RA-19-0154, Response to NRC Request for Additional Information (RAI) Regarding Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (Sscs) for Nuclear Power Reactors

Response to NRC Request for Additional Information (RAI) Regarding Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (Sscs) for Nuclear Power Reactors

ML19126A143
Person / Time
Site:	Robinson
Issue date:	05/06/2019
From:	Kapopoulos E Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA-19-0154
Download: ML19126A143 (24)

Text

Ernest J. Kapopoulos, Jr.

( ., DUKE H. B. Robinson Steam Electric Plant Unit 2 ENERGY Site Vice President Duke Energy 3581 West Entrance Road Hartsville, SC 29550 O: 843 951 1701 F: 843 857 1319 10 CFR 50.90 May 6, 2019 Serial: RA-19-0154 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 H.B. Robinson Steam Electric Plant, Unit 2 Renewed Facility Operating License No. DPR-23 Docket No. 50-261

Subject:

Response to NRC Request for Additional Information (RAI) Regarding Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Reactors

References:

1. Duke Energy letter, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors, dated April 5, 2018 (ADAMS Accession No. ML18099A130).

2. Duke Energy letter, Supplement to the Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors, dated June 6, 2018 (ADAMS Accession No. ML18162A147).

3. Duke Energy letter, Response to NRC Request for Additional Information (RAI)

Regarding Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors, dated November 13, 2018 (ADAMS Accession No. ML18317A026).

4. NRC letter, Request for Additional Information Regarding License Amendment Request to Adopt 10 CFR 50.69 Risk-Informed Categorization of Structures, Systems, and Components, dated March 12, 2019 (ADAMS Accession No. ML19072A024).

Ladies and Gentlemen:

By letter dated April 5, 2018 (Reference 1), as supplemented by letters dated June 6, 2018 (Reference 2) and November 13, 2018 (Reference 3), Duke Energy Progress, LLC (Duke Energy) submitted a license amendment request (LAR) for H.B. Robinson Steam Electric Plant Unit 2 (HBRSEP2). The proposed amendment would modify the licensing basis, by the addition of a License Condition, to allow for the implementation of the provisions 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.

U.S. Nuclear Regulatory Commission Serial RA-19-0154 Page 2 By letter dated March 12, 2019 (Reference 4), the Nucle ar Regulatory Commission (NRC) staff requested additional information from Duke Energy that is needed to complete the LAR review.

The enclosure to this letter provides Duke Energy's response to the Reference 4 RAI related to this amendment request. Attachment 1 contains PRA implementation items which must be completed prior to implementation of 10 CFR 50.69 at HBRSEP2. Attachment 2 contains proposed markups of the HBRSEP2 Renewed Facili ty Operating License. The markups supersede those provided in Reference 3.

The conclusions of the original No Significant Hazards Consideration and Environmental Consideration in the original LAR are unaffected by this RAI response.

There are no regulatory commitments contained in this letter.

In accordance with 10 CFR 50.91, Duke Energy is notify ing the State of South Carolina of this LAR by transmitting a copy of this letter and enclosure to the designated State Official.

Should you have any questions concerning this letter and its enclosure, or require additional information, please contact Art Zaremba at (980) 373-2 062.

I declare under penalty of perjury that the foregoing is true and correct. Executed on May 6, 2019.

Sincerely, Ernest J. Kapopoulos, Jr.

Site Vice President EJK/jlv

Enclosure:

Response to NRC Request for Additional Information Attachments:

1. HBRSEP2 50.69 PRA Implementation Items

2. Markup of Proposed Renewed Facility Operating License cc: Ms. C. Haney, NRC Regional Administrator, Region II Mr. N. Jordan, NRC Project Manager, HBRSEP2 (Elect ronic Copy Only)

Mr. M. Fannon, NRC Senior Resident Inspector, HBRS EP2 Ms. L. Garner, Manager, S.C. DHEC (Electronic Copy Only)

Ms. A. Nair-Gimmi, S.C. DHEC (Electronic Copy Only)

Mr. A. Wilson, Attorney General (S.C.) (Electronic Copy Only)

Serial: RA-19-0154 H.B. Robinson Steam Electric Plant, Unit No. 2 Docket No. 50-261 / Renewed License No. DPR-23 Response to NRC Request for Additional Information (RAI) Regarding Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors Enclosure Response to NRC Request for Additional Information

U.S. Nuclear Regulatory Commission Page 2 of 15 Serial RA-19-0154 Enclosure NRC Request for Additional Information By letter dated April 5, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18099A130), as supplemented by letter dated June 6, 2018 (ADAMS Accession No. ML18162A147), Duke Energy Progress, LLC., (the licensee) submitted a license amendment request (LAR) regarding the H.B. Robinson Steam Electric Plant, Unit 2 (Robinson). The licensee proposed to add a new license condition to the Renewed Facility Operating Licenses to allow the implementation 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. The provisions of 10 CFR 50.69 allow adjustment of the scope of structures, systems, and components (SSCs) subject to special treatment requirements (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation) based on a method of categorizing SSCs according to their safety significance.

The NRC staff has determined the following request for additional information (RAI) is needed to complete its review.

Regulatory Basis Nuclear Energy Institute (NEI) 00-04, Revision 0, 10 CFR 50.69 SSC Categorization Guideline (ADAMS Accession No. ML052910035), describes a process for determining the safety-significance of SSCs and categorizing them into the four RISC categories defined in 10 CFR 50.69. This categorization process is an integrated decision-making process that incorporates risk and traditional engineering insights.

NUREG-1855, Revision 1, Guidelines on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making (ADAMS Accession No. ML17062A466), provides guidance on how to treat uncertainties associated with probabilistic risk assessment (PRA) in risk-informed decision making.

Regulatory Guide (RG) 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities (ADAMS Accession No. ML090410014) describes an acceptable approach for determining whether the quality of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision making for light-water reactors. It endorses, with clarifications, the American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) PRA Standard ASME/ANS RA-Sa-2009 (ASME/ANS 2009 Standard or PRA Standard) (ADAMS Accession No. ML092870592).

RAI 3.01 Identifying Key Assumptions and Uncertainties that Could Impact the Application:

The April 5, 2018, LAR states:

The process to categorize each system will be consistent with the guidance in NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline, as endorsed by RG 1.201. RG 1.201 states that the implementation of all processes described in NEI 00-04 (i.e., Sections 2 through 12) is integral to providing reasonable confidence and that all aspects of NEI 00-04 must be followed to achieve reasonable confidence in the evaluations required by

§50.69(c)(1)(iv).

U.S. Nuclear Regulatory Commission Page 3 of 15 Serial RA-19-0154 Enclosure NEI 00-04 references RG 1.200 as the primary basis for evaluating the technical adequacy of the PRA. RG 1.200 references the ASME/ANS RA-Sa-2009 Standard which requires the identification and documentation of assumptions and source of uncertainty during a peer review.

RG 1.200 also references NUREG-1855 as one acceptable means of identifying key assumptions and key sources of uncertainty. RG 1.200, Rev. 2 defines a key uncertainty as one that is related to an issue in which there is no consensus approach or model and where the choice of the approach or model is known to have an impact on the risk profile such that it influences a decision being made using the PRA. RG 1.200, Rev. 2 defines a key assumption as one that is made in response to a key source of modeling uncertainty in the knowledge that a different reasonable alternative assumption would produce different results. The term reasonable alternative is also defined in RG 1.200, Rev. 2.

RAI 3 requested that the licensee clarify how key assumptions and (key) uncertainties that could impact the results are identified and included in the evaluation. In addition to this general RAI, two additional RAIs (RAI 5a and RAI 6) further question specific entries in Attachment 6 of the LAR. RAI 5a referenced entry 10 in Attachment 6 of the LAR, and requested additional information to justify that updated success criteria for feed and bleed would have an insignificant effect on the categorization. RAI 6 referenced entries 1, 2, and 3 in Attachment 6 of the LAR, and requested additional information explaining how the NEI 00-04 sensitivity studies in Tables 5-2, 5-3, 5-4, and 5-5 adequately address uncertainties in reactor coolant pump seal failure models, uncertainties in loss of off-site power frequencies, and uncertainties associated with several fire PRA modelling assumptions respectively.

The licensees response to RAI 3 (dated November 13, 2018) refers to the integrated risk sensitivity, as described in Section 8 of NEI 00-04. For this integrated risk sensitivity study, the unreliability of all low safety significant (LSS) SSCs is increased by a factor of three (consistent with NEI 00-04) and the subsequent total risk increase is compared to the 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. ML17317A256) acceptable risk increase guidelines. The licensee stated that this integrated risk sensitivity study, and the subsequent performance monitoring of LSS SSCs, could be used directly to address assumptions and sources of uncertainty instead of identifying and evaluating key assumptions and key uncertainties as described in NUREG-1855. The response to RAI 3 also included a table titled Uncertainties and Assumptions Not Addressed by 10 CFR 50.69 Factor of 3 Sensitivity/Performance Monitoring with 22 entries and stated that this table supersedes the table in Attachment 6 of the LAR. The licensee responses to RAIs 5a and 6, referred to the proposed methodology in the response to RAI 3 as being sufficient to address all 4 entries in of the LAR and, therefore, these entries are no longer key sources of uncertainty and were removed.

The licensee recognized that assumptions and uncertainties that cause SSCs to be excluded from the PRA cannot be addressed by the integrated risk sensitivity. The entries in the Table are apparently identified and included because they cause SSCs to be excluded. The dispositions in the Table include dispositions consistent with the NUREG-1855 options of (1) refining the PRA if needed, (2) redefine the application (e.g., add a sensitivity study), or (3) add compensatory measure and monitoring specific to that assumption of uncertainty.

However, the title of the table implies that all the unreported assumptions and uncertainty are evaluated and dispositioned as not being key solely using the factor of 3. Furthermore, most

U.S. Nuclear Regulatory Commission Page 4 of 15 Serial RA-19-0154 Enclosure dispositions included in the Table also include the phrase [a]ny impact of the exclusion of these scenarios on acceptance criteria for categorizations of other components is addressed by the factor of 3 sensitivity and performance monitoring.

The NRC staff determined that the licensees proposed method is a deviation from the guidance of NEI 00-04 and NUREG-1855, Revision 1 for the following reasons. Figure 1-2 in Section 1.5, Categorization Process Summary, of NEI 00-04 illustrates the available paths through the accepted categorization process. The categorization provides the appropriate LSS/HSS category. The integrated risk sensitivity study is only performed after all steps in the categorization have been completed and it is not intended to be a change in the risk estimate.

The study simply verifies that the combined impact of any postulated simultaneous degradation in reliability of all LSS SSCs would not result in significant increases in core damage frequency (CDF) and large early release frequency (LERF). Therefore, the aggregate risk sensitivity study is intended to capture the uncertainty from relaxation of special treatment for candidate LSS SSCs. Other assumptions and uncertainties are related to models and methods used in the PRA and the impact of these assumptions and uncertainties is not considered or included in the integrated risk sensitivity study.

NUREG-1855 identifies that one key source of uncertainty is the unknown increase in unreliability associated with the reduced special treatment requirements on LSS SSCs allowed by 10 CFR 50.69. The NUREG states that one acceptable technique to address this specific key source of uncertainty is to increase the unreliability of LSS SSCs by a multiplicative factor in an integrated risk sensitivity study. NEI 00-04 discusses using a factor of 3 to 5 as an acceptable multiplicative factor to address this uncertainty and the licensee selected to use the factor of 3 In contrast, addressing key assumptions and key sources of uncertainty in the PRA might require that SSCs be added to the PRA, might require changes to the model logic, or might require changes in the unreliability (e.g., unreliability increases for unusual uses of SSCs and for consequential failures) greater than the factor of 3 used in the integrated risk sensitivity study. Even for components that are modeled, the integrated risk sensitivity study only addresses the impact of SSCs as they are included in the PRA logic models without addressing any changes to the logic model itself that might be needed to address the key assumption (i.e.,

because of limitations in scope or level of detail). In addition, the use of the integrated risk sensitivity will result in the licensee identifying potential categorization of a LSS SSC as HSS only if the RG 1.174 risk acceptance guidelines are exceeded. However, addressing key assumptions and source of uncertainty, can result in a change in categorization even if the RG 1.174 guidelines are not exceeded. NEI 00-04 guidance in Tables 5-2 through 5-5 recognizes such occurrences and Figure 7-2 in NEI 00-04, Example Risk-Informed SSC Assessment Worksheet, captures such a change in categorization due to the sensitivity studies recommended in Tables 5-2 through 5- 5.

The licensees response simply states and does not justify that the use of the factors in the integrated risk sensitivity study are sufficient to capture the impact of all assumptions and uncertainties on the categorization of SSCs modeled in the current PRA. The approach proposed by the licensee represents a substantial deviation from the endorsed guidance for categorization in NEI 00-04 and the RAI response does not provide sufficient justification for the appropriateness of the deviation. It is unclear to the NRC staff whether the evaluation of assumptions and uncertainties proposed by the licensee can determine the effect of the key assumptions and uncertainties on the categorization of an indeterminate number of components. Therefore, the staff is unable to conclude that the components placed in LSS

U.S. Nuclear Regulatory Commission Page 5 of 15 Serial RA-19-0154 Enclosure accurately reflect the approved risk-informed process. Based on the above, provide the following information:

RAI 3.01.a:

a. Clarify which process is used and is meant by the RAI 5 Table tittle Uncertainties and Assumptions Not Addressed by 10 CFR 50.69 Factor of 3 Sensitivity/Performance Monitoring. , i.e., which types of uncertainties and assumptions have been addressed by the factor of three.

Duke Energy Response to RAI 3.01.a:

The following RAI responses in parts b through f supersede the response to RAI 3 (ADAMS Accession No. ML18317A026). Accordingly, the table titled Uncertainties and Assumptions Not Addressed by 10 CFR 50.69 Factor of 3 Sensitivity/Performance Monitoring that was provided in response to RAI 3 is also being superseded by the following response. Additionally, this response supersedes Attachment 6 of the original LAR.

RAI 3.01.b:

b. Describe the approach used to identify the assumptions and uncertainties that are used in the base PRA models.

Duke Energy Response to RAI 3.01.b:

To identify the assumptions and uncertainties used in the Internal Events and Internal Flood base PRA models supporting the categorization, the generic issues identified in Table A.1 of EPRI 1016737 were reviewed, as well as the PRA documentation for plant-specific assumptions and uncertainties. This identification process is consistent with NUREG-1855 Revision 1 Stage E.

To identify the assumptions and uncertainties used in the Fire base PRA model supporting the categorization, the generic issues identified in EPRI 1026511 were reviewed, as well as the PRA documentation for plant-specific assumptions and uncertainties. This identification process is consistent with NUREG-1855 Revision 1 Stage E.

RAI 3.01.c:

c. Describe the approach(s) used to evaluate each assumption and uncertainty to determine whether each assumption and uncertainty is key or not for this application.

Duke Energy Response to RAI 3.01.c:

To determine whether each assumption or uncertainty is key or not for this application, the assumption or uncertainty was individually assessed based on the definitions in RG 1.200 Revision 2, NUREG-1855 Revision 1, and related references (i.e. EPRI 1016737, EPRI 1013491, and EPRI 1026511). These documents provide definitions and guidance to identify if a specific assumption or uncertainty is key for an application and requires further consideration of the impact to the application.

U.S. Nuclear Regulatory Commission Page 6 of 15 Serial RA-19-0154 Enclosure This assessment was applied to all uncertainties and assumptions identified via the methods in part b for the internal hazards (including fire).

RAI 3.01.d:

d. Provide a summary of the different types of dispositions used for those assumptions and uncertainties determined not to be key for this application.

Duke Energy Response to RAI 3.01.d:

Assumptions or uncertainties determined not to be key are those that do not meet the definitions of key uncertainty or key assumption in RG 1.200 Revision 2, NUREG-1855 Revision 1, or related references. Specifically, the following considerations were used to determine those assumptions and uncertainties that do not require further consideration as key to the application:

- The uncertainty or assumption is implementing a consensus model as defined in NUREG 1855 Rev 1.

- The uncertainty or assumption will have no impact on the PRA results and therefore no impact on the decision of HSS or LSS for any SSCs.

- There is no different reasonable alternative to the assumption which would produce different results and/or there is no reasonable alternative that is at least as sound as the assumption being challenged. (RG1.200 Rev 2)

- The uncertainty or assumption implements a conservative bias in the PRA model, and that conservatism does not influence the results. These conservatisms are expected to be slight and only applied to minor contributors to the overall model. EPRI 1013491 uses the term realistic conservatisms. Thus, uncertainties/assumptions that implement realistic [slight] conservativisms can be screened from further consideration.

- EPRI 1013491 elaborates on the definition of a consensus model to include those areas of the PRA where extensive historical precedence is available to establish a model that has been accepted and yields PRA results that are considered reasonable and realistic. Thus, uncertainties/assumptions where there is extensive historical precedence that produces reasonable and realistic results can be screened from further consideration.

If the assumption or uncertainty does not meet one of the considerations above, then it is retained as key for the application and is presented in part e.

This assessment was applied to all uncertainties and assumptions identified via the methods in part b for the internal hazards (including fire).

U.S. Nuclear Regulatory Commission Page 7 of 15 Serial RA-19-0154 Enclosure RAI 3.01.e:

a. Provide a summary list of the key assumptions and uncertainties that have been identified for the application, and discuss how each identified key assumptions and uncertainty will be dispositioned in the categorization process. The discussion should clarify whether the licensee is following NEI 00-04 guidance by performing sensitivity analysis or other accepted guidance such as NUREG-1855 Stages A through F.

Duke Energy Response to RAI 3.01.e:

Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Assumption/ Disposition Discussion Index Uncertainty

1. Requirement to Operator failure to isolate the depressurized The action to isolate the accumulators is part of isolate accumulators accumulators (potential nitrogen injection) is not the action to cooldown and depressurize the after injection included in LOCA modeling. This is not an issue for RCS for transients and SGTRs, which is modeled large and medium LOCAs where any N2 is likely to be in the PRA via HEP events OPER-33 Failure to Model: Internal swept out of the break. However, for small LOCAs or depressurize the RCS and OPER-47 Failure to Events/Flood/Fire transients in which the RCS must be depressurized to establish shutdown cooling. However, the get to shutdown conditions, the insertion of N2 into specific execution steps to isolate the the RCS could be an issue. accumulators are not included in the development of the HEPs. The execution steps to isolate the accumulators will be added to these HEP event calculations, and the failures of the isolation valves will be added to the model, prior to implementation of 50.69.

Additionally, any uncertainty from these operator actions will also be addressed by the NEI 00-04 Table 5-2 sensitivity to evaluate human error basic events to their 5th and 95th percentile for all system categorizations under 50.69, and the results are presented to the IDP.

U.S. Nuclear Regulatory Commission Page 8 of 15 Serial RA-19-0154 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Assumption/ Disposition Discussion Index Uncertainty Implementation of this model change and sensitivity study is consistent with NUREG-1855 Rev. 1 Stage F (i.e., update the PRA model) and NEI 00-04 guidance (i.e., HRA sensitivity).

2. Emergency Diesel EDGs are modeled to fail without EDG room HVAC. A sensitivity was performed assuming that EDG Generators (EDGs) Loss of EDG room HVAC analysis indicates elevated HVAC is not required for the EDGs to operate for are dependent upon temperature, but not so high that failure is assured. their mission time. This resulted in a minor EDG room HVAC air EDGs may be able to meet mission time without room reduction in CDF and LERF. However, no flow for success cooling. component basic events changed from LSS to HSS.

Model: Internal Events/Flood/Fire As such, this sensitivity study shows 10 CFR 50.69 categorization is not sensitive to this uncertainty.

Implementation of this sensitivity is consistent with the guidance in NUREG-1855 Rev. 1 Stage E to quantify the impact of an uncertainty with respect to the application acceptance criteria.

3. Exposure time of These manual valves are modeled with an exposure A sensitivity was performed assuming and Deepwell (DW) to time of 2 years as they are cycled every 2 years for exposure time of 10 years for these valves. This Auxiliary Feedwater in-service testing (IST). However, there is no actual resulted in a minor increase in CDF and LERF.

(AFW) manual confirmation of flow through these valves during the However, no component basic events changed supply valves IST testing, so a longer exposure time could from LSS to HSS.

potentially be used.

Model: Internal Events/Flood/Fire

U.S. Nuclear Regulatory Commission Page 9 of 15 Serial RA-19-0154 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Assumption/ Disposition Discussion Index Uncertainty As such, this sensitivity study shows 10 CFR 50.69 categorization is not sensitive to this uncertainty.

Implementation of this sensitivity is consistent with the guidance in NUREG-1855 Rev. 1 Stage E to quantify the impact of an uncertainty with respect to the application acceptance criteria.

4. Number of PORVs One High Head Safety Injection (HHSI) pump and two A sensitivity study was completed in response to required for Feed PORVs are assumed to be required for bleed-and- RAI 5.01a to investigate the PORV success and Bleed Cooling feed. However, thermal-hydraulic analysis using criteria. The sensitivity investigated the effect (or MAAP determined the requirement for 2 PORVs may lack) of the excluded success criteria of the Model: Internal be reduced to 1 PORV for those sequences where PORVS on components by changing the model Events/Flood/Fire AFW operated successfully for a minimum period of to a success criteria of 1 PORV being allowable time (approximately 40 minutes). for all transient events, instead of 2 being required.

The results from the sensitivity show no basic events that were LSS in the base case (i.e., 2 PORVs required) became HSS in the sensitivity (one PORV required). As such, this sensitivity study shows 10 CFR 50.69 categorization is not sensitive to this uncertainty.

Implementation of this sensitivity is consistent with the guidance in NUREG-1855 Stage E to quantify the impact of an uncertainty with respect to the application acceptance criteria.

U.S. Nuclear Regulatory Commission Page 10 of 15 Serial RA-19-0154 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Assumption/ Disposition Discussion Index Uncertainty

5. Flood propagation It is assumed that the contribution to flood A detailed evaluation of flow through HVAC ducts through HVAC and propagation through vent and HVAC openings is and openings was performed. The results of this vents small compared to that of open doorways, stairwells, evaluation have been incorporated into the RNP and grating, unless otherwise noted. flooding model documentation.

Model: Internal Flood Implementation of this change is consistent with the guidance in NUREG-1855 Rev. 1 Stage F to refine the PRA model.

6. Flood flow under or The potential for leakage of flood water through small A detailed evaluation of flow through door gaps around doors gaps underneath doors and around the perimeter of was performed. The results of this evaluation doors is possible; however, it is assumed that the have been incorporated into the RNP flooding Model: Internal amount of flood water that leaks through the door model documentation.

Flood perimeter is insignificant and able to be mitigated by the floor drains. Implementation of this change is consistent with the guidance in NUREG-1855 Rev. 1 Stage F to refine the PRA model.

7. Credit for incipient Incipient detection credit at RNP is similar to NUREG- The RNP Fire PRA model was reviewed against detection 2180. NUREG-2180 guidance. There are no differences between the method used at RNP Model: Fire and the method described in NUREG-2180. The wording in the LAR with regards to additional credit for operator actions is not implemented in the model. Thus, treatment of incipient detection at RNP fully aligns with NUREG-2180 guidance.

Based on the above, the method used to incipient detection is a consensus method and eliminates the need to explore an alternative hypothesis.

U.S. Nuclear Regulatory Commission Page 11 of 15 Serial RA-19-0154 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Assumption/ Disposition Discussion Index Uncertainty This approach is consistent with the guidance in NUREG-1855 Rev. 1 Stage E, section 7.2.4.

8. Probability of hot A bounding probability of 0.06 for failure of hot shorts The bounding probability used in the RNP Fire short clearing to clear within the required time has been applied to model is higher than probabilities for failure of hot both AC and DC circuits as compared to NUREG/CR- shorts to clear within the required time given in Model: Fire 7150 Vol 2 values. This is conservative from an NUREG/CR-7150. Therefore, a sensitivity was overall CDF/LERF perspective but could impact performed using the most beneficial (lowest) hot 10CFR 50.69 categorization. short clearing probability of 7.1E-03 from NUREG/CR-7150 for both AC and DC hot shorts, and assuming all scenarios have 15 minutes for the short to clear prior to detrimental impact on the plant. NUREG/CR-7150 includes probabilities for failure of hot shorts to clear based on timing considerations, but all of these values are between the base model (0.06) and the value used in the sensitivity (7.1E-03). Thus, these bounding values were used for sensitivity.

The sensitivity resulted in minor reductions in fire CDF and LERF. Based on the sensitivity results, no component basic events changed from LSS to HSS.

As such, this sensitivity study shows 10 CFR 50.69 categorization is not sensitive to this uncertainty.

U.S. Nuclear Regulatory Commission Page 12 of 15 Serial RA-19-0154 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Assumption/ Disposition Discussion Index Uncertainty Implementation of this sensitivity is consistent with the guidance in NUREG-1855 Rev. 1 Stage E to quantify the impact of an uncertainty with respect to the application acceptance criteria.

U.S. Nuclear Regulatory Commission Page 13 of 15 Serial RA-19-0154 Enclosure RAI 3.01.f:

b. If NEI 00-04 or NUREG-1855 guidance is not used (e.g. completing all the Stages A through F in NUREG-1855, Revision 1) provide justification that the licensees approach is adequate to identify, capture the impact, and disposition key assumptions and uncertainties to support the categorization process.

Duke Energy Response to RAI 3.01.f:

The response provided in subparts b through e of this RAI are consistent with the guidance in NUREG-1855 Rev 1 and NEI 00-04.

RAI 05.01.a Feed and Bleed Success Criteria for loss of secondary heat removal: of the LAR (page 48) states that the current PRA model of record success criteria for Feed and Bleed includes one high pressure safety injection (HPSI) and two power operated relief valves (PORVs), but the thermal hydraulic analysis concludes that only one PORV is required. The disposition states that this could result in certain SSCs having higher risk significance and, therefore, is considered conservative. The NRC staff is aware that conservative modeling choices have the potential to artificially lower other components risk importance values to below the safety significance threshold criteria (i.e. masking). In the Response to RAI 05.a (submitted in November 13, 2018), the licensee stated that two PORVs are required assuming all loss of secondary cooling at time zero, while only one PORV is needed if all loss of secondary cooling is lost after 50 minutes or more. The response also states that since this assumption does not result in any components being excluded from the PRA, it is addressed by the integrated risk sensitivity study proposed in response to RAI 3a. As discussed in RAI 3.01 above, the staff disagrees with the position that the integrated risk sensitivity study can be used, unless components are excluded for the PRA. Based on the above, provide the following information:

RAI 05.01.a.i:

i. Provide justification, such as a sensitivity study, that the exclusion of the updated success criteria does not affect any of the SSC risk categorizations.

Duke Energy Response to RAI 05.01.a.i:

A sensitivity study was completed to investigate the PORV success criteria. The sensitivity investigated the effect (or lack) of the excluded success criteria of the PORVS on other components by changing the model to a success of 1 PORV being allowable for transient events, instead of 2 being required.

The results from the sensitivity show no basic events that were LSS in the base case (i.e., 2 PORVs required) became HSS in the sensitivity (one PORV required). As such, this sensitivity study shows 10 CFR 50.69 categorization is not sensitive to this uncertainty.

Implementation of this sensitivity is consistent with the guidance in NUREG-1855 Stage E to quantify the impact of an uncertainty with respect to the application acceptance criteria.

U.S. Nuclear Regulatory Commission Page 14 of 15 Serial RA-19-0154 Enclosure RAI 05.01.a.ii:

ii. Alternatively, propose a mechanism to incorporate the updated success criteria into the PRA model of record prior to implementation of the 10 CFR 50.69 categorization program.

Duke Energy Response to RAI 05.01.a.ii:

The response to item i fully addresses the uncertainty/assumption, thus an update to the success criteria in the model is not required.

RAI 06.01 Key Assumptions and Uncertainties Subject to Sensitivity Studies:

In LAR Attachment 6, assumptions 1, 2, and 3 address reactor coolant pump seal failure, loss of offsite power frequencies, and fire modelling respectively. Each of these assumptions is dispositioned with, In accordance with NEI 00-04, sensitivity studies will be used to determine whether other conditions might lead to the component being safety significant. The assessment of the uncertainties, therefore, is appropriately addressed by the sensitivity studies required by this risk-informed application.

NEI 00-04 sensitivity studies in Tables 5-2, 5-3, 5-4, and 5-5 all include human error probabilities, CCF probabilities, and maintenance unavailabilities. The uncertainties in assumptions 1, 2 and 3 are not related to these issues or parameters and therefore the sensitivity studies in the Tables do not resolve the effect of the assumptions. However, each Table also has provision for [a]ny applicable sensitivity studies identified in the characterization of PRA adequacy but these PRA specific studies need to be identified.

The November 13, 2018, response to RAI 06 states, [t]he updated assessment of key sources of uncertainty and assumptions performed in response to RAI-03.b supersedes the contents of to the original LAR. That updated assessment summarizes the strategy to address key assumptions and uncertainties for this application.

As discussed in RAI 3.01 above, the NRC staff disagrees with the position that the integrated risk sensitivity study can be used to address the impact of all assumptions and uncertainties on the categorization of SSCs modeled in the current PRA. Based on the above, provide the following information for assumptions 1, 2, and 3:

RAI 06.01.i:

i. Describe the applicable sensitivity study that will be undertaken to address each uncertainty, or otherwise resolve the effect of the assumption on the categorization process. In addition, propose a mechanism that ensures that the identified sensitivity studies will be included in the categorization evaluations, or

U.S. Nuclear Regulatory Commission Page 15 of 15 Serial RA-19-0154 Enclosure Duke Energy Response to RAI 06.01.i:

In LAR Attachment 6, assumptions 1, 2, and 3 address reactor coolant pump (RCP) seal failure, loss of offsite power (LOOP) frequencies, and fire modelling respectively. The LAR Attachment 6 is superseded by the response in RAI 3.01. The specific assumptions noted here are within the scope of those assumptions and uncertainties considered as part of RAI 3.01. Based on the criteria described in RAI 3.01.d (e.g., the uncertainty or assumption is implementing a consensus model as defined in NUREG 1855 Rev 1, etc.), these assumptions screened as not being key for this application.

RAI 06.01.ii:

ii. Provide justification, such as a sensitivity study, to support the licensees claim that not addressing each of the three entries (entries 1, 2, and 3) does not affect any of the SSC risk categorizations.

Duke Energy Response to RAI 06.01.ii:

See response to RIA 6.01.i.

Serial: RA-19-0154 H.B. Robinson Steam Electric Plant, Unit 2 Docket No. 50-261 / Renewed License No. DPR-23 Response to NRC Request for Additional Information (RAI) Regarding Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Reactors Attachment 1 HBRSEP2 50.69 PRA Implementation Items

U.S. Nuclear Regulatory Commission Page 1 of 2 Serial RA-19-0154 The table below identifies the items that are required to be completed prior to implementation of 10 CFR 50.69 at H.B. Robinson Steam Electric Plant, Unit 2 (HBRSEP2). The issues identified below will be addressed and any associated changes made, focused scope peer reviews performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2), and findings resolved and reflected in the PRA of record prior to implementation of 10 CFR 50.69.

Robinson 50.69 PRA Implementation Items Description U Resolution U

i. The HBRSEP2 internal flood model Duke Energy will update the HBRSEP2 does not account for generic human internal flood model to account for generic induced flooding data as described in human induced flooding events using an industry accepted methodology described response to RAI 1.b. in Duke letter in the response to RAI 1.b. in Duke letter dated November 13, 2018. If this dated November 13, 2018.

update is determined to be a PRA model upgrade per the 2009 ASME/ANS PRA standard, then a focused scope peer review will be conducted. Any findings from the focused scope peer review will be resolved and closed per an NRC approved process prior to implementing 50.69.

ii. Human Failure Events (HFEs) related to Duke Energy will update the HBRSEP2 isolating a ruptured SG following a internal events model to include these steam generator tube rupture (SGTR) operator actions.

are not represented in the internal events model as described in response to RAI 1.a. in Duke letter dated November 13, 2018. If this update is determined to be a PRA model upgrade per the 2009 ASME/ANS PRA standard, then a focused scope peer review will be conducted. Any findings from the focused scope peer review will be resolved and closed per an NRC approved process prior to implementing 50.69.

U.S. Nuclear Regulatory Commission Page 2 of 2 Serial RA-19-0154 Attachment 1 iii. Update the HBRSEP2 Internal Events, Duke Energy will update the Robinson Internal Flood, and Fire models to PRA model to account for isolation of the resolve uncertainties. RCS accumulators as indicated in response to RAI 3.01 contained in Duke

a. The execution steps to isolate letter dated May 6, 2019.

RCS accumulators as detailed in the EOPs will be added to the appropriate HEP event calculations, and the failures of the isolation valves will be added to the model.

These conditions are described in response to RAI 3.01 in Duke letter dated May 6, 2019.

Serial: RA-19-0154 H.B. Robinson Steam Electric Plant, Unit 2 Docket No. 50-261 / Renewed License No. DPR-23 Response to NRC Request for Additional Information (RAI) Regarding Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Reactors Attachment 2 Markup of Proposed Renewed Facility Operating License

APPENDIX B ADDITIONAL CONDITIONS FACILITY OPERATING LICENSE NO. DPR-23 Duke Energy Progress, LLC. (the term licensee in Appendix B refers to Duke Energy Progress, LLC.) shall comply with the following conditions on the schedules noted below:

Amendment Number Additional Conditions Implementation Date 176 The licensee is authorized to relocate This amendment is certain requirements included in effective immediately Appendix A and the former Appendix B and shall be to licensee-controlled documents. implemented within Implementation of this amendment 90 days of the date of shall include the relocation of these this amendment.

requirements to the appropriate documents, as described in the licensees letters dated September 10, 1997, and October 13, 1997, evaluated in the NRC staffs Safety Evaluation enclosed with this amendment.

219 Upon implementation of the amendment This amendment is adopting TSTF-448, Revision 3, the effective immediately determination of control room envelope and shall be (CRE) unfiltered air inleakage as required implemented as by TS 5.5.17.c.(i), the assessment of CRE specified habitability as required by TS 5.5.17.c.(ii),

and the measurement of CRE pressure as required by TS 5.5.17.d, shall be considered met. Following implementation:

(a) The first performance of TS 5.5.17.c.(i), shall be within the specified Frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from January 27,2003, the HBRSEP, Unit No. 2 1 Amendment No. 246

APPENDIX B ADDITIONAL CONDITIONS date of the most recent successful tracer gas test, or within the next 18 months if the time period since the most recent successful tracer gas test is greater than 6 years.

(b) The first performance of the periodic assessment of CRE habitability, TS 5.5.17.c.(ii), shall be within the next 9 months.

(c) The first performance of the periodic measurement of CRE pressure, TS 5.5.17.d, shall be within 18 months, plus the 138 days allowed by SR 3.0.2, as measured from the date of the most recent successful pressure measurement test.

INSERT 1 HBRSEP, Unit No. 2 2 Amendment No. 219

INSERT 1 Amendment Additional Conditions Implementation Number Date

[NUMBER] Duke Energy is approved to implement 10 Upon implementation of CFR 50.69 using the processes for Amendment No. [XXX].

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, internal fire, high winds, and external flood; 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; and the results of non PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e., seismic margin analysis (SMA) to evaluate seismic risk, 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 No. [XXX] dated

[DATE].

Duke Energy will complete the implementation items list in Attachment 1 of Duke letter to NRC dated May 6, 2019 prior to implementation of 10 CFR 50.69. All issues identified in the attachment will be addressed and any associated changes will be made, focused-scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to implementation of the 10 CFR 50.69 categorization process.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).