Informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 and 3 Moderate and High Energy Systems," (TAC NO. MD5250) (ML090930246), April 22, 2009.

6.

Vogtle Electric Generating Plant, Units 1 and 2- "Issuance of Amendments Re: Use Of 10 CFR 50.69" (TAC NOS. ME9472 AND ME9473), December 17, 2014.

7.

Hatch Units 1 and 2: Issuance of Amendments 266/210 re: Adoption of TSTF-425, Revision 3, *Relocate Surveillance Frequencies to Licensee Controi-RITSTF Initiative 5b*

(TAC Nos. ME5016 and ME5017). (ML11108A129), January3, 2012.

8.

Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, March 2009..

9.

R. Correia, NRC Research, to 0. Martinez, ASME,.. U.S. Nuclear Regulatory Commission (NRC) Comments On.. Addenda to A Current ANS: ASME RA-SB-20XX, Standard for Levei1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," July 6, 2011, NRC ADAMS Accession Number ML111720067.

10. ASME/ANS RA-Sb-2013,.. Standard for Levei1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications,.. Addendum 8 to RAS-2008, ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, December 2013.

11. NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," March 2009.

12. EPRI TR-1 016737, Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments," December 2008.

13. NEI Letter to USNRC, "Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations (F&Os)," February 21, 2017, Accession Number ML17086A431.

14. USNRC Letter to Mr. Greg Krueger (NEI), "U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 7-12, and 12-13, Close Out of Facts and Observations (F&Os)," May 3, 2017, ADAMS Accession Number ML17079A427.

15. Pilot Aviation Information www.airnav.com.

E-17

Enclosure to NL-17-1962 Evaluation of the Proposed Change

16. Hatch Unit 1 & 2 Final Safety Analysis Report (FSAR), Rev. 35 January 2017.

17. Tornado Climatology of the Contiguous United States (NUREG/CR-4461 ), Revision 2; PNNL-15112.

18. NRC Regulatory Issue Summary 2015-06, "Tornado Missile Protection," June 10, 2015.

19. Regulatory Guide 1.78, "Evaluating the Habitability of a Nuclear Power Plant Control Room during a Postulated Hazardous Chemical Release," USNRC, Revision 1, 2001.

20. Regulatory Guide 1.91, "Evaluations of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plants," USNRC, Revision 1, February 1978.

21. Regulatory Guide 1.147, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1," USNRC, Revision 15, October 2007.

22. NEI 05-04, "Process for Performing Follow-on PRA Peer Reviews Using the ASME PRA Standard," Revision 0, Nuclear Energy Institute, January 2005.

23. NEI 07-12, "Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines,"

Revision 1, Nuclear Energy Institute, June 2010.

24. NRC Regulatory Issue Summary 2007-06," Regulatory Guide 1.200 Implementation,"

March 22, 2007.

25. Vogtle Electric Generating Plant, Units 1 and 2- "License Amendment Request to Modify Approved 10 CFR 50.69 Categorization Process," June 22, 2017, ADAMS Accession Number ML17173A875.

26. Hatch Nuclear Plant, Units 1 and 2- "License Amendment Request to Adopt NFPA-805 Performance Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition)," April 4, 2018, ADAMS Accession Number ML18096A955 E-18

Enclosure to NL-17-1962 Evaluation of the Proposed Change : List of Categorization Prerequisites The fire PRA model to be used for categorization credits modifications listed in Table 1-1 to achieve an overall CDF and LEAF consistent with NRC Regulatory Guide 1.174 risk limits.

Use of the alternative treatment on a categorized plant system will only occur after the modifications have been completed. No modifications are assumed in the internal event, internal flooding, or seismic PRA models.

The following legend should be used when reviewing Table 1-1:

High-Modification would have an appreciable impact on reducing overall fire CDF.

Medium - Modification would have a measurable impact on reducing overall fire CDF.

Low-Modification would have either an insignificant or no impact on reducing overall fire CDF E-19

Enclosure to NL-17-1962 Evaluation of the Proposed Change Table 1*1 Plant Modifications Item Rank 1

High Unit 1,2 Problem Statement Circuit configurations have been identified where a postulated fire could impact cables resulting in bypass of both the motor operated valve (MOV) limit and torque switches as described in NRC Information Notice (IN) 92-18. These postulated fires could result in physical damage to credited valves, preventing them from being operated locally or from the remote shutdown panel.

Proposed Modification Modify the MOV control circuits to prevent the postulated IN 92-18 failure mode or modify the valve weak-link components to prevent the following valves from being damaged for fire scenarios where operation of the MOV(s) may be required.

The affected Fire Areas and MOVs are as follows:

Unit 1 Fire Area 0024:

• 1 E11-F003B - RHR Heat Exchanger B Discharge Valve

• 1 E11-F0048 - RHR Pump 28 Torus Suction Valve

• 1 E11-F0068 - RHR Pump 28 SOC Suction Valve

• 1 E11-F007B - RHR Pumps B and 0 Minimum Flow Bypass Valve

• 1 E11-F011 B - RHR Heat Exchanger B Drain Valve To Suppression Pool

• 1 E11-F015B - RHR LPCI Inboard Discharge Valve

• 1 E11-F016B - Containment Spray Outboard Isolation Valve

• 1 E11-F017B - RHR LPCI Outboard Discharge Valve

• 1 E11-F028B -Torus Spray/RHR Test Line Outboard Isolation Valve E-20 Rlek Informed Characterization This modification implements an electrical configuration at FPRA-credited MOVs to mitigate IN 92-18 type failures.

This modification reduces the risk of fire induced MOV failures that could prevent operation of the valves locally or from the remote shutdown panel.

Enclosure to NL-17-1962 Evaluation of the Proposed Change Table 1*1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification

• 1 E11-F047B-RHR Heat Exchanger B Inlet Valve

• 1 E11-F048B - RHR Heat Exchanger B Bypass Valve

• 1 E11-F049-RHR to Radwaste Discharge Isolation Valve

• 1 E11-F068B - RHR Heat Exchanger B Service Water Flow Control Valve

• 1 E11-F073B - RHR Service Water to RHR Cross-tie Valve

• 1 E11-F1 04B - RHR Heat Exchanger B Vent Valve

• 1 E51-F007-RCIC Steam Supply Inboard Isolation Valve

• 1 E51-F008 - RCIC Steam Supply Outboard Isolation Valve

• 1 E51-F012-RCIC Pump Outboard Discharge Valve

• 1 E51-F013-RCIC Pump Inboard Discharge Valve

• 1 E51-F019-RCIC Minimum Flow Bypass to Suppression Pool Valve

• 1 E51-F022 - RCIC Pump Test Bypass Valve to Condensate Storage Tank

• 1 E51-F045 - RCIC Turbine Steam Supply Valve

• 1 E51-F046-RCIC Barometric Condenser and Lube Oil Cooler Cooling Water Supply Valve

• 1 E51-F524-RCIC Turbine Trip and Throttle Valve E-21 Risk Informed Characterization

Enclosure to NL-17-1962 Evaluation of the Proposed Change Table 1-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification Unit 2 Fire Area 0024:

• 2E11-F003B - RHR Heat Exchanger B Discharge Valve

• 2E11-F004B - RHR Pump 2B Torus Suction Valve

• 2E11-F006B - RHR Pump 2B SOC Suction Valve

• 2E11-F007B - RHR Pumps 2B and 20 Minimum Flow Bypass Valve

• 2E11-F011 B - RHR Heat Exchanger B Drain Valve To Suppression Pool

• 2E11-F015B - RHR LPCI Inboard Discharge Valve

• 2E11-F016B - Containment Spray Outboard Isolation Valve

• 2E11-F017B - RHR LPCI Outboard Discharge Valve

• 2E11-F028B -Torus Spray/RHR Test Line Outboard Isolation Valve

• 2E11-F047B - RHR Heat Exchanger B Inlet Valve

• 2E11-F048B - RHR Heat Exchanger B Bypass Valve

• 2E11-F049 - RHR to Radwaste Isolation Valve

• 2E11-F068B - RHR Heat Exchanger B Service Water Flow Control Valve

• 2E11-F073B - RHR Service Water to RHR Cross-tie Valve

• 2E11-F1 04B - RHR Heat Exchanger B Vent Valve E-22 Risk Informed Characterization

Enclosure to NL-17-1962 Evaluation of the Proposed Change Table 1*1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification

• 2E51-F007-RCIC Steam Supply Inboard Isolation Valve

• 2E51-F008-RCIC Steam Supply Outboard Isolation Valve

• 2E51-F01 0 - RCIC Pump Suction Valve From Condensate Storage Tank

• 2E51-F012-RCIC Pump Outboard Discharge Valve

• 2E51-F013 - RCIC Pump Inboard Discharge Valve

• 2E51-F019 - RCIC Minimum Flow Bypass Valve to Suppression Pool

• 2E51-F022 - RCIC Pump Test Bypass Valve to Condensate Storage Tank

• 2E51-F029-RCIC Pump Suction Valve From Suppression Pool

• 2E51-F031 - RCIC Pump Suction From Suppression Pool Outboard Isolation Valve

• 2E51-F045 - RCIC Turbine Steam Supply Valve

• 2E51-F046 - RCIC Barometric Condenser and Lube Oil Cooler Cooling Water Supply Valve

• 2E51-F524-RCIC Turbine Trip and Throttle Valve E-23 Risk Informed Characterization

Enclosure to NL-17-1962 Evaluation of the Proposed Change Table 1*1 Plant Modifications Item Rank Unit 2

Low 2

3 Low 2

Problem Statement Breakers between distribution panels 2R25-S021 and 2R25-S022 and the associated upstream power supply breakers (2R23-S021 Frames 4M and 4B) lack adequate coordination.

Breakers between the 208V section of MCC 2R24-S048 and the upstream power supply breaker in distribution panel 2 R25-S035 lack adequate coordination.

Proposed Modification Adjust the settings on the applicable breakers on 208V Switchgear 2R23-S021 Frames 4M and 4B to allow proper coordination between the switchgear breakers and the downstream distribution panel breakers.

Remove the 1 OOA breaker in distribution panel 2R25-S035 and replace it with a 1 OOA molded case switch. In addition, install a 1 OOA dual element time-delay fuse and fuse block in a qualified junction box between the distribution panel and the 208V section of MCC 2R24-S048.

E-24 Risk Informed Characterization This modification establishes adequate coordination between distribution panel breakers and upstream power supply breakers credited within the FPRA. Proper coordination ensures that equipment not fire affected are also not affected by a fire-induced cable failure on a separate load circuit supplied from the same power source.

This modification establishes adequate coordination between load breakers and upstream power supply breakers credited within the FPRA. Proper coordination ensures that equipment not fire affected are also not affected by a fire-induced cable failure on a separate load circuit supplied from the same power source.

Enclosure to NL-17-1962 Evaluation of the Proposed Change Table 1*1 Plant Modifications Item Rank Unit 4

Medium 1

5 Medium 1,2 Problem Statement A fire in Fire Area 0014 can result in failure of MCC 1 C (1 R24-S027). Failure of 1 R24-S027 could prevent automatic isolation of plant service water (PSW) to the Turbine Building resulting in a PSW flow diversion from the emergency diesel generators (EDG) and loss of power to EDG support components.

A fire in Fire Area 0014 can result in failure of MCC 1 C (1 R24-S003). Failure of 1 R24-S003 could cause loss of all three Control Room air handling Units 1 Z41-8003A, 1 Z41-B0038 and 1 Z41-B003C.

Proposed Modification Re-route or protect cables R23-S004-ES8-C22 and R23-S004-ES8-C09 to prevent fire induced failure of MCC 1 C (1 R24-S027) in Fire Area 0014.

Re-route or protect cable R23-S004-ES8-C04 to prevent fire induced failure of MCC 1 C (1 R24-S003} in Fire Area 0014.

E-25 Risk Informed Characterization This modification will eliminate the need for a Recovery Action associated with ensuring that onsite power remains available in the event of a fire in Fire Area 0014.

This modification will eliminate the need for a Recovery Action associated with ensuring that control room ventilation remains available in the event of a fire in Fire Area 0014.

Enclosure to N L 1962 Evaluation of the Proposed Change Table 1*1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification 6

Medium 1

A fire in Fire Area Re-route or protect cables R23-S004-0014 can result in ES8-C23 and R23-S004-ES8-C 14 to failure of MCC 1 B prevent fire induced failure of MCC 1 C (1 R24-S012). Failure (1 R24-S012) in Fire Area 0014.

of 1 R24-S012 could result in a loss of motive power to RHR Loop B components.

7 Medium 1,2 A fire in Fire Area Re-route or protect cable R23-S004-C 17 0014 can result in to prevent fire-induced failure of MCC 1 G failure of MCC 1 G (1 R24-S031) in Fire Area 0014.

(1 R24-S031 ). Failure of 1 R24-S031 could result in a loss of Control Building Ventilation exhaust.

E-26 Risk Informed Characterization This modification will eliminate the need for a Recovery Action associated with ensuring decay heat removal capability is maintained in the event of a fire in Fire Area 0014.

This modification will eliminate the need for a Recovery Action associated with ensuring that Control Building ventilation remains available in the event of a fire in Fire Area 0014.

Enclosure to NL-17-1962 Evaluation of the Proposed Change Table 1*1 Plant Modifications Item Rank 8

Medium Unit Problem Statement 1, 2 Degraded voltage protection system setpoints and analytical limits for the 4.16kV emergency busses must be adequate to ensure the proper operation of all safety-related equipment necessary to mitigate the consequences of the worst-case design basis event (DBE), while under postulated degraded grid conditions without any credit for administratively controlled bus voltage levels, i.e.,

manual operator actions.

Proposed Modification Implement degraded voltage modifications to eliminate the manual actions in lieu of automatic degraded voltage protection to assure adequate voltage to safety-related equipment.

(Reference HNP Unit 1 and Unit 2 Operating Licensing, Degraded Grid License Conditions 2.C.(11) and 2.C.(3)(i) respectively).

E-27 Risk Informed Characterization Operating License Compliance

Enclosure to NL-17-1962 Evaluation of the Proposed Change SNC has already established procedures for the use of the categorization process on plant systems. By letter dated December 12, 2014, the Southern Nuclear Fleet Categorization process was approved for Vogtle Electric Generating Station (ADAMS Accession Number ML14237A034) (Reference 6). This application for Plant Hatch is identical with only one exception. The currently approved process for Vogtle utilizes a Seismic Margins Approach rather than SPRA for categorization. Because Hatch has a peer reviewed seismic model, specific instructions for Hatch will use the SPRA model.

The current Southern Nuclear Fleet procedures contain the elements/steps listed below.

IDP member qualification requirements.

Qualitative assessment of system functions. System functions are qualitatively categorized as preliminary HSS of LSS based on the seven criteria in Section 9 of NEI 00-04 (see Section 3.2). Any component supporting an HSS function is categorized as preliminary HSS. Components supporting an LSS function are categorized as preliminary LSS.

Component safety significance assessment. Safety significance of active components is assessed through a combination of PRA and non-PRA methods, covering all hazards. Safety significance of passive components is assessed using a methodology for passive components.

Assessment of defense-in-depth (DID) and safety margin. Components that are categorized as preliminary LSS are evaluated for their role in providing defense-in-depth and safety margin and, if appropriate, upgraded to HSS.

Review by the IDP. The categorization results are presented to the IDP for review and approval. The I DP reviews the categorization results and makes the final determination on the safety significance of system functions and components.

Risk sensitivity study. For PRA-modeled components, an overall risk sensitivity study is used to confirm that the population of preliminary LSS components results in acceptably small increases to core damage frequency (CDF) and large early release frequency (LERF) and meets the acceptance guidelines of RG 1.17 4.

Periodic reviews are performed to ensure continued categorization validity and acceptable performance for those SSCs that have been categorized.

Documentation requirements per Section 3.1.1.

E-28

Enclosure to NL-17-1962 Evaluation of the Proposed Change A

h ttac f

ment 2: Descnpt1on o PRA Models Used in Cat~gonzation Units Model Baseline CDF Baseline LERF Comments i

Internal Events /Internal Flood Model Hatch Unit 1 The Rev. 4 model Revision 5 peer review applies 1

3.2E-06 2.8E-07 since the Rev. 5 Peer Reviewed model is an Against RG 1.200 R2 update.

in November 2009 Hatch Unit 2 The U1 Rev. 4 Revision 5 V2 model peer review 2

3.5E-06 2.2E-07 applies since the Peer Reviewed Rev. 5 model is an Against RG 1.200 R2 update.

in November 2009 Internal Fire Model Hatch Unit 1 Revision 1 1

Peer Reviewed 3.4E-05 1.4E-06 Against RG 1.200 R2 in June 201-6 Hatch Unit 2 Revision 1 2

Peer Reviewed 4.4E-05 2.1 E-06 Against RG 1.200 R2

• in June 2016 E-29

Enclosure to NL-17-1962 Evaluation of the Proposed Change I

Units I

Model I

I I

Hatch Units 1 I

Revision 3 1

Peer Reviewed Against RG 1.200 R2 in June 2016 Hatch Units 2 Revision 3 2

Peer Reviewed Against RG 1.200 R2 in June 2016 i

I Baseline CDF Baseline LERF Comments I

I Seismic Model I

2.7E-07 1.1 E-07 2.1 E-07 1.0E-07 E-30

Enclosure to NL-17-1962 Evaluation of the Proposed Change : Disposition and Resolution of Open Peer Review Findings Note: The items listed are for the Internal Events Internal Flooding model. There are no open F&Os associated with Fire and Seismic PRA Models Finding Supporting Capability Description Disposition for 50.69 Number Requirement(s)

Category (CC)

-~-------~~-- ---- ---~- --*-----

1-9 AS-83, Not Met Reviewed the AS Notebook. Generally, OPEN.

AS-C2 Cat 1/11/111 the discussion of the accident sequence modeling is adequate. However, the Per the Finding Closure Review:

following information was not present:

Reviewed Section 3 of the (1) Discussion of environmental Accident Sequence Notebook conditions associated with sequences.

and the revised sections regarding environmental (2) Interface between accident sequences conditions and the associated and plant damage states.

sequences. Reviewed Section 4.3 of L2 Notebook. Section 4.3 Basis: SR AS-83 requires the of L2 Notebook describes identification of phenomenological interface between accident conditions expected from each accident sequences and PDS binning.

sequence.

In AS Notebook Subsections Possible Resolution: Include additional 3.#.3, Environmental Conditions detail for each accident sequence.

have been evaluated for each Particularly, there was no mention of the accident sequence. In generation of harsh environments subsections 3.#.5, each affecting temperature, pressure, debris, sequence has a detailed water levels or humidity that could impact description of the the success of the system.

characteristics, which also describes the interface between accident sequences and the plant damage states.

The detail required bv this E-31 I

I

Enclosure to N L 1962 Evaluation of the Proposed Change Finding Supporting Number Requirement(s)

Capability Description Category (CC)

E-32

• • • -****~ **-

Disposition for 50.69 finding has.bee*n--aaded to the accident sequence notebook.

The sequence descriptions have a discussion of Environmental Conditions. The Hatch PRA does not typically rely on equipment or operator actions in an area where a severe environment is encountered.

However, no information has been provided on why no environmental impact. One Accident Sequence that does have an adverse impact, is based upon an assumption of equipment qualifications. No listing of Environmental considerations is provided. For SBO sequences and uses of Fire Water no discussion about access or other issues.

SNC disposition for 50.69:

Hatch credits equipment qualification for equipment located inside containment for LOCA and other sequences that cause harsh conditions in containment. For all other areas, the models do not credit use of equipment in the area of events that cause adverse

Enclosure to NL-17-1962 Evaluation of the Proposed Change Finding Supporting Number Requirement(s)

Capability Description Category (CC)

E-33 Disposition for 50.69 environmentar events, such as ISLOCA events and steam line breaks outside containment.

The Internal flooding analysis evaluates the susceptibility of components to spray and flooding separately. This finding is a documentation issue and does not affect the 50.69 categorization process.

Enclosure to NL-17-1962 Evaluation of the Proposed Change Finding Supporting Number Requirement(s) 1-15 IFEV-81, IFQU-81, IFQU-82 Capability Category (CC)

CAT 1/11/111 Description Reviewed the IF Notebook. The documentation of the initiating events is complete but does not facilitate reviews.

The documentation in the IF Notebook does not represent the approach currently used in the modeling; it must be supplemented by the.. FLOODING MODEL CHANGES.. document. The combination of the two documents constitutes the model documentation.

Basis; The documentation for IF covers two revisions of the IF model and the documentation is conflicting in some areas.

Possible Resolution: Combine the current IF Notebook document with the information contained in the.. FLOODING MODEL CHANGES.. document into a new IF document representing the current modeling approach.

E-34 Disposition for 50.69 PARTIALLY CLOSED Per the Finding Closure Review:

This F&O identifies a documentation issue. No new method is introduced.

Reviewed PRA-BC-H-1 0-002 Rev. 1, Hatch PRA Flooding Analysis for Units 1.

The documentation for IF model changes have been combined into the IF notebook PRA-BCH-1 0-002 Rev. 1. However, the Hatch provided document does not appear to be the final notebook yet.

SNC Disposition for 50.69:

The internal flooding evaluations were revised by the addition of appendices instead of revisions to the actual main body.

Although this makes the document difficult to follow, all of the technical information is correct and present. This is a documentation issue with no impact on this application.

Enclosure to NL-17-1962 Evaluation of the Proposed Change Finding Supporting Number Requirement(s) 4-5 IFSN-A10, IFQU-A5 Capability Category (CC)

Not Met (Both SRs)

Description In the development process of the flooding scenarios there is no credit taken for the manual isolation of floods. This approach was assumed to be conservative; however, the propagation of flood water would be expanded if no operator action was taken, therefore, affecting more areas and SSCs than initially projected.

Basis: The particular scenario investigated was o/oFL-42 {1203K-FWMP-

1) which is a Plant Service Water leak affecting room 1203C. Using the documented data this room would fill up and overflow into other rooms containing PSA equipment required for accident mitigation.

Possible Resolution: Operator actions should be developed and added to the scenario development and the PRA model to reflect how the plant would be operated in the event of this scenario. It may be beneficial to consider use of mitigation event trees to assure that all mitigation issues are considered.

E-35

~'

Disposition for 50.69 OPEN.

Per the Finding Closure Review:

Reviewed PRA-BC-H-1 0-002 Rev. 1, Hatch PRA Flooding Analysis for Units 1. Also reviewed the integrated PRA model documented in calculation PRA-BC-H-1 0-008.

This F&O identifies a major modeling issue with internal flooding on the credit for operator manual isolation.

Although no new method has been introduced with the new model changes, the results have been confirmed to have significant changes from the previously peer reviewed models based on the inputs provided by Hatch. The resolutions to F&O 4-5 issues require new model changes beyond the flooding isolation HFEs modeled in the original ABS notebook, which was not the model of record reviewed by the original peer review team.

The resolutions also require updates to the technical bases supporting the current PRA model of record. Moreover_~_

Enclosure to NL-17-1962 Evaluation of the Proposed Change Finding Supporting Number Requirement{s)

Capability Description Category {CC)

E-36 Disposition for 50.69 detailed H RA may have been performed for these flooding isolation H FEs that have contributed to the significant reduction of CDF.

Hatch provided significant amount of information and justification for the model changes during the F&O closure process, which gave reasonable confidence that the current Hatch internal flooding model is developed appropriately and the changes can be classified as maintenance. However, the model changes have been confirmed to be significant and the current documentation supporting the internal flooding model is assessed to be not adequate to support this determination.

SNC Disposition for 50.69:

The original flooding evaluation credited manual isolation of floods using some screening values and some detailed HRA analysis. A subsequent revision removed all credit for isolation but performed a flooding screening analysis. Then a third revision re-applied the previous

Enclosure to NL-17-1962 Evaluation of the Proposed Change Finding Supporting Number Requirement(s) 6-8 HR-G6 Capability Category (CC)

Cat 1/11/111 Description Consistency check did not include comparison of HEPs in regards to scenario context, plant history, procedures, operational practices, and experience.

Basis: Section 8.3 of the HRA notebook evaluated consistency of HEPs based on the ratio of execution time and recovery time to check the reasonableness of the*

derived probability. This checks the reasonableness of the derivation of the probability. However, there was no check E-37 Disposition for 50.69 HRA analysist6the scenarios -..

that passed the screening. The internal flooding evaluations were revised by the addition of appendices instead of revisions to the main body of the evaluation. Although this makes the document difficult to follow, all the technical information is correct and present. This is a documentation issue with no impact on this application. When the final model is compared to the original model, the overall internal flooding risk changes but not the important flooding contributors or the risk insights.

OPEN.

Per the Finding Closure Review:

Reviewed Hatch Human Reliability Analysis Update, dated Oct. 26, 2009. The evidence presented in the HRA notebook does not appear to be adequate, comparing with the documentation for a similar BWR plant. The model owner at the time of the Peer Review felt that the brief comparison of action versus overall timinQ in

Enclosure to NL-17-1962 Evaluation of the Proposed Change Finding Supporting Number Requirement(s)

Capability Category (CC)

Description of consistency based on scenario context, plant history, procedures.

Possible Resolution: Compare HEPs and determine if the HEP values, relati.ve to each other, are representative of applicable scenario context, plant history, procedures, operational practices, and experience E-38 au..... ~..

~

n Disposition for 50.69 Section

• a~3 of' the.HRA.

• -***~-***~*~-*-~**** ***--**

notebook was adequate to comply with HR-G6 and that no additional information needed to be added. Because of this approach there were no revisions made to the HRA notebook after the peer review I

and there are no added words I

or amplifying discussion to further support this view in the calculation.

SNC Disposition for 50.69:

The internal events HRA documentation was revised to incorporate a better consistency analysis. This revised analysis was reviewed and incorporated into the Hatch seismic HRA calculations, thus the documentation associated with this issue has been revised.

Enclosure to NL-17-1962 Evaluation of the Proposed Change A

h ttac ment External Hazard Screened?

(V/N)

Aircraft Impact y

Avalanche y

Biological Event y

Coastal Erosion y

Drought y

External Flooding y

4 E IH d s

. xterna azar s creen1ng Screening Result Screening Criterion Comment (Note a)

There are no airports within 10 miles of the plant. There are no military facilities or military training routes close to the plant. Aircraft hazard is not a design PS2 basis hazard event for the plant and the present review using the most recent PS4 data (Reference 15) confirms this conclusion. The analysis shows that the CDF is less than 1 e-06, using a bounding analysis. Therefore, aircraft hazard is screened out.

C3 Topography is such that no avalanche is possible.

Hazard is slow to develop and there is cs sufficient time to eliminate the source of the threat or to provide an adequate response.

C3 Not applicable to the site because of location.

The river flow past the site during periods of low flow is controlled by the C1 upstream dams: Sinclair Dam, Wallace Dam and Lloyd Shoals Dam as described in Section 2.4.11.3 of the cs UFSAR (Reference 16).

A calculation was performed to re-evaluate the flooding depths and mitigating strategies resulting from river C1 flooding and local maximum precipitation in response to Near Term Task Force (NTTF) Recommendation 2.1. The calculation demonstrates that external flooding events cannot affect safety related systems, structures or E-39

Enclosure to NL-17-1962 Evaluation of the Proposed Change External Hazard Screened?

{Y/N)

Extreme Wind or y

Tornado Fog y

Forest or Range Fire y

Screening Criterion (Note a)

PS3 C4 C1 E-40 Screening Result Comment components. Hatch is not vulnerable to river flooding, dam failures, or local maximum precipitation and external flooding events.

Hatch has been designed for extreme winds and tornado loadings that are substantially higher than the design basis events presently required. The design basis for extreme winds and tornadoes for Hatch is described in UFSAR Section 3.3 (Reference 16). Per the updated discussions in NUREG/CR-4461 (Reference 17), extreme winds at the Hatch site from thunderstorms and hurricanes are bounded by the design basis tornado winds.

for tornado missiles, in response to Regulatory Issue Summary 2015-06 (Reference 18), a plant specific Torn?tdo Missile Vulnerability Evaluation was completed and all non-conformances were resolved.

Calculations show that the initiator probability is 3.3e-06 and the CCDP is 1.7e-03, thus tornadoes and high winds can be screened out.

Fog affects frequency of occurrence of other hazards, e.g., highway accidents, aircraft landing and take-off accidents; and is indirectly considered. Fog is rare in the site region. UFSAR Section 2.3.2.2 (Reference 16) states that visibility of less than 1A mile occurs a maximum of 4 days during each winter month and 1 day a month during April, May, June and July.

UFSAR Section 2.2.3.1 (Reference 16) states that the terrain and ground cover

Enclosure to NL-17-1962 Evaluation of the Proposed Change External Hazard Screened?

(YIN)

Frost y

Hail y

High Summer y

Temperature Screening Criterion (Note a)

C3 C1 C1 C1 cs E-41 Screening Result Comment surrounding Hatch are such that they are not conducive to forest or brush fires.

Frost is of limited occurrence because of arid climate. Snow and ice cover bounds this hazard.

Hail may occur but there are no openings in the walls or roofs of safety related buildings through which hail may enter and damage essential equipment.

Tornado missile protection features, structural walls and roofs are adequate to withstand the impact of hail.

At Savannah, the record maximum was 1 05°F in July 1879, and the record minimum was 8°F in February 1899. At Macon a record maximum temperature of 1 06°F was recorded in June 1954 and a record low of 3°F in Janua_ry 1966, (UFSAR Section 2.3.2.2 (Reference 16)).

To maintain operator comfort, the HVAC systems are designed to maintain prescribed building temperatures during outside temperature variations between 20°F and 95°F. Even if the maximum temperature exceeds the design limits for HV AC systems, such exceedance lasts only for a brief period, and given the thermal inertia of the concrete structures where safety-related equipment is located, equipment functionality is maintained.

High summer temperatures are of negligible impact on the plant. This phenomenon provides a large amount of time for preparation (weather forecast) with time for implementation of

Enclosure to NL-17-1962 Evaluation of the Proposed Change External Hazard Screened?

(Y/N)

High Tide, Lake Level, y

or River Stage Hurricane y

Ice Cover y

Industrial or Military Facility Accident y

Internal Flooding N

Screening Criterion (Note a)

C3 C5 C1 C3 C3 C3 Detailed PRA E-42 Screening Result Comment appropriate mitigation actions (e.g. plant power reduction or shutdown).

Site is located inland. See External Flooding evaluation.

The plant is not on the coast; the hurricane wind effects are enveloped by those of tornadoes.

Icing does not normally occur on the Altamaha River at Hatch.

The UFSAR Section 2.2.2.1 (Reference

16) states that within a 5-mile radius of Hatch, there are no manufacturing plants, chemical plants, refineries, storage facilities, mining and quarrying operations, military bases, missile sites, transportation facilities, oil and gas wells, or underground gas storage facilities.

Also, there are no known military firing or bombing ranges or aircraft low-level flight holding or landing patterns near the site area. No military bases or firing ranges, oil pipelines, or tank farms are located within a 1 0 mile radius of the plant site. Therefore, the hazards from industrial and military facility accidents are screened.

The Hatch internal events PRA includes evaluation of risk from internal flooding events.

Enclosure to NL-17-1962 Evaluation of the Proposed Change External Hazard Screened?

(Y/N)

Internal Fire N

Landslide y

Lightning y

Low Lake Level or River Stage y

Low Winter y

Temperature Screening Criterion (Note a)

Detailed PRA C3 C1 C1 cs C2 cs E-43 Screening Result Comment The Hatch internal events PRA includes evaluation or risk from internal fire events.

In the immediate vicinity of the plant, there are no steep hills.

Lightning protection is provided as part of the plant design.

Operational procedures are in place to monitor and mitigate this hazard. There are procedures to monitor low river level and designs to install a sandbag weir across the river if needed.

At Savannah, the record maximum was 1 05°F in July 1879, and the record minimum was 8°F in February 1899. At Macon a record maximum temperature of 1 06°F was recorded in June 1954 and a record low of 3°F in January 1966, (UFSAR Section 2.3.2.2 (Reference 16)).

To maintain operator comfort, the HVAC systems are designed to maintain prescribed building temperatures during outside temperature variations between 20°F and 95°F. Even if the minimum temperature exceeds the design limits for HV AC systems, such exceedance lasts only for a brief period and, given the thermal inertia of the concrete structures where safety-related equipment is located, will not have any impact. Systems subject to cold temperatures are insulated and heat traced. Temperatures inside the plant

Enclosure to NL-17-1962 Evaluation of the Proposed Change External Hazard Screened?

(V/N)

Meteorite or Satellite y

Impact Pipeline Accident y

Release of Chemicals y

in Onsite Storage River Diversion y

Sand or Dust Storm y

Seiche y

Screening Criterion (Note a)

C2 C3 PS4 C3 C3 C3 E-44 Screening Result Comment buildings are expected to be higher than 17°F.

This event has a very low frequency of occurrence for any site in the US.

A Southern Natural Gas Company pipeline is located within-4 1/2 miles of Hatch. The pipeline is sufficiently distant from Hatch such that any potential leak or detonation would not affect Hatch.

Based on the evaluations reported in the UFSAR 2.2.3.1 of Unit 2 on storage and handling of toxic chemicals near the site (UFSAR Reference 16), and the results of a plant specific calculation, the probability of this event is less than 1 e-07 and this hazard group does not pose a credible threat to Hatch Units 1 & 2.

UFSAR Section 2.4.9 (Reference 16) states that the river channel is relatively straight for a distance of 1.5 miles below US Highway No. 1 and that there are no meanders at the plant site which could cut across and divert the flow. The US No 1 bridge and highway fill serve to control the channel alignment to its present location.

This is not relevant for this region.

There is no large body of water close the site for this event.

Enclosure to NL-17-1962 Evaluation of the Proposed Change External Hazard Screened?

{Y/N)

Seismic Activity N

Snow y

Soil Shrink-Swell Consolidation y

Storm Surge y

Toxic Gas y

Transportation y

Accident Tsunami y

Turbine-Generated y

Missiles Screening Criterion (Note a)

Detailed PRA C1 C1 C5 C3 C4 C3 C3 PS4 E-45 Screening Result Comment Hatch has a detailed seismic PRA as discussed in Section 3.2.3 of this application.

The 50 year snow load is estimated as less than 5 pounds per square foot (psf).

The design basis roof live load for seismic Category I structures is 20 psf (UFSAR Table 3.8-13) (Reference 16).

Procedures are in place to monitor differential settlement per UFSAR Section 2A.5.3 (Reference 16).

Hatch is inland and is not affected by storm surge.

Included in transportation accident, on-site chemical release, and industrial and military facilities accidents.

Analysis of postulated accidents on the transportation routes has shown that they do not pose a credible threat to Hatch since these routes are farther than the safe distances specified in RG 1. 78 (Reference 19) and RG 1.91 (Reference 20). Therefore, this hazard class can be screened out from the Hatch PRA.

Hatch is inland and is not exposed to the tsunami threat.

Per vendor documentation, a probabilistic analysis performed for postulated failures of turbines in Units 1

& 2 has shown that the overall probability of turbine missile damage is

Enclosure to NL-17-1962 Evaluation of the Proposed Change External Hazard Screened?

(Y/N)

Volcanic Activity y

Waves y

Screening Criterion (Note a)

C3 C3 Screening Result Comment less than the NRC accepted value of 1x10-7 per year.

The site is not close to any active volcanoes.

Hatch is inland and is not affected by any wave activity.

Note a) - See Attachment 5 for descriptions of the screening criteria.

E-46

Enclosure to NL-17-1962 Evaluation of the Proposed Change : Progressive Screen1ng Approach for Addressing External Hazards Event Analysis Criterion Source Comments C1. Event damage NUREG/CR-2300 and Initial Preliminary potential is < events ASMEIANS Standard Screening for which plant is RA-Sa-2009 designed.

C2. Event has lower mean frequency and NUREG/CR-2300 and no worse ASME/ ANS Standard consequences than RA-Sa-2009 other events analyzed.

C3. Event cannot NUREG/CR-2300 and occur close enough to AS ME/ ANS Standard the plant to affect it.

RA-Sa-2009 C4. Event is included NUREG/CR-2300 and Not used to screen.

in the definition of ASMEIANS Standard Used only to include another event.

RA-Sa-2009 within another event.

C5. Event develops slowly, allowing ASME/ANS Standard adequate time to RA-Sa-2009 eliminate or mitigate the threat.

PS 1. Design basis Progressive hazard cannot cause AS ME/ ANS Standard Screening a core damage RA-Sa-2009 accident.

PS2. Design basis for the event meets the NUREG-1407 and criteria in the NRC ASMEIANS Standard 1975 Standard RA-Sa-2009 Review Plan (SRP).

PS3. Design basis event mean NUREG-1407 as frequency is < 1 E-5/y modified in and the mean ASMEIANS Standard conditional core RA-Sa-2009 damage probability is

< 0.1.

E-47

Enclosure to NL-17-1962 Evaluation of the Proposed Change Event Analysis Criterion PS4. Bounding mean CDF is < 1 E-6/y.

Screening not successful. PRA Detailed PRA modeling needed to meet requirements in the ASMEIANS PRA Standard.

Source NUREG-1407 and ASME/ ANS Standard RA-Sa-2009 NUREG-1407 and ASMEIANS Standard RA-Sa-2009 E-48 Comments

Enclosure to NL-17-1962 Evaluation of the Proposed Change : Disposition of Key Assumptions/Sources of Uncertainty Table 6-1 Internal Events/Flooding PRA Model Sources of Uncertainty Assumption/Uncertainty 50.69 Impact Model Sensitivity and Disposition If all other containment heat Low pressure ECCS pumps do Loss of NPSH following removal methods fail the not immediately lose NPSH if emergency venting is Hardened Vent can be used to containment venting thru the assumed to have a probability remove heat from the hardened vent is required.

of 1 E-2. Using an assumed suppression pool. Opening the value for loss of NPSH to vent will result in pressure allow credit for injection decrease in containment.

systems after containment Calculations show that a small venting is judged to be a amount of containment potential source of pressure is needed to uncertainty. However, maintain NPSH for the ECCS because the low pressure pumps at full pump flow.

ECCS pumps are already risk Because only a certain significant, there is no impact specific set of circumstances from this assumption and no might lead to loss of NPSH, additional sensitivity cases an adjustment factor is applied will be required.

to these cases in the model.

The Automatic The PRA models cannot be The PRA models the ADS Depressurization System used to assess the importance system as being bypassed (ADS) actuation logic is not of non-modeled ADS by the operators for all modeled in detail. A pseudo-components.

scenarios. The existing event is used instead to generic 50.69 sensitivity approximate the reliability of cases for operator actions the logic.

adequately assess this assumption.

The Analog Transmitter Trip The PRA model cannot be The RPS channel failure System (A TTS) panels used to assess the importance probabilities have little containing Reactor Protection of non-modeled A TTS impact on the Hatch PRA System (RPS) input logic are components in the RPS risk metrics. The addition of not modeled. The RPS logic is cabinets.

events to represent the de-energize to actuate and it is A TTS panels would have assumed that.failure of the little impact as well. No panels or power supplies would additional sensitivity cases lead to RPS actuation. The four will be required.

channels of the two RPS divisions are identified as 1 A, 2A, 1 B, and 28, with hardware for each channel housed in separate cabinets. Surrogate common cause events are used instead of detailed modeling.

E-49

Enclosure to NL-17-1962 Evaluation of the Proposed Change Table 6-1 Internal Events/Flooding PRA Model Sources of Uncertainty Assumption/Uncertainty 50.69 Impact Model Sensitivity and Disposition To simplify the model, the The simplified modeling will Additional modeling would drywall coolers are assumed impact use of the model to be needed to assess drywall to have only one fan each; no evaluate drywall cooling cooling elements, but no credit is taken for the standby system importance.

other functions are affected fan and no additional sensitivity cases will be required.

The Hatch PRA assumes that if The assumption increases the No credit for these systems the containment is failed, the importance of the containment after containment failure is injection sources contained in boundary.

conservative. No additional the reactor building would fail sensitivity cases will be due to environmental required.

conditions, steam binding of pumps, or disruption of flow paths due to catastrophic containment failure.

In the Level 2 PRA, injection sources external to the reactor building are considered after containment failure. Of these, only condensate/ RHRSW are credited (fire protection and LPCI/CS from CST are not).

E-50

Enclosure to NL-17-1962 Evaluation of the Proposed Change Table 6-2 Seismic PRA Sources of Uncertainty Assumption/Uncertainty 50.69 Impact Model Sensitivity and Disposition Credit for FLEX equipment Operation of FLEX equipment A sensitivity case to determine impact from the model.

requires successful the impact of not crediting FLEX deployment and actuation by equipment during a seismic operators.

event was evaluated. The result was a 41 °/o increase in CDF, and a 57°/o increase in LEAF.

Therefore, crediting FLEX equipment is important from a seismic perspective. Although this impact is significant, it is realistic. The 50.69 categorization process includes a required evaluation of sensitivity to all modeled human error probabilities, which will address the impact of this issue on categorization results.

Further, this issue has minimal impact on this application given the relatively low seismic risk compared to total plant risk.

The shared diesel Alignment affects response of No impact on 50.69 evaluations.

generator is assumed to diesel during dual unit LOSP Diesel will show as equally be aligned to each unit events.

important for both units.

50°/o of the time.

E-51

Enclosure to NL-17-1962 Evaluation of the Proposed Change Table 6-3 Fire PRA Sources of Uncertainty Assumption/Uncertainty 50.69 Impact Model Sensitivity and Disposition Most secondary side Assuming failure of non-safety The Fire PRA evaluated this as a equipment is failed in all fire components increases worth of sensitivity.

scenarios because the safety-related components.

exact location of electrical raceways and cables was not walked down.

Conduits that were not Results in higher risk than if Model results are conservative.

walked down are assumed every field routed conduit was No additional sensitivity studies failed by all fires in the fire located.

required.

zone.

A minimum floor value is This could make some JHEP The existing HRA sensitivity assigned to Joint HEP events appear more important.

requirements address this event probabilities.

adequately for 50.69.

E-52