CNL-19-108, Response to NRC Second Request for Additional Information Regarding Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors

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Response to NRC Second Request for Additional Information Regarding Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors
ML19302D625
Person / Time
Site: Watts Bar  Tennessee Valley Authority icon.png
Issue date: 10/28/2019
From: Polickoski J
Tennessee Valley Authority
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
CNL-19-108, EPID L-2018-LLA-0493, WBN-TS-17-24
Download: ML19302D625 (22)
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Text

Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 CNL-19-108 October 28, 2019 10 CFR 50.69 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Watts Bar Nuclear Plant, Units 1 and 2 Facility Operating License Nos. NPF-90 and NPF-96 NRC Docket Nos. 50-390 and 50-391

SUBJECT:

Response to NRC Second Request for Additional Information Regarding Watts Bar Nuclear Plant, Units 1 and 2, Application to Adopt 10 CFR 50.69, "Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors (WBN-TS-17-24)

(EPID L-2018-LLA-0493)

References:

1. TVA letter to NRC, CNL-18-068, Watts Bar Nuclear Plant, Units 1 and 2, Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors (WBN-TS-17-24), dated November 29, 2018 (ML18334A363)
2. NRC Electronic Mail to TVA, Watts Bar Nuclear Plant - Final Request for Additional Information Related to Application to Adopt 10 CFR 50.69 (EPID L-2018-LLA-0493), dated June 18, 2019 (ML19169A359)
3. TVA letter to NRC, CNL-19-065, Partial Response to NRC Request for Additional Information Regarding Watts Bar Nuclear Plant, Units 1 and 2, Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors (WBN-TS-17-24) (EPID L-2018-LLA-0493), dated July 15, 2019 (ML19196A362)
4. TVA letter to NRC, CNL-19-069, Final Response to NRC Request for Additional Information Regarding Watts Bar Nuclear Plant, Units 1 and 2, Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors (WBN-TS-17-24) (EPID L-2018-LLA-0493), dated July 29, 2019 (ML19210D430)
5. NRC Electronic Mail to TVA, Watts Bar Nuclear Plant - Request for Additional Information Related to the Application to Adopt 10 CFR 50.69 (EPID L-2018-LLA-0493), dated September 13, 2019 (ML19259A006)

U.S. Nuclear Regulatory Commission CNL-19-108 Page 2 October 28, 2019 In Reference 1, Tennessee Valley Authority (TVA) submitted to the Nuclear Regulatory Commission (NRC) a request for an amendment to Facility Operating License Nos. NFP-90 and NPF-96 for the Watts Bar Nuclear Plant (WBN) Units 1 and 2 to allow for the implementation of the provisions of 10 CFR, Part 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors." This License Amendment Request (LAR) included in Attachment 1 to Enclosure 1, a list of categorization pre-requisites, that were proposed to be controlled under a proposed license condition. In Reference 2, the NRC provided a request for additional information (RAI). In Reference 3, TVA submitted responses to NRC DRA RAls 01, 02, 06, 07, 09, and 11 of Reference 2. In that RAI submittal, TVA identified additional changes to Attachment 1 and made changes to the proposed license condition to refer to Reference 3. In Reference 4, TVA submitted the remaining RAI responses. In Reference 5, the NRC provided an additional RAI and requested that TVA respond by October 28, 2019. to this letter provides the response to the additional RAI. As a result of this RAI response, additional changes were identified for Attachment 1 as provided in Reference 3.

Accordingly, Enclosure 2 of this submittal provides the Attachment 1 revision. Enclosure 3 provides the existing WBN Unit 1 and Unit 2 Facility Operating Licenses marked-up to show the revision of the proposed license condition to refer to Enclosure 2, Attachment 1, of this submittal. Enclosure 4 to this letter provides the existing WBN Unit 1 and Unit 2 Facility Operating Licenses re-typed pages to show the proposed changes. Enclosures 3 and 4 supersede those Operating License changes provided in References 1 and 3.

The enclosures to this letter do not change the no significant hazards consideration nor the environmental considerations contained in Reference 1. Additionally, in accordance with 10 CFR 50.91 (b)(1 ), TVA is sending a copy of this letter and the enclosures to the Tennessee Department of Environment and Conservation.

There are no new regulatory commitments made in this letter. Please address any questions regarding this submittal to Kimberly D. Hulvey, TVA Fleet Licensing Manager, at (423) 751-3275.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 28th day of October 2019.

Respectfully,

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Director, Nuclear Regulatory Affairs Enclosures cc: See Page 3

U.S. Nuclear Regulatory Commission CNL-19-108 Page 3 October 28, 2019

Enclosures:

1. Response to NRC Second Request for Additional Information Regarding Watts Bar Nuclear Plant, Units 1 and 2, Application to Adopt 10 CFR 50.69, "Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors (WBN-TS-17-24) (EPID L-2018-LLA-0493)
2. Attachment 1: List of Categorization Prerequisites
3. WBN Units 1 and 2 Facility Operating Licenses Changes Markup
4. WBN Units 1 and 2 Facility Operating Licenses Changes Retyped Copy cc: NRC Regional Administrator - Region II NRC Senior Resident Inspector - Watts Bar Nuclear Plant NRC Project Manager - Watts Bar Nuclear Plant Division of Radiological Health - Tennessee State Department of Environment and Conservation

Enclosure 1 Response to NRC Second Request for Additional Information Regarding Watts Bar Nuclear Plant, Units 1 and 2, Application to Adopt 10 CFR 50.69, "Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors (WBN-TS-17-24) (EPID L-2018-LLA-0493)

NRC Introduction By application dated November 29, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18334A363), as supplemented by letters dated July 15 and July 29, 2019 (ADAMS Accession Nos. ML19196A362 and ML19210D430, respectively),

Tennessee Valley Authority (the licensee), submitted a license amendment request (LAR) to adopt Title 10 of the Code of Federal Regulations (10 CFR) 50.69, Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors, for the Watts Bar Nuclear Plant (WBN), Units 1 & 2.

The U.S. Nuclear Regulatory Commission (NRC) staff previously transmitted requests for additional information (RAIs) to the licensee via electronic mail on June 18, 2019 (ADAMS Accession No. ML19169A359). The licensee submitted responses to the RAIs in the supplemental letters dated July 15, 2019 and July 29, 2019.

The NRC Probabilistic Risk Assessment (PRA) Licensing Branch A (APLA) and Risk-Informed Licensing Initiatives Team (RILIT) staff have reviewed the application, as supplemented, and have identified areas where additional information is necessary for the staff to complete its technical review.

DRA RAI 01 Appendix X, Close-out of Facts and Observations (APLA)

The process to close finding-level facts and observations (F&Os) is documented in Appendix X to Nuclear Energy Institute (NEI) 05-04, 07-12, and 12-13 Close-out of Facts and Observations (F&Os) 1, as accepted, with conditions by the NRC in letter dated May 3, 2017 2. Section 3.3 of the LAR states that a finding closure review was conducted on the internal events (including internal floods) PRA (IEPRA) model in June 2017 and for the seismic PRA (SPRA) in April 2017. In the LAR, the licensee further confirms that the closed findings were reviewed and closed using the process documented in Appendix X to NEI 05-04, NEI 0-12, and NEI 12-13, as accepted by the NRC.

In DRA RAI 01, the NRC staff requested the licensee to either confirm that the Independent Assessment team was provided with and performed an independent written assessment that included justification of whether the resolution for each F&O constituted a PRA upgrade or maintenance update, as defined in the American Society for Mechanical Engineers/American Nuclear Society (ASME/ANS) RA-Sa-2009 PRA Standard and endorsed by Regulatory Guide (RG) 1.200, Revision 2 or perform a subsequent Independent Assessment for F&O closure or 1 Errata in title: Anderson, V. K., Nuclear Energy Institute, letter to Stacey Rosenberg, U.S. Nuclear Regulatory Commission, Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations, dated February 21, 2017 (ADAMS Package Accession No. ML17086A431).

2 Giitter, J., and Ross-Lee, M. J., U.S. Nuclear Regulatory Commission, letter to Mr. Greg Krueger, Nuclear Energy Institute, U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 07-12, and 12-13, Close-Out of Facts and Observations (F&Os), dated May 3, 2017 (ADAMS Accession No. ML17079A427).

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Enclosure 1 addendum to the report to address the inconsistency with Appendix X, as accepted, with conditions, by the NRC staff.

In response to DRA RAI-01.a, the licensee stated, in part:

[T]he information provided did not include a written assessment and justification of whether the resolution of each F&O, within the scope of the IA, constitutes a PRA upgrade or maintenance update as required by NEI 05-04 Appendix X Section X.1.3.

The absence of this update/upgrade self-assessment did not negatively impact the ability of the IA team in performance of their review because the team based their conclusions on the merits for each F&O resolution. The teams assessment of whether the resolution constituted a PRA upgrade or maintenance update is based on consensus of the Independent Assessment team.

Without a written assessment and proper justification of whether the resolution of each F&O constitutes a PRA upgrade or maintenance update, the process performed for closure of the F&Os is not consistent with the Appendix X process as accepted, with conditions by the NRC staff. Material that is provided to a reviewer that is not complete in its entirety can lead to decisions made prematurely, in error, and are not transparent or traceable. To address the discrepancy identified, provide either of the following:

i. Perform a subsequent Independent Assessment for F&O(s) closure to the Independent Assessment report to address the inconsistency with Appendix X, as accepted, with conditions, by the NRC staff via letter dated May 3, 2017.

OR ii. Provide all F&Os (i.e., intended to be closed out by the Independent Assessment) along with a disposition for each F&O that describes the impact on the 10 CFR 50.69 categorization process and justification for why it is appropriate.

TVA Response to DRA RAI 01-01 TVA is responding to DRA RAI 01-01.i, which requested TVA to perform a subsequent independent assessment (IA) for F&O closure to the Independent Assessment Report to address the inconsistency with Appendix X with respect to the licensee-provided assessment of upgrade/update for each resolved F&O.

In response to the NRC request, TVA re-constituted the IA team and provided the IA team with the licensees written characterization of whether the resolutions for addressing the F&Os constituted an upgrade or an update activity, as defined in the ASME/ANS RA-Sa-2009 PRA Standard. The re-constituted IA team consisted of the same individuals that performed the initial F&O closure review in 2017. The purpose of this review was to consider the TVA input concerning the upgrade/update characterization. The revised F&O Closure Report acknowledges receipt of TVAs self-assessment. Furthermore, the IA teams rationale for the upgrade/update characterization has been refined to improve the documentation. The documentation for each closed F&O was enhanced to more clearly show that the F&O resolution met the Capability Category II requirements of the ASME/ANS PRA Standards Supporting Requirements (SRs) that were referenced in the F&O. The performance of a PRA upgrade or the use of new methods would require that a peer review be performed instead of a findings closure review. None of the changes made to the Watts Bar PRA were considered by CNL-19-108 E1-2 of 9

Enclosure 1 the review team to constitute a PRA upgrade or the usage of a new PRA method. Therefore, the activities described above close the gap identified by RAI 01-01.

DRA RAI 05 Dispositions of Key Assumptions and Sources of Uncertainties (APLA/RILIT)

The regulations at 10 CFR 50.69(e) require periodic updates and, if necessary, changes to the categorization process and treatment of structures, systems, and components (SSCs). Section 3.3.2 of RG 1.200, Revision 2, states that [f]or each application that calls upon this regulatory guide, the applicant identifies the key assumptions and approximations relevant to that application. This will be used to identify sensitivity studies as input to the decision-making associated with the application. Section 5 of NEI 00-04 identifies sensitivity studies related to the technical adequacy of the PRA as part of the categorization process.

The key assumptions and sources of uncertainties identified as part of the LAR may change because updates to the PRAs supporting this application (i.e., internal events, including internal floods, and the SPRA) could affect the significance of those assumptions or create new key assumptions or sources of uncertainties for this application. DRA RAI 05.b requested the licensee to describe how the evaluation for key assumptions and sources of uncertainty is updated and modified for the PRAs supporting this application.

In response to DRA RAI 05.b, the licensee stated that a key assumption or source of uncertainty would be updated if the estimated cumulative impact exceeds the threshold of a 25 percent change to the baseline core damage frequency (CDF) or large early release frequency (LERF).

The risk-informed categorization of SSCs in accordance with 10 CFR 50.69 is based on importance measures, not changes in CDF or LERF (i.e., delta CDF and delta LERF). Changes to importance measures, and consequently SSC categorization, as well as treatment of SSCs, can be impacted before reaching the threshold of a 25 percent change to the baseline CDF or LERF. To address the discrepancy identified, provide either of the following:

i. Justify how a threshold of a 25 percent change to the baseline CDF or LERF is appropriate and bounding for determining changes to the importance measures for SSCs arising from changes to key assumptions and sources of uncertainty, and therefore, not adversely impacting the categorization and treatment of SSCs.

OR ii. Alternatively, describe an approach that will continue to evaluate assumptions and sources of uncertainty for categorization of SSCs, including the identification of those key to the application, when WBN PRA models are updated in the future for the PRAs supporting this application (i.e., IEPRA (includes internal floods), and the SPRA) consistent with the guidance in NEI 00-04.

TVA Response to DRA RAI 05-01 TVA is responding to DRA RAI 05-01.ii. In order to establish an approach that will continue to evaluate assumptions and sources of uncertainty for categorization of SSCs, TVA proposes to revise Attachment 1 to Enclosure 1 of the referenced letter, to state:

Consistent with the guidance in NEI 00-04, TVA shall establish in the categorization procedure requirements for periodic re-assessment. The PRA staff will reassess all previously categorized systems whenever a change is made to any key source of CNL-19-108 E1-3 of 9

Enclosure 1 uncertainty or assumption listed in Attachment 6 of the original LAR (CNL-18-068 dated November 29, 2018). Additionally, the procedure will require evaluation of model changes to identify changes that introduce new key sources of uncertainty and assumptions with respect to the application.

The purpose of the reassessment is to confirm SSCs having a Low Safety Significance do not meet the criteria for High Safety Significance.

Reference TVA letter to NRC, CNL-19-065, Partial Response to NRC Request for Additional Information Regarding Watts Bar Nuclear Plant, Units 1 and 2, Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors (WBN-TS-17-24) (EPID L-2018-LLA-0493), dated July 15, 2019 (ML19196A362)

DRA RAI 07 Integrated PRA Hazards Model (APLA/RILIT)

The categorization of SSCs, including those categorized using the SPRA, is based on importance measures and corresponding numerical criteria, as described in Sections 5.1 and 5.3 of NEI 00-04. Further, Section 5.6 of NEI 00-04 discusses the "integral assessment" wherein the hazard specific importance measures are weighted by the hazards contribution to the plant risk. DRA RAI 07.c requested information on how the importance measures are determined for the SPRA, considering that the seismic hazard is discretized into bins. DRA RAI 07c further requested the licensee to describe and justify how the same basic events, which were discretized by binning during the development of the SPRA, are then combined to develop representative importance measures, as well as to discuss and justify how the importance measures are compared to the numerical criteria consistent with the guidance in NEI 00-04.

The licensees response to DRA RAI 07.c stated that the licensee will not assess any importance measures based on a PRA one-top all hazards model. However, DRA RAI 07.c included a request for information on how importance measures are derived from the SPRA considering that the seismic hazard is discretized into bins, including discussion of how the same basic events, which were discretized by binning during the development of the SPRA, are then combined (i.e., combined across bins as well as across failure modes such as seismic and random failures) to develop representative importance measures. The licensees response to DRA RAI-07.c did not provide any relevant information and the approach is not discussed in the LAR. Therefore, the NRC staff remains unclear on the approach that will be followed by the licensee and its alignment with the guidance in NEI 00-04.

a. Describe how the importance measures (i.e., Fussell-Vesely [FV] and Risk Achievement Worth [RAW]) are derived from the SPRA considering that the seismic hazard is discretized into bins. The discussion should include how the same basic events, which were discretized by binning during the development of the SPRA, are then combined (i.e., combined across bins as well as across failure modes such as seismic and random failures) to develop representative importance measures. Further, discuss how they are compared to the importance measure thresholds in NEI 00-04. Provide justification to support the determined impact on the categorization results and describe how the approach is consistent with the guidance in NEI 00-04.

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Enclosure 1 Paragraph (c)(1)(ii) of 10 CFR 50.69 requires that the SSC functional importance be determined using an integrated, systematic process. NEI 00-04, Section 5.6, Integral Assessment, discusses the need for an integrated computation using available importance measures. The licensees response to DRA RAI 07d stated that the importance evaluations are performed in accordance with NEI 00-04 and that some components in the Internal Events PRA may not be explicitly modeled in the SPRA. However, DRA RAI 07d included a request for information on how the integrated importance measures are calculated for certain components where corresponding basic events, which represent different failure modes for a component, in the SPRA may not align with basic events in other PRA modeled hazards. The licensees response to DRA RAI-07.d did not provide any information relevant to that request and the approach is not apparent from the LAR. Therefore, the NRC staff remains unclear on the approach that will be followed by the licensee.

b. Provide details and justification to support how the integrated importance measures will be calculated for the SPRA modeled basic events that may not align directly with basic events modeled in other PRA hazards. Include discussion for any mapping that will be performed across the SPRA basic events and those in other PRA modeled hazards where additional modelling is determined to be necessary.

TVA Response to DRA RAI 07-01 TVA Response to DRA RAI 07-01a The following information describes how the importance measures (i.e., FV and RAW) are derived from the SPRA considering that the seismic hazard being discretized into bins.

TVA uses the FRANX computer code to quantify the WBN SPRA. For the WBN SPRA, the seismic hazard curve is subdivided into eight bins, covering the peak ground acceleration range from 0.09 g (the WBN operational basis earthquake or OBE) to > 3.0 g. The FRANX code automatically assigns a representative mean peak ground acceleration (PGA) for each bin based on the bounds of the bin. Note that top bin (%G08) is unbounded. The FRANX code automatically assigns a representative mean PGA for bin %G08 to 10% higher than the lower boundary, so that bin mean PGA is 3.3 g. The quantification of seismic risk (CDF and LERF) is performed individually for each bin for each of the two units.

Once all of the bins for a given unit for a particular risk measure (CDF or LERF) are quantified, the cutsets for a given risk measure and unit are grouped together using the FRANX code, which weights each cutset with the bin frequency (a basic event representing the respective bin frequency is appended to the cutset). After the quantification and grouping calculations are complete, there are four cutset files (U1 CDF, U2 CDF, U1 LERF, and U2 LERF). These cutset files are then post-processed using ACUBE 2.0 to determine the true value of the risk measure.

The SYSIMP code, which also uses ACUBE 2.0, is used to calculate the risk importance measures (i.e., FV and RAW) of each component for CDF and LERF for each unit separately.

Within SYSIMP, all failure modes are mapped to the components unique identifier. The mapping includes all random failures (e.g., fail to run, fail to start, fail to open, fail to close, spurious operation) and all seismic failures (e.g., seismic failure in bin %G01, %G02, %G03).

This process accounts for every failure mode in the SPRA and the frequency of the discretized bins. The risk measures for non-seismic (random) common cause failures events are calculated separately since these failure modes have a different threshold for High Safety Significance (HSS).

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Enclosure 1 The weighted SPRA importance measures for individual components and random common cause failures events developed above are compared to the following criteria which provides margin to the NEI 00-04 criteria. The NEI 00-04 criteria is provided in brackets following the values used for the SPRA importance measure.

  • individual component failures events with FV > 0.004 [0.005] are initially categorized as HSS;
  • individual components with a RAW > 1.8 [2.0] are initially categorized as HSS;
  • random common cause failures events with FV > 0.004 [0.005] are initially categorized as HSS;
  • random common cause failures events with RAW > 18 [20] are initially categorized as HSS.

A margin to the NEI 00-04 criteria is recommended by NEI 16-09 "Risk-Informed Engineering Programs (10 CFR 50.69) Implementation Guidance" to minimize the possibility of future PRA model changes resulting in critical changes where the categorization of an SSC is moved from Low Safety Significance (LSS) to HSS. The process for SPRA is both a conservative and consistent approach when compared to the NEI 00-04 process, since lower risk importance thresholds are used to account for key sources of uncertainty.

TVA Response to DRA RAI 07-01b As discussed in the response to DRA RAI 07-01a, the importance evaluations performed in accordance with the process in NEI 00-04 are determined on a component basis. It is not essential that there be complete alignment among the basic events that are pertinent to a given component from one hazard PRA to another, i.e., there may be hazard-specific basic events whose importance contributions are captured within the component importance calculations for that hazard.

A large majority of SPRA basic events is directly aligned with the basic events in other PRA models, and are combined using the formulae in Section 5.6 of NEI 00-04. However, as noted in this RAI, there are a few SSCs in the SPRA that are not directly included in the other PRA models.

Subcomponents The importance of a subcomponent that was not directly modeled in other PRAs will be accounted for in the importance calculation for the component to which it is associated because it can be treated as another failure mode of that component. For example, the steam admission valve for the auxiliary feedwater pump is modeled in the IEPRA, Internal Flooding (IF) PRA, and the SPRA. Seismic-induced relay chatter, a failure mechanism unique to the SPRA, could cause the valve to close, stopping the flow of steam to the pump, which can cause the pump to fail. In other PRAs, this relay was considered part of the valve boundary and its failures are inherently accounted for in the valve failure probability and associated component importance.

For the SPRA, the relay was directly modeled to spuriously close the valve. The SPRA importance of the relay would be considered as a contributor to the valve failure and accounted for appropriately within the valve's importance measures for the integrated importance measures assessment, following the process in NEI 00-04. The decision on the need to treat seismic basic events as representing subcomponents within the importance calculations for another modeled component will be made based on the modeling in each of the PRAs, as part of the PRA basic event-to-component mapping within the categorization process.

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Enclosure 1 SSCs Not in Other PRA Models While most of the SSCs in the SPRA are directly aligned with SSCs in the other PRAs (internal events, internal flooding), there are some SSCs that are unique to the SPRA. These SSCs may have been screened out of the other PRAs, following the PRA modeling requirements in the ASME/ANS PRA Standard, based on their having no credible failure mode (or an extremely low probability of failure). If these SSCs are HSS for the SPRA, then their integrated safety significance computation is not necessary. The safety significance would be presented to the Integrated Decisionmaking Panel (lDP) for their consideration in the decision-making process.

The NEI 00-04 process allows the IDP to adjust significance of a SPRA modeled SSCs using proper justification. The quantitative integrated importance measure assessment is only one portion of the categorization process.

The following examples demonstrate how the SSCs that are only in the SPRA would be treated for the importance analysis. Some components only appear in the SPRA, because they do not have a credible failure mode in other PRAs, or have been screened for other reasons.

  • Structures are not directly included in the other WBN PRA models because there is no credible failure mode, but some structures are included in the SPRA. If these structures are HSS in the SPRA, then their integrated safety significance computation is not necessary. The safety significance would be presented to the IDP for their consideration in the decision-making process. The NEI 00-04 process allows the IDP to adjust significance of a SPRA modeled SSCs using proper justification.
  • SPRA specific SSCs: Some components only appear in the SPRA, because they do not have a credible failure mode in other PRAs, or have been screened for other reasons.

These components will be treated as separate components for the integral importance measure assessment. For example, components such as cable trays, conduits, motor control centers, electrical cabinets and panels, Heating, Ventilating, and Air Conditioning (HVAC) Cable Tray ducting, and piping were not included in the IE PRA because they are passive components. However, for the SPRA, their seismic anchorage failure could potentially fail modeled components. For 50.69 categorization, the associated seismic failure events would be categorized based on their impact on modeled function. The integral importance assessment would not change this categorization.

In summary, most of the seismic basic event importance measures can be directly aligned with components in the other PRAs. Those seismic basic events that are not explicitly modeled in other PRAs, but function as subcomponents of components modeled in other PRAs, will have their seismic importance measures combined with the other PRA importance measures using the NEI 00-04 formulae for the integral assessment. An integrated safety significance computation is not necessary for other seismic basic events that are not explicitly modeled in the IE or IF PRAs (such as structures and SPRA unique components). The safety significance would be presented to the IDP for their consideration in the decision-making process. The NEI 00-04 process allows the IDP to adjust significance of a SPRA modeled SSCs using proper justification.

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Enclosure 1 DRA RAI 12 Propagation of Closed and Open/Partially Open Findings from DRA RAI 08 (RILIT)

The regulations at 10 CFR 50.69(e) require periodic updates and, if necessary, changes to the categorization process and treatment of SSCs. According to Sections 7-1.2 and 8-1.2 of the 2009 ASME/ANS PRA Standard, it is assumed that a full-scope internal-events at-power Level 1 and Level 2 LERF PRAs exist and that those PRAs are used as the basis for the SPRA.

DRA RAI 12.b requested the licensee to describe how changes to the IEPRA (which includes internal floods) arising from the review of this application, as part of any implementation item resulting from this application, or as part of routine maintenance and updating of the IEPRA (includes internal floods) will be propagated to the SPRA used to support this application. In its response to DRA RAI 12b, the licensee stated that changes to the PRAs supporting this application would be performed if the estimated cumulative impact exceed the threshold of a 25 percent change to the baseline CDF or LERF. The risk-informed categorization of SSCs in accordance with 10 CFR 50.69 is based on importance measures, not changes in CDF or LERF (i.e., delta CDF and delta LERF). Changes to importance measures, and, consequently, SSC categorization, as well as treatment of SSCs, can be impacted before reaching the threshold of a 25 percent change to the baseline CDF or LERF. To address the discrepancy identified, provide either of the following:

i. Justify how a threshold of a 25 percent change to the baseline CDF or LERF is appropriate and bounding for determining the impact of changes to PRAs supporting this application on the importance measures and therefore, the categorization and treatment of SSCs consistent with the requirements in 10 CFR 50.69(e) and the guidance in NEI 00-04.

OR ii. Alternately, describe an approach that will propagate changes in the internal events, including internal flooding, PRA to the SPRA that is consistent with the requirements in 10 CFR 50.69(e) and the guidance in NEI 00-04 for appropriate categorization of SSCs.

TVA Response to DRA RAI 12-01 TVA is responding to DRA RAI 12-01.ii. TVA maintains a living model that assesses the change in risk due to changes in the as-built, as-operated plant. In order to establish an approach that will appropriately propagate changes in the IEPRA and internal flooding PRA to the SPRA, TVA proposes to revise Attachment 1 to Enclosure 1 of the referenced letter, to state:

TVA shall assess the impact on the internal events with internal flooding living model with respect to the risk importance measures used to assign the safety classification (high or low) from pending model changes to be compared to previously categorized system SSCs to confirm that the criteria for Low Safety and High Safety Significance is still applicable, and reclassify, if necessary, in accordance with NEI 00-04 (i.e., PRA model update, and at least once per two fuel cycles in a unit).

This approach would be covered by procedure and presented to the IDP for concurrence. As such, this routine periodicity would be independent to the 25% threshold for an off-cycle model update. This approach is judged to be in alignment with 10 CFR 50.69(e) and the guidance of NEI 00-04.

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Enclosure 1 Reference TVA letter to NRC, CNL-19-065, Partial Response to NRC Request for Additional Information Regarding Watts Bar Nuclear Plant, Units 1 and 2, Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors (WBN-TS-17-24) (EPID L-2018-LLA-0493), dated July 15, 2019 (ML19196A362)

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Enclosure 2 Attachment 1: List of Categorization Prerequisites TVA will establish procedure(s) prior to the use of the categorization process on a plant system. The procedure(s) will contain the elements/steps listed below.

  • IDP member qualification requirements.
  • Qualitative assessment of system functions. System functions are qualitatively categorized as preliminary HSS or LSS based on the seven criteria in Section 9 of NEI 00-04 (see Section 3.2 of this enclosure). Any component supporting an HSS function is categorized as preliminary HSS. Components supporting, an LSS function are categorized as preliminary LSS.
  • Component safety significance assessment. Safety significance of active components is assessed through a combination of PRA and non-PRA methods, covering all hazards.

Safety significance of passive components is assessed using a methodology for passive components.

  • Assessment of DID and safety margin. Safety-related components that are categorized as preliminary LSS are evaluated for their role in providing DID and safety margin and, if appropriate, upgraded to HSS.
  • Review by the IDP. The categorization results are presented to the lDP for review and approval. The lDP reviews the categorization results and makes the final determination on the safety significance of system functions and components.
  • Risk sensitivity study. For PRA-modeled components, an overall risk sensitivity study is used to confirm that the population of preliminary LSS components results in acceptably small increases to CDF and LERF and meets the guidelines of Regulatory Guide 1.174.
  • Periodic reviews are performed to ensure continued categorization validity and acceptable performance for those SSCs that have been categorized. This includes:

- Consistent with the guidance in NEI 00-04, TVA shall establish in the categorization procedure requirements for periodic re-assessment. The PRA staff will reassess all previously categorized systems whenever a change is made to any key source of uncertainty or assumption listed in Attachment 6 of the original LAR (CNL-18-068 dated November 29, 2018). Additionally, the procedure will require evaluation of model changes to identify changes that introduce new key sources of uncertainty and assumptions with respect to the application.

  • TVA shall assess the impact on the internal events with internal flooding living model with respect to the risk importance measures used to assign the safety classification (high or low) from pending model changes to be compared to previously categorized system SSCs to confirm that the criteria for Low Safety and High Safety Significance is still applicable, and reclassify, if necessary, in accordance with NEI 00-04 (i.e., PRA model update, and at least once per two fuel cycles in a unit).
  • Documentation requirements per Section 3.1.1 of the enclosure to CNL-18-068.

In addition to the procedure changes above, TVA will also perform the following actions.

CNL-19-108 E2-1 of 2

Enclosure 2

  • As documented in the F&O Closure Report, all changes initiated by the F&O resolutions were confirmed by the Integrated Assessment Team to have been incorporated into the living model and associated documentation. TVA shall update the Model of Record (MOR) with this information prior to system categorization.
  • TVA shall re-introduce the State of Knowledge Correlation (SOKC) into the MOR prior to using the PRA model to support categorization of SSCs under 10 CFR 50.69.
  • With respect to the external flooding hazards, TVA shall re-confirm that there is sufficient time to eliminate the source of the threat or to provide an adequate response in accordance with screening criterion C5, prior to 50.69 categorization.

CNL-19-108 E2-2 of 2

Enclosure 3 WBN Units 1 and 2 Facility Operating Licenses Changes Markup CNL-19-108

4b (10) By May 31, 2018, TVA shall ensure that a listing organization acceptable to the NRC (as the Authority Having Jurisdiction) determines that the fire detection monitoring panel in the main control room either meets the appropriate designated standards or has been tested and found suitable for the specified purpose.

(11) The licensee shall replace the WBN, Unit 1 upper compartment cooler cooling coils with safety-related cooling coils to eliminate a potential source of containment sump dilution during design basis events prior to increasing the number of Tritium Producing Burnable Absorber Rods (TPBARs) loaded in the reactor core above 704.

(12) Adoption of 10 CFR 50.69, Risk-Informed categorization and treatment of structures, systems and components for nuclear power plants (a) TVA 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) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and seismic hazards; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards; fire hazards by use of the fire protection program (FPP) safe shutdown equipment list (SSEL), and a screening of other external hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009, as specified in Unit 1 License Amendment [Number].

(b) Prior to implementation of the provisions of 10 CFR 50.69, TVA shall complete the implementation items in Enclosure 2, Attachment 1, "List of Categorization Prerequisites," to TVA letter CNL-19-108, Response to NRC Second Request for Additional Information Regarding Watts Bar Nuclear Plant, Units 1 and 2, Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors (WBN-TS-17-24) (EPID L-2018-LLA-0493), dated October 28, 2019.

(13) Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from using the FPP SSEL approach to an internal fire probabilistic risk assessment approach).

D. The following exemptions are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security. Therefore, these exemptions are granted pursuant to 10 CFR 50.12.

(1) Deleted Facility License No. NPF-90 Amendment 107, XXX

TVA may make changes to the approved fire protection program without prior approval of the Commission, only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

(9) By May 31, 2018, TVA shall report that a listing organization acceptable to the NRC (as the Authority Having Jurisdiction) has determined that the fire detection monitoring panel in the main control room either meets the appropriate designated standards or has been tested and found suitable for the specified purpose.

(10) TVA will verify for each core reload that the actions taken if FQW(Z) is not within limits will assure that the limits on core power peaking FQ(Z) remain below the initial total peaking factor assumed in the accident analyses.

(11) TVA will implement the compensatory measures described in Section 3.4, Additional Compensatory Measures, of TVA Letter CNL-18-012, dated January 17, 2018, during the timeframe the temperature indicator for RCS hot let 3 is not required to be operable for the remainder of Cycle 2. If the RCS hot leg 3 temperature indicator is returned to operable status prior to the end of Cycle 2, then these compensatory measures are no longer required.

(12) Adoption of 10 CFR 50.69, Risk-Informed categorization and treatment of structures, systems and components for nuclear power plants (a) TVA 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) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and seismic hazards; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards; fire hazards by use of the fire protection program (FPP) safe shutdown equipment list (SSEL), and a screening of other external hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009, as specified in Unit 2 License Amendment [Number].

(b) Prior to implementation of the provisions of 10 CFR 50.69, TVA shall complete the implementation items in Enclosure 2, Attachment 1, "List of Categorization Prerequisites " to TVA letter CNL-19-108, Response to NRC Second Request for Additional Information Regarding Watts Bar Nuclear Plant, Units 1 and 2, Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors (WBN-TS-17-24)

(EPID L-2018-LLA-0493), dated October 28, 2019.

Amendment 19, XXX

(13) Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from using the FPP SSEL approach to an internal fire probabilistic risk assessment approach).

D. The licensee shall have and maintain financial protection of such types and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.

E. This license is effective as of the date of issuance and shall expire at midnight on October 21, 2055.

FOR THE NUCLEAR REGULATORY COMMISSION William M. Dean, Director Office of Nuclear Reactor Regulation Appendices: 1. Appendix A -

Technical Specifications

2. Appendix B -

Environmental Protection Plan Date of Issuance: October 22, 2015 Amendment 19, XXX

Enclosure 4 WBN Units 1 and 2 Facility Operating Licenses Changes Retyped Copy CNL-19-108

4b (10) By May 31, 2018, TVA shall ensure that a listing organization acceptable to the NRC (as the Authority Having Jurisdiction) determines that the fire detection monitoring panel in the main control room either meets the appropriate designated standards or has been tested and found suitable for the specified purpose.

(11) The licensee shall replace the WBN, Unit 1 upper compartment cooler cooling coils with safety-related cooling coils to eliminate a potential source of containment sump dilution during design basis events prior to increasing the number of Tritium Producing Burnable Absorber Rods (TPBARs) loaded in the reactor core above 704.

(12) Adoption of 10 CFR 50.69, Risk-Informed categorization and treatment of structures, systems and components for nuclear power plants (a) TVA 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) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and seismic hazards; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards; fire hazards by use of the fire protection program (FPP) safe shutdown equipment list (SSEL), and a screening of other external hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009, as specified in Unit 1 License Amendment [Number].

(b) Prior to implementation of the provisions of 10 CFR 50.69, TVA shall complete the implementation items in Enclosure 2, Attachment 1, "List of Categorization Prerequisites," to TVA letter CNL-19-108, Response to NRC Second Request for Additional Information Regarding Watts Bar Nuclear Plant, Units 1 and 2, Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors (WBN-TS-17-24) (EPID L-2018-LLA-0493), dated October 28, 2019.

(13) Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from using the FPP SSEL approach to an internal fire probabilistic risk assessment approach).

D. The following exemptions are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security. Therefore, these exemptions are granted pursuant to 10 CFR 50.12.

(1) Deleted Facility License No. NPF-90 Amendment 107, XXX

TVA may make changes to the approved fire protection program without prior approval of the Commission, only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

(9) By May 31, 2018, TVA shall report that a listing organization acceptable to the NRC (as the Authority Having Jurisdiction) has determined that the fire detection monitoring panel in the main control room either meets the appropriate designated standards or has been tested and found suitable for the specified purpose.

(10) TVA will verify for each core reload that the actions taken if FQW(Z) is not within limits will assure that the limits on core power peaking FQ(Z) remain below the initial total peaking factor assumed in the accident analyses.

(11) TVA will implement the compensatory measures described in Section 3.4, Additional Compensatory Measures, of TVA Letter CNL-18-012, dated January 17, 2018, during the timeframe the temperature indicator for RCS hot let 3 is not required to be operable for the remainder of Cycle 2. If the RCS hot leg 3 temperature indicator is returned to operable status prior to the end of Cycle 2, then these compensatory measures are no longer required.

(12) Adoption of 10 CFR 50.69, Risk-Informed categorization and treatment of structures, systems and components for nuclear power plants (a) TVA 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) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and seismic hazards; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards; fire hazards by use of the fire protection program (FPP) safe shutdown equipment list (SSEL), and a screening of other external hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009, as specified in Unit 2 License Amendment [Number].

(b) Prior to implementation of the provisions of 10 CFR 50.69, TVA shall complete the implementation items in Enclosure 2, Attachment 1, "List of Categorization Prerequisites " to TVA letter CNL-19-108, Response to NRC Second Request for Additional Information Regarding Watts Bar Nuclear Plant, Units 1 and 2, Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors (WBN-TS-17-24)

(EPID L-2018-LLA-0493), dated October 28, 2019.

Amendment 19, XXX

(13) Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from using the FPP SSEL approach to an internal fire probabilistic risk assessment approach).

D. The licensee shall have and maintain financial protection of such types and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.

E. This license is effective as of the date of issuance and shall expire at midnight on October 21, 2055.

FOR THE NUCLEAR REGULATORY COMMISSION William M. Dean, Director Office of Nuclear Reactor Regulation Appendices: 1. Appendix A -

Technical Specifications

2. Appendix B -

Environmental Protection Plan Date of Issuance: October 22, 2015 Amendment 19, XXX

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