ML21041A543
| ML21041A543 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 02/05/2021 |
| From: | Michael Wentzel NRC/NRR/DORL/LPL2-2 |
| To: | Orf T Tennessee Valley Authority |
| References | |
| L-2020-LLA-0162 | |
| Download: ML21041A543 (10) | |
Text
From: Wentzel, Michael Sent: Friday, February 5, 2021 7:06 AM To: Orf, Tracy J
Subject:
Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Draft Request for Additional Information Regarding 10 CFR 50.69 LAR (EPID L-2020-LLA-0162)
Attachments: Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Draft Request for Information Regarding LAR to Adopt 10 CFR 50.69 (L-2020-LLA-0162).docx
- Trace, By letter dated July 17, 2020 (Agencywide Documents Access and Management System Accession No. ML20199M373), , the Tennessee Valley Authority (TVA) submitted a license amendment request for Browns Ferry Nuclear Plant, Units 1, 2, and 3. The proposed amendments would allow for the voluntary adoption of the requirements of Title 10 of the Code of Federal Regulations, Section 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors.
The NRCs Probabilistic Risk Assessment Licensing Branches A and C (APLA and APLC) staff have reviewed the application and identified areas where they need additional information to support its review. The draft request for additional information (RAI) is provided in the attached.
Please let me know by February 12, 2021, if a clarification call is needed. If a clarification call is not needed, please let me know if TVA can respond to the RAI within 30 days from the date of final issuance.
Sincerely, Michael Wentzel, Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Hearing Identifier: NRR_DRMA Email Number: 1014 Mail Envelope Properties (BLAPR09MB726771C002A3A4277DA6836AE7B29)
Subject:
Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Draft Request for Additional Information Regarding 10 CFR 50.69 LAR (EPID L-2020-LLA-0162)
Sent Date: 2/5/2021 7:06:06 AM Received Date: 2/5/2021 7:06:06 AM From: Wentzel, Michael Created By: Michael.Wentzel@nrc.gov Recipients:
"Orf, Tracy J" <tjorf@tva.gov>
Tracking Status: None Post Office: BLAPR09MB7267.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 1169 2/5/2021 7:06:06 AM Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Draft Request for Information Regarding LAR to Adopt 10 CFR 50.69 (L-2020-LLA-0162).docx 32299 Options Priority: Normal Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:
REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUEST TO ADOPT 10 CFR 50.69, RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURE, SYSTEMS AND COMPONENTS FOR NUCLEAR POWER REACTORS TENNESSEE VALLEY AUTHORITY BROWNS FERR NUCLEAR PLANT, UNITS 1, 2, AND 3 DOCKET NOS. 50-259, 50-260, and 50-296 By letter dated July 17, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20199M373), the Tennessee Valley Authority (TVA, the licensee) submitted a license amendment request (LAR) for Browns Ferry Nuclear Plant, Units 1, 2, and 3 (Browns Ferry). The proposed amendments would allow for the voluntary adoption of the requirements of Title 10 of the Code of Federal Regulations (10 CFR),
Section 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors, at Browns Ferry. The NRC staff is reviewing the LAR and has identified where additional information is needed to complete its review. The NRC staffs request for additional information (RAI) is below.
RAI 01 - Open PRA Facts and Observations (F&O)
Regulatory Guide (RG) 1.200, Revision 2 (ADAMS Accession No. ML090410014) provides guidance for addressing probabilistic risk assessment (PRA) acceptability. RG 1.200, Revision 2, describes a peer-review process using the ASME/ANS PRA standard ASME/ANS-RA-Sa-2009, Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, as one acceptable approach for determining the technical acceptability of the PRA. The primary results of peer review are the facts and observations (F&Os) recorded by the peer review team and the subsequent resolution of these F&Os. Appendix X to the Nuclear Energy Institute (NEI) guidance documents NEI 05-04, NEI 07-12, and NEI 12-13, titled NEI 05-04/07-12/12-06 Appendix X: Close-out of Facts and Observations (F&Os)
(ADAMS Package Accession No. ML17086A431), which was accepted by the NRC in a letter dated May 3, 2017 (ADAMS Accession No. ML17079A427), describes a process to close finding-level F&Os.
LAR Attachment 3 presents all findings not closed by the F&O closure review for both internal events and fire PRA (IEPRA and FPRA). The NRC staff reviewed the open findings and associated dispositions and has identified that the licensees stated disposition for each F&O is to resolve the finding and have it closed in accordance with an Independent Assessment. The NRC staff notes that the last item in the Categorization Prerequisite List regards a commitment to close all open F&Os listed in [LAR] Attachment 3 and incorporate changes into the MOR
[model of record] prior to system categorization. Thus, the license condition proposed in LAR does not appear to commit to close the open findings using an Independent Attachment
Assessment, as stated in the F&O dispositions. In addition, the NRC staff notes that the resolutions to many of the internal events F&Os appear to have the potential to impact the FPRA. However, the LAR does not indicate that the FPRA will be updated to incorporate the resolutions to the internal events findings. Additionally, the NRC staff observed the TVAs F&O Closure Review conducted December 7-10, 2020, which appeared to close many of the F&Os provided in the LAR.
In light of these observations, address the following:
a) Provide any necessary updates to the status of open F&Os.
b) Confirm that the license condition is meant to commit to resolving all the open IEPRA and FPRA F&Os and to closing them with an Appendix X Independent Assessment process, as accepted by the NRC in letter dated May 3, 2017, and, as applicable, update the wording in the license condition or in LAR Attachment 1, accordingly.
c) Alternatively, if the intention is not to commit to resolving the open IEPRA and FPRA F&Os and closing them with an Independent Assessment, then update the dispositions for the open F&Os and provide justification that the existing PRA modeling associated with the open F&Os does not impact this application.
d) Confirm that the resolutions to the internal events findings with the potential to impact the FPRA modeling will be incorporated into the FPRA.
e) If it is confirmed in the response to part (c), above, that the resolutions to the internal events findings with the potential to impact the FPRA modeling will be incorporated into the FPRA, then propose a mechanism that ensures that the internal events findings with the potential to impact the FPRA modeling will be incorporated into the FPRA prior to the implementation of the risk categorization program.
f) If it is not confirmed in the response to part (c), above, that the resolutions to the internal events findings with the potential to impact the FPRA modeling will be incorporated into the FPRA, then justify that incorporation of the IEPRA finding resolutions into the FPRA would have no impact on the risk categorization program.
RAI 02 - Identification of Key Assumptions and Sources of Uncertainties Paragraphs (c)(1)(i) and (ii) of 10 CFR 50.69 require that a licensees PRA be of sufficient quality and level of detail to support the structure, system, and component (SSC) categorization process, and that all aspects of the integrated, systematic process used to characterize SSC importance must reasonably reflect the current plant configuration and operating practices, and applicable plant and industry operational experience.
Section 5 of NEI 00-04, Revision 0, 10 CFR 50.69 SSC Categorization Guideline (ADAMS Accession No. ML052900163), provides guidance for performing sensitivity studies for each PRA model to address the uncertainty associated with those models. Specifically, Sections 5.1, 5.2, and 5.3 provide guidance for such sensitivities for the internal events, fire and seismic PRA, respectively. The sensitivity studies are performed to ensure that assumptions and sources of uncertainty (e.g., human error, common cause failure, and maintenance probabilities) do not mask importance of components.
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LAR Section 3.1.10 explains that TVA used the detailed process of identifying, characterizing and qualitative screening of model uncertainties found in Section 5.3 of NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making (ADAMS Accession No. ML17062A466), and Section 3.1.1 of Electric Power Research Institute (EPRI) Technical Report (TR) 1016737, Treatment of Parameter and Modeling Uncertainty for Probabilistic Risk Assessments. LAR Attachment 6 presents a total of six FPRA key assumptions and sources of uncertainty and LAR Section 3.1.10 explains that no internal events or seismic PRA modeling uncertainties key to this application were identified. The LAR states that a list of assumptions and sources of uncertainty were reviewed to identify those which would be significant to this application and that if a non-conservative treatment or a method not commonly accepted were used then it was reviewed for its impact in application. The LAR does not explain how the initial list of assumptions and sources of uncertainty was developed nor does it indicate whether plant-specific issues, generic industry concerns, and modeling choice concerns (e.g., level of detail) were all reviewed to compile this list. It is also not clear to the NRC staff whether other screening criteria beyond identification of non-conservative treatments and uncommon practices were used to screen sources of uncertainty.
Section 3.2.3 of RG 1.200, Revision 2, as well as NUREG-1855, Revision 1, provide guidance on how to identify, characterize, and treat key sources of uncertainty relevant to a risk-informed application. Revision 1 of NUREG-1855 additionally cites EPRI TR 1026511, Practical Guidance on the Use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty. Furthermore, Section 1.3 of NUREG-1855, Revision 1, states, in part, that [a]lthough assumptions and approximations made on the level of detail in a PRA can influence the decision-making process, they are generally not considered to be model uncertainties because the level of detail in the PRA model could be enhanced, if necessary.
Therefore, methods for identifying and characterizing issues associated with level of detail are not explicitly included in NUREG-1855; they are, however, addressed in EPRI TRs 1016737 and 1026511. Additionally, Section 3.3.2 of RG 1.200, Revision 2, defines key assumptions and sources of uncertainty. Therefore, the NRC staff requests the following information to confirm that the key assumptions and sources of uncertainty provided in Attachment 6 of the LAR were properly assessed from the base PRAs that have received peer reviews:
a) Provide a description of the process used to determine the key sources of uncertainty and assumptions for each PRA model used to support this application. The discussion should be provided separately for the IEPRA, FPRA, and seismic PRA (SPRA) and include:
- i. A description of how the key assumptions and sources of uncertainties provided in Attachment 6 were identified from the initial comprehensive list of PRA model(s) (i.e.,
base model) source of uncertainties and assumptions, including those associated with plant-specific features, modeling choices, and generic industry concerns. This can include an identification of the sources of plant-specific and applicable generic modeling uncertainties identified in the uncertainty analyses for the base internal events and internal flooding PRA.
ii. Discussion and justification that the evaluation criteria used to identify an assumption or source of uncertainty as key is consistent with RG 1.200, and/or NUREG-1855, Revision 1, Revision 2, or other NRC-accepted method.
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b) If the process of identifying key assumptions or sources of uncertainty for the PRA models used to support this application cannot be justified for use in the 10 CFR 50.69 categorization process, provide the results of an updated assessment that includes a description of each key assumption or source of uncertainty identified.
RAI 03 - Dispositions of Key Assumptions and Sources of Uncertainties Paragraph (c)(1)(i) of 10 CFR 50.69 requires the licensee to consider the results and insights from the PRA during categorization. The guidance in NEI 00-04 specifies sensitivity studies to be conducted for each PRA model. The sensitivity studies are performed to ensure that assumptions and sources of uncertainty (e.g., human error, common cause failure, and maintenance probabilities) do not mask importance of components. NEI 00-04 guidance states that additional applicable sensitivity studies from characterization of PRA adequacy should be considered.
The NRC staff notes that modeling conservatisms (i.e., assumptions and sources of uncertainty) can mask the importance measures of other SSCs. Sections 5.1 and 5.3 of NEI 00-04 provide guidance on performing individual sensitivity studies for key assumptions and sources of uncertainties as part of the categorization process. Section 3.1.10 of the LAR states that [t]he conclusion of this review is that no additional sensitivity analyses are required to address Browns Ferry PRA model specific assumptions or sources of uncertainty. It is unclear to the NRC staff if any sensitivity studies will be performed for each of the key assumptions and sources of uncertainties provided in Attachment 6 of the LAR and how the determination to either perform or not perform sensitivities was made. Considering these observations, address the following:
a) For any additional key assumptions/sources of uncertainty identified as a result of the response to RAI 02, discuss how each identified key assumption and uncertainty will be dispositioned in the categorization process. The discussion should clarify whether TVA is following the guidance in Section 5 of NEI 00-04 by performing sensitivity analysis or other accepted guidance such as NUREG-1855. The summaries and descriptions should be provided separately for the identified key assumptions and uncertainties related to the IEPRA (includes internal floods), internal FPRA, and SPRA.
LAR Attachment 6 identifies the key assumptions and sources of PRA modeling uncertainty for this application. The NRC staff notes that the LAR presents six key sources of uncertainty for the FPRA and no key sources of uncertainty for the internal events or seismic PRAs. As part of the audit, a comprehensive uncertainty analysis was provided for internal, fire and seismic events that consisted of (1) identification of plant specific assumptions from the PRA notebooks and identification of applicable generic sources of modeling uncertainty from EPRI TRs 1016737 and 1026511, and (2) evaluation and screening of these assumptions and sources of uncertainty to identify key sources of uncertainty. This comprehensive uncertainty analysis was specifically performed for the Browns Ferry TSTF-425 LAR (ADAMS Accession No. ML20087P262). It not clear to the NRC staff whether the conclusions of this analysis are meant to (or do) apply to the 10 CFR 50.69 LAR. The NRC staff notes that the sensitivity of an application to sources of uncertainty can be different for different applications. Therefore, address the following:
b) Clarify whether the uncertainty analysis performed for the TSTF-425 LAR is also the basis for the uncertainty analysis performed for the 10 CFR 50.69. If so, provide justification that disposition of the identified sources uncertainties (especially those 4
identified as potential key sources of uncertainty) are also applicable to the 10 CFR 50.69 LAR.
c) If the uncertainty analysis performed for the TSTF-425 LAR is not the basis for the 10 CFR 50.69 uncertainty analysis, then describe the uncertainty analysis that was performed for the 10 CFR 50.69 LAR and justify why no key sources of uncertainty were identified for the IEPRA or SPRA.
RAI 04 - Total Risk Consideration Revision 3 of RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (ADAMS Accession No. ML17317A2560, provides the risk acceptance guidance in terms of change-in-risk in combination with total core damage frequency defined by regions. These regions are shown in Table 4 and 5 as Region I (No changes allowed), II (Small changes), and III (Very Small Changes and More Flexibility with Respect to Baseline core damage frequency (CDF)/large early release frequency (LERF). NEI 00-04 includes an overall risk sensitivity study for all the Low Safety Significant (LSS) components to assure that if the unreliability of the components was increased, the increase in risk would be small (i.e., meet the acceptance guidelines of RG 1.174, Revision 3).
RG 1.174 and Section 6.4 of NUREG-1855, Revision 1, for a Capability Category II risk evaluation, indicate that the mean values of the risk metrics (total and incremental values) need to be compared against the risk acceptance guidelines. The mean values referred to are the means of the probability distributions that result from the propagation of the uncertainties on the PRA input parameters and model uncertainties explicitly reelected in the PRA models. In general, the point estimate CDF and LERF obtained by quantification of the cutset probabilities using mean values for each basic event probability does not produce a true mean of the CDF/LERF. Under certain circumstances, a formal propagation of uncertainty may not be required if it can be demonstrated that the state of knowledge correlation (SOKC) is unimportant (i.e., the risk results are well below the acceptance guidelines).
The NRC staff notes that the LAR does not stipulate whether the total CDF and LERF values presented in LAR Attachment 2 are mean values and notes there is a small margin between the LERF for Units 1 and 2 of 9.3E-06 and 9.4E-06 per year, respectively, and the RG 1.174, Revision 3 LERF threshold of 1E-05 per year. Accordingly, the risk increase due to consideration of the SOKC and the possible risk increase associated IEPRA and FPRA model updates committed to in the license condition to resolve open F&Os could impact the conclusions of the NEI 00-04 Section 8 overall sensitivity study results by increasing the Browns Ferry LERF values above 1E-05 per year.
In light of the observations above, address the following:
a) Demonstrate that for the NEI 00-04 Section 8 overall sensitivity study results, Browns Ferry will be in conformance with the RG 1.174 risk acceptance guidance after the IEPRA and FPRA models are updated to include the increase associated with SOKC (if needed) and potential increases due to committed PRA updates to resolve F&Os.
b) Alternatively, propose a mechanism that ensures that, for NEI 00-04 Section 8 overall sensitivity study results, Browns Ferry will be in conformance with the RG 1.174 risk acceptance guidance after the IEPRA and FPRA models are updated 5
to include the increase associated with SOKC (if needed) and potential increases due to updates to PRA models performed to resolve F&Os.
c) Explain for the update that was performed or will be performed to address the impact of the SOKC which fire PRA parameters are assumed to be correlated beyond the component failure modes frequencies. Include justification that consideration of the identified parameters is sufficient to estimate the impact of the SOKC on fire risk.
RAI 05 - Categorization of Interfacing SSCs Section 7.1 of NEI 00-04, Revision 0, 10 CFR 50.69 SSC Categorization Guideline (ADAMS Accession No. ML052910035), states, [d]ue to the overlap of functions and components, a significant number of components support multiple functions. In this case, the SSC, or part thereof, should be assigned the highest risk significance for any function that the SSC or part thereof supports. Section 4 of NEI 00-04 states that a candidate low-safety-significant SSC that supports an interfacing system should remain uncategorized until the interfacing system is considered.
The NRC staff notes an apparent inconsistency in LAR Section 3.1.14 concerning treatment of interfacing systems. The sections states that TVA considers its approach to be the same as approved by in the Calvert Cliffs 10 CFR 50.69 Safety Evaluation (SE) (ADAMS Accession No ML19330D909). However, the following partial description of the approach provided in LAR Section 3.1.14 is not consistent with the Calvert Cliffs 10 CFR 50.69 SE:
In some cases, impacts that an interfacing component could have on an interfacing system can be fully determined and the interface component can be categorized (and alternative treatment implemented) without categorizing the entire interfacing system.
In this event, an assessment of interface component risk associated with uncategorized systems will be limited to:
- 1. Cases where an interface component failure cannot prevent performance of interface system functions, or
- 2. The risk is limited to passive failures assessed as low safety-significant following the passive categorization process for the applicable pressure boundary segments.
In the Calvert Cliffs 10 CFR 50.69 SE, the NRC-approved approach specifies that both limitations 1 and 2 above should be true before the interfacing SSC can be categorized without categorizing the entire interfacing system.
Therefore, provide a clarification that reconciles this inconsistency regarding TVAs approach to categorizing interfacing SSCs. If TVAs proposed approach differs from that referenced and approved for Calvert Cliffs, provide further explanation and justification for the proposed approach.
RAI 06 - Integrated PRA Hazards Model 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.
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Section 5.6 further states that the integrated importance measure essentially weighs the importance from each risk contributor (e.g., internal events, fire, seismic PRAs) by the fraction of the total core damage frequency [or large early release frequency] contributed by that contributor. The guidance provides formulas to compute the integrated Fussel-Vesely (FV),
and integrated Risk Achievement Worth (RAW).
LAR Section 3.3 states that the weighted average importance method presented in NEI 00-04, Section 1.5 will be used to integrate seismic PRA results into the overall importance measures.
The licensee cited the response to Watts Bar 10 CFR 50.69 RAI-07 (ADAMS Accession No. ML19196A362) for the integration of risk importance measures across all hazards. The NRC staff notes that SPRA basic events, such as structural failures, may often not align with basic events in other PRA models. The licensee did not mention whether the same approach for Watts Bar 10 CFR 50.69 RAI 07-01 (ADAMS Accession No. ML19302D625) will be applicable to the Browns Ferry 10 CFR 50.69 LAR.
a) Confirm that the response to RAI 07-01 is applicable to Browns Ferry 10 CFR 50.69 LAR.
b) If question a) cannot be confirmed, provide responses to Watts Bar 10 CFR 50.69 RAI 07-01 applicable for Browns Ferry.
RAI 07 - Overall Use of NEI 00-04 Figure 5-6 and Use for External Floods and High Wind The guidance in NEI 00-04 Figure 5-6 provides guidance to be used to determine SSC safety significance. The guidance in NEI 00-04 states, in part, that if it can be shown that the component either did not participate in any screened scenarios or, even if credit for the component was removed, the screened scenario would not become unscreened, then it is considered a candidate for the LSS category.
LAR Section 3.1.7, states that [n]o SSCs were explicitly credited to allow a scenario to screen, and therefore, [s]creened hazards are considered insignificant for every SSC and [] will not be considered during the categorization process. The NRC staff notes that LAR Attachment 4 screens all other external events (besides internal flood, internal fire and seismic events). It appears to NRC staff that, based on this description, at the time an SSC is categorized, it will not be evaluated using the guidance in NEI 00-04, Figure 5-6 to confirm that the SSC is not credited in screening an external hazard because that evaluation has already been made. The NRC staff notes that plant changes, plant or industry operational experience, and identified errors or limitations in the PRA models could potentially impact the conclusion that an SSC is not needed to screen an external hazard.
Also, concerning the external flooding, the NRC staff provided an assessment of the Browns Ferry flood hazard mitigating strategy assessment (MSA) dated September 5, 2017 (ADAMS Accession No. ML17222A328). That assessment discusses SSCs that would be relied upon to mitigate the impact of an extreme flooding event such as Local Intense Precipitation (LIP). The report discusses passive features such as external doors but also refers to credit for active components. Section 3.2 of the NRC staffs assessment discusses use of FLEX strategies against external flooding events such as a LIP. Section 3.2.1 of the NRC staffs assessment states that, regarding the Intake Pump Station, each compartment contains sump pumps to remove rainwater that accumulates from openings at the roof. These passive and active SSCs appear to be credited in screening of the external flooding hazard.
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In light of these observations, address the following:
a) Clarify whether or not an SSC will be evaluated during categorization of the SSC using the guidance in NEI 00-04, Figure 5-6, to confirm that the SSC is not credited in screening an external hazard.
b) Identify any active and passive SSCs that are credited for screening the external flooding hazard and discuss how those SSCs will be included and considered in the proposed categorization process.
c) Identify any active and passive SSCs that are credited for screening the high winds and tornado hazard, including tornado-generated missiles, and discuss how those SSCs will be included and considered in the proposed categorization process.
RAI 08 - Propagation of Closed and Open/Partially Open Findings from Internal Events Open Finding Level F&O According to Section 5-1.2 of the 2009 ASME/ANS PRA Standard it is assumed that 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. Therefore, the acceptability of the IEPRA model used as the foundation for the SPRA is an important consideration. Section 3.2 of the LAR states that the internal events and seismic hazards findings were reviewed and closed using the process documented in Appendix X to NEI 05-04, NEI 07-12, and NEI 12-13. Further, Attachment 3 of the LAR provides an evaluation of internal events open finding level F&Os that impact the 10 CFR 50.69 applications. However, the LAR does not provide information about the propagation of changes made to the IEPRA (includes internal floods) for (1) resolving the finding level F&Os that are closed, and (2) addressing the open/partially open finding level F&Os. The NRC staff's concerns in this RAI are similar to those discussed in Watts Bar 10 CFR 50.69 RAI 12-01 (ADAMS Accession No. ML19302D625).
- a. Clarify whether changes made to the internal events model to close finding level F&Os or to disposition the open/partially open finding level F&Os that are applicable to the SPRA, have been implemented in the SPRA used to support this application or justify not implementing the changes in the context of impact on this application.
- b. Describe an approach that is consistent with the requirements in 10 CFR 50.69(e) and the guidance in NEI 00-04 for appropriate categorization of SSCs to propagate changes in the IEPRA (includes internal floods) to the SPRA 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).
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