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

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License Amendment Request to Modify Approved 10 CFR 50.69 Categorization Process
ML17173A875
Person / Time
Site: Vogtle  Southern Nuclear icon.png
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)
v • d • e


Text

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Bmmngh.un. AI  :\5242 205 992 5S72 tel 205 992 760 I fax jjhutlo(n*.,outhcmco wm JUN Z 2 2011 Docket Nos.: 50-424 NL-16-2382 50-425 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Vogtle Electric Generating Plant- Units 1 and 2 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,

. I rc_:;>

J. J. Hutto '

J Regulatory Affairs Director )

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Sworn to and subscribed before me this 2..2_ day of ~ '2017.

~2/~t Notary Public 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

Vogtle Electric Generating Plant- Units 1 and 2 License Amendment Request to Modify Approved 10 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 10 of the Code of Federal Regulations (1 0 CFR}, Part 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs} for Nuclear Power 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-16-2382 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 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance, alternative treatment requirements can be implemented in accordance with the regulation. For equipment determined to be of high safety significance, requirements will not be changed or will be enhanced. This allows improved focus on equipment that has safety significance resulting in improved plant safety.

The rule contains requirements on how a licensee categorizes SSCs using a risk-informed process, adjusts treatment requirements consistent with the relative significance of the 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 NL-16-2382 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 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) specified in the 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-16-2382 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 10 numbered topics in that attachment:

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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.

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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 10 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).

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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.

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-16-2382 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 10 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 NL-16-2382 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 10 of the Code of Federal Regulations (1 0 CFR), Part 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power 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.

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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 NL-16-2382 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 Attachment 1 : Seismic PRA Model Summary Information Unit Model Baseline CDF Baseline LERF Comments one model 1 VEGP-SPAA-1, 2.8E-6/yr 3.3E-7/yr applicable to 12/31 /16 both units VEGP-SPAA-1, one model 2 2.8E-6/yr 3.3E-7/yr 12/31 /16 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 1E-05/year, and therefore meets AG 1.174 total risk criteria.

E-15

Enclosure to NL 2382 Basis of the Proposed Change Attachment 2: Disposition and Resolution of SPRA Peer Review Findings 1 Supporting Finding Finding Suggested Finding Finding Number Require- cc Description Basis Resolution Disposition for 50.69 and other applications ment(s) 11-3 SHA-E2 IIIII I While variability in the mean To maintain hazard-consistent Expand documentation to There is an abundance base-case Vs profile is ground motion hazard at the demonstrate that a single of site-specific Vs data incorporated in the site control point, the site response base-case Vs profile from VEGP Units 3&4, response analysis, no analysis needs to incorporate adequately represents the which reduces epistemic epistemic uncertainty in the appropriate epistemic uncertainty Units 1&2 site. Or if that is not uncertainty to an base-case profile is and aleatory variability in its the case, include epistemic insignificant level.

represented. inputs. The Vs profile for the uncertainty in the Documentation of the Vogtle Units 1&2 site is characterization of Vs profile Additional discussion of justification for this represented by a single Vs profile, and evaluate the impact on the rationale for use of a assessment should be indicating there is no epistemic control point ground motions. single base-case Vs expanded. uncertainty in the mean base-case profile for the site has profile. Documentation of this been included in the (This F&O originated from assessment needs to be documentation. The SA SHA-E2) expanded. added discussion demonstrates that a Discussion with staff indicates that single base-case shear-consideration of the combined wave velocity (Vs) profile data for the Vogtle site (Units 1&2, adequately represents Units 3&4, ISFSI) provides - the Vogtle site, based on sufficient confidence that a single the availability of Vs mean base-case profile data, which reduces the characterizes the site. This epistemic uncertainty for conclusion is based on the this particular parameter.

quantity and quality of the combined data and an evaluation This finding has been showing the site is relatively resolved with no uniform with respect to Vs. For significant impact to the some depth ranges, data from the SPRA results or nearby Savannah River Site conclusions.

(SRS) are used to support the profile interpretation.

Bechtel Document 23162-000-G65-GEK-0001 0 (SNC #SVO-GB-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.

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Disposition for 50.69 Number Require- cc Description Basis Resolution and other applications ment(s)

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.]

11-8 SHA-E2 II/III Upper crustal site Calculation X2CFS129 Ver2 notes Provide a basis in the A discussion of the range attenuation of ground motion that the damping associated with documentation for of possible values of (kappa) is, generally, an the base-case profile corresponds representing base-case kappa deep soil damping has uncertain parameter. Thus, to a total kappa value for the soil at the site by a single value. been included in the to maintain hazard- column of 0.01 sec. The report The basis might include documentation.

consistent ground motion at does not address epistemic sensitivity analyses to show the control point, this uncertainty in kappa. the impact of epistemic A sensitivity study on the uncertainty should be uncertainty in kappa. epistemic uncertainty of incorporated in the site In discussion with staff during the deep soil damping has response analysis, or the peer review, it was noted that been performed using basis for not including it randomization of the damping median, lower range, and should be provided. In associated with the profile layers upper range alternatives either case, the technical represents both random variability for deep rock damping.

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

Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Number Require- cc Description Basis Resolution Disposition for 50.69 and other applications ment(s)

Vogtle site and uncertainties in motion. The resulting that small value would not be amplification functions (This F&O originated from expected to have a significant and log-standard SR SHA-E2) impact on site amplification. Staff deviation were weight-also noted that the approach used averaged and compared had been reviewed by the NRC to the original base case for the Vogtle ESP and COLA. for each of BBM High PI and BBM Low PI soil The SPID (EPRI, 2013) provides columns. It was guidance accepted by the NRC for concluded that the response to NTTF 2.1 inclusion of alternative Recommendation: Seismic that base cases for deep soil indicates kappa is difficult to damping to account measure and thus subject to large explicitly for the uncertainty (SPID Section B- - epistemic uncertainty 5.1.3.2). associated with site kappa does not have any Documentation of the technical significant effects on the basis for kappa characterization resulting seismic hazard should be expanded. 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.

12-1 SHA-J1 Not As part of the PSHA The approach that was taken to Documentation should be A PSHA report has been Met implementation, the analyst model earthquakes in the PSHA provided that describes how prepared that describes has different alternatives for calculation was not identified. seismic sources are modeled how earthquake events modeling the earthquake There are two basic alternatives in the PSHA (i.e., how the SSG were modeled for area occurrences in the that can be used to model and GMMs) were implemented sources in the PSHA calculations. The PSHA earthquake events; as extended in the Vogtle PSHA. calculations. This was documentation does not fault ruptures, or as point sources. by modeling each earth-describe the approach that The approach that is used quake as a point source, was used to model influences how the CEUS ground and using correction earthquakes. motion model is implemented. factors for distance and

- Qround motion uncertain-E-18

Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Disposition for 50.69 Finding Number Require- cc Description Basis Resolution and other applications ment(s)_

No documentation is provided on ty that modify the ground either of these subjects motion estimate to in- I (This F&O originated from (earthquake source modeling and elude the effect of a SA SHA-J1) use of the ground motion closer distance to a fault attenuation models). From rupture (because the questions posed to the PSHA rupture may be closer to analysts, it is our understanding the site than the single that earthquakes were modeled point used to represent as point sources and the that event) and the un-appropriate ground motion certainty in ground aleatory uncertainty was used in

- motion because the the calculation. 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.

12-11 SHA-J1 Not As part of the PSHA The PSHA analysts were asked to Provide a description of the A PSHA report has been Met implementation, the analyst describe the approach that was earthquake modeling approach prepared that describes has alternatives for modeling used to model earthquakes in the that was used to model the how pseudo-faults were the earthquake occurrences Charleston ALME seismic source. Charleston ALME seismic implemented to in the calculations. The The response indicated that source and how the approach represent the Charleston PSHA documentation does earthquakes in the Charleston was implemented. ALME source. This not describe the approach ALME source were modeled using includes: 1. A description that was used to model 'pseudo faults'. of the pseudo-faults. 2. A earthquakes in ALME definition of pseudo-sources. The PSHA report does not: faults as constructed

1. Describe that a 'pseudo fault' faults that represent approach was used to model possible sources of earthquakes in the Charleston future large earthquakes.

(This F&O originated from ALME source. 3. Implementation of the SA SHA-J1) 2. Provide a definition of 'pseudo pseudo-faults including faults'. - spacing and limits at the

3. Describe how the 'pseudo fault' borders of the Charleston approach was implemented for source. 4.Documentation the Charleston ALME seismic of the rupture area, source (e.g., what was the fault length, and width that E-19

Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Disposition for 50.69 Finding Number Require- cc Description Basis Resolution and other applications ment(s) spacing that was used; how was were estimated for the earthquake rate distributed to possible future the faults, etc.). earthquakes. 5. A

4. Document the fault rupture description of how model that was used. earthquake ruptures are
5. Describe how earthquake distributed on the faults.

events are distributed on the faults. This finding has been I resolved with no significant impact to the SPRA results or conclusions.

12-15 SHA-11, Not A screening assessment A screening analysis was not A screening analysis for other This evaluation was done SHA-12 Met was performed for soil performed for hazards such as seismic hazards should be for the Vogtle 3&4 COLA liquefaction and is described settlement, fault displacement, performed and documented as [9] and is noted in the in seismic fragility tsunami, seiche, etc. part of the PSHA and SPRA. ESP SAR [1 0]. The calculation (PRA-BC-V Vogtle 3&4 evaluation is 025). It is anticipated these other It is expected that information applicable to, and has seismic hazards will be screened in the FSAR for Vogtle 1 & 2 been cited in, the Vogtle A screening assessment out. and in the COLA for Units 3 & 1&2 SPRA Qualification was not performed for other 4 can be used to support this report.

potential seismic hazards. requirement.

This finding has been (This F&O originated from resolved with no SR SHA-11) significant impact to the SPRA results or conclusions.

12-16 SHA-J1 Not The Vogtle PSHA has gone The documentation of the PSHA Prepare a complete and up-to- A PSHA report has been Met through a number of is provided in a collection of date PSHA document that prepared that includes changes and revisions since documents that were prepared in includes all results, sensitivity hazard results, 2012 due to changes in the 2012-2014 time frame. There calculations, deaggregation uncertainties in hazard, models, input data, etc. As does not exist a single document results, etc. that is based on and sensitivities to input new calculations were that contains a set of results that the current model. uncertainties; this performed and reports is based on the current PSHA summarizes hazard generated, sensitivity model. results for the Vogtle site.

results, were not carried forward. As a result, there This finding has been does not exist a current resolved with no report that includes all PSHA significant impact to the results, deaggregations, etc. SPRA results or that is based on the current conclusions. - -

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Disposition for 50.69 Number Require- cc Description Basis Resolution and other applications ment(s)

PSHA model.

(This F&O originated from SA SHA-J1) 12-18 SHA-82, 1111 The Vogtle PSHA is based As part of a site-specific PSHA, A data gathering effort should A detailed study of new SHA-C4, Not on the CEUS SSC seismic there is a need to gather, review be undertaken to identify new geological, SHA-H1 Met source model which was and evaluate new geological, information that post-dates the seismological, and Not completed in 2012. The SSC seismological, or geophysical CEUS SSC data collection geophysical information Met model was a developed at a information or information that is effort. The data gathering effort was conducted, to regional scale that was defined at a scale that was not should also look for determine if any based on data gathered up considered in the development of information local to the Vogtle information subsequent until about 2010. (Note, the the CEUS sse model. As part of site region that was not to the EPAI sse model date when data was the Vogtle SPAA, no effort was considered, or at a scale that (EPAI, 2012 [11]) is gathered varied; for example made to gather up-to-date and was not addressed as part of available that should be the earthquake catalog was local (local to the Vogtle site) the CEUS SSC regional incorporated into the complete through 2008.) In information to evaluate whether evaluation. seismic hazard results the sense that the CEUS any new information has become for Vogtle. This study is sse model was not available on active faulting and/or Some of this information may described in the SPAA specifically performed as a the development new seismic be available in the COLA for documentation. While site-specific PSHA for the sources or the revision of sources Vogtle Units 3 & 4. the area around the site Vogtle site. in the CEUS SSC model in the continues to be studied vicinity of the Vogtle plant. by many earth scientists, (This F&O originated from there was no new SA SHA-82) Since up-to-data was not information identified that gathered, consideration of would change the alternatives could not be estimate of seismic addressed. hazard for Vogtle.

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

12-2 SHA-J1 Not The method that is used in For soil sites, the soil hazard is The documentation should The methodology used Met the Vogtle PSHA to estimate generally (though not exclusively, include a description of the for the surface hazard the soil site hazard is not since other methods could be methodology that is used to calculation has been described or referenced. used) determined in two steps; combine the rock hazard described in detail, and a probabilistic rock hazard results results and the site comparison made are estimated which are then amplification factors to between the GMAS combined with probabilistic determine the soil hazard at using the two (This F&O originated from estimates of the site response. the Vogtle site. approaches 2A and 3.

SA SHA..J1)_ The method used in the Vogtl~ Approach 2A was used_

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Disposition for 50.69 Finding Number Require- cc Description Basis Resolution and other applications ment(s)

PSHA to estimate the soil hazard for the calculation of SSI is not described. 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.

12-22 SHA-E2 IIIII I The site response The site response calculation A framework and approach for A description of the Calculation X2CFS129 Ver. does not present a clear evaluating and modeling methodology used to 1 (2012) and Ver. 2 (2014) description of how aleatory and uncertainties in the site account for epistemic does not describe a epistemic uncertainties are response should be developed and aleatory framework for evaluating identified and evaluated. As a and implemented. The site uncertainties in soil and characterizing sources result it is difficult to track the response calculation hazard has been added of aleatory and epistemic propagation of uncertainties is documentation should fully to the documentation.

uncertainty and how the carried out in the site response describe the methodology and approach was implemented. analysis. its implementation. This finding has been resolved with no (This F&O originated from It is worth noting that there is significant impact to the SR SHA-E2) some epistemic site response SPRA results or I conclusions.

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Disposition for 50.69 Number Require- cc Description Basis Resolution and other applications ment(s) uncertainty that is accounted for in the rock GMPEs.

12-23 SPA-E5 II The quantification process Table 5-1 presents the results of Develop and document an Additional detail has has included the three different uncertainty understanding of the earlier been added to the SPAA uncertainties in the seismic calculations for CDF and LEAF. In point estimate results for CDF Quantification report to hazard, fragility and addition, point estimates for CDF and LEAF (as reported in document the systems-analysis elements and LEAF are calculated and Sections 3 and 4) and of uncertainty, importance, of the SPAA. The results in reported in Section 5.1.1. Thus uncertainty results. and sensitivity analyses Table 5.1 are internally the table reports two estimates of and relate the uncertainty inconsistent and are the mean CDF and LEAF analysis mean CDF and inconsistent with the results respectively from different LEAF to the point reported in Sections 3 and 4 uncertainty calculations and a estimate values.

for CDF and LEAF, 'Point Estimates' result for each.

respectively. All of these results are different This finding has been than the point estimate resolved with no (This F&O originated from (approximate mean) reported in significant impact to the SA SPA-E5) Sections 3 and 4 for CDF and SPAA results or LEAF, respectively. The conclusions. I' documentation in the report does  !

not describe the basis (inputs) for these calculations, or offer an '

interpretation of the results.

12-24 SPA-E5 II The Quantification report The uncertainty analysis is Provide documentation of the Additional detail has I does not provide presented in Section 5.1 with the uncertainty analysis that been added to the documentation of the results reported in Table 5.1 . The describes the results, how they documentation of the uncertainty analysis results. report provides limited discussion are being interpreted and the seismic plant response of the results and the insights that insights that are derived from model, model might be gained from them. them. implementation, and quantification in the au (This F&O originated from The two sets of results that are report. In addition, the SA SPA-E5) reported in Table 5-1 are not uncertainty, importance, discussed in terms of their and sensitivity analyses relationship to each other. For are described in more E-23

Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Number Require- cc Description Basis Resolution Disposition for 50.69 and other applications ment(s) instance the mean values should detail.

be the same (but are not). The uncertainty estimates provide This finding has been insight to the total uncertainty and resolved with no the contribution of the basic event significant impact to the uncertainty to the total. SPAA results or conclusions.

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.

12-26 SPA-E5 II There are differences in the The report does not present the Document the results of Updated Monte Carlo results for CDF and LEAF results of sensitivity calculations sensitivity calculations on the uncertainty runs have that are reported in Table with regard to the number of number of Monte Carlo been performed with 5.1. A possible contributor to Monte Carlo simulations that are simulations required to 20,000 iterations for these differences may be needed to produce stable results. produce stable results. SCDF and SLEAF. This due to the number of Monte is a sufficiently high Carlo simulations that were It is our understanding from number of simulations to performed. discussion with the PAA staff that produce a stable result.

these types of sensitivity The SPAA (This F&O originated from calculations were performed. documentation has been SA SPA-E5) updated to clearly indicates the results.

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

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Disposition for 50.69 Finding Number Require- cc Description Basis Resolution and other applications ment(s) 12-27 SPR-F2 Met Documentation should be The current quantification Provide clear and complete The QU report provided that describes how document does not provide a documentation of the approach documentation has been the plant model analysis is clear description of the how the used to quantify the seismic updated to describe the quantified. plant model is quantified. For plant response model, to quantification process, example the discussion does not perform the risk quantification, including the technique identify how calculations are uncertainty analysis, and for combining cutsets performed, what the limitations of importance analysis. over the 14 acceleration (This F&O originated from these quantifications are and how intervals, and obtaining SR SPR-F2) they affect the results. the importance measures.

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

12-29 SPR-E2 Met The Quantification report There is limited documentation of Document the process and Additional detail has provides limited the process and the numerical methods that were used to been added to the QU documentation of the methods that were used to perform the uncertainty report to document the process and methods that perform the uncertainty analysis. analysis. Where appropriate uncertainty, importance, were used to perform the Based on the documentation that document where consistencies and sensitivity analyses uncertainty analysis. is provided and discussions with and potential inconsistencies and relate the uncertainty the PRA staff there is limited but in results might be expected. analysis mean SCDF not complete understanding of the and SLERF to the point methods that were used and the estimate values.

(This F&O originated from relationship of these methods to SR SPR-E2) the results were obtained This finding has been (reported in Table 5.1 ). resolved with no significant impact to the In some cases (as described in SPRA results or the documentation) the results conclusions.

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.

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Number Require- cc Description Basis Resolution Disposition for 50.69 and other applications ment(s) 12-31 SPR-F1 Not The standard requires a There is limited documentation Documentation should be Additional detail has Met level of documentation that that describes the seismic plant provided in sufficient detail that been added to the provides an understanding response analysis and describes the seismic plant documentation of the of the seismic plant quantification; how the model was model, how it is implement and seismic plant response response model and the implemented, how the quantified. model, model quantification. This quantification was performed and implementation, and requirement is not met. a discussion of the analysis quantification in the QU results. report. In addition, the uncertainty, importance, To meet this requirement, the and sensitivity analyses

{This F&O originated from documentation must be in are described in more SR SPR-F1) 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.

12-32 SPR-F3 Met The documentation of the The purpose of this supporting Document and discuss the The documentation of sources of model uncertainty requirement is that documentation contribution of the different the uncertainty analysis and a description of the should be presented that sources of uncertainty that are has been expanded in analysis assumptions is not addresses the sources of modeled in the SPRA. the Quantification report.

complete in the SPRA epistemic {knowledge) uncertainty A discussion of sources quantification report. In that are modeled and their of model uncertainty has addition, there is not a clear contribution to the total uncertainty been added to the report, description of the uncertainty in CDF and LERF. and potentially important analysis and the contributors sources have been to the total uncertainty In addition, the documentation addressed in the beyond a simple report from should discuss elements of the sensitivity analysis.

UNCERT. seismic plant model where there may be latent sources of This finding has been

{This F&O originated from uncertainty that are not modeled resolved with no SR SPR-F3) and assumptions that are made in significant impact to the performing the analysis. SPRA results or conclusions.

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Disposition for 50.69 Number Require- cc Description Basis Resolution and other applications ment(s) 12-36 SHA-83, 1/11, As part of a site-specific As part of the Vogtle PSHA an An up-to-date earthquake An update to the SHA-C4, Not PSHA, an up-to-date effort was not made to gather data catalog for the Vogtle site earthquake catalog was SHA-H1 Met earthquake catalog should on earthquakes that occurred region should be developed to prepared from the time of Not be used. The CEUS sse since 2008. As such, the analysts assess whether modifications the CEUS SSG catalog Met study involved the did not assess whether more to the seismic source (through 2008) through development of a recent seismicity is consistent with recurrence parameters or February 2016. The rate comprehensive earthquake the characterization parameters required. The updated catalog, of occurrence of catalog based on data estimated as part of the CEUS resources used in compiling earthquakes within 320 through 2008. The Vogtle sse study (NRC, 2012). the update and the results of km of the Vogtle site was site-specific PSHA should the evaluation should be compared to the rate of consider the impact sse of We note that as part of the Vogtle documented as part of the earthquakes represented any additional seismicity PSHA, calculations were PSHA. If more recent by the CEUS sse since 2008 up to the time performed to recompute the seismicity is not consistent seismic source model for the study started. seismic hazard at the site to take with the existing CEUS SSG that same area, this into account changes in the CEUS seismic source parameters, comparison being made (This F&O originated from sse earthquake catalog through the parameters should be for M>2.9. It was found SR SHA-C4) 2008 that were made following the updated and the PSHA should that the updated catalog completion of the CEUS SSG be updated. implied a rate of study. These changes reflect the earthquakes that is lower identification of reservoir induced than the mean rate from seismicity earthquakes and the re- the CEUS sse seismic interpretation of the location of sources. Therefore, some earthquakes in the incorporating the effects Charleston, SC area that occurred of a updated catalog on in the 1880's (EPRI, 2014). the hazard at Vogtle would decrease the References hazard slightly, and was not undertaken. This EPRI (2014). Review of EPRI comparison is 1021 097 Earthquake Catalog for documented in the SPRA RIS Earthquakes in the documentation.

Southeastern U. S. and Earthquakes in South Carolina This finding has been Near the Time of the 1886 resolved with no Charleston Earthquake Sequence, significant impact to the transmitted by letter from J. SPRA results or Richards to R. McGuire on March conclusions.

5, 2014.

12-8 SHA-J3 Not A foundational element of The documentation of the sources The resolution to this finding Sources of uncertainty in Met PSHA as it has evolved over of model uncertainty analysis and could involve: the seismic hazard the past 30 years is the a description of the analysis analysis for Vogtle are E-27

Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Number Require- cc Description Basis Resolution Disposition for 50.69 and other applications ment(s) development and implemen- assumptions is not complete in 1. Documentation and discussed in the updated tation of methods to identify, the PSHA report in its current form discussion of the contribution SPRA documentation.

evaluate, and model sources such that a clear understanding of of different sources of These include of epistemic (model and the contribution of individual uncertainty that are modeled in uncertainty in seismic parametric) uncertainty in sources of uncertainty to the the PSHA. The documentation source model (for the estimate of ground estimate of hazard are of the contribution of different background earthquake motion hazards. As such understood. Limited information sources of uncertainty can be sources and for the fairly rigorous analyses are on the contribution of seismic shown by means of 'tornado Charleston RLME), in carried out (SSHAC studies) sources to the total mean hazard plots' that quantify the maximum magnitude for to quantitatively address is presented, but information on sensitivity of the hazard at background seismic model uncertainties. the contributors to the uncertainty different ground motion levels sources and for the is not provided. to the various branches in the Charleston RLME, in At the same time there is logic tree. These plots show ground motion prediction within any analysis sources With respect to addressing model which sources of epistemic equation, in smoothing of uncertainty that are not uncertainties and associated uncertainty are most important. assumptions for directly modeled and assumptions there are some It should include the source seismicity parameters in assumptions that are made examples that can be identified in model uncertainty, ground background sources, and for pragmatic or other the Vogtle PSHA. For example, in motion model uncertainty, and in site amplification reasons. There are also the site response analysis the site response uncertainty. model. "Tornado plots" sources of model uncertainty assumption is made that the 1D Currently, the total uncertainty are included in the that are embedded in the equivalent linear model (SHAKE is shown by the hazard updated SPRA context of current practice type) to estimate the site fractiles, but it is not broken documentation that show that are 'accepted' and amplification and ground motion down to provide understanding the contribution to total typically not subject to input to plant structures is as to what is most important. uncertainty in seismic critical review. For instance, appropriate. hazard from source in the PSHA it is standard 2. Identification and discussion model uncertainty, practice to assume that the of model assumptions that are maximum magnitude temporal occurrence of made. uncertainty, ground earthquakes is defined by a motion prediction Poisson process. This equation uncertainty, assumption is well accepted smoothing assumptions despite the fact that it for seismicity parameters violates certain funda- in background sources, mentally understanding of and site response tectonic processes (strain uncertainty. These plots accumulation). A second are presented for 10 Hz practice is the fact that and 1 Hz spectral earthquake aftershocks are acceleration, for ground not modeled in the PSHA, motion amplitudes even though they may be corresponding to mean sianificant events -------

annual frequencies of E-28

Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Number Require- cc Description Basis Resolution Disposition for 50.69 and other applications ment(s)

(depending on the size of exceedance of 10-4 and the main event). 10-5. These "tornado plots" show that ground In the spirit of the standard it motion prediction seems appropriate that equation is the major sources of model uncertainty contributor to seismic that are modeled as well as hazard uncertainty for sources of uncertainty and both 10Hz and 1 hz associated assumptions as spectral acceleration, they relate to the site- and maximum magnitude specific analysis should be of the Charleston RLME identified/ discussed and source is an important their influence on the results contributor for 1 Hz discussed. spectral acceleration.

The use of equivalent As SPRA reviews and the linear one-dimensional use of the standard has site response analysis, evolved, it would seem the and its associated former interpretation is assumptions, and its reasonable, but potentially adequacy for the Vogtle incomplete. It is reasonable site are documented in from the perspec-tive that the hazard calculation.

document-tation of the This finding has been sources of model uncer- resolved with no tainty and their contribution significant impact to the to the site-specific hazard SPRA results or results is a valuable product conclusions.

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 E-29

Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Number Require- cc Description Basis Resolution Disposition for 50.69 and other applications ment(s) 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) 14-1 SFA-A2 I The conservatisms that exist SFA-A2 requires that seismic Account for conservatism in Evaluation of anchorage in structural demand were fragilities be based on plant- the building response analyses has been updated to not properly accounted for in specific data and that they are in the structure response factor include clipping of in-the estimation of component realistic and median centered with for component fragility structure response and structure fragilities. 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.

(This F&O originated from used in all component fragilities capacities. Structure response is SA SFR-A2). 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 -

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Disposition for 50.69 Number Require- cc Description Basis Resolution and other applications ment(s) the structural analysis. Because cause higher damping at of this, the component and lower hazard frequency structural fragilities are biased levels and lead to stress low. similar to the stress calculated for the The fragilities developed for buildings at 1E-4. As a structures and components that result the structural are mounted in those structures response factor is close will be biased low because the to 1 and is accounted for input structural demands include appropriately in the conservatisms. Time histories fragility evaluations.

used for the SSI analysis have The input time history been processed such that each motion at the control record envelopes the target point in the SSI UHRS. This will introduce some analysishas been level of conservatism. The input modified to reasonably motion at the control point has match the corresponding been scaled to produce resultant 1E-4 UHRS from the FIRS that envelopes the FIRS site-consistent input coming out of the site-consistent motion analysis.

input motion analysis. In structure response spectra coming out of This finding has been the SSI analyses were not peak resolved.

clipped when computing anchorage demands. Structure response at the calculated equipment fragility levels is considerably higher than the 1E-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.

14-10 SFR-A2 I Significant conservatisms In the fragility calculations of heat Realistic nozzle loads should The CCW and ACCW were noted in several exchangers (PRA-BC-V-14-009 be determined for fragility heat exchanger sampled fragility Appendix A), nozzle loads evaluation of heat exchangers. capacities have been calculations. significantly contribute to the updated to reflect seismic demands which form the realistic nozzle loads.

basis for the median capacities. The equipment fragilities Based o_11_lr!-pl(int_walkdowns by The equipment capacity factor have been updated to -

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Disposition for 50.69 Number Require- cc Description Basis Resolution and other applications ment(s)

(This F&O originated from the peer review teams and also should be based on the account for appropriate SA SFR-A2) noted in the walkdown report, the frequency range of interest. frequency, and piping is well supported in all That frequency range of uncertainty has been directions and will not impose interest is centered at the considered in these significant nozzle loads during a fundamental frequency of the updates.

seismic event. The CCW and pump, and considers some ACCW capacities are below the uncertainty in that frequency. This finding has been 2.5g screening level and are resolved.

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 1g, 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. I 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.

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Number Require- cc Description Basis Resolution Disposition for 50.69 and other applications ment{s) 14-14 SFR-G2 Met The iterative process used In review of the seismic fragility Add a description of the The description of the for developing realistic calculation for the safety features iterative process for computing iterative process for fragilities is not well sequencer (11821 U3001 ), it was the component fragilities in the computing fragilities has documented. discovered that an iterative SPRA documentation been documented.

process was used. The initial fragility is based on EPRI 6041 This finding has been screening methodology and an resolved with no (This F&O originated from equipment capacity factor that is significant impact to the SA SFR-G2) equal to the EPA I 6041 median SPRA results or capacity divided by the peak in conclusions.

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.

14-17 SFR-02 Met Inconsistencies and errors in Fragilities for the Vogtle 1&2 Update SNC calculation no. The following changes NSSS fragility development. Nuclear Steam Supply System PRA-BC-V-14-015 to have been made: NSSS (NSSS) are based on the results incorporate corrections and fragility calculations have of the Westinghouse analysis of enhancements. been updated to reflect record (AOR} associated with the Westinghouse-provided (This F&O originated from safe shutdown earthquake (SSE}. critical loads and support SA SFR-02} In general, fragilities are capacities represented in developed through scaling of the the critical failure modes; SSE demands to the ALE and the effect of inelastic using the AOR seismic margins. energy absorption is Various deficiencies were noted in factored in and the development of the fragilities documented in fragility associated with these calculation as components. appropriate; the Reactor E-33

Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Number Require- cc Description Basis Resolution Disposition for 50.69 and other applications ment(s)

Basis: The NSSS Seismic fragility Coolant Pump fragility evaluation (SNC calculation no. has been updated to PRA-BC-V-14-015) includes detail reflect the failure of the calculations for each of the major pump associated with NSSS components. It indicates LOCA; the reactor that the critical failure modes for internals fragility has the components are controlled by been updated in the the support capacities. calculation; and the new During the Peer Review, the team fragilities have been members discussed these issues reflected in the updated with SNC staff to obtain insights SPRA model.

and develop potential resolution I

paths. Key issues included: This finding has been (a) Basis for assumption that the resolved. I I

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.

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Disposition for 50.69 Finding Number Require- cc Description Basis Resolution and other applications ment(s)

(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.

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Number Require- cc Description Basis Resolution Disposition for 50.69 and other applications I

ment(s) 14-20 SFR-E4, Met Seismic induced fire The only mention for seismic Seismic induced fire is an The seismic-induced fire SPR-89 Met evaluations are not induced fire evaluation is important element of the and flood evaluations documented in the contained in the quantification fragility evaluation process and have been updated, and walkdown report or fragility notebook. Based on discussions this should be clearly documented in the calculations. during the peer review, it is documented. fragility and quantification understood that seismic induced report. This includes the fire was a key consideration details of the walkdown during the walkdowns. However, procedure used to (This F&O originated from detail of the walkdown procedure evaluate the potential for SR SFR-E4) for fire following earthquake is seismically induced fires, missing. The write up should including the include team composition, methodology, screening methodology, screening criteria, criteria and results.

and results, This finding has been resolved with no significant impact to the SPRA results or conclusions. I 14-4 SFR-01 Met A potential for sloshing SFR-01 requires that realistic Evaluate the potential for flood The evaluation for induced inundation of the failure modes of structures and induced failure of the NSCW potential flood induced NSCW Pumps equipment that interfere with the Pumps or NSCW discharge failure of the NSCW (11202P4007, 11202P408) operation of that equipment be MOVs. pumps or the NSCW and associated discharge identified. discharge MOVs has motor operated valves been performed and (1HV11600, 11606,11607, The potential for earthquake documented in the 11613) in the NSCW exists induced sloshing of the water fragility calculation for the and was not identified either within the NSCW tower exists. NSCW tower. There was in the walkdowns or From field walkdowns of the no significant impact on subsequent analysis. NSCW it was observed that there the pump or MOV is a potential for sloshing of fragilities.

(This F&O originated from contents to potentially splash onto SR SFR-01) or flood the pumps and or motor This finding has been operated valves on the attached resolved with no discharge piping. significant impact to the SPRA results or

_ conclusions.

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Disposition for 50.69 Finding Number Require- cc Description Basis Resolution and other applications I

1 ment(s) 14-5 SFR-D1 Met The potential for seismically- Vogtle 1&2 is a soil site, with Develop estimates of the Documentation has been induced differential engineered fill from the rock differential settlements updated to include the settlements between interface to the finished grade. between adjacent structures effects of earthquake structures was not The in-scope Seismic Category I and assess the fragility of induced settlement; no addressed. structures have foundations with commodities based on their significant differential varying embedment depths, ability to accommodate the settlements were ranging from surface founded associated differential computed between the (elev. 220ft.) to a foundation displacements. structures.

(This F&O originated from embedment of 11 0 ft. (elev. 11 0 SR SFR-D1) ft.). Since soils, including This finding has been engineered fill, will resolved with no consolidate/settle to some extent significant impact to the when subjected to high level SPRA results or earthquake ground motion, and conclusions.

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.

14-6 SFR-G2 Met The results of the seismic The walkdown guidance provided Provide documentation of the As noted in the Finding gap/shake space walkdowns in Appendix F (Checklists and results of the seismic gap basis, inspection of the are not documented. Walkdown Data Sheets) of EPRI walkdowns. seismic gaps was NP-6041 includes attributes of included in the seismic E-37

Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Number Require- cc Description Basis Resolution Disposition for 50.69 and other applications ment(s) seismic gaps between structures walkdowns. Piping which should be addressed in the across seismic gaps is (This F&O originated from performance of the walkdowns. designed with adequate SR SFR-G2) These include the clearance flexibility to between adjacent structures and accommodate building the ability of any subsystems motions, and pipe (e.g., piping, cable trays, HVAC sleeves provide ducts) spanning the gap to adequate gaps for piping accommodate the differential movement. The seismic displacements. documentation has been updated to reflect the The Seismic Walkdown Report inspections performed (PRA-BC-V-14-005) does not during the walkdowns.

include documentation of the results/findings/observations This finding has been associated with the inspection of resolved with no the seismic gaps between significant impact to the structures or the subsystems SPRA results or spanning the gap. During the conclusions.

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.

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Number Require- cc Description Basis Resolution Disposition for 50.69 and other applications ment{s) 14-7 SFR-A2, I, The fragility evaluation for The determination of the Update the fragility evaluation The fragility evaluation of SFR-F4 Met the Containment Polar fundamental frequency of for the polar crane to address the polar crane has been Crane (in fragility notebook) structures and components potential uncertainty in the updated to address did not address the impact involves a certain degree of fundamental frequency and the potential uncertainty in of variation in the uncertainty. This uncertainty must contribution of higher modes. the fundamental fundamental frequency on be accounted for in the frequency and the applicable seismic determination of the seismic contribution of higher demand. accelerations from the applicable modes.

in-structure response spectra (ISRS). This finding has been resolved.

(This F&O originated from Section 7.4 (Vogtle 1 and 2 Polar SR SFR-A2) 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%

decrease in the median capacity of the polar crane. I E-39

Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Number Require- cc Description Basis Resolution Disposition for 50.69 and other applications ment(s) 14-8 SFR-F3 IIIII I Relay fragility calculations The relay evaluation for the Perform more realistic relay The relay fragilities have include conservative turbine driven auxiliary feedwater fragility evaluations. been updated using the assumptions. 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.

(This F&O originated from control centers) and an SA SFR-F3) 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.

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Number Require- cc Description Basis Resolution Disposition for 50.69 and other applications ment(s) 14-9 SFR-D2 Met The seismic walkdown The summary of the seismic Perform resolution of open The noted walkdown report includes a number of walkdowns documents a number items and provide issues have been open items that are not are of issues identified during the documentation of the evaluated and reflected not traceable to a resolution performance of the walkdowns resolution associated with in the revised that required follow-up actions each of the issues, either in documentation:

(31). These include spatial the Fragility Notebook or the - potential piping interaction issues, housekeeping SPRA Database. interaction; (This F&O originated from issues, anchorage issues, valves - the difference in SR SFR-D2) having configurations that do not inverter anchorage meet the EPRI guidelines, configuration; configuration issues, installation - potential interaction errors, etc. concerns with the overhead heater; this The Seismic Walkdown Report evaluation is in the (PRA-BC-V-14-005) does not fragility notebook in document how the issues section 3.4.2.

identified during the walkdowns Valve operator heights &

have been addressed, either in weights that were outside the field (e.g., correction of EPRI guidelines have installation errors, resolution of been taken into account housekeeping issues) or in the in the fragility analysis for fragility evaluations (e.g., valve these components.

configurations, anchorage issues). The Diesel Generator During the performance of the Exhaust Silencer was re-Peer Review, the Peer Review evaluated to the as-Team provided a list of the operated condition.

walkdown issues to SNC The fragility analysis for personnel, and SNC provided a these components has summary of how they were been completed for the addressed. Most issues had been as built condition.

adequately addressed during the development of the SPRA, but it This finding has been was determined that the following resolved.

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 ~ -~

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Disposition for 50.69 Number Require- cc Description Basis Resolution and other applications ment(s) 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 105) -

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 107) -

further evaluation required.

16-1 SFR-F3, 111111, The model presented for Relay chatter is consistently being Complete the analysis and The approach to SPR-84, Not peer review did not observed as a significant incorporate the effects of relay screening and modeling SPR-E5 Met, incorporate the effects of contributor to risk profile in chatter and similar devices in of seismically-induced II relay chatter as the analysis recently peer reviewed S-PRAs the PRA logic model. relay failures and chatter was not yet complete. 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; (This F&O originated from discussed that the SPRA team only the incorporation SR SPR-84) 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_

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Number Require- cc Description Basis Resolution Disposition for 50.69 and other applications ment(s) combination of relays and a chatter fragilities and number of HEP may not be. impacts have been incorporated into the seismic model, in a manner consistent with that used for other failures.

This finding has been resolved.

16-10 SPR-86 Met The documentation about There is only a short sentence More detailed documentation Walkdown the walkdowns in support to supporting the discussion on is suggested to support the documentation on seismic impact on HRA alternative access pathways. conclusion on accessibility, accessibility for operator appear limited. 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 (This F&O originated from and documented.

SR SPR-86) 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 -

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Disposition for 50.69 Number Require- cc Description Basis Resolution and other applications ment(s)

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.

16-11 SPR-E2 Met Missing review of the It is understood that the As this exercise was A detailed quantitative potential for additional investigation performed in internal apparently performed for the HRA dependency dependencies introduced by events to identify potential H FE Fire PRA (as discussed during analysis based on using the SPRA models (QU- dependency has been relied upon the peer review), it is the HRA calculator was C1&2) in the Vogtle SPRA. suggested that a review of the performed and potential for unforeseen documented. There was The SPRA logic may identify dependencies trends is no significant impact on additional dependencies trends performed. results since human (This F&O originated from that were not identified in the actions are not significant SR SPR-E2) internal events. 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.

16-12 SPR-E2 Met Missing documentation of It is an industry expectation (as It is understood that the SPRA The QU report has been the review of non significant discussed in NEI peer review task documentation will be revised updated to document the cutsets QU-05. force meetings) that review of the to incorporate explicitly the two review of both dominant non significant cutsets is explicitly reviews discussed in the basis cutsets and non-documented. for this F&O. It is also significant cutsets for recommended to document both CDF and LEAF.

(This F&O originated from Based on discussion during the the review of cutsets following SR SPR-E2) peer review, two reviews were guidance from the NEI peer This finding has been performed to validate the overall review task force. resolved with no model and cutsets. The first was significant impact to the a random review of cutsets at SPRA results or midpoints and low significance for conclusions. i each of the %Gxx initiators to verify that the cutsets are valid cutsets, and that the patterns are appropriate. That is, if one cutset --

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Number Require- cc Basis Resolution Disposition for 50.69 and other applications ment(s) Description 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 ATWT cutsets were generated.

16-15 SPA-E6, Met, Documentation of LEAF The current documentation does Expand the documentation to The LEAF SPA-F2 Met model applicability review. not explain what are the basis for ensure that the criteria used to documentation in the QU retaining the LEAF logic and retain the LEAF analysis in the report was expanded to analysis unchanged within the SPAA is explained so that the describe the review of SPAA logic. same applicability review can applicability of the (This F&O originated from be performed following future internal events PAA SA SPA-F2) During the peer review the potential revisions of the LEAF LEAF analysis to the following explanation was modeling. seismic PAA.

provided by the SPAA team:

This finding has been "The internal events Level 2 resolved with no notebook (Chapter 9) was significant impact to the reviewed to ensure that the SPAA results or definition of LEAF would be conclusions.

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 -

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Number Require- cc Description Basis Resolution Disposition for 50.69 and other applications ment(s) 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 I straightforward, with sequences from the CDF model ANDed with the appropriate LERF fault tree.

This logic is also appropriate for seismic events."

16-18 SPR-88 Ill Very small LOCA have been The DB has a specific entry for To the peer review team Additional information on screened from the analysis the incore thermocouples and knowledge Vogtle is the only the walkdown for very based on walkdowns but provides pictures of them. Still, in- plant that has elected to small LOCA has been I little documentation exists of core thermocouple tubing is not perform dedicated walkdowns added to fragility report such walkdowns. the only possible source of very in support of not modeling very to provide the basis for small LOCA that is envisioned and small LOCA. This would be a the VSLOCA screening.

the only documentation of best practice but it also I addressing the other potential behooves to the SPRA team to This finding has been (This F&O originated from sources is in section 2.3.3 of the provide detailed resolved with no SR SPR-88) quantification notebook: documentation of such significant impact to the walkdowns and how they SPRA results or "For Vogtle 1&2, the seismic supported a systematic conclusions.

walkdowns inspected and evaluation of the potential photographed a large sample of sources of very small LOCA.

the small piping and tubing lines connected to the primary system in order to identify any weaknesses. The piping was iudQed to be ruQQed."

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Number Require- cc Description Basis Resolution Disposition for 50.69 and other applications ment(s}

16-2 SFR-C1, 1/11, Fragilities were not The 2014 hazard was only used During the peer review the The fragilities have been SPR-E1 Met corrected to reflect the 2014 as input to FRAN X for the final SNC staff answered a question recalculated based on hazard used for quantification. It is understood that on this topic by performing an the 2014 hazard [13] and quantification. (This F&O the fragility estimates have been initial limited investigation of the new values originated from SR SPR-E1) performed based on the 2012 the effect on fragilities incorporated into the hazard. While it is not expected correction to reflect the 2014 SPRA model and nor recommended to regenerate hazard and concluded that the quantification.

all the fragility work with the new effect of this scaling is not hazard, some consideration on insignificant (especially for This finding has been the possible change in fragility LEAF). It is recommended to resolved.

due to the use of the newer continue and expand this hazard should be made. investigation to make the quantification fully consistent with the fragility values.

16-4 SPR-82 Not The effect of seismic impact There is no assessment of the While it is recognized that the The methodology used Met on performance shaping effect of changing the breaking industry is still developing for the seismic HRA factors is considered in the points in the Surry method. The methods in support to this analysis is based on analysis by the usage of the Surry method is based on particular topic (e.g., recently defining PSFs as a Surry method. methods used in the past at published EPRI HRA method function of seismic SONGS and Diablo Canyon and for external events), some hazard level (bins), which the O.Bg breaking point was additional considerations is consistent with the developed for California should be done to understand EPRI seismic HRA (This F&O originated from earthquakes. In the Vogtle the effect of HEPs in the model guidance in EPRI SR SPR-82) analysis there is no indications on rather than simply 3002008093 [14]. The whether the breaking point at O.Bg implementing the Surry Integrated PSFs and bins is also applicable to Vogtle. There method as is. (breaking points) have are also no sensitivity analyses been updated with that would support whether a Three examples for addressing additional breaking change in the breaking points is this finding may be the points and integrated significant or not. 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.

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding I Finding Disposition for 50.69 Number Require- cc Basis Resolution and other applications ment(s) Description There was no significant impact on the SPRA results.

This finding has been resolved.

16-5 SPR-81, Met, LOCA modeling and fragility The selection of the fragility data Documentation on the use of LOCA basis has been re-SPR-F1 Not selection not clearly used for all LOCA is discussed in fragility in support to LOCA evaluated and updated.

Met documented. Appendix 8.2 of the quantification should be clarified to better This was partially due to notebook but is confusing in the represent the rationale seismic fragility update mapping of selected fragilities with selected and potentially and partially a matter of specific failures. addresses the modeling adding amplifying (This F&O originated from uncertainties associated with information to the LOCA SA SPR-F1) It appears that the fragility this selection. basis. The quantification selected to represent LOCA report includes updated sequences are coming from While this finding is expected documentation. Although specific components but then they to be addressed via LOCAs are a significant are used to represents sort of documentation, some contributor to the SPRA surrogate events for potential additional suggestions are results, the VEGP SCOF failures along the piping network. provided, such as: and SLERF are sufficiently small that Using localized events as 1. Perform a sensitivity to further LOCA modeling surrogate for pipe network failure show that the modeling sensitivity beyond what is probably conservative and may approach described is not has been provided in the not be fully consistent with the significantly skew the results updated model system success criteria and for seismic; quantification is not modeling in the internal events warranted. '

modeling. For example, the 2. Modify the logic by mapping seismic-induced MLOCA fragility the seismic-induced MLOCA to This finding has been seems to be based on failure of a different position in the logic resolved.

the pressurizer surge line, which (e.g., a dummy event can be is a localized failure. The seismic- entered in the model to provide induced MLOCA initiator is a target for the FRANX mapped to the internal events injection).

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

Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Number Require- cc Description Basis Resolution Disposition for 50.69 and other applications ment(s) 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.

16-6 SPR-82 Not The effect of seismic impact The Vogtle SPRA elected to use Expand the IPSF approach to The methodology used Met on performance shaping Integrated Performance Shaping all the operator actions for the seismic HRA factors is not considered for Factors (IPSF) multipliers. While credited in the SPRA. analysis is based on any action that was explicitly this approach was used for the defining PSFs as a added for the SPRA (e.g., HEPs that were carried over from function of seismic flood isolation or DG output internal events, it was hazard level (bins), which breaker closure). systematically not done for all the is consistent with the actions explicitly added for EPRI seismic HRA seismic. guidance in EPRI 3002008093 [14). The Based on discussion during the Integrated PSFs and bins (This F&O originated from peer review, the analyst believed (breaking points) have SR SPR-82) that having designed these been updated to reflect actions for specific scenarios seismic binning following a seismic event, the applicable to Vogtle, in impact of seismic specific PSF is accordance with this already included. finding and consistent with the EPRI The objection to this conclusion is guidance. The updated that the seismic specific PSF values have been should realistically change with applied to both internal the magnitude of the event. This events HFEs and change addresses the change in seismic-unique HFEs the overall context of the plant within the plant response when a small seismic event model.

happens as opposed to when a There was no significant very large seismic event happens. impact on the SPRA This seems not to be captured by results.

the approach selected for the Vogtle SPRA. One example of this E-49

Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Number Require- cc Description Basis Resolution Disposition for 50.69 and other applications ment(s) is that an action that has a 30 This finding has been minute Tsw (S-OA-BKR-LOCAL) resolved.

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.

16-7 SPR-E2 Met Base case seismic LEAF Both CDF and LEAF are LEAF at 1E-11 truncation LEAF truncation, which does not meet the truncation truncated at 1.0E-09 with 1000 meets the QU-B3 truncation was already considered requirements from QU-B3. cutsets managed by ACUBE. This requirement. Rename LEAF at in sensitivity studies, has meets the QU-B3 requirement for 1E-11 as the base case for been revised CDF but not for LEAF. LEAF. appropriately to meet QU-B3. A new LEAF (This F&O originated from truncation limit has been SR SPR-E2) established consistent with the LEAF results.

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

This finding has been resolved.

16-8 SPR-E2 Met Missing documentation of Section 3.1 is the only description While it is understood that the The QU report has been cutsets review (cfr. QU-01) of the most important scenarios Draft. B version of the updated to document the but there is no cutset-by-cutset quantification notebook is still review of both dominant review. somewhat a work in process, it cutsets and non-is expected that when the significant cutsets for (This F&O originated from model reaches a more stable both CDF and LEAF.

SR SPR-E2) 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.

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Number Require- cc Description Basis Resolution Disposition for 50.69 and other applications I

ment(s) 16-9 SPR-81, Met, Screening values used for At the time when the An appropriate resolution of The seismic HRA SPR-84b Met the HEPs that (at the time of documentation was provided for this F&O is pending the current analysis has been the provided documentation) peer review, the most significant evolution of the model and the revised to be consistent were in the most significant operator actions (i.e., flood importance of operator actions with the EPRI seismic I cutsets. isolation of ACCW HX) were all in the SPRA. Given the HRA guidance in EPRI screening values, which would expectation that operator 3002008093 [14]. The only meet CCI for HR-G1 (directly actions will be needed to original screening HEPs called through SPR-81 ). mitigate the importance of have been updated using (This F&O originated from relay chatter (not yet included the HRA Calculator, SR SPR-81) In addition, there is little in the SPRA logic model) this consistent with the documentation or supporting F&O was provided to ensure approach used in the evidence to justify screening care is used in the generation VEGP internal events values as low as 3.00E-2 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.

17-1 SPR-81 Met The documentation does not The modeling approach injected A separate section in the The discussion of specifically address the seismic fragilities into fault trees documentation that specifically accident sequences and applicability of the internal that were modified from the addresses accident sequences success criteria has been events accident sequences internal events PRA model. It can and success criteria is needed expanded, and specific and success criteria to the be inferred from this approach, to collect the information in descriptions of the SPRA model, and does not and it was verified by discussions one logical place, and is flooding scenarios has properly document the with the staff, that the internal needed to support effective been added. This finding accident sequences created events sequences and success peer reviews and future model is documentation only specifically for the SPRA criteria were considered to be updates. and does not impact model. applicable to the SPRA model. Seismic PRA model This was not specifically stated in results.

the documentation.

This finding has been (This F&O originated from Further, several additional seismic resolved with no SR SPR-81) flooding sequences were added to significant impact to the the fault tree. These sequences SPRA results or are not discussed from an conclusions.

accident sequence and success _ - - - - -

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Number Require- cc Description Basis Resolution Disposition for 50.69 and other applications mentes) 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.

17-2 SPR-E2, Met, The processes used to Examples include: Expand the documentation to Documentation for QU SPR-F2 Met create the presented clearly explain the post- results has been im-quantification results are not The top cutsets shown in table 3-1 processing of the results proved to describe the fully documented. of the quantification report are generated by CAFTA and processes used to ag-produced by combining the FRANX. Examples include: gregate results over the cutsets from all the seismic 14 hazard intervals. The interval cutsets in a process that is - Explain how the cutsets importance calculations (This F&O originated from not documented. generated by FRANX are have been re-quantified SR SPR-F2) combined into g-level- and the method for While the process used to obtain independent cutsets. presentation the importance measures in documented.

section 5.2 of the quantification - Explain the post-processing notebook is documented in that used to generate importance This finding has been section, discussions with the PRA measures, especially focusing resolved with no staff indicated that importances for on the deviation from a normal significant impact to the some of the basic events were practice that is currently only SPRA results or obtained in a different manner mentioned in the notebook. conclusions.

(setting to one or zero and requantifying). This is not documented in the notebook.

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Disposition for 50.69 Finding Number Require- cc Description Basis Resolution and other applications ment(s) 17-3 SPR-83, 1/11, Subdividing correlation To account for similar equipment The impact of the retention of The non-minimal cutsets SPR-E4 1/11 groups based on that has different fragilities due to these non-minimal cutsets on in the peer reviewed weaker/stronger different building locations, certain CDF/LERF and importance model were identified components resulted in correlation groups were measures should be assessed and reviewed for impact, retention of non-minimal subdivided to assign a seismic and the results documented, or and determined to be cutsets in some cases, capacity to a weaker component a method to remove the non- non-significant to risk.

which could impact that only failed that component. minimal cutsets should be The results were very CDF/LERF results as well as The higher capacity was then devised. Each subdivided slightly conservative due model importance assigned to both components, and correlation group should be to these non-minimal measures. The magnitude was effectively the correlated investigated for similar effects. cutsets. The issue has and acceptability of these failure of both components. This been addressed in the impacts was not can result in the retention of non- updated model, such documented. 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 (This F&O originated from core damage. Inspection of the SPRA results or SA SPR-E4) cutsets shows that two otherwise conclusions.

identical cutsets are retained: one in which the 1Fan '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.

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Enclosure to NL-16-2382 Basis of the Proposed Change Supporting Finding Finding Suggested Finding Finding Disposition for 50.69 Number Require- cc Description Basis Resolution and other applications ment(s) 17-4 SPR-E6 Met No quantitative analysis of A quantitative analysis is required Perform the analysis and The quantitative analysis the relative contribution to to meet CCII for LE-F1 & LE-G3, include the results in the of significant LERF plant LERF from Plant Damage which are directly called from quantification notebook. damage states and States and Significant LERF SPR-E6. contributors has been contributors from Table 2- performed. A table and 2.8-9 was presented in the associated discussion of quantification results. plant damage states and significant contributors (This F&O originated from has been added to the SR SPR-E6) LERF QU documentation i to resolve this finding. i This finding has been resolved.

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