NL-16-2382, License Amendment Request to Modify Approved 10 CFR 50.69 Categorization Process

License Amendment Request to Modify Approved 10 CFR 50.69 Categorization Process

ML17173A875
Person / Time
Site:	Vogtle
Issue date:	06/22/2017
From:	Hutto J Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-16-2382
Download: ML17173A875 (58)

Text

~ Southern Nuclear JUN Z 2 2011 Docket Nos.: 50-424 50-425 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 J. J. Hutto Regulatory AHa irs Dtrector Vogtle Electric Generating Plant-Units 1 and 2 40 lll\\cmcs~ Center P~rk\\\\ay Po'l Office Bu\\ 129'>

Bmmngh.un. AI :\\5242 205 992 5S72 tel 205 992 760 I fax jjhutlo(n*.,outhcmco wm NL-16-2382 License Amendment Request to Modify Approved 10 CFR 50.69 Categorization Process Ladies and Gentlemen:

Pursuant to 10 CFR 50.69 and 10 CFR 50.90, Southern Nuclear Operating Company (SNC) hereby requests amendments to the Vogtle Electric Generating Plant (VEGP) Units 1 and 2 Renewed License Numbers NPF-68 and NPF-81. The proposed amendment would modify the licensing basis to implement a change to the previously approved 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Plants." 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. By letter dated December 17, 2014, the Nuclear Regulatory Commission (NRC) approved VEGP use of 10 CFR 50.69.

The proposed amendment would incorporate the use of the peer reviewed plant-specific VEGP seismic probabilistic risk assessment (SPRA) into the previously approved 10 CFR 50.69 categorization process, as allowed by the NRC endorsed industry guidance. Amendments 173 (Unit 1) and 155 (Unit 2) specify that NRC prior approval, under 10 CFR 50.90, is required for a change to a categorization process that is outside the bounds specified (e.g., change from a seismic margins approach (SMA) to an SPRA approach). The scope of this request is limited to the change from SMA to SPRA. No other changes to the categorization process are being requested by this amendment. provides the basis for the proposed change to the VEGP Unit 1 and Unit 2 Operating Licenses. The categorization process being implemented through this change is consistent with NEI 00-04, "10 CFR 50.69 SSC Categorization Guideline," Revision 0 dated July 2005 which was endorsed by the NRC in Regulatory Guide 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance". SNC requests approval of the proposed license amendments by June 30, 2018.

U.S. Nuclear Regulatory Commission NL-16-2382 Page 2 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 NRC commitments. If you have any questions, please contact Ken McElroy at 205.992.7369.

Mr. J. J. Hutto states he is the Regulatory Affairs Director for Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and, to the best of his knowledge and belief, the facts set forth in this letter are true.

Respectfully submitted, rc_:;>

J. J. Hutto Regulatory Affairs Director Sworn to and subscribed before me this 2..2_ day of ~

~2/~t Notary Public

'2017.

My commission expires: / () ~ )i' - d-O I I

Enclosure:

Basis of the Proposed Change Attachments:

1.

Seismic PRA Model Summary Information

2.

Disposition and Resolution of SPRA Peer Review Finding cc:

U.S. Nuclear Requlatorv Commission Regional Administrator, Region II NRR Project Manager -Vogtle 1 & 2 Senior Resident Inspector-Vogtle 1 & 2 NRR Project Manager-Vogtle 1 & 2 RType: CVC7000 State of Georgia Director-Environmental Protection Division I

)

I J

(

Vogtle Electric Generating Plant-Units 1 and 2 License Amendment Request to Modify Approved 1 0 CFR 50.69 Categorization Process Enclosure Basis of the Proposed Change

Enclosure Basis of the Proposed Change Table Of Contents 1

SUMMARY

DESCRIPTION................................................................................................. 1 2

DETAILED DESCRIPTION................................................................................................. 1 2.1 Current Regulatory Requirements.............................................................................. 1 2.2 Reason for Proposed Change.................................................................................... 2 2.3 Description of the Proposed Change.......................................................................... 3 3

TECHNICAL EVALUATION................................................................................................ 3 3.1 Seismic PRA Technical Adequacy Evaluation (1 0 CFR 50.69(b)(2)(ii))...................... 4 3.1.1 Seismic Hazards................................................................................................ 4 3.1.2 PRA Maintenance and Updates......................................................................... 4 3.1.3 PRA Uncertainty Evaluations............................................................................. 5 3.2 PRA Review Process Results (1 0 CFR 50.69(b)(2)(iii)).............................................. 6 3.3 Risk Evaluations (10 CFR 50.69(b)(2)(iv))................................................................. 9 4

REGULATORY EVALUATION........................................................................................... 9 4.1 Applicable Regulatory Requirements/Criteria............................................................. 9 4.2 No Significant Hazards Consideration Analysis........................................................ 11 4.3 Conclusions.............................................................................................................. 12 5

ENVIRONMENTAL CONSIDERATION............................................................................ 12 6

REFERENCES................................................................................................................. 13 LIST OF ATTACHMENTS : Seismic PRA Model Summary Information................................................ 15 : Disposition and Resolution of SPRA Peer Review Findings.......................... 16 E-i

Enclosure to NL-16-2382 Basis of the Proposed Change 1

SUMMARY

DESCRIPTION The proposed amendment would modify the licensing basis to implement a change to the approved voluntary implementation of the provisions of 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 (SSCs} for Nuclear Power Plants." The proposed amendment would incorporate the use of the peer reviewed, plant-specific Vogtle Electric Generating Plant (VEGP} seismic probabilistic risk assessment (SPRA} into the previously approved 10 CFR 50.69 categorization process, as allowed by the Nuclear Regulatory Commission (NRC} endorsed industry guidance. 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 10 CFR 50.69 categorization process has been reviewed and approved by the NRC for VEGP. Categorization includes an integrated assessment of total risk and the regulations and categorization guidance allows licensees to implement different approaches depending on the scope of their probabilistic risk assessment (PRA} models. The currently approved risk assessment tools are:

1. Internal event PRA for internal risk

2. Fire PRA for fire risk

3. Seismic margin analysis (SMA} for seismic risk

4. Individual Plant Examination of External Events (IPEEE} screening to asses risk from other external hazards (high winds external floods}

5. Assess shutdown risk This proposed amendment request only substitutes a peer reviewed seismic PRA in place of the SMA to assess seismic risk. This type of change was envisioned by the regulations and guidance as new PRA tools became available. All other aspects of the program remain as the NRC approved in Reference 6. It is important to note the VEGP program was approved by the NRC using a detailed pilot plant process taking over two years. As recently as the summer of 2016, the NRC conducted a pilot inspection of the process with favorable results (Reference 7}.

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 E-1

Enclosure to N L 2382 Basis of the Proposed Change 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 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.

Special treatment includes, but is not limited to, quality assurance, testing, inspection, condition monitoring, assessment, evaluation, and resolution of deviations. The distinction between "treatment" 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.

The previously approved VEGP 50.69 categorization process conforms to the guidance in NRC RG 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance," Revision 1 dated May 2006 (Reference 3). The categorization process also conforms to the guidance in NEI 00-04, "1 0 CFR 50.69 SSC Categorization Guideline," Revision 0 dated July 2005 (Reference 4), as endorsed by RG 1.201.

With this change, to utilize the Seismic PRA model rather than the seismic margins approach, the VEGP categorization process will continue to conform to these guidance documents.

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, PRA addresses credible initiating events by assessing the event frequency. Mitigating system reliability is then assessed, 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 1 0 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 SSC, 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 E-2

Enclosure to N L 2382 Basis of the Proposed Change 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 how SSG is categorized. Finally, assessment activities are conducted to make adjustments to the categorization and treatment processes as needed so that SSCs continue to meet all applicable requirements.

The rule does not allow for the elimination of sse 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 by restructuring the regulations to allow an alternative risk-informed approach to special treatment. Conversely, for SSCs that do not significantly contribute to plant safety on an individual basis, the rule allows a reasonable, though reduced, level of confidence that these SSCs will satisfy functional requirements. Implementation of 10 CFR 50.69 allows an improved focus on equipment that has safety significance resulting in improved plant safety.

The VEGP 10 CFR 50.69 categorization process has previously been reviewed and approved by NRC (Reference 6). The proposed change implements a modification to the process, as allowed by the 10 CFR 50.69 guidance endorsed by NRC in Regulatory Guide 1.201 (Reference 2), to incorporate use of the peer reviewed plant-specific VEGP SPRA.

2.3 DESCRIPTION

OF THE PROPOSED CHANGE SNC proposes the addition of the following condition to the operating license[s] of VEGP Unit 1 and Unit 2 to document the NRC's approval of the use 10 CFR 50.69.

SNC is approved to implement 1 0 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 license amendments No. 173 (Unit 1) and No.

155 (Unit 2) using Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events including internal flooding, internal fire, and seismic events. SNC is approved to utilize the SPRA model for use in the categorization process rather than the previously approved seismic margin approach.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above.

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 probabilistic risk assessment (PRA), margins-type approaches, or other systematic evaluation techniques E-3

Enclosure to NL 2382 Basis of the Proposed Change 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).

The above information was previously provided to NRC as part of the VEGP pilot application of 10 CFR 50.69 (Reference 5). The VEGP 10 CFR 50.69 categorization process (overall process, including active and passive categorization elements) has been reviewed and approved by NRC (Reference 6) and NRC has performed an audit of its implementation (Reference 7) and found it to be in conformance with the criteria specified in the rule and in RG 1.201 (Reference 2}.

In its review and approval of that application, NRC reviewed the technical adequacy of the VEGP internal events at power and internal fire PRA models and approved their use for 10 CFR 50.69 categorization (Reference 6). The VEGP 50.69 process addresses seismic risk through the use of the IPEEE SMA results, following the process defined in NEI 00-04 (Reference 1) and endorsed in RG 1.201 (Reference 2). The purpose of this license amendment request is to replace, within the approved VEGP 50.69 program, the use of the SMA process with use of the VEGP SPRA, also in accordance with NEI 00-04 (Reference 1) and RG 1.201 (Reference 2) guidance. Therefore, the remainder of this technical evaluation is focused on establishing the technical adequacy of the VEGP SPRA for this application.

3.1 SEISMIC PRA TECHNICAL ADEQUACY EVALUATION (10 CFR 50.69(b}(2}(11))

The following sections demonstrate that the quality and level of detail of the processes used in categorization of SSCs are adequate. The SPRA model described below has been peer reviewed and there are no PRA upgrades that have not been peer reviewed.

3.1.1 Seismic Hazards The approved VEGP categorization process uses the SMA performed for the IPEEE in response to GL 88-20 (Reference 4) for evaluation of safety significance related to seismic hazards. Through this requested change, the VEGP categorization process will instead use the peer reviewed plant-specific VEGP seismic PRA model. The SNC risk management process ensures that the SPRA model used in this application reflects the as-built and as-operated plant for each of the VEGP units. No plant specific approaches were utilized in development of the seismic hazards for the SPRA model. Attachment 1 at the end of this enclosure identifies the current applicable Seismic PRA model.

3.1.2 PRA Maintenance and Updates The SNC risk management process, which was previously reviewed by NRC as part of the VEGP 50.69 approval (Reference 6), ensures that the applicable PRA models, including the SPRA model, used in this application continue to reflect the as-built and as-operated plant for each of the VEGP 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 E-4

Enclosure to NL-16-2382 Basis of the Proposed Change 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 sse categorization will be re-evaluated.

In addition, SNC has implemented a process that addresses the requirements in NEI 00-04 (Reference 1 ), Section 11, "Program Documentation and Change Control." The process reviews 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 to any of the PRA models used in support of the VEGP 50.69 process will be peer reviewed prior to implementing those changes in the PRA model used for categorization.

3.1.3 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 peer review processes as discussed in Section 3.2 of this enclosure.

Uncertainty evaluations associated with the risk categorization process are addressed using the processes discussed in Section 8 and in the prescribed sensitivity studies discussed in Section 5 of NEI 00-04 (Reference 1 ).

The VEGP1 0 CFR 50.69 categorization process is described in NMP-ES-065 (Reference 16) and subordinate documents. That process follows the guidance in NEI 00-04 (reference 1) as endorsed in RG 1.201 (Reference 2). Within this process, when a PRA model is used to address the contribution to sse risk significance due to a given hazard, a set of sensitivity evaluations is required to be performed. NMP-ES-065-001 (Reference 17) describes the active categorization process that encompasses this requirement. Specifically, Section 4.11 (Table 4-6) of NMP-ES-065-001 includes the recommended set of sensitivity studies to be included in the SPRA portion of the active categorization. (Note that the current version of the procedures describes both use of SPRA and SMA, and specifies use of the VEGP SMA; this will be revised after approval is received to use the SPRA.) The last item on the list of sensitivities is "Any applicable sensitivity studies identified in the characterization of PRA adequacy and identification of important assumptions and sources of uncertainty." For the current SPRA model, no additional SPRA-specific sensitivities have been identified that would be expected to have an important impact on categorization results. As the model is updated, the sources of uncertainty will be re-evaluated and, if appropriate, additional sensitivities may be added to the process.

In the overall risk sensitivity studies SNC utilizes a factor of 3 to increase the unavailability or unreliability of low safety significance (LSS) components. Consistent with the NEI 00-04 guidance (Reference 1 ), SNC performs 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, including the SPRA once this amendment request is approved, 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 E-5

Enclosure to NL-16-2382 Basis of the Proposed Change increases in failure rates of categorized components are detected and addressed before reaching the rate assumed in the sensitivity study.

The SPRA assumptions and sources of uncertainty were reviewed to identify those which would be significant for the evaluation of this application. If the VEGP SPRA model used a potentially 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 configuration risk calculations were considered key for this application.

Key VEGP SPRA model specific assumptions and sources of uncertainty for this application have been identified and dispositioned within the VEGP PRA documentation, which is available for NRC audit. The conclusion of this review is that the only additional sensitivity analysis required to address VEGP SPRA model specific assumptions or sources of uncertainty beyond those specified in NEI 00-04 is to evaluate the impact of the possibility that, following actuation of the reactor coolant pump shutdown seals (RCP SDS), there may be some scenarios where cold leg temperatures could exceed the rated temperature in a timeframe insufficient to credit operator action following a seismic event, leading to RCP seal LOCA not currently included in the SPRA. This issue is still under investigation by Westinghouse.

3.2 PRA REVIEW PROCESS RESULTS (1 0 CFR 50.69(8)(2)(111))

The VEGP SPRA model has 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 4). Specifically, the model was subject to a self-assessment and a peer review conducted in November 2014. The VEGP SPRA peer review report (Reference 19) states that the "peer review was performed using the process defined in Nuclear Energy Institute (NEI) 12-13." No exceptions to use of NEI 12-13 (Reference 20) are noted in the peer review report.

Regulatory guide 1.200 (Reference 4) endorses ASMEIANS PRA Standard Addendum A (Reference 21) but, as noted in an NRC letter to ASME (Reference 22), does not endorse PRA Standard Addendum B (Reference 23). The VEGP SPRA peer review was performed using the SPRA requirements in Addendum B. The following discussion addresses the differences relative to establishing the technical capability of the VEGP SPRA.

Requirements for SPRA are provided in Part 5 of the PRA Standard. A comparison of the Part 5 requirements between Addendum A and Addendum B is included in Tables 6-4, 6-5, and 6-6 of EPRI 1025287 (Reference 24, referred to as the SPID). These SPID tables provide, in columns labeled "Relevant Intent of Guidance in SPID," notes regarding the differences between the requirements in Addendum A and Addendum B and an assessment of how the requirements in the PRA Standard relate to those in the SPID. The notes confirm that the changes in Addendum B are primarily focused on clarification of existing requirements or revision of "action verbs" used in the PRA Standard, and do not introduce substantive changes that would affect the assessment of PRA technical capability by a peer review team.

In the Summary of NRC Comments on Addendum B attached to the letter to ASME (Reference 22), NRC commented on numerous issues across the Parts of the PRA Standard. Of the 1 0 numbered topics in that attachment:

E-6

Enclosure to N L-16-2382 Basis of the Proposed Change Topics 1, 4, 5, 6, 7, 9, and 10 are generic, i.e., NRC comments on issues with the PRA Standard that are not specific to Part 5 and not pertinent to the assessment of VEGP SPRA technical capability; Topic 2 is labeled "Clarity and Understandability of the Requirements", and includes 21 sub-topics. Of those 21 sub-topics, 12 make no reference to the requirements in Part 5 of the Standard. Sub-topics a, b, f, h, i, j, I, r, and u do make reference to specific requirements in Part 5, as examples of NRC observations of places where additional clarity should be provided in Part 5. However, any such lack of clarity is a generic issue with Part 5 of the PRA Standard rather than an issue that would affect the VEGP SPRA in particular. There are various industry peer review consistency initiatives in place through the Owners Groups and Nuclear Energy Institute that focus on achieving consistency across peer reviews regardless of such issues with the PRA Standard.

These comments are therefore not pertinent to the assessment of VEGP SPRA technical capability.

Topic 3 is labeled "Inconsistency in the Standard", and includes nine sub-topics, four of which make no reference to the requirements in Part 5. Sub-topics a, c, d, g, and h include paragraphs that focus on specific NRC comments dealing with examples of inconsistent treatment of specific technical parameters, definitions, lack of parallel structure, and so forth, either across the Parts of the Standard, or in some examples within Part 5. As is the case for Topic 2, these are not issues that would affect the VEGP SPRAin particular, and there are industry initiatives to achieve consistency across peer reviews despite such issues. These comments are therefore not pertinent to the assessment of VEGP SPRA technical capability.

Topic 8 is labeled "Application". The comment is that certain requirements in Part 5 of the PRA Standard are stated with regard to applications rather than relative to the base SPRA. Since the VEGP 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 issue is not pertinent to the assessment of VEGP SPRA technical capability.

Based on the above assessment, it is determined that the seismic PRA also meets the technical adequacy of Addendum A. summarizes the SPRA model results and provides the date of the industry peer review performed against RG1.200 (Reference 4). Attachment 2 provides a summary of:

VEGP SPRA peer review Fact and Observation findings and disposition relative to the 50.69 application.

Identification of and basis for any sensitivity analysis performed to address issues identified in the peer review findings, as part of the noted dispositions.

Of the peer review finding-level Facts and Observations (F&Os) listed in Attachment 2, most were associated with PRA Standard supporting requirements (SRs) that were deemed by the peer reviewers to be either Met or met at capability category II. This indicates, as can be seen from the finding details, that these findings deal with relatively focused issues that have been adequately dispositioned within the reviewed methodologies, for the SPRA and for the 50.69 application. Many of these were documentation-related.

E-7

Enclosure to NL-16-2382 Basis of the Proposed Change The remaining finding-level F&Os are associated with SRs deemed by the peer reviewers to be not met, or not met at capability category II. These are as summarized in the list below.

As this list indicates, there are only 1 0 not met I capability category I SRs associated with the finding F&Os.

Of these, 6 are seismic hazard-related SRs, for which the findings are associated with:

(a) inadequate documentation of the hazard analysis performed; (b) demonstration that sufficient consideration has been given to more recent geologic events and associated modeling; or (c) sensitivity calculations for the models and parameters used in the site hazard. The documentation items have been addressed, as noted in the dispositions for the affected findings in Attachment 2.

Findings associated with Not Met or Capability Category I SRs SR Findings Summary of Issue Not Impact on SPRA Results Fully Resolved SHA-C4 12-18, 12-36 Finding issues are resolved.

No impact on SPRA results.

SHA-H1 12-18, 12-36 Finding issues are resolved.

No impact on SPRA results.

SHA-11 12-15 Finding issues are resolved.

No impact on SPRA results.

SHA-12 12-15 Finding issues are resolved.

No impact on SPRA results.

SHA-J1 12-1, 12-2, 12-11, 12-16 Finding issues are resolved.

No impact on SPRA results.

SHA-J3 12-8 Finding issues are resolved.

No impact on SPRA results.

SFR-A2 14-1 ' 14-7' 14-1 0 Finding issues are resolved.

No impact on SPRA results.

SPR-82 16-4, 16-6 Finding issues are resolved.

No impact on SPRA results.

SPR-84 16-1 Finding issues are resolved.

No impact on SPRA results.

SPR-F1 12-31 ' 16-5 Finding issues are resolved.

No impact on SPRA results.

One of the SRs is fragilities-related. Two of the 3 findings associated with this SR deal with conservatisms that the reviewers noted, which have now been addressed within the analytical methodology that the peer reviewers found acceptable. The remaining finding is associated with a specific polar crane fragility issue, which has also been addressed within the reviewed methodology.

Three of the SRs are PRA modeling-related. Three of the findings associated with this SR are related to implementation of the seismic performance shaping factor approach in the human reliability analysis. The comments in those findings have been addressed and implemented in the SPRA model, within the reviewed methodology, without significant impact on the results. One finding was related to the relay chatter evaluation, for which the model update resolves the finding. The last finding was related to the SPA documentation, which has been updated to resolve the finding.

The information in the table identified above demonstrates 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-8

Enclosure to NL-16-2382 Basis of the Proposed Change 3.3 RISK EVALUATIONS (10 CFR 50.69(8)(2)(1V))

The VEGP 10 CFR 50.69 categorization process will implement the guidance in NEI 00-04 (Reference 1 ). 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 Section 8 of the guidance will be used to confirm that the categorization process results in acceptably small increases to core damage frequency (CDF) and large early release frequency (LEAF). 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.

The VEGP SPRA model does not credit portable FLEX or offsite FLEX capabilities or any associated operator actions. The SPRA model is built on the VEGP internal events at power model, which includes the installed new reactor coolant pump (RCP) low leakage seals, which have been identified as part of the VEGP installed FLEX capability. Thus, the SPRA includes the impact of the new RCP seals. No additional operator actions are required to implement the new seals relative to the original seals.

The VEGP SPRA reflects the current seismic hazard applicable to the plant. SNC will follow industry guidance and common practice in determining whether an update of the SPRA may be warranted due to new availability of new consensus seismic hazard information.

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

The regulations at Title 10 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 1.17 4, "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.

In addition, a review of the Federal Register Notice (FAN) announcing 10 CFR 50.69 (Reference 15) was performed to identify any specific requirements or expectations regarding impact on compliance if substituting SPRA for an SMA safe shutdown equipment list (SSEL).

The Statements of Consideration in the FAN, in 111.4.3 "Section 50.55a(f), (g), and (h) Codes and Standards," clarifies that "The Commission will not remove the repair and replacement provisions of the ASME Code required by § 50.55a(g) for ASME Class 1 SSCs, even if they are E-9

Enclosure to NL 2382 Basis of the Proposed Change categorized as RISC-3, because those SSCs constitute principal fission product barriers as part of the reactor coolant system or containment" and notes that "the Commission has not removed the requirements for fracture toughness specified for ASME Class 2 and Class 3 SSCs because fracture toughness is a significant design parameter for the material used to construct the SSG. Fracture toughness is a property of the material that prevents premature

• failure of an SSG at abrupt geometry changes, or at small undetected flaws. Adequate fracture toughness of SSCs is necessary to prevent common cause failures due to design basis events, such as earthquakes." These considerations affect treatment of RISC-3 SSCs, but not the categorization process. Substitution of a SPRA for the SMA SSEL does not have a direct bearing on treatment. Section 111.4.8 of the SOC, "Appendix A to 10 CFR Part 100 (and Appendix S to 10 CFR Part 50 (Seismic Requirements))" clarifies the scope of seismic requirements considered to be special treatment requirements; however there are no requirements imposed in this section on categorization.

In the Section by Section Analysis in the FRN, in V.5.2 "Section 50.69(d)(2) RISC-3 Treatment", it is noted that "Section 50.69(d)(2) requires that the licensee or applicant must ensure with reasonable confidence that RISC-3 SSCs remain capable of performing their safety-related functions under design basis conditions, including seismic conditions and environmental conditions and effects throughout their service life." The FRN further states that "Under§ 50.69, RISC-3 SSCs would continue to be required to function under design basis seismic conditions (such as design load combinations of normal and accident conditions with earthquake motions), but would not be required to be qualified by testing or specific engineering methods in accordance with the requirements stated in 1 0 CFR part 100,... The rule does not remove the design requirements related to the capability of RISC-3 SSCs to remain functional considering Safe Shutdown Earthquake and Operating Basis Earthquake seismic loads, including applicable concurrent loads. The rule does not change the design input earthquake loads (magnitude of the loads and number of events) or the required load combinations used in the design of RISC-3 SSCs. For example, for the replacement of an existing safety related SSG that is subsequently categorized as RISC-3, the same seismic design loads and load combinations would still apply." These are treatment considerations that are neither affected by substituting use of a SPRA for the SMA SSEL in the categorization process, nor of issue to the approved VEGP 50.69 categorization process.

Section V.5.2 does provide language noting the link between categorization and treatment, as follows: "Section 50.69(d)(2) requires that the treatment of RISC-3 SSCs be consistent with the categorization process. This rule language means that, when establishing the treatment for RISC-3 SSCs, the licensee or applicant must take into account the assumptions in the categorization process regarding the design basis capability and reliability of RISC-3 SSCs to perform their safety related functions throughout their service life. The evaluation by the licensee or applicant of the consistency of the treatment of RISC-3 SSCs with the categorization process may be qualitative so long as it provides reasonable confidence of the design basis capability of RISC-3 SSCs, based on plant-specific and industry-wide operational experience and vendor information. In establishing treatment for RISC-3 SSCs, the licensee or applicant is responsible for addressing applicable vendor recommendations and operational experience such that the treatment established for RISC-3 SSCs provides reasonable confidence for design basis capability. For example, operational experience might be described in NRC information notices or identified in responses to NRC bulletins, generic letters, or other licensee commitment documents. The treatment applied to RISC-3 SSCs must also support the assumptions used in justifying the removal of requirements applicable to those SSCs." The existing SNC 10 CFR 50.69 procedure (NMP-ES-065, Reference 16) and the treatment procedure (NMP-ES-065-004, Reference 18) already address this, e.g., Section 4.1, Note 8 of E-10

Enclosure to N L 2382 Basis of the Proposed Change NMP-ES-065-004 states: "8. Alternative treatments must be consistent with and maintain the validity of the SSC categorization basis. For example, alternative treatments for RISC-3 SSCs should maintain any risk increase (that could result from an application of alternate treatment) to an acceptably small level (i.e., below that identified in risk sensitivity results, if any)."

Given the above discussion, there is no direct impact on the NRC approved VEGP 10 CFR 50.69 program with regard to treatment that will result from use of the SPRA instead of the SMA SSEL, in accordance with the existing categorization process.

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 amend the approved voluntary implementation of the provisions of 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 (SSCs) for Nuclear Power Plants" to include use of the Vogtle Electric Generating Plant (VEGP) Seismic Probabilistic Risk Assessment (SPRA) in place of the VEGP Individual Plant Examination Of External Events Seismic Margins Analysis (SMA).

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 amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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

Response: No.

The proposed change replaces the use of the VEGP SMA with use of the peer reviewed VEGP SPRA within the NRC approved 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 use of an SPRAin place of an SMA is allowed by the 50.69 process guidance defined in NEI 00-04 (Reference 1) as endorsed by NRC in RG 1.201 (Reference 2}. 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 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.

E-11

Enclosure to NL-16-2382 Basis of the Proposed Change 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 continues to 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 continue to 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 "no 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 Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5 ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined E-12

Enclosure to N L 2382 Basis of the Proposed Change in 10 CFR 20, or would change an inspection or suNeillance 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.

6 REFERENCES

1.

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

2.

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.

3.

Generic Letter 88-20, "Individual Plant Examination of External Events (I PEEE) for Severe Accident Vulnerabilities - 10 CFR 50.54(f), Supplement 4," USNRC, June 1991.

4.

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.

5.

Ajluni, M. J. (Southern Nuclear Operating Company) to U. S. Nuclear Regulatory Commission, Vogtle Electric Generating Plant Pilot 10 CFR 50.69 License Amendment Request, August 31, 2012 (ADAMS Accession No. ML12248A035).

6.

U.S. Nuclear Regulatory Commission to C.R Pierce. (Southern Nuclear Operating Company), 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 (ADAMS Accession No. ML14237A034).

7.

U.S. Nuclear Regulatory Commission to B.K Taber (Southern Nuclear Operating Company) Vogtle Electric Generating Plant-NRC Evaluation Of Risk Informed Categorization And Treatment Of Systems, Structures, And Components, Inspection Report 05000424/2016008 AND 05000425/2016008, August 10, 2016 [ADAMS Accession No. ML12061A245]

8.

EPRI 1009684, "CEUS Ground Motion Project Final Report," Electric Power Research Institute, Palo Alto, CA, December 2004.

9.

Southern Nuclear Operating Company, Vogtle Electric Generating Plant Units 3 and 4, COL Application Part 2, Final Safety Analysis Report, Revision 5, Section 19.55.6.3, Site Specific Seismic Margin Analysis, March 2011.

10. Southern Nuclear Operating Company, "Vogtle Early Site Permit Application Part 2, Site Safety Analysis Report (SSAR)," Revision 5, December 2008.

11. EPRI, USDOE, USNRC, 2012, "Central and Eastern United States Seismic Source Characterization for Nuclear Facilities," U.S. Nuclear Regulatory Commission Report NUREG-2115.

12. McGuire, R.K., W. J. Silva, and C. J. Costantino. Technical Basis for Revision of E-13

Enclosure to NL-16-2382 Basis of the Proposed Change Regulatory Guidance on Design Ground Motions, Hazard-and Risk-Consistent Ground Motion Spectra Guidelines", prepared for Nuclear Regulatory Commission, NUREG/CR-6728, 2001.

13.. Southern Nuclear Operating Company, NL-14-0344, "Vogtle Electric Generating Plant-Units 1 and 2, Seismic Hazard and Screening Report for CEUS Sites," March 31, 2014.

NRC Adams ML14092A019.

14. EPRI 3002008093, "An Approach to Human Reliability Analysis for External Events with a Focus on Seismic," Electric Power Research Institute, Palo Alto, CA, December 2016.

15. Federal Register, Vol. 69, No. 224, Monday, November 22, 2004, Rules and Regulations.

16. NMP-ES-065, "1 0 CFR 50.69 Program," Version 2.0, Southern Nuclear Operating Company, April 2016.

17. NMP-ES-065-001, "1 0 CFR 50.69 Active Component Risk Significance Insights," Version 2.0, Southern Nuclear Operating Company, February 2015.

18. NMP-ES-065-004, "1 0 CFR 50.69 Alternative Treatment Requirements," Southern Nuclear Operating Company, June 2016.

19. PWROG-15004-P, "Peer Review of the Vogtle Units 1 & 2 Seismic Probabilistic Risk Assessment," Westinghouse Electric Company for PWR Owners Group, February 2015.

20. NEI12-13, "External Hazards PRA [Probabilistic Risk Assessment] Peer Review Process Guidelines", Nuclear Energy Institute, Rev. 0, August 2012.

21. ASMEIANS RA-Sa-2009, "Standard for Levei1/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.

22. 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 ML111720067.

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

24. EPRI 1025287, "Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic," Electric Power Research Institute, Palo Alto, CA, February 2013

25. EPRI TR-1 016737, Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments, December 2008 E-14

Enclosure to NL-16-2382 Basis of the Proposed Change : Seismic PRA Model Summary Information Unit Model Baseline CDF Baseline LERF 1

VEGP-SPAA-1, 2.8E-6/yr 3.3E-7/yr 12/31/16 2

VEGP-SPAA-1, 2.8E-6/yr 3.3E-7/yr 12/31 /16 Comments one model applicable to both units one model applicable to both units The pe+.r review was performed in November 2014 against AGI1.200 A2. The SPAA was updated to address peer review findings and refine fragilities. The update did not involve changes in methods or scope that would require a focused scope peer review. Though not shown in the table above, combining the Internal Events, Fire, Seismic, and Other External Events CDF values yields a total CDF estimate of 4.39E-05/yr (Unit 1) and 5.05E-05/yr (Unit 2).

The total LEAF estimates are 1. 73E-06/yr (Unit 1) and 1.90E-06/yr (Unit 2). For Units 1 and 2, the value for Total Internal and External events CDF is less than 1.0E-04/year and the Total LEAF is less than 1 E-05/year, and therefore meets AG 1.17 4 total risk criteria.

E-15

Enclosure to NL 2382 Basis of the Proposed Change : Disposition and Resolution of SPRA Peer Review Findings 1 Finding Supporting Finding Finding Suggested Finding Require-cc Number ment(s)

Description Basis Resolution 11-3 SHA-E2 IIIII I While variability in the mean To maintain hazard-consistent Expand documentation to base-case Vs profile is ground motion hazard at the demonstrate that a single incorporated in the site control point, the site response base-case Vs profile response analysis, no analysis needs to incorporate adequately represents the epistemic uncertainty in the appropriate epistemic uncertainty Units 1 &2 site. Or if that is not base-case profile is and aleatory variability in its the case, include epistemic represented.

inputs. The Vs profile for the uncertainty in the Documentation of the Vogtle Units 1 &2 site is characterization of Vs profile justification for this represented by a single Vs profile, and evaluate the impact on assessment should be indicating there is no epistemic control point ground motions.

expanded.

uncertainty in the mean base-case profile. Documentation of this (This F&O originated from assessment needs to be SA SHA-E2) expanded.

Discussion with staff indicates that consideration of the combined data for the Vogtle site (Units 1 &2, Units 3&4, ISFSI) provides sufficient confidence that a single mean base-case profile characterizes the site. This conclusion is based on the quantity and quality of the combined data and an evaluation showing the site is relatively uniform with respect to Vs. For some depth ranges, data from the nearby Savannah River Site (SRS) are used to support the profile interpretation.

Bechtel Document 23162-000-G65-GEK-0001 0 (SNC #SVO-GB-Disposition for 50.69 and other applications There is an abundance of site-specific Vs data from VEGP Units 3&4, which reduces epistemic uncertainty to an insignificant level.

Additional discussion of the rationale for use of a single base-case Vs profile for the site has been included in the documentation. The added discussion demonstrates that a single base-case shear-wave velocity (Vs) profile adequately represents the Vogtle site, based on the availability of Vs data, which reduces the epistemic uncertainty for this particular parameter.

This finding has been resolved with no significant impact to the SPRA results or conclusions.

1 In Attachment 2, all but the last column are extracted directly from the Peer Review report. The last column provides the disposition for the Findings.

E-16

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s) 11-8 SHA-E2 II/III Finding Description Upper crustal site attenuation of ground motion (kappa) is, generally, an uncertain parameter. Thus, to maintain hazard-consistent ground motion at the control point, this uncertainty should be incorporated in the site response analysis, or the basis for not including it should be provided. In either case, the technical basis and justification should be documented.

Finding Suggested Finding Disposition for 50.69 Basis Resolution and other applications X?R-011-001) presents summaries of velocity data, but does not provide sufficient information to support the lack of epistemic uncertainty at the Units 1 &2 site over the complete depth range of the Vs profile. This would typically require multiple measurements throughout the depth range that provide a consistent picture of natural variability about a single mean base-case profile. The technical basis and justification that a single base-case profile is appropriate should be provided in more detail.

This should include the basis for applying conclusions from other Vogtle locations to the Units 1 &2 site.

[A related Suggestion 11-2 addresses specifically potential epistemic uncertainty in the Blue Bluff Marl stratum.]

Calculation X2CFS129 Ver2 notes Provide a basis in the A discussion of the range that the damping associated with documentation for of possible values of the base-case profile corresponds representing base-case kappa deep soil damping has to a total kappa value for the soil at the site by a single value.

been included in the column of 0.01 sec. The report The basis might include documentation.

does not address epistemic sensitivity analyses to show uncertainty in kappa.

the impact of epistemic A sensitivity study on the uncertainty in kappa.

epistemic uncertainty of In discussion with staff during the deep soil damping has peer review, it was noted that been performed using randomization of the damping median, lower range, and associated with the profile layers upper range alternatives represents both random variability for deep rock damping.

and epistemic uncertainty. It was Site response analysis also noted that kappa was was performed using 1 E-expected to be small for the 4 HF and LF rock input E-17

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s) 12-1 SHA-J1 Not Met Finding Description (This F&O originated from SR SHA-E2)

As part of the PSHA implementation, the analyst has different alternatives for modeling the earthquake occurrences in the calculations. The PSHA documentation does not describe the approach that was used to model earthquakes.

Finding Basis Vogtle site and uncertainties in that small value would not be expected to have a significant impact on site amplification. Staff also noted that the approach used had been reviewed by the NRC for the Vogtle ESP and COLA.

The SPID (EPRI, 2013) provides guidance accepted by the NRC for response to NTTF 2.1 Recommendation: Seismic that indicates kappa is difficult to measure and thus subject to large uncertainty (SPID Section B-5.1.3.2).

Documentation of the technical basis for kappa characterization should be expanded.

The approach that was taken to model earthquakes in the PSHA calculation was not identified.

There are two basic alternatives that can be used to model earthquake events; as extended fault ruptures, or as point sources.

The approach that is used influences how the CEUS ground motion model is implemented.

E-18 Suggested Finding Disposition for 50.69 Resolution and other applications motion. The resulting amplification functions and log-standard deviation were weight-averaged and compared to the original base case for each of BBM High PI and BBM Low PI soil columns. It was concluded that the inclusion of alternative base cases for deep soil damping to account explicitly for the epistemic uncertainty associated with site kappa does not have any significant effects on the resulting seismic hazard curves and UHRS.

The sensitivity study has been added to the SPRA documentation.

This finding has been resolved with no significant impact to the SPRA results or conclusions.

Documentation should be A PSHA report has been provided that describes how prepared that describes seismic sources are modeled how earthquake events in the PSHA (i.e., how the SSG were modeled for area and GMMs) were implemented sources in the PSHA in the Vogtle PSHA.

calculations. This was by modeling each earth-quake as a point source, and using correction factors for distance and Qround motion uncertain-

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s)_

12-11 SHA-J1 Not Met Finding Description (This F&O originated from SA SHA-J1)

As part of the PSHA implementation, the analyst has alternatives for modeling the earthquake occurrences in the calculations. The PSHA documentation does not describe the approach that was used to model earthquakes in ALME sources.

(This F&O originated from SA SHA-J1)

Finding Basis No documentation is provided on either of these subjects (earthquake source modeling and use of the ground motion attenuation models). From questions posed to the PSHA analysts, it is our understanding that earthquakes were modeled as point sources and the appropriate ground motion aleatory uncertainty was used in the calculation.

The PSHA analysts were asked to describe the approach that was used to model earthquakes in the Charleston ALME seismic source.

The response indicated that earthquakes in the Charleston ALME source were modeled using

'pseudo faults'.

The PSHA report does not:

1. Describe that a 'pseudo fault' approach was used to model earthquakes in the Charleston ALME source.

2. Provide a definition of 'pseudo faults'.

3. Describe how the 'pseudo fault' approach was implemented for the Charleston ALME seismic source (e.g., what was the fault E-19 Suggested Finding Disposition for 50.69 Resolution and other applications ty that modify the ground motion estimate to in-I elude the effect of a closer distance to a fault rupture (because the rupture may be closer to the site than the single point used to represent that event) and the un-certainty in ground motion because the azimuth of the rupture is unknown. These correc-tion factors were pub-lished in EPAI (2004) (8].

This finding has been resolved with no significant impact to the SPAA results or conclusions.

Provide a description of the A PSHA report has been earthquake modeling approach prepared that describes that was used to model the how pseudo-faults were Charleston ALME seismic implemented to source and how the approach represent the Charleston was implemented.

ALME source. This includes: 1. A description of the pseudo-faults. 2. A definition of pseudo-faults as constructed faults that represent possible sources of future large earthquakes.

3. Implementation of the pseudo-faults including spacing and limits at the borders of the Charleston source. 4.Documentation of the rupture area, length, and width that

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s) 12-15 SHA-11, Not SHA-12 Met 12-16 SHA-J1 Not Met Finding Description A screening assessment was performed for soil liquefaction and is described in seismic fragility calculation (PRA-BC-V 025).

A screening assessment was not performed for other potential seismic hazards.

(This F&O originated from SR SHA-11)

The Vogtle PSHA has gone through a number of changes and revisions since 2012 due to changes in models, input data, etc. As new calculations were performed and reports generated, sensitivity results, were not carried forward. As a result, there does not exist a current report that includes all PSHA results, deaggregations, etc.

that is based on the current Finding Basis spacing that was used; how was the earthquake rate distributed to the faults, etc.).

4. Document the fault rupture model that was used.

5. Describe how earthquake events are distributed on the faults.

A screening analysis was not performed for hazards such as settlement, fault displacement, tsunami, seiche, etc.

It is anticipated these other seismic hazards will be screened out.

The documentation of the PSHA is provided in a collection of documents that were prepared in the 2012-2014 time frame. There does not exist a single document that contains a set of results that is based on the current PSHA model.

E-20 Suggested Finding Disposition for 50.69 Resolution and other applications were estimated for possible future earthquakes. 5. A description of how earthquake ruptures are distributed on the faults.

This finding has been I

resolved with no significant impact to the SPRA results or conclusions.

A screening analysis for other This evaluation was done seismic hazards should be for the Vogtle 3&4 COLA performed and documented as

[9] and is noted in the part of the PSHA and SPRA.

ESP SAR [1 0]. The Vogtle 3&4 evaluation is It is expected that information applicable to, and has in the FSAR for Vogtle 1 & 2 been cited in, the Vogtle and in the COLA for Units 3 &

1 &2 SPRA Qualification 4 can be used to support this report.

requirement.

This finding has been resolved with no significant impact to the SPRA results or conclusions.

Prepare a complete and up-to-A PSHA report has been date PSHA document that prepared that includes includes all results, sensitivity hazard results, calculations, deaggregation uncertainties in hazard, results, etc. that is based on and sensitivities to input the current model.

uncertainties; this summarizes hazard results for the Vogtle site.

This finding has been resolved with no significant impact to the SPRA results or conclusions.

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s) 12-18 SHA-82, 1111 SHA-C4, Not SHA-H1 Met Not Met 12-2 SHA-J1 Not Met Finding Description PSHA model.

(This F&O originated from SA SHA-J1)

The Vogtle PSHA is based on the CEUS SSC seismic source model which was completed in 2012. The SSC model was a developed at a regional scale that was based on data gathered up until about 2010. (Note, the date when data was gathered varied; for example the earthquake catalog was complete through 2008.) In the sense that the CEUS sse model was not specifically performed as a site-specific PSHA for the Vogtle site.

(This F&O originated from SA SHA-82)

The method that is used in the Vogtle PSHA to estimate the soil site hazard is not described or referenced.

(This F&O originated from SA SHA.. J1)_

Finding Basis As part of a site-specific PSHA, there is a need to gather, review and evaluate new geological, seismological, or geophysical information or information that is defined at a scale that was not considered in the development of the CEUS sse model. As part of the Vogtle SPAA, no effort was made to gather up-to-date and local (local to the Vogtle site) information to evaluate whether any new information has become available on active faulting and/or the development new seismic sources or the revision of sources in the CEUS SSC model in the vicinity of the Vogtle plant.

Since up-to-data was not gathered, consideration of alternatives could not be addressed.

For soil sites, the soil hazard is generally (though not exclusively, since other methods could be used) determined in two steps; probabilistic rock hazard results are estimated which are then combined with probabilistic estimates of the site response.

The method used in the Vogtl~

E-21 Suggested Finding Disposition for 50.69 Resolution and other applications A data gathering effort should A detailed study of new be undertaken to identify new geological, information that post-dates the seismological, and CEUS SSC data collection geophysical information effort. The data gathering effort was conducted, to should also look for determine if any information local to the Vogtle information subsequent site region that was not to the EPAI sse model considered, or at a scale that (EPAI, 2012 [11]) is was not addressed as part of available that should be the CEUS SSC regional incorporated into the evaluation.

seismic hazard results for Vogtle. This study is Some of this information may described in the SPAA be available in the COLA for documentation. While Vogtle Units 3 & 4.

the area around the site continues to be studied by many earth scientists, there was no new information identified that would change the estimate of seismic hazard for Vogtle.

This finding has been resolved with no significant impact to the SPAA results or conclusions.

The documentation should The methodology used include a description of the for the surface hazard methodology that is used to calculation has been combine the rock hazard described in detail, and a results and the site comparison made amplification factors to between the GMAS determine the soil hazard at using the two the Vogtle site.

approaches 2A and 3.

Approach 2A was used_

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s) 12-22 SHA-E2 IIIII I Finding Description The site response Calculation X2CFS129 Ver.

1 (2012) and Ver. 2 (2014) does not describe a framework for evaluating and characterizing sources of aleatory and epistemic uncertainty and how the approach was implemented.

(This F&O originated from SR SHA-E2)

Finding Basis PSHA to estimate the soil hazard is not described.

The site response calculation does not present a clear description of how aleatory and epistemic uncertainties are identified and evaluated. As a result it is difficult to track the propagation of uncertainties is carried out in the site response analysis.

It is worth noting that there is some epistemic site response E-22 Suggested Finding Disposition for 50.69 Resolution and other applications for the calculation of SSI input motions at foundation elevations and Approach 3 was used for the calculation of surface hazard and GMRS at the ground surface, as defined in NUREG/CR-6728 [12). It I

was concluded that the use of Approach 2A USHRS as input to the SSI analysis of the Vogtle plant is considered acceptable and does not present any significant inconsistency with the seismic hazard curve and GMRS at the ground surface, which were calculated using Approach 3.

This finding has been resolved with no significant impact to the SPRA results or conclusions.

A framework and approach for A description of the evaluating and modeling methodology used to uncertainties in the site account for epistemic response should be developed and aleatory and implemented. The site uncertainties in soil response calculation hazard has been added documentation should fully to the documentation.

describe the methodology and its implementation.

This finding has been resolved with no significant impact to the SPRA results or conclusions.

I

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Number Require-cc ment(s) 12-23 SPA-E5 II 12-24 SPA-E5 II Finding Description The quantification process has included the uncertainties in the seismic hazard, fragility and systems-analysis elements of the SPAA. The results in Table 5.1 are internally inconsistent and are inconsistent with the results reported in Sections 3 and 4 for CDF and LEAF, respectively.

(This F&O originated from SA SPA-E5)

The Quantification report does not provide documentation of the uncertainty analysis results.

(This F&O originated from SA SPA-E5)

Finding Basis uncertainty that is accounted for in the rock GMPEs.

Table 5-1 presents the results of three different uncertainty calculations for CDF and LEAF. In addition, point estimates for CDF and LEAF are calculated and reported in Section 5.1.1. Thus the table reports two estimates of the mean CDF and LEAF respectively from different uncertainty calculations and a

'Point Estimates' result for each.

All of these results are different than the point estimate (approximate mean) reported in Sections 3 and 4 for CDF and LEAF, respectively. The documentation in the report does not describe the basis (inputs) for these calculations, or offer an interpretation of the results.

The uncertainty analysis is presented in Section 5.1 with the results reported in Table 5.1. The report provides limited discussion of the results and the insights that might be gained from them.

The two sets of results that are reported in Table 5-1 are not discussed in terms of their relationship to each other. For E-23 Suggested Finding Disposition for 50.69 Resolution and other applications Develop and document an Additional detail has understanding of the earlier been added to the SPAA point estimate results for CDF Quantification report to and LEAF (as reported in document the Sections 3 and 4) and of uncertainty, importance, uncertainty results.

and sensitivity analyses and relate the uncertainty analysis mean CDF and LEAF to the point estimate values.

This finding has been resolved with no significant impact to the SPAA results or conclusions.

' I Provide documentation of the Additional detail has I

uncertainty analysis that been added to the describes the results, how they documentation of the are being interpreted and the seismic plant response insights that are derived from model, model them.

implementation, and quantification in the au report. In addition, the uncertainty, importance, and sensitivity analyses are described in more

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s) 12-26 SPA-E5 II Finding Description There are differences in the results for CDF and LEAF that are reported in Table 5.1. A possible contributor to these differences may be due to the number of Monte Carlo simulations that were performed.

(This F&O originated from SA SPA-E5)

Finding Basis instance the mean values should be the same (but are not). The uncertainty estimates provide insight to the total uncertainty and the contribution of the basic event uncertainty to the total.

In addition, neither Table 5.1 or the discussion identifies what is the 'final' uncertainty result that includes the propagation of uncertainties of all elements of the SPAA to the estimates of CDF and LEAF.

The report does not present the results of sensitivity calculations with regard to the number of Monte Carlo simulations that are needed to produce stable results.

It is our understanding from discussion with the PAA staff that these types of sensitivity calculations were performed.

E-24 Suggested Finding Disposition for 50.69 Resolution and other applications detail.

This finding has been resolved with no significant impact to the SPAA results or conclusions.

Document the results of Updated Monte Carlo sensitivity calculations on the uncertainty runs have number of Monte Carlo been performed with simulations required to 20,000 iterations for produce stable results.

SCDF and SLEAF. This is a sufficiently high number of simulations to produce a stable result.

The SPAA documentation has been updated to clearly indicates the results.

This finding has been resolved with no significant impact to the SPAA results or conclusions.

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s) 12-27 SPR-F2 Met 12-29 SPR-E2 Met Finding Description Documentation should be provided that describes how the plant model analysis is quantified.

(This F&O originated from SR SPR-F2)

The Quantification report provides limited documentation of the process and methods that were used to perform the uncertainty analysis.

(This F&O originated from SR SPR-E2)

Finding Basis The current quantification document does not provide a clear description of the how the plant model is quantified. For example the discussion does not identify how calculations are performed, what the limitations of these quantifications are and how they affect the results.

There is limited documentation of the process and the numerical methods that were used to perform the uncertainty analysis.

Based on the documentation that is provided and discussions with the PRA staff there is limited but not complete understanding of the methods that were used and the relationship of these methods to the results were obtained (reported in Table 5.1 ).

In some cases (as described in the documentation) the results from the uncertainty analysis (Table 5.1) are not the same as the results reported in Sections 3 and 4 for CDF and LEAF (though this connection is not clearly stated in the report). However, it would seem the results in Table 5.1 should be internally consistent.

E-25 Suggested Finding Disposition for 50.69 Resolution and other applications Provide clear and complete The QU report documentation of the approach documentation has been used to quantify the seismic updated to describe the plant response model, to quantification process, perform the risk quantification, including the technique uncertainty analysis, and for combining cutsets importance analysis.

over the 14 acceleration intervals, and obtaining the importance measures.

This finding has been resolved with no significant impact to the SPRA results or conclusions.

Document the process and Additional detail has methods that were used to been added to the QU perform the uncertainty report to document the analysis. Where appropriate uncertainty, importance, document where consistencies and sensitivity analyses and potential inconsistencies and relate the uncertainty in results might be expected.

analysis mean SCDF and SLERF to the point estimate values.

This finding has been resolved with no significant impact to the SPRA results or conclusions.

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s) 12-31 SPR-F1 Not Met 12-32 SPR-F3 Met Finding Description The standard requires a level of documentation that provides an understanding of the seismic plant response model and the quantification. This requirement is not met.

{This F&O originated from SR SPR-F1)

The documentation of the sources of model uncertainty and a description of the analysis assumptions is not complete in the SPRA quantification report. In addition, there is not a clear description of the uncertainty analysis and the contributors to the total uncertainty beyond a simple report from UNCERT.

{This F&O originated from SR SPR-F3)

Finding Suggested Finding Disposition for 50.69 Basis Resolution and other applications There is limited documentation Documentation should be Additional detail has that describes the seismic plant provided in sufficient detail that been added to the response analysis and describes the seismic plant documentation of the quantification; how the model was model, how it is implement and seismic plant response implemented, how the quantified.

model, model quantification was performed and implementation, and a discussion of the analysis quantification in the QU results.

report. In addition, the uncertainty, importance, To meet this requirement, the and sensitivity analyses documentation must be in are described in more considerable detail in order to detail.

support the review process and future updates. Part of the This finding has been documentation should include a resolved with no detailed discussion of the results, significant impact to the sensitivity calculations, and the SPRA results or uncertainty analysis.

conclusions.

The purpose of this supporting Document and discuss the The documentation of requirement is that documentation contribution of the different the uncertainty analysis should be presented that sources of uncertainty that are has been expanded in addresses the sources of modeled in the SPRA.

the Quantification report.

epistemic {knowledge) uncertainty A discussion of sources that are modeled and their of model uncertainty has contribution to the total uncertainty been added to the report, in CDF and LERF.

and potentially important sources have been In addition, the documentation addressed in the should discuss elements of the sensitivity analysis.

seismic plant model where there may be latent sources of This finding has been uncertainty that are not modeled resolved with no and assumptions that are made in significant impact to the performing the analysis.

SPRA results or conclusions.

E-26

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s) 12-36 SHA-83, 1/11, SHA-C4, Not SHA-H1 Met Not Met 12-8 SHA-J3 Not Met Finding Description As part of a site-specific PSHA, an up-to-date earthquake catalog should be used. The CEUS sse study involved the development of a comprehensive earthquake catalog based on data through 2008. The Vogtle site-specific PSHA should consider the impact sse of any additional seismicity since 2008 up to the time the study started.

(This F&O originated from SR SHA-C4)

A foundational element of PSHA as it has evolved over the past 30 years is the Finding Basis As part of the Vogtle PSHA an effort was not made to gather data on earthquakes that occurred since 2008. As such, the analysts did not assess whether more recent seismicity is consistent with the characterization parameters estimated as part of the CEUS sse study (NRC, 2012).

We note that as part of the Vogtle PSHA, calculations were performed to recompute the seismic hazard at the site to take into account changes in the CEUS sse earthquake catalog through 2008 that were made following the completion of the CEUS SSG study. These changes reflect the identification of reservoir induced seismicity earthquakes and the re-interpretation of the location of some earthquakes in the Charleston, SC area that occurred in the 1880's (EPRI, 2014).

References EPRI (2014). Review of EPRI 1021 097 Earthquake Catalog for RIS Earthquakes in the Southeastern U. S. and Earthquakes in South Carolina Near the Time of the 1886 Charleston Earthquake Sequence, transmitted by letter from J.

Richards to R. McGuire on March 5, 2014.

The documentation of the sources of model uncertainty analysis and a description of the analysis E-27 Suggested Finding Disposition for 50.69 Resolution and other applications An up-to-date earthquake An update to the catalog for the Vogtle site earthquake catalog was region should be developed to prepared from the time of assess whether modifications the CEUS SSG catalog to the seismic source (through 2008) through recurrence parameters or February 2016. The rate required. The updated catalog, of occurrence of resources used in compiling earthquakes within 320 the update and the results of km of the Vogtle site was the evaluation should be compared to the rate of documented as part of the earthquakes represented PSHA. If more recent by the CEUS sse seismicity is not consistent seismic source model for with the existing CEUS SSG that same area, this seismic source parameters, comparison being made the parameters should be for M>2.9. It was found updated and the PSHA should that the updated catalog be updated.

implied a rate of earthquakes that is lower than the mean rate from the CEUS sse seismic sources. Therefore, incorporating the effects of a updated catalog on the hazard at Vogtle would decrease the hazard slightly, and was not undertaken. This comparison is documented in the SPRA documentation.

This finding has been resolved with no significant impact to the SPRA results or conclusions.

The resolution to this finding Sources of uncertainty in could involve:

the seismic hazard analysis for Vogtle are

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s)

Finding Description development and implemen-tation of methods to identify, evaluate, and model sources of epistemic (model and parametric) uncertainty in the estimate of ground motion hazards. As such fairly rigorous analyses are carried out (SSHAC studies) to quantitatively address model uncertainties.

At the same time there is within any analysis sources of uncertainty that are not directly modeled and assumptions that are made for pragmatic or other reasons. There are also sources of model uncertainty that are embedded in the context of current practice that are 'accepted' and typically not subject to critical review. For instance, in the PSHA it is standard practice to assume that the temporal occurrence of earthquakes is defined by a Poisson process. This assumption is well accepted despite the fact that it violates certain funda-mentally understanding of tectonic processes (strain accumulation). A second practice is the fact that earthquake aftershocks are not modeled in the PSHA, even though they may be sianificant events Finding Basis assumptions is not complete in the PSHA report in its current form such that a clear understanding of the contribution of individual sources of uncertainty to the estimate of hazard are understood. Limited information on the contribution of seismic sources to the total mean hazard is presented, but information on the contributors to the uncertainty is not provided.

With respect to addressing model uncertainties and associated assumptions there are some examples that can be identified in the Vogtle PSHA. For example, in the site response analysis the assumption is made that the 1 D equivalent linear model (SHAKE type) to estimate the site amplification and ground motion input to plant structures is appropriate.

E-28 Suggested Finding Disposition for 50.69 Resolution and other applications

1. Documentation and discussed in the updated discussion of the contribution SPRA documentation.

of different sources of These include uncertainty that are modeled in uncertainty in seismic the PSHA. The documentation source model (for of the contribution of different background earthquake sources of uncertainty can be sources and for the shown by means of 'tornado Charleston RLME), in plots' that quantify the maximum magnitude for sensitivity of the hazard at background seismic different ground motion levels sources and for the to the various branches in the Charleston RLME, in logic tree. These plots show ground motion prediction which sources of epistemic equation, in smoothing uncertainty are most important.

assumptions for It should include the source seismicity parameters in model uncertainty, ground background sources, and motion model uncertainty, and in site amplification site response uncertainty.

model. "Tornado plots" Currently, the total uncertainty are included in the is shown by the hazard updated SPRA fractiles, but it is not broken documentation that show down to provide understanding the contribution to total as to what is most important.

uncertainty in seismic hazard from source

2. Identification and discussion model uncertainty, of model assumptions that are maximum magnitude made.

uncertainty, ground motion prediction equation uncertainty, smoothing assumptions for seismicity parameters in background sources, and site response uncertainty. These plots are presented for 10 Hz and 1 Hz spectral acceleration, for ground motion amplitudes corresponding to mean annual frequencies of

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s)

Finding Description (depending on the size of the main event).

In the spirit of the standard it seems appropriate that sources of model uncertainty that are modeled as well as sources of uncertainty and associated assumptions as they relate to the site-specific analysis should be identified/ discussed and their influence on the results discussed.

As SPRA reviews and the use of the standard has evolved, it would seem the former interpretation is reasonable, but potentially incomplete. It is reasonable from the perspec-tive that document-tation of the sources of model uncer-tainty and their contribution to the site-specific hazard results is a valuable product that supports the peer review process and assessments in the future as new information becomes available). Similarly, documenting assumptions provides similar support for peer reviews and future updates.

The notion that model uncertainties and related assumptions that are not addressed in the PSHA is at Finding Suggested Finding Disposition for 50.69 Basis Resolution and other applications exceedance of 1 0-4 and 1 0-5. These "tornado plots" show that ground motion prediction equation is the major contributor to seismic hazard uncertainty for both 10Hz and 1 hz spectral acceleration, and maximum magnitude of the Charleston RLME source is an important contributor for 1 Hz spectral acceleration.

The use of equivalent linear one-dimensional site response analysis, and its associated assumptions, and its adequacy for the Vogtle site are documented in the hazard calculation.

This finding has been resolved with no significant impact to the SPRA results or conclusions.

E-29

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s) 14-1 SFA-A2 I

Finding Description a certain level an extreme requirement that may not be readily met and may not be particularly supportive of the analysis that is performed.

For purposes of this review, the following approach is taken with regard to this supporting requirement:

1. The documentation should present quantitative results and discussion the sources of epistemic uncertainty that are modeled and their contribu-tion to the total uncertainty in the seismic hazard.

2. The documentation should discuss elements of the PSHA model where their may be latent sources of model uncertainty that are not modeled and assumptions that are made in performing the analysis.

(This F&O originated from SA SHA-J3)

The conservatisms that exist in structural demand were not properly accounted for in the estimation of component and structure fragilities.

(This F&O originated from SA SFR-A2).

Finding Suggested Finding Disposition for 50.69 Basis Resolution and other applications SFA-A2 requires that seismic Account for conservatism in Evaluation of anchorage fragilities be based on plant-the building response analyses has been updated to specific data and that they are in the structure response factor include clipping of in-realistic and median centered with for component fragility structure response reasonable estimates of evaluations.

spectra, and the uncertainty.

methodology is Use clipped spectra for documented in the The structural response factor assessing anchorage fragility notebook.

used in all component fragilities capacities.

Structure response is reviewed is reported as 1.0. This dominated by the soft factor will be greater than 1.0 soil on which Vogtle 1 because of the conservatism and 2 structures are introduced in the demand through founded. This would E-30

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s) 14-10 SFR-A2 I

Finding Description Significant conservatisms were noted in several sampled fragility calculations.

Finding Basis the structural analysis. Because of this, the component and structural fragilities are biased low.

The fragilities developed for structures and components that are mounted in those structures will be biased low because the input structural demands include conservatisms. Time histories used for the SSI analysis have been processed such that each record envelopes the target UHRS. This will introduce some level of conservatism. The input motion at the control point has been scaled to produce resultant FIRS that envelopes the FIRS coming out of the site-consistent input motion analysis. In structure response spectra coming out of the SSI analyses were not peak clipped when computing anchorage demands. Structure response at the calculated equipment fragility levels is considerably higher than the 1 E-4 UHRS considered in the building response analyses. The structure will have additional cracked shear walls and higher associated levels of damping at these higher ground motions.

In the fragility calculations of heat exchangers (PRA-BC-V-14-009 Appendix A), nozzle loads significantly contribute to the seismic demands which form the basis for the median capacities.

Based o_11_lr!-pl(int_walkdowns by E-31 Suggested Finding Disposition for 50.69 Resolution and other applications cause higher damping at lower hazard frequency levels and lead to stress similar to the stress calculated for the buildings at 1 E-4. As a result the structural response factor is close to 1 and is accounted for appropriately in the fragility evaluations.

The input time history motion at the control point in the SSI analysishas been modified to reasonably match the corresponding 1 E-4 UHRS from the site-consistent input motion analysis.

This finding has been resolved.

Realistic nozzle loads should The CCW and ACCW be determined for fragility heat exchanger evaluation of heat exchangers.

capacities have been updated to reflect realistic nozzle loads.

The equipment fragilities The equipment capacity factor have been updated to

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s)

Finding Description (This F&O originated from SA SFR-A2)

Finding Basis the peer review teams and also noted in the walkdown report, the piping is well supported in all directions and will not impose significant nozzle loads during a seismic event. The CCW and ACCW capacities are below the 2.5g screening level and are significant contributors to risk so more realistic fragilities are required.

Battery rack 11806B3BN3 in calculation PRA-BC-V-14-010 Appendix J2 is governed by GERS capacity. The GERS capacity is taken to be 1 g, which corresponds to a frequency of 1 Hz. This is not realistic. The actual capacity is about 4g. The median capacity reported in the calculation is well below the 2.5g screening level and is not realistic.

The median capacity reported for the Turbine Driven Auxiliary Feedwater Pump is reported in Calculation PRA-BC-V-14-008 as 1.56g. This fragility is based on the seismic qualification document. The frequency range of interest for the fragility evaluation should be centered around the fundamental frequency of the assembly and not consider the entire frequency range.

E-32 Suggested Finding Disposition for 50.69 Resolution and other applications should be based on the account for appropriate frequency range of interest.

frequency, and That frequency range of uncertainty has been interest is centered at the considered in these fundamental frequency of the updates.

pump, and considers some uncertainty in that frequency.

This finding has been resolved.

I

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment{s) 14-14 SFR-G2 Met 14-17 SFR-02 Met Finding Description The iterative process used for developing realistic fragilities is not well documented.

(This F&O originated from SA SFR-G2)

Inconsistencies and errors in NSSS fragility development.

(This F&O originated from SA SFR-02}

Finding Basis In review of the seismic fragility calculation for the safety features sequencer (11821 U3001 ), it was discovered that an iterative process was used. The initial fragility is based on EPRI 6041 screening methodology and an equipment capacity factor that is equal to the EPA I 6041 median capacity divided by the peak in structure demand. If this value is Jess than the screening capacity (2.5g), then the fragility may be refined by examining the component fundamental frequency. The fragility may be further refined by examining component specific qualification test reports. However, the fragility used in the logic tree by the systems analyst is generally the highest of these computed. This is reasonable and appropriate, however, this process is not described in the fragility notebook or fragility calculations.

Fragilities for the Vogtle 1 &2 Nuclear Steam Supply System (NSSS) are based on the results of the Westinghouse analysis of record (AOR} associated with the safe shutdown earthquake (SSE}.

In general, fragilities are developed through scaling of the SSE demands to the ALE and using the AOR seismic margins.

Various deficiencies were noted in the development of the fragilities associated with these components.

E-33 Suggested Finding Disposition for 50.69 Resolution and other applications Add a description of the The description of the iterative process for computing iterative process for the component fragilities in the computing fragilities has SPRA documentation been documented.

This finding has been resolved with no significant impact to the SPRA results or conclusions.

Update SNC calculation no.

The following changes PRA-BC-V-14-015 to have been made: NSSS incorporate corrections and fragility calculations have enhancements.

been updated to reflect Westinghouse-provided critical loads and support capacities represented in the critical failure modes; the effect of inelastic energy absorption is factored in and documented in fragility calculation as appropriate; the Reactor

Enclosure to NL 2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s)

-~

Finding Description Finding Basis Basis: The NSSS Seismic fragility evaluation (SNC calculation no.

PRA-BC-V-14-015) includes detail calculations for each of the major NSSS components. It indicates that the critical failure modes for the components are controlled by the support capacities.

During the Peer Review, the team members discussed these issues with SNC staff to obtain insights and develop potential resolution paths. Key issues included:

(a) Basis for assumption that the support capacities represented the critical failure mode was not documented. SNC indicated that this was based on input from Westinghouse and NUREG-3360 and will update the fragility evaluation of provide this information.

(b) Inelastic energy absorption was not credited to increase the median capacities - this does not result in realistic median capacities (overly conservative).

(c) Reactor Coolant Pump fragility was based on consideration of the failure of the attached CCW piping, due to an assumption that a small-break/RCP seal LOCA was critical. It was learned during the Peer Review that failure in the system model was linked to a large-break LOCA, so the failure mode considered in the fragility evaluation is not consistent with the system model - SNC indicated that they will revise the fragility evaluation.

E-34 Suggested Finding Disposition for 50.69 Resolution and other applications Coolant Pump fragility has been updated to reflect the failure of the pump associated with LOCA; the reactor internals fragility has been updated in the calculation; and the new fragilities have been reflected in the updated SPRA model.

I This finding has been I

resolved.

I

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s)

Finding Description Finding Suggested Finding Disposition for 50.69 Basis Resolution and other applications (d) Reactor Internal fragility evaluation determined the demand based an average spectral acceleration over the range of 2 to 3 Hz, rather than using the peak acceleration in this range of the ISRS, and did not consider the contribution of higher modes. SNC indicated that this was done to avoid an overly conservative capacity, but agreed that the contribution of higher modes should be addressed, and will revise the calculation.

(f) Control Rod Drive Mechanism fragility evaluation assumed that material stresses were the critical failure mode, and did not address the potential impact of deflections on rod drop. SNC indicated that information provided by Westinghouse (based on a Japanese testing program) indicated that the deflection levels associated with seismic loading does not impact rod drop, and agree to add this discussion to the calculation.

E-35

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s) 14-20 SFR-E4, Met SPR-89 Met 14-4 SFR-01 Met Finding Description Seismic induced fire evaluations are not documented in the walkdown report or fragility calculations.

(This F&O originated from SR SFR-E4)

A potential for sloshing induced inundation of the NSCW Pumps (11202P4007, 11202P408) and associated discharge motor operated valves (1HV11600, 11606,11607, 11613) in the NSCW exists and was not identified either in the walkdowns or subsequent analysis.

(This F&O originated from SR SFR-01)

Finding Basis The only mention for seismic induced fire evaluation is contained in the quantification notebook. Based on discussions during the peer review, it is understood that seismic induced fire was a key consideration during the walkdowns. However, detail of the walkdown procedure for fire following earthquake is missing. The write up should include team composition, methodology, screening criteria, and results, SFR-01 requires that realistic failure modes of structures and equipment that interfere with the operation of that equipment be identified.

The potential for earthquake induced sloshing of the water within the NSCW tower exists.

From field walkdowns of the NSCW it was observed that there is a potential for sloshing of contents to potentially splash onto or flood the pumps and or motor operated valves on the attached discharge piping.

E-36 Suggested Finding Disposition for 50.69 I

Resolution and other applications Seismic induced fire is an The seismic-induced fire important element of the and flood evaluations fragility evaluation process and have been updated, and this should be clearly documented in the documented.

fragility and quantification report. This includes the details of the walkdown procedure used to evaluate the potential for seismically induced fires, including the methodology, screening criteria and results.

This finding has been resolved with no I

significant impact to the SPRA results or conclusions.

Evaluate the potential for flood The evaluation for induced failure of the NSCW potential flood induced Pumps or NSCW discharge failure of the NSCW MOVs.

pumps or the NSCW discharge MOVs has been performed and documented in the fragility calculation for the NSCW tower. There was no significant impact on the pump or MOV fragilities.

This finding has been resolved with no significant impact to the SPRA results or conclusions.

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Number Require-cc ment(s) 14-5 SFR-D1 Met 14-6 SFR-G2 Met Finding Description The potential for seismically-induced differential settlements between structures was not addressed.

(This F&O originated from SR SFR-D1)

The results of the seismic gap/shake space walkdowns are not documented.

Finding Basis Vogtle 1 &2 is a soil site, with engineered fill from the rock interface to the finished grade.

The in-scope Seismic Category I structures have foundations with varying embedment depths, ranging from surface founded (elev. 220ft.) to a foundation embedment of 11 0 ft. ( elev. 11 0 ft.). Since soils, including engineered fill, will consolidate/settle to some extent when subjected to high level earthquake ground motion, and the amount of settlement is proportional to the thickness of the soil layer under the foundation, the settlement of one structure relative to another structure is dependent on the depth of the foundation embedment.

The Fragility Notebook (PRA-BC-V-14-025) does not address the potential differential settlement between buildings, or the potential effect on commodities (e.g.,

piping, electrical raceways, HVAC ducts, etc.) that cross the separation between adjacent structures. During the performance of the Peer Review, SNC personnel indicated that the consideration of differential settlements was not required, since the structures were founded on engineered fill.

The walkdown guidance provided in Appendix F (Checklists and Walkdown Data Sheets) of EPRI NP-6041 includes attributes of E-37 Suggested Finding Disposition for 50.69 I

Resolution and other applications 1

Develop estimates of the Documentation has been differential settlements updated to include the between adjacent structures effects of earthquake and assess the fragility of induced settlement; no commodities based on their significant differential ability to accommodate the settlements were associated differential computed between the displacements.

structures.

This finding has been resolved with no significant impact to the SPRA results or conclusions.

Provide documentation of the As noted in the Finding results of the seismic gap basis, inspection of the walkdowns.

seismic gaps was included in the seismic

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s)

Finding Description (This F&O originated from SR SFR-G2)

Finding Basis seismic gaps between structures which should be addressed in the performance of the walkdowns.

These include the clearance between adjacent structures and the ability of any subsystems (e.g., piping, cable trays, HVAC ducts) spanning the gap to accommodate the differential seismic displacements.

The Seismic Walkdown Report (PRA-BC-V-14-005) does not include documentation of the results/findings/observations associated with the inspection of the seismic gaps between structures or the subsystems spanning the gap. During the performance of the Peer Review, SNC personal indicated that inspection of the seismic gaps was included in the seismic walkdowns, but not explicitly described in the report. The ability of components to accommodate potential differential movement at the building separations is implied in the discussion of rugged components (piping, cable trays, and HVAC ducts) in Section 2.1 (Rationale for Screening) of the report. In addition, information from the Vogtle IPEEE Report (page 3.1-

37) indicated that the seismic gaps had been inspected during the IPEEE.

E-38 Suggested Finding Disposition for 50.69 Resolution and other applications walkdowns. Piping across seismic gaps is designed with adequate flexibility to accommodate building motions, and pipe sleeves provide adequate gaps for piping movement. The documentation has been updated to reflect the inspections performed during the walkdowns.

This finding has been resolved with no significant impact to the SPRA results or conclusions.

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment{s) 14-7 SFR-A2, I,

SFR-F4 Met Finding Description The fragility evaluation for the Containment Polar Crane (in fragility notebook) did not address the impact of variation in the fundamental frequency on the applicable seismic demand.

(This F&O originated from SR SFR-A2)

Finding Suggested Finding Disposition for 50.69 Basis Resolution and other applications The determination of the Update the fragility evaluation The fragility evaluation of fundamental frequency of for the polar crane to address the polar crane has been structures and components potential uncertainty in the updated to address involves a certain degree of fundamental frequency and the potential uncertainty in uncertainty. This uncertainty must contribution of higher modes.

the fundamental be accounted for in the frequency and determination of the seismic contribution of higher accelerations from the applicable modes.

in-structure response spectra (ISRS).

This finding has been resolved.

Section 7.4 (Vogtle 1 and 2 Polar Crane) of the Fragility Notebook (SNC calculation no. PRA-BC-V-14-025) evaluates the polar crane as a potential seismic interaction source relative to the reactor vessel and other NSSS components inside the containment structure. In the determination of the vertical spectral acceleration applicable to the polar crane, the computed fundamental frequency falls within a valley in the applicable ISRS, on the low frequency side of the primary spectral peak.

Uncertainty in the calculated frequency, and the contribution of high modes, could result in an increase in the applied vertical acceleration. During the performance of the Peer Review, SNC personnel provided a written response indicating that it is appropriate to increase the applied acceleration by 50%,

which will result in a 20%

I decrease in the median capacity of the polar crane.

E-39

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s) 14-8 SFR-F3 IIIII I Finding Description Relay fragility calculations include conservative assumptions.

(This F&O originated from SA SFR-F3)

Finding Suggested Finding Disposition for 50.69 Basis Resolution and other applications The relay evaluation for the Perform more realistic relay The relay fragilities have turbine driven auxiliary feedwater fragility evaluations.

been updated using the pump control panel in calculation appropriate response PRA-BC-V-14-008 is based on a and in-cabinet generic capacity for motor starters amplification factors, and and contactors (intended for motor are realistic.

control centers) and an amplification factor associated This finding has been with center of door panel resolved.

response. Based on walkdown observations the relay is not mounted on the door panel so is likely on an internal bracket. The median capacity of 0.627g is well below the screening level and is not realistic.

The relay evaluations in calculation PRA-BC-V-14-009 are governed by response in the vertical direction, and the in-cabinet amplification factors used in the calculation are associated with horizontal response. The resulting median capacities of 0.762g (Appendix M1) and 1.026g (Appendix M2) are well below the screening level and are not realistic.

E-40

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s) 14-9 SFR-D2 Met Finding Description The seismic walkdown report includes a number of open items that are not are not traceable to a resolution (This F&O originated from SR SFR-D2)

Finding Basis The summary of the seismic walkdowns documents a number of issues identified during the performance of the walkdowns that required follow-up actions (31). These include spatial interaction issues, housekeeping issues, anchorage issues, valves having configurations that do not meet the EPRI guidelines, configuration issues, installation errors, etc.

The Seismic Walkdown Report (PRA-BC-V-14-005) does not document how the issues identified during the walkdowns have been addressed, either in the field (e.g., correction of installation errors, resolution of housekeeping issues) or in the fragility evaluations (e.g., valve configurations, anchorage issues).

During the performance of the Peer Review, the Peer Review Team provided a list of the walkdown issues to SNC personnel, and SNC provided a summary of how they were addressed. Most issues had been adequately addressed during the development of the SPRA, but it was determined that the following would require further effort for resolution:

(a) Potential interaction between piping and deluge valve (page 19)

- follow-up walkdowns required.

(b) Anchorage configuration on inverter (page 40) - follow-up revision to fragilit'r' evaluation

~

E-41 Suggested Finding Disposition for 50.69 Resolution and other applications Perform resolution of open The noted walkdown items and provide issues have been documentation of the evaluated and reflected resolution associated with in the revised each of the issues, either in documentation:

the Fragility Notebook or the

- potential piping SPRA Database.

interaction;

- the difference in inverter anchorage configuration;

- potential interaction concerns with the overhead heater; this evaluation is in the fragility notebook in section 3.4.2.

Valve operator heights &

weights that were outside EPRI guidelines have been taken into account in the fragility analysis for these components.

The Diesel Generator Exhaust Silencer was re-evaluated to the as-operated condition.

The fragility analysis for these components has been completed for the as built condition.

This finding has been resolved.

-~

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s) 16-1 SFR-F3,

111111, SPR-84, Not SPR-E5

Met, II Finding Description The model presented for peer review did not incorporate the effects of relay chatter as the analysis was not yet complete.

(This F&O originated from SR SPR-84)

Finding Suggested Finding Disposition for 50.69 Basis Resolution and other applications required (c) Overhead heater poses potential interaction issue (page

60) - follow-up walkdown required.

(d) Valve operator heights/weights outside of EPRI guidelines (page 74)- follow-up walkdown required.

(e) Diesel Generator exhaust silencer anchor bolt nuts (page

96) - not addressed in fragility evaluation, further evaluation required.

(f) Valve operator heights outside of EPRI guidelines and potential lack of yoke support (page 1 05) -

these valves are part of the unfinished scope described in the Fragility Notebook, which will be completed in the future.

I (g) Valve operator heights outside of EPRI guidelines (page 1 07) -

further evaluation required.

Relay chatter is consistently being Complete the analysis and The approach to observed as a significant incorporate the effects of relay screening and modeling contributor to risk profile in chatter and similar devices in of seismically-induced recently peer reviewed S-PRAs the PRA logic model.

relay failures and chatter and it is therefore realistic to was provided to the peer expect that relay chatter is a review team and potential significant contributor.

determined to have been During the peer review it was performed appropriately; discussed that the SPRA team only the incorporation does not believe relays will be a into the model of the significant contributors but it was impacts of relay chatter also said that this conclusion/

from unscreened relays expectation is based on potentially was not complete. The crediting operator actions. Thus, final screening resulted the effects of relay chatter per se in only 2 relays being may be significant (and provide incorporated into the some insights) while the model, with one having an operator action. Relay_

E-42

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s) 16-10 SPR-86 Met Finding Finding Description Basis combination of relays and a number of HEP may not be.

The documentation about There is only a short sentence the walkdowns in support to supporting the discussion on seismic impact on HRA alternative access pathways.

appear limited.

(This F&O originated from SR SPR-86)

E-43 Suggested Finding Disposition for 50.69 Resolution and other applications chatter fragilities and impacts have been incorporated into the seismic model, in a manner consistent with that used for other failures.

This finding has been resolved.

More detailed documentation Walkdown is suggested to support the documentation on conclusion on accessibility, accessibility for operator alternative route, availability of actions, including photos, tools/keys, clear identification has been improved.

of equipment manipulated in Potential failure of block each local action.

walls has been reviewed and documented.

Obviously, the goal of the Required tools and enhanced documentation is equipment, such as not to convince the peer ladders, have been reviewer that the walkdowns identified with locations were performed but rather to when needed. The ensure that the analyst is fully documentation supports convinced of the conclusions.

the seismic HRA assumptions and Past SPRAs have shown modeling.

examples of equipment needed for the HFE that was This finding has been not in the SEL, or that has resolved with no different actuators when significant impact to the manually actuated, or that SPRA results or needed ladders that were not conclusions.

easily accessible or that were close to block walls (or under ceiling that could collapse) that were not considered an issue because the block walls were not near safety related equipment (and therefore not addressed in the rest of the

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s) 16-11 SPR-E2 Met 16-12 SPR-E2 Met Finding Description Missing review of the potential for additional dependencies introduced by the SPRA models (QU-C1&2)

(This F&O originated from SR SPR-E2)

Missing documentation of the review of non significant cutsets QU-05.

(This F&O originated from SR SPR-E2)

Finding Basis It is understood that the investigation performed in internal events to identify potential H FE dependency has been relied upon in the Vogtle SPRA.

The SPRA logic may identify additional dependencies trends that were not identified in the internal events.

It is an industry expectation (as discussed in NEI peer review task force meetings) that review of the non significant cutsets is explicitly documented.

Based on discussion during the peer review, two reviews were performed to validate the overall model and cutsets. The first was a random review of cutsets at midpoints and low significance for each of the %Gxx initiators to verify that the cutsets are valid cutsets, and that the patterns are appropriate. That is, if one cutset E-44 Suggested Finding Disposition for 50.69 Resolution and other applications SPRA work). In this perspective, a more systematic documentation of the feasibility and accessibility analysis for each of the HFE credited in the SPRA is suggested.

As this exercise was A detailed quantitative apparently performed for the HRA dependency Fire PRA (as discussed during analysis based on using the peer review), it is the HRA calculator was suggested that a review of the performed and potential for unforeseen documented. There was dependencies trends is no significant impact on performed.

results since human actions are not significant As it is understood that the contributors in the Vogtle plan is to transition to a SPRA.

different dependency analysis method (based on HRA This finding has been calculator), this may be resolved.

addressed within the same transition as it is realistic to expect that not too many (if any) new dependencies would be identified.

It is understood that the SPRA The QU report has been documentation will be revised updated to document the to incorporate explicitly the two review of both dominant reviews discussed in the basis cutsets and non-for this F&O. It is also significant cutsets for recommended to document both CDF and LEAF.

the review of cutsets following guidance from the NEI peer This finding has been review task force.

resolved with no significant impact to the SPRA results or conclusions.

i

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s) 16-15 SPA-E6,

Met, SPA-F2 Met Finding Description Documentation of LEAF model applicability review.

(This F&O originated from SA SPA-F2)

Finding Basis is valid, then another cutset with slightly different seismic failures (or random failures) should also be nearby.

The second review, more importantly, lowered the median seismic capacity for each of the seismic initiators and some of the other seismic failures to ensure that the model would properly generate valid cutsets. For example, the LLOCA fragility was reduced to 0.5g to generate LLOCA cutsets. For ATWT, the fragility of the CADs and AV internals were reduced to 0.5g to verify that valid A TWT cutsets were generated.

The current documentation does not explain what are the basis for retaining the LEAF logic and analysis unchanged within the SPAA logic.

During the peer review the following explanation was provided by the SPAA team:

"The internal events Level 2 notebook (Chapter 9) was reviewed to ensure that the definition of LEAF would be appropriate for seismic events.

Section 9.2 provides the LEAF definition, including the use of a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> time period for release after event initiation, to allow for evacuation. This time period is considered to be valid for Vogtle seismic events, particularly due to E-45 Suggested Finding Disposition for 50.69 Resolution and other applications Expand the documentation to The LEAF ensure that the criteria used to documentation in the QU retain the LEAF analysis in the report was expanded to SPAA is explained so that the describe the review of same applicability review can applicability of the be performed following future internal events PAA potential revisions of the LEAF LEAF analysis to the modeling.

seismic PAA.

This finding has been resolved with no significant impact to the SPAA results or conclusions.

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s) 16-18 SPR-88 Ill Finding Description Very small LOCA have been screened from the analysis based on walkdowns but little documentation exists of such walkdowns.

(This F&O originated from SR SPR-88)

Finding Basis the very low population density in the area. Other characteristics, such as bypass and scrubbing, are the same for seismic as for internal events.

The logic for the internal events LERF model is very straightforward, with sequences from the CDF model ANDed with the appropriate LERF fault tree.

This logic is also appropriate for seismic events."

The DB has a specific entry for the incore thermocouples and provides pictures of them. Still, in-core thermocouple tubing is not the only possible source of very small LOCA that is envisioned and the only documentation of addressing the other potential sources is in section 2.3.3 of the quantification notebook:

"For Vogtle 1 &2, the seismic walkdowns inspected and photographed a large sample of the small piping and tubing lines connected to the primary system in order to identify any weaknesses. The piping was iudQed to be ruQQed."

E-46 Suggested Finding Disposition for 50.69 Resolution and other applications I

To the peer review team Additional information on knowledge Vogtle is the only the walkdown for very plant that has elected to small LOCA has been I

perform dedicated walkdowns added to fragility report I

in support of not modeling very to provide the basis for small LOCA. This would be a the VSLOCA screening.

best practice but it also behooves to the SPRA team to This finding has been provide detailed resolved with no documentation of such significant impact to the walkdowns and how they SPRA results or supported a systematic conclusions.

evaluation of the potential sources of very small LOCA.

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s}

16-2 SFR-C1, 1/11, SPR-E1 Met 16-4 SPR-82 Not Met Finding Description Fragilities were not corrected to reflect the 2014 hazard used for quantification. (This F&O originated from SR SPR-E1)

The effect of seismic impact on performance shaping factors is considered in the analysis by the usage of the Surry method.

(This F&O originated from SR SPR-82)

Finding Basis The 2014 hazard was only used as input to FRAN X for the final quantification. It is understood that the fragility estimates have been performed based on the 2012 hazard. While it is not expected nor recommended to regenerate all the fragility work with the new hazard, some consideration on the possible change in fragility due to the use of the newer hazard should be made.

There is no assessment of the effect of changing the breaking points in the Surry method. The Surry method is based on methods used in the past at SONGS and Diablo Canyon and the O.Bg breaking point was developed for California earthquakes. In the Vogtle analysis there is no indications on whether the breaking point at O.Bg is also applicable to Vogtle. There are also no sensitivity analyses that would support whether a change in the breaking points is significant or not.

E-47 Suggested Finding Disposition for 50.69 Resolution and other applications During the peer review the The fragilities have been SNC staff answered a question recalculated based on on this topic by performing an the 2014 hazard [13] and initial limited investigation of the new values the effect on fragilities incorporated into the correction to reflect the 2014 SPRA model and hazard and concluded that the quantification.

effect of this scaling is not insignificant (especially for This finding has been LEAF). It is recommended to resolved.

continue and expand this investigation to make the quantification fully consistent with the fragility values.

While it is recognized that the The methodology used industry is still developing for the seismic HRA methods in support to this analysis is based on particular topic (e.g., recently defining PSFs as a published EPRI HRA method function of seismic for external events), some hazard level (bins), which additional considerations is consistent with the should be done to understand EPRI seismic HRA the effect of HEPs in the model guidance in EPRI rather than simply 3002008093 [14]. The implementing the Surry Integrated PSFs and bins method as is.

(breaking points) have been updated with Three examples for addressing additional breaking this finding may be the points and integrated following:

PSFs to reflect seismic 1. Perform sensitivities on the binning applicable to values of the multipliers and Vogtle, in accordance the g levels where the with this finding and breaking point happens.

consistent with the EPRI

2. Use a different multipliers guidance. The updated method with more breaking values have been points.

applied to both internal

3. Apply the impact of seismic events H FEs and specific PSF at the individual seismic-unique HFEs PSF level (i.e., timing, stress, within the plant response

~c.) in the HRA calculator.

model.

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s) 16-5 SPR-81,

Met, SPR-F1 Not Met Finding Description LOCA modeling and fragility selection not clearly documented.

(This F&O originated from SA SPR-F1)

Finding Basis The selection of the fragility data used for all LOCA is discussed in Appendix 8.2 of the quantification notebook but is confusing in the mapping of selected fragilities with specific failures.

It appears that the fragility selected to represent LOCA sequences are coming from specific components but then they are used to represents sort of surrogate events for potential failures along the piping network.

Using localized events as surrogate for pipe network failure is probably conservative and may not be fully consistent with the system success criteria and modeling in the internal events modeling. For example, the seismic-induced MLOCA fragility seems to be based on failure of the pressurizer surge line, which is a localized failure. The seismic-induced MLOCA initiator is mapped to the internal events MLOCA initiator. The internal events logic for MLOCA has a split fraction that divides MLOCA (and LLOCA) in four 25%

contributors impacting all four CUHL. Since the seismic-induced MLOCA is a localized failure, the E-48 Suggested Finding Disposition for 50.69 I

Resolution and other applications There was no significant impact on the SPRA results.

This finding has been resolved.

Documentation on the use of LOCA basis has been re-fragility in support to LOCA evaluated and updated.

should be clarified to better This was partially due to represent the rationale seismic fragility update selected and potentially and partially a matter of addresses the modeling adding amplifying uncertainties associated with information to the LOCA this selection.

basis. The quantification report includes updated While this finding is expected documentation. Although to be addressed via LOCAs are a significant documentation, some contributor to the SPRA additional suggestions are results, the VEGP SCOF provided, such as:

and SLERF are sufficiently small that

1. Perform a sensitivity to further LOCA modeling show that the modeling sensitivity beyond what approach described is not has been provided in the significantly skew the results updated model for seismic; quantification is not warranted.

2. Modify the logic by mapping the seismic-induced MLOCA to This finding has been a different position in the logic resolved.

(e.g., a dummy event can be entered in the model to provide a target for the FRANX injection).

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s) 16-6 SPR-82 Not Met Finding Description The effect of seismic impact on performance shaping factors is not considered for any action that was explicitly added for the SPRA (e.g.,

flood isolation or DG output breaker closure).

(This F&O originated from SR SPR-82)

Finding Basis internal events logic is not fully applicable (probably slightly conservative).

Because the documentation is potentially leading to a misunderstanding of the selected approach (thus impacting ease on update), this F&O is considered a finding against the documentation SR.

The Vogtle SPRA elected to use Integrated Performance Shaping Factors (IPSF) multipliers. While this approach was used for the HEPs that were carried over from internal events, it was systematically not done for all the actions explicitly added for seismic.

Based on discussion during the peer review, the analyst believed that having designed these actions for specific scenarios following a seismic event, the impact of seismic specific PSF is already included.

The objection to this conclusion is that the seismic specific PSF should realistically change with the magnitude of the event. This change addresses the change in the overall context of the plant when a small seismic event happens as opposed to when a very large seismic event happens.

This seems not to be captured by the approach selected for the Vogtle SPRA. One example of this E-49 Suggested Finding Disposition for 50.69 Resolution and other applications Expand the IPSF approach to The methodology used all the operator actions for the seismic HRA credited in the SPRA.

analysis is based on defining PSFs as a function of seismic hazard level (bins), which is consistent with the EPRI seismic HRA guidance in EPRI 3002008093 [14). The Integrated PSFs and bins (breaking points) have been updated to reflect seismic binning applicable to Vogtle, in accordance with this finding and consistent with the EPRI guidance. The updated values have been applied to both internal events HFEs and seismic-unique HFEs within the plant response model.

There was no significant impact on the SPRA results.

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s) 16-7 SPR-E2 Met 16-8 SPR-E2 Met Finding Description Base case seismic LEAF does not meet the truncation requirements from QU-B3.

(This F&O originated from SR SPR-E2)

Missing documentation of cutsets review (cfr. QU-01)

(This F&O originated from SR SPR-E2)

Finding Basis is that an action that has a 30 minute Tsw (S-OA-BKR-LOCAL) maintains an HEP of 1.60E-03 at all g levels, including the %G14 interval (i.e., >2g}.

It is understood that this is not expected to be quantitatively significant because failure of the recovered equipment is taken care by the loQic model.

Both CDF and LEAF are truncated at 1.0E-09 with 1 000 cutsets managed by ACUBE. This meets the QU-B3 requirement for CDF but not for LEAF.

Section 3.1 is the only description of the most important scenarios but there is no cutset-by-cutset review.

E-50 Suggested Finding Disposition for 50.69 Resolution and other applications This finding has been resolved.

LEAF at 1 E-11 truncation LEAF truncation, which meets the QU-B3 truncation was already considered requirement. Rename LEAF at in sensitivity studies, has 1 E-11 as the base case for been revised LEAF.

appropriately to meet QU-B3. A new LEAF truncation limit has been established consistent with the LEAF results.

Quantification is at I

1 E-12, which is a suitably low value.

This finding has been resolved.

While it is understood that the The QU report has been Draft. B version of the updated to document the quantification notebook is still review of both dominant somewhat a work in process, it cutsets and non-is expected that when the significant cutsets for model reaches a more stable both CDF and LEAF.

state documentation of the review of the cutsets is going This finding has been to be part of the resolved with no documentation.

significant impact to the SPRA results or conclusions.

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Number Require-cc ment(s) 16-9 SPR-81,

Met, SPR-84b Met 17-1 SPR-81 Met Finding Description Screening values used for the HEPs that (at the time of the provided documentation) were in the most significant cutsets.

(This F&O originated from SR SPR-81)

The documentation does not specifically address the applicability of the internal events accident sequences and success criteria to the SPRA model, and does not properly document the accident sequences created specifically for the SPRA model.

(This F&O originated from SR SPR-81)

Finding Basis At the time when the documentation was provided for peer review, the most significant operator actions (i.e., flood isolation of ACCW HX) were all screening values, which would only meet CCI for HR-G1 (directly called through SPR-81 ).

In addition, there is little documentation or supporting evidence to justify screening values as low as 3.00E-2 The modeling approach injected seismic fragilities into fault trees that were modified from the internal events PRA model. It can be inferred from this approach, and it was verified by discussions with the staff, that the internal events sequences and success criteria were considered to be applicable to the SPRA model.

This was not specifically stated in the documentation.

Further, several additional seismic flooding sequences were added to the fault tree. These sequences are not discussed from an accident sequence and success _

E-51 Suggested Finding Disposition for 50.69 I

Resolution and other applications An appropriate resolution of The seismic HRA this F&O is pending the current analysis has been evolution of the model and the revised to be consistent I

importance of operator actions with the EPRI seismic in the SPRA. Given the HRA guidance in EPRI expectation that operator 3002008093 [14]. The actions will be needed to original screening HEPs mitigate the importance of have been updated using relay chatter (not yet included the HRA Calculator, in the SPRA logic model) this consistent with the F&O was provided to ensure approach used in the care is used in the generation VEGP internal events of HEPs if they appear in PRA. The important cutsets and also to Documentation has been provide more justification for updated. Operator screening values less than response to relay chatter 1.00E-1 because a low has been addressed and screening value may indeed evaluated within the skew the actual importance of same process, and not the newly generated HEP.

found to be important.

This finding has been resolved.

A separate section in the The discussion of documentation that specifically accident sequences and addresses accident sequences success criteria has been and success criteria is needed expanded, and specific to collect the information in descriptions of the one logical place, and is flooding scenarios has needed to support effective been added. This finding peer reviews and future model is documentation only updates.

and does not impact Seismic PRA model results.

This finding has been resolved with no significant impact to the SPRA results or conclusions.

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number mentes) 17-2 SPR-E2,

Met, SPR-F2 Met Finding Description The processes used to create the presented quantification results are not fully documented.

(This F&O originated from SR SPR-F2)

Finding Basis criteria perspective. Inspection of the fault tree and discussions with the staff indicate that the sequences were appropriately developed with specific success criteria that is different from other internal events sequences. The development of these sequences needs to be included in the documentation. Including event trees for these sequences would also aid in a reader's understandinQ.

Examples include:

The top cutsets shown in table 3-1 of the quantification report are produced by combining the cutsets from all the seismic interval cutsets in a process that is not documented.

While the process used to obtain the importance measures in section 5.2 of the quantification notebook is documented in that section, discussions with the PRA staff indicated that importances for some of the basic events were obtained in a different manner (setting to one or zero and requantifying). This is not documented in the notebook.

E-52 Suggested Finding Disposition for 50.69 Resolution and other applications Expand the documentation to Documentation for QU clearly explain the post-results has been im-processing of the results proved to describe the generated by CAFT A and processes used to ag-FRANX. Examples include:

gregate results over the 14 hazard intervals. The

- Explain how the cutsets importance calculations generated by FRANX are have been re-quantified combined into g-level-and the method for independent cutsets.

presentation documented.

- Explain the post-processing used to generate importance This finding has been measures, especially focusing resolved with no on the deviation from a normal significant impact to the practice that is currently only SPRA results or mentioned in the notebook.

conclusions.

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s) 17-3 SPR-83, 1/11, SPR-E4 1/11 Finding Description Subdividing correlation groups based on weaker/stronger components resulted in retention of non-minimal cutsets in some cases, which could impact CDF/LERF results as well as model importance measures. The magnitude and acceptability of these impacts was not documented.

(This F&O originated from SA SPR-E4)

Finding Suggested Finding Disposition for 50.69 Basis Resolution and other applications To account for similar equipment The impact of the retention of The non-minimal cutsets that has different fragilities due to these non-minimal cutsets on in the peer reviewed different building locations, certain CDF/LERF and importance model were identified correlation groups were measures should be assessed and reviewed for impact, subdivided to assign a seismic and the results documented, or and determined to be capacity to a weaker component a method to remove the non-non-significant to risk.

that only failed that component.

minimal cutsets should be The results were very The higher capacity was then devised. Each subdivided slightly conservative due assigned to both components, and correlation group should be to these non-minimal was effectively the correlated investigated for similar effects.

cutsets. The issue has failure of both components. This been addressed in the can result in the retention of non-updated model, such minimal cutsets in some cases.

non-minimal cutsets no For example, for the Containment longer appear.

Fan Cooler Units there are cutsets in which, due to other failures, This finding has been only one containment fan cooler resolved with no needs to seismically fail to cause significant impact to the core damage. Inspection of the SPRA results or cutsets shows that two otherwise conclusions.

identical cutsets are retained: one in which the 1 Fan 'group' occurs, and one in which the 4Fans group occurs. The 4Fans cutset is not minimal, and should not be included in the results.

Discussions with the staff indicated that these non minimal cutsets were noted during the quantification review process, but were thought to not greatly impact overall results. No formal assessment was done, however, and no record of the informal assessment was included in the documentation.

E-53

Enclosure to NL-16-2382 Basis of the Proposed Change Finding Supporting Require-cc Number ment(s) 17-4 SPR-E6 Met Finding Description No quantitative analysis of the relative contribution to LERF from Plant Damage States and Significant LERF contributors from Table 2-2.8-9 was presented in the quantification results.

(This F&O originated from SR SPR-E6)

Finding Suggested Finding Disposition for 50.69 Basis Resolution and other applications A quantitative analysis is required Perform the analysis and The quantitative analysis to meet CCII for LE-F1 & LE-G3, include the results in the of significant LERF plant which are directly called from quantification notebook.

damage states and SPR-E6.

contributors has been performed. A table and associated discussion of plant damage states and significant contributors has been added to the LERF QU documentation i to resolve this finding.

i This finding has been resolved.

E-54