RA-19-0153, 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 ..

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 ..

ML19113A285
Person / Time
Site:	Harris
Issue date:	04/23/2019
From:	Snider S Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA-19-0153
Download: ML19113A285 (29)

Text

Steve Snider

( ., DUKE Vice President ENERGY Nuclear Engineering 526 South Church Street, EC-07H Charlotte, NC 28202 980-373-6195 Steve.Snider@duke-energy.com 10 CFR 50.90 April 23, 2019 Serial: RA-19-0153 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400 / Renewed License No. NPF-63

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 (SSCs) for Nuclear Power Reactors, dated February 1, 2018 (ADAMS Accession No. ML18033B768).

2. 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 (SSCs) for Nuclear Power Reactors, dated October 18, 2018 (ADAMS Accession No. ML18291A606).

3. NRC letter, Shearon Harris Nuclear Power Plant, Unit 1 - Request for Additional Information Regarding License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, dated March 18, 2019 (ADAMS Accession No.

ML19060A091).

Ladies and Gentlemen:

By letter dated February 1, 2018 (Reference 1), as supplemented by letter dated October 18, 2018 (Reference 2), Duke Energy Progress, LLC (Duke Energy) submitted a license amendment request (LAR) for Shearon Harris Nuclear Power Plant (HNP), Unit No. 1. 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.

By letter dated March 18, 2019 (Reference 3), the Nuclear Regulatory Commission (NRC) staff requested additional information from Duke Energy that is needed to complete the LAR review.

U.S. Nuclear Regulatory Commission Page2 Serial RA-19-0153 The enclosure to this letter provides Duke Energy's response to the Reference 3 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 HNP. Attachment 2 contains proposed markups of the HNP Renewed Facility Operating License. The markups supersede those provided in Reference 2.

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 notifying the State of North 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, Manager - Fleet Licensing, at (980) 373-2062.

I declare under penalty of perjury that the foregoing is true and correct. Executed on April 23, 2019.

Sincerely, Steve Snider Vice President - Nuclear Engineering JLV

Enclosure:

Response to NRC Request for Additional Information Attachments:

1. HNP 50.69 PRA Implementation Items

2. Markup of Proposed Renewed Facility Operating License cc: Ms. C. Haney, NRC Regional Administrator, Region II Ms. M. Barillas, NRC Project Manager, HNP (Electronic Copy Only)

Mr. J. Zeiler, NRC Sr. Resident Inspector, HNP Mr. W. L. Cox, Ill, Section Chief, N.C. DHSR (Electronic Copy Only)

Serial: RA-19-0153 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400 / Renewed License No. NPF-63 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 Enclosure Response to NRC Request for Additional Information

U.S. Nuclear Regulatory Commission Page 1 of 20 Serial RA-19-0153 Enclosure NRC Request for Additional Information By letter dated February 1, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18033B768), as supplemented by letter dated October 18, 2018 (ADAMS Accession No. ML18291A606), Duke Energy Progress, LLC (Duke Energy, the licensee), submitted a license amendment request (LAR) for Shearon Harris Nuclear Power Plant, Unit 1. The proposed amendment would modify the licensing basis 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. 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 Nuclear Regulatory Commission (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 Risk Informed Safety Class categories defined in 10 CFR 50.69. This categorization process is an integrated decisionmaking process that incorporates risk and traditional engineering insights.

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

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 decisionmaking 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).

U.S. Nuclear Regulatory Commission Page 2 of 20 Serial RA-19-0153 Enclosure RAI 5.01:

The February 1, 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).

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 sources of uncertainty during a peer review. RG 1.200 also references NUREG-1855 as one acceptable means to identify key assumptions and key sources of uncertainty. RG 1.200, Revision 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, Revision 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, Revision 2.

RAI 5 requested the licensee to clarify how key assumptions and (key) uncertainties that could impact the results are identified and included in the evaluation. In a letter dated October 18, 2018, in the licensees response to RAI 5, the licensee 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 3 (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 most of the in excess of 1000 assumptions and sources of uncertainty instead of identifying and evaluating key assumptions and key uncertainties as described in NUREG-1855, Revision 1. The response also included a table titled Uncertainties and assumptions not addressed by 10 CFR 50.69 factor of 3 sensitivity/performance monitoring with 28 entries. 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, Revision 1 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 dispositions included in the Table also include the phrase [a]ny impact of the exclusion of these scenarios on acceptance criteria for

U.S. Nuclear Regulatory Commission Page 3 of 20 Serial RA-19-0153 Enclosure categorizations of other components is addressed by the factor of 3 sensitivity and performance monitoring.

The NRC staff finds 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/high safety significant (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 and large early release frequency. 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 sources 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 place in LSS

U.S. Nuclear Regulatory Commission Page 4 of 20 Serial RA-19-0153 Enclosure accurately reflect the approved risk-informed process. Based on the above, provide the following information:

RAI 5.01.a:

a. Clarify which process is used and is meant by the RAI 5 Table title 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 3).

Duke Energy Response to RAI 5.01.a:

The following RAI responses in parts b through f supersede the response to RAI 5 (ADAMS Accession No. ML18291A606). 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 5 is also being superseded by the following response. Additionally, this response supersedes Attachment 6 of the original LAR.

RAI 5.01.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 5.01.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 5.01.c:

c. Describe the approach(es) 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 5.01.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 5 of 20 Serial RA-19-0153 Enclosure This assessment was applied to all uncertainties and assumptions identified via the methods in part b for the internal hazards (including fire).

RAI 5.01.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 5.01.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. (RG 1.200 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 6 of 20 Serial RA-19-0153 Enclosure RAI 5.01.e:

Provide a summary list of the key assumptions and uncertainties that have been identified for the application, and discuss how each identified key assumption 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 and F.

Duke Energy Response to RAI 5.01.e:

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

Index Assumption/ Disposition Discussion Uncertainty

1. Assumptions In the HNP internal events model, there are These uncertainties associated with HRA within HEP several assumptions made when developing the development will be addressed by the calculations calculation of human error probabilities (HEP) that NEI 00-04 Table 5-2 and Table 5-3 have the potential to have a more than negligible sensitivity to evaluate human error basic Model : Internal impact on the HEP values. These include the events at their 5th and 95th percentile for Events/Flood/Fire following 5 items: all system categorizations under 50.69 and presented to the IDP. There is no

1. Based on the diversity of the additional sensitivity required to evaluate instrumentation, the unavailability of the these uncertainties.

condensate storage tank (CST) level The sensitivity shows the impact on SSC indication is not modeled in the fault tree.

importance in light of unknowns regarding These level transmitters are only required human error probabilities. As such, SSC for a single human reliability analysis (HRA) importance with respect to the 50.69 event to align emergency service water application is assessed in light of this (ESW) to the auxiliary feedwater (AFW) uncertainty.

pumps when the CST drains.

2. It assumed that the operators have 10 Implementation of this sensitivity study is consistent with NEI 00-04 guidance.

minutes to respond to spurious opening of

U.S. Nuclear Regulatory Commission Page 7 of 20 Serial RA-19-0153 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Index Assumption/ Disposition Discussion Uncertainty a component cooling water (CCW) relief valve and would receive indications such as low CCW surge tank level and low pump suction pressure. An operator action is included for the potential to isolate the non-essential header flow prior to pump failure.

3. Potential common restoration error events involving errors within a single procedure were judged to have moderate dependence when calculating the pre-initiator HEP, as the activities typically are performed by a single operator (or pair of operators) within a single day.

4. An 8-inch loss of coolant accident (LOCA) was selected to establish the time available for operators to manually start the residual heat removal (RHR) pumps. The range established for medium LOCAs is 5 inches to 13 inches, such that the time could be longer or shorter.

5. A 3-inch LOCA was selected to establish the time available for operators to establish an alternate high head safety injection (HHSI) path. The range established for the

U.S. Nuclear Regulatory Commission Page 8 of 20 Serial RA-19-0153 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Index Assumption/ Disposition Discussion Uncertainty small (S2) LOCA is 3 inches to 5 inches, such that the time could be shorter.

2. Common Cause In the HNP internal events model, there are These uncertainties associated with CCF modeling several assumptions made when developing the development will be addressed by the calculation of common cause events. These three NEI 00-04 Table 5-2 and Table 5-3 Model : Internal have the potential to have a more than negligible sensitivity to evaluate CCF basic events Events/Flood/Fire impact on the results. at their 5th and 95th percentile for all system categorizations under 50.69 and

1. In the calculations, multiple greek letter presented to the IDP. There is no (MGL) factors for group size four were additional sensitivity required to evaluate available, so the fifth valve was these uncertainties.

conservatively assumed to fail in common The sensitivity shows the impact on SSC cause. importance in light of unknowns regarding

2. Common cause failures (CCFs) were CCF probabilities. As such, SSC considered only for those combinations of importance with respect to the 50.69 components which would disable both application is assessed in light of this trains of the emergency safety features uncertainty.

actuation system (ESFAS), since the Implementation of this sensitivity study is probability of a lesser CCF disabling one consistent with NEI 00-04 guidance.

train in conjunction with another random failure is considered probabilistically insignificant.

3. The method used to determine the common cause factor for the CCW pumps,

U.S. Nuclear Regulatory Commission Page 9 of 20 Serial RA-19-0153 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Index Assumption/ Disposition Discussion Uncertainty while necessary, has a number of assumptions.

3. Requirement to After the accumulators have emptied, the operator The action to isolate the accumulators is isolate is required by emergency operating procedures to part of the action to cooldown and accumulators after close the three accumulator discharge valves and depressurize the RCS for transients and injection lock the breakers open in order to prevent injecting SGTRs, which is modeled in the PRA via nitrogen into the RCS. This action is not assumed HEP events OPER-9 Failure to initiate Model : Internal to be required in the PSA. RCS cooldown to use LPSI/RHR and Events/Flood/Fire OPER-41 Failure to initiate RCS This is not an issue for large and medium LOCAs cooldown to use LPSI/RHR (SGTR).

where any N2 is likely to be swept out of the break. However, the specific execution steps to However, for small LOCAs or transients in which isolate the accumulators are not included the RCS must be depressurized to get to shutdown in the development of the HEP. The conditions, the insertion of N2 into the RCS could execution steps to isolate the be an issue. accumulators will be added to these HEP event calculations 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 and Table 5-3 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.

Implementation of this model change and sensitivity study is consistent with

U.S. Nuclear Regulatory Commission Page 10 of 20 Serial RA-19-0153 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Index Assumption/ Disposition Discussion Uncertainty NUREG-1855 Rev. 1 Stage F (i.e.,

update the PRA model) and NEI 00-04 guidance (i.e., HRA sensitivity).

4. CFC system is not The Containment Fan Cooler (CFC) system is A spatial analysis was performed which impacted by assumed to be protected from damage due to the shows that two of the CFCs (AH-3 and LOCAs LOCA initiator. Although failure of the CFCs due AH-4) are on the 286 ft level of to LOCA effects for small LOCAs can be containment, while the other two (AH-1 Model : Internal discounted, no specific spatial analysis has been and AH-2) are on the 236 ft level, such Events/Flood/Fire performed for larger LOCAs. that a LOCA event would not impact CFCs on both floors. On the 286 ft level, the two CFCs are on the containment wall approximately 60 degrees apart such that a single large LOCA would not impact both CFCs. Similarly, on the 236 ft level the two CFCs are on the containment wall approximately 60 degrees apart such that again, a single large LOCA would not impact both CFCs.

Based on this, a sensitivity was performed in which all large and medium LOCA events fail a single CFC, while the other 3 are unaffected. LERF was then calculated (CFCs do not impact CDF since they are only credited to prevent containment overpressure), and importance measures were generated.

No basic events increased from LSS in

U.S. Nuclear Regulatory Commission Page 11 of 20 Serial RA-19-0153 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Index Assumption/ Disposition Discussion Uncertainty the base case to HSS in the sensitivity case.

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.

5. Modeling of S/G Although there are five safety relief valves (SRVs) The HNP models will be updated to credit SRVs on each of three steam generators (S/G) for a total all SRVs, and appropriate common cause of 15 valves, any one of which can perform the groups will be added to the model to Model : Internal steam relief function to remove reactor decay heat, include all relief valves that are intended Events/Flood/Fire the model conservatively assumes that if any relief to open at the same pressure, prior to valve on a steam generator fails, then all relief implementation of 50.69.

Table # 128 valves on that steam generator also fail. A common cause failure of all 15 SRVs is also Any uncertainty from the new CCF events included. will also be addressed by the NEI 00-04 Table 5-2 and Table 5-3 sensitivity to This is a very conservative assumption that evaluate CCF basic events to their 5th overstates the likelihood of losing steam and 95th percentile for all system relief/decay heat removal. Taking credit for more categorizations under 50.69, and the of the SRVs, and consideration of more results are presented to the IDP.

appropriate CCF groupings (such as all relief valves that are intended to open at the same

U.S. Nuclear Regulatory Commission Page 12 of 20 Serial RA-19-0153 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Index Assumption/ Disposition Discussion Uncertainty pressure - on each of the steam lines) would Implementation of this model change and provide a more realistic result. sensitivity study is consistent with NUREG-1855 Rev. 1 Stage F (i.e.,

update the PRA model) and NEI 00-04 guidance (i.e., CCF sensitivity).

6. Modeling of During steam generator tube rupture (SGTR) A sensitivity has been performed to pressurizer sprays scenarios requiring reactor coolant system (RCS) address the impact of including of cooldown/depressurization, the pressurizer power pressurizer sprays to mitigate SGTR Model : Internal operated relief valves (PORVs) are required to events. The sensitivity showed inclusion Events/Flood/Fire reduce RCS pressure. The PRA model assumes of sprays would decrease the CDF by that the spray valves and/or the reactor coolant approximately 0.3%. This extremely small pumps are unavailable, and the RCS PORVs are change in CDF will have a negligible always required to function. impact on component importance measures, and 10 CFR 50.69 categorization is not sensitive to this uncertainty.

This approach is consistent with NUREG-1855 Rev. 1, Stage E (sensitivity study).

7. Failure of ESW A failure of Emergency Service Water (ESW) due Further evaluation of the interconnection due to backflow to backflow through the Normal Service Water between the ESW and NSW systems through NSW (NSW) system if NSW fails to isolate is not shows that additional failures, beyond the postulated since a motor-operated valve (MOV) MOV and check valve failures noted in Model : Internal and a check valve would both need to fail to close the assumption, would be required to get Events/Flood/Fire if the NSW pump is unavailable or fails to run. backflow through the NSW system. When an NSW pump trips or is stopped, its discharge MOV automatically closes.

Additionally, when an ESW pump starts,

U.S. Nuclear Regulatory Commission Page 13 of 20 Serial RA-19-0153 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Index Assumption/ Disposition Discussion Uncertainty the ESW cross-tie MOV (1SW-39 or 1SW-40) between the NSW supply and the ESW supply automatically closes.

Thus, to get backflow through the NSW system on ESW start would require the running NSW pump to fail to run, failure of its discharge MOV to close, failure of the common NSW supply check valve (1SW-59) to close, and failure of an ESW cross-tie valve (1SW-39 or 1SW-40) to close. In the HNP PRA model the failure rate for an MOV to close on demand is 3.5E-03. Since there is no common cause between the NSW pump discharge MOV and the ESW cross-tie MOV, the probability of failure of both is 1.2E-05.

The probability of the running NSW pump failing over the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time is 1.4E-04. Therefore, even ignoring the check valve, the likelihood of this event is approximately 1.7E-09. Therefore, this assumption has a negligible impact on component importance measures, and 10 CFR 50.69 categorization is not sensitive to this uncertainty.

This sensitivity study is consistent with NUREG-1855 Rev. 1, Stage E (sensitivity study), to quantify the impact of an

U.S. Nuclear Regulatory Commission Page 14 of 20 Serial RA-19-0153 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Index Assumption/ Disposition Discussion Uncertainty uncertainty with respect to the application acceptance criteria.

8. Use of Generic The model uses generic screening values for the A sensitivity was performed to evaluate data for PAL personnel air lock (PAL) failures. Plant-specific the impact of this uncertainty by information may significantly vary from these increasing the PAL mechanical failure Model : Internal values. rate and leakage failure rate by a factor of Events/Flood/Fire 2. LERF was then calculated (the PAL has no impact on CDF) and importance measures were generated. Only one component basic event out of all component basic events in the model increased from LSS in the base case to HSS in the sensitivity case (the basic event for the PAL was already HSS in the base case). A similar sensitivity was performed by setting the mechanical failure rate and leakage failure rate for the PAL to 0, calculating LERF, and generating importance measures. In this case, no component basic events changed from LSS in the base case to HSS in the sensitivity case.

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

This sensitivity is consistent with the guidance in NUREG-1855 Rev. 1 Stage

U.S. Nuclear Regulatory Commission Page 15 of 20 Serial RA-19-0153 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Index Assumption/ Disposition Discussion Uncertainty E, to quantify the impact of an uncertainty with respect to the application acceptance criteria.

9. Exclusion of There are breakers connecting the non safety- Exclusion of a common cause failure of common cause related 6.9 kV bus supply to the emergency buses the breakers to open (124-SB and 125-failure of breakers which open to allow the emergency diesel SB or 104-SA and 105-SA) as a failure of generators (EDGs) to close onto the bus. Since the EDG supply to the bus may impact Model: Internal there are two in-series breakers, both of which the acceptance criteria for 10 CFR 50.69 Events/Flood/Fire would have to fail to trip, failure of these breakers categorization for these breakers. The was not included in the model due to low independent failures of each breaker, and probability. The impact of a common cause event appropriate CCF events will be added to on the model was not determined. the model prior to implementation of 50.69.

Any uncertainty from the new CCF events will also be addressed by the NEI 00-04 Table 5-2 and Table 5-3 sensitivity to evaluate CCF basic events to their 5th and 95th percentile for all system categorizations under 50.69, and the results are presented to the IDP.

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., CCF sensitivity).

U.S. Nuclear Regulatory Commission Page 16 of 20 Serial RA-19-0153 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Index Assumption/ Disposition Discussion Uncertainty

10. Number of SI One high volume, low-pressure injection pump and A sensitivity was performed by modifying Accumulators two passive safety injection accumulators are the fault tree to require only 1 required for large required for successful RCS makeup after a large Accumulator to inject during a large LOCAs break LOCA during injection to meet the licensing LOCA, and re-running the CDF and LERF design basis. This assumes a double-ended pipe results. No basic events increased from Model: Internal break at the upper range of the possible large LSS in the base case to HSS in the Events/Flood/Fire LOCA events. Lesser events would not be as sensitivity case for either CDF or LERF.

extreme and would not necessarily require accumulator injection. 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.

11. Floor value for For those cutsets with three or more HEPs a lower Table 4-3 of EPRI TR 1021081, HRA combinations bound of 1E-06 was used. This lower bound was Establishing Minimum Acceptable decreased one decade [from the NUREG-1972 Values for Probabilities of Human Failure Model: Internal recommendation] to account for the fact that many Events, October 2010, provides a lower Events/Flood of the third and fourth HEPs are actions that occur limiting value of 1E-6 for sequences with many hours after the initiating event and thus new a very low level of dependence.

individuals are there to not only respond but they Assigning joint HEPs that are less than a also perform an independent review of the actions minimum value should be individually and diagnosis of the event. reviewed for timing, cues, etc., to check the dependency between all the operator actions in the cutset. All HEP

U.S. Nuclear Regulatory Commission Page 17 of 20 Serial RA-19-0153 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Index Assumption/ Disposition Discussion Uncertainty combinations which were determined to be below 1.0E-05 were evaluated to determine the level of dependency between the actions and it was confirmed that the dependencies were low enough to support the lower value. Additionally, HEP combination events were never assigned a value of less than 1.0E-06.

Based on the above, the method used to develop dependent HEP values is a consensus method and eliminates the need to explore an alternative hypothesis.

This approach is consistent with the guidance in NUREG-1855 Rev. 1 Stage E, section 7.2.4.

Additionally, this will be addressed by the NEI 00-04 Table 5-2 sensitivity to evaluate human error basic events at their 5th and 95th percentile for all system categorizations under 50.69 and presented to the IDP.

12. Manipulation time The Tm (manipulation time) is based on FSAR This uncertainty associated with HRA for flood Isolation analysis that states a 30 minute time for rupture development will be addressed by the HRAs is 30 location determination, isolation, and securing of NEI 00-04 Table 5-2 sensitivity to minutes the applicable pump. The training observed on evaluate human error basic events at internal flooding referenced this document as the their 5th and 95th percentile for all system

U.S. Nuclear Regulatory Commission Page 18 of 20 Serial RA-19-0153 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Index Assumption/ Disposition Discussion Uncertainty Model: Internal source of the 30 minutes criteria. This value could categorizations under 50.69 and Flood be more or less depending on the particular flood presented to the IDP. There is no scenario. additional sensitivity required to evaluate these uncertainties.

The sensitivity shows the impact on SSC importance in light of unknowns regarding human error probabilities. As such, SSC importance with respect to the 50.69 application is assessed in light of this uncertainty.

13. HRAs not Blocked HRAs were only considered for flood and This uncertainty associated with HRA considered HELB events. Spray events were assumed not to development will be addressed by the blocked in spray result in conditions that would prevent operator NEI 00-04 Table 5-2 sensitivity to areas actions from being performed, since it is very likely evaluate human error basic events at an operator would be able to complete actions in a their 95th percentile failure rate for all Model: Internal spray area. However, no actions involving system categorizations under 50.69 and Flood electrical equipment were credited in spray areas. presented to the IDP. There is no additional sensitivity required to evaluate these uncertainties.

The sensitivity shows the impact on SSC importance in light of unknowns regarding human actions in spray areas. As such, SSC importance with respect to the 50.69 application is assessed in light of this uncertainty.

14. Credit for Incipient Incipient detection in low voltage cabinets provides See response to RAI 06 (ADAMS Detection additional 60 minutes for manual suppression. Accession No. ML18291A606)

U.S. Nuclear Regulatory Commission Page 19 of 20 Serial RA-19-0153 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Index Assumption/ Disposition Discussion Uncertainty Model: Fire

15. Assumption In the HNP fire model, it is assumed that if no This uncertainty associated with manual related to manual detection system is installed in an area, manual fire suppression will be addressed by the detection detection will occur in 15 minutes. Although this NEI 00-04 Table 5-3 sensitivity to take no assumption is probably realistic, some fire credit for manual suppression for all Model: Fire compartments may have a relatively low potential system categorizations under 50.69 and of fire detection, especially if they are closed and presented to the IDP. There is no have low occupancy levels. additional sensitivity required to evaluate these uncertainties.

The sensitivity shows the impact on SSC importance in light of unknowns regarding manual suppression. As such, SSC importance with respect to the 50.69 application is assessed in light of this uncertainty.

U.S. Nuclear Regulatory Commission Page 20 of 20 Serial RA-19-0153 Enclosure RAI 5.01.f:

If NEI 00-04 or NUREG-1855 guidance is not used (e.g. all of 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 5.01.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.

Serial: RA-19-0153 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400 / Renewed License No. NPF-63 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 HNP 50.69 PRA Implementation Items

U.S. Nuclear Regulatory Commission Page 1 of 2 Serial RA-19-0153 The table below identifies the items that are required to be completed prior to implementation of 10 CFR 50.69 at Shearon Harris Nuclear Power Plant (HNP), Unit No. 1. 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.

Harris 50.69 PRA Implementation Items Description Resolution

i. In the Fire PRA model, detailed analysis Duke Energy will perform detailed analysis is needed for four significant HFEs in accordance with current methods for the identified in open finding HRC-C1-3. four significant HFEs identified and incorporate the analysis into the Harris This condition is described in response Fire PRA model as indicated in the Duke to RAI 02.e in Duke letter dated October letter dated October 18, 2018.

18, 2018.

ii. Update the HNP Fire PRA model to The Fire PRA model will be updated to incorporate NUREG-2180 or other NRC credit incipient detection per NUREG-2180 acceptable methodology for incipient or other NRC acceptable methodology, as described in Duke letter dated October 18, detection credit. If this update is 2018.

determined to be a PRA model upgrade per the 2009 ASME/ANS PRA standard, then conduct a focused scope peer review. Any findings from the focused scope peer review will be resolved and closed per an NRC approved process, or the findings will be dispositioned for the application and submitted for NRC review and approval prior to implementing 10 CFR 50.69.

This condition is described in response to RAI 06 in Duke letter dated October 18, 2018.

U.S. Nuclear Regulatory Commission Page 2 of 2 Serial RA-19-0153 Attachment 1 iii. Update the HNP Fire PRA model to The fire PRA model will be updated to address finding FSS-F3-01 to meet account for scenarios to address fire Capability Category II of the ASME/ANS induced failure of structural steel in the Turbine Building, as indicated in response RA-Sa-2009 as endorsed by RG 1.200, to RAI 02.f contained in Duke letter dated Revision 2. If this update is determined October 18, 2018.

to be a PRA model upgrade per the 2009 ASME/ANS PRA standard, then conduct a focused scope peer review.

Any findings from the focused scope peer review will be resolved and closed per an NRC approved process, or the findings will be dispositioned for the application and submitted for NRC review and approval prior to implementing 10 CFR 50.69.

This condition is described in response to RAI 02.f in Duke letter dated October 18, 2018.

iv. Update the HNP Internal Events, The HNP PRA models will be updated to Internal Flood and Fire PRA models to account for isolation of the RCS resolve uncertainties. accumulators and steam generator SRVs, as indicated in response to RAI 5.01 of

a. The execution steps to isolate Duke Energy letter dated April 23, 2019.

RCS accumulators as detailed in the EOPs will be added to the appropriate HEP event calculations.

b. The HNP models will be updated to credit all safety relief valves (SRVs) and appropriate common cause groups will be added to the model to include all relief valves that are intended to open at the same pressure.

c. The independent failures of breakers 124-SB and 125-SB or 104-SA and 105-SA to open as a failure of the EDG supply to the emergency buses along with their common cause failure events will be added to the model.

These conditions are described in response to RAI 5.01 in Duke Energy letter dated April 23, 2019.

Serial: RA-19-0153 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400 / Renewed License No. NPF-63 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

L. This license is effective as of the date of issuance and shall expire at midnight on Octobe r 24, 2046.

FOR THE NUCLE AR REGUL ATORY COMMI SSION

/RA/

Eric J. Leeds, Director Office of Nuclear Reactor Regulation Renewed License No. NPF-63 Amendm ent No.4Se -

APPENDIX D ADDITIONAL CONDITIONS RENEWED LICENSE NO. NPF-63 Duke Energy Progress, LLC shall comply with the following conditions on the schedule noted below:

Amendment Additional Conditions Implementation Number Date Duke Energy 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 internal fire; 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 1 License Amendment No. [XXX] dated [DATE]. Prior to implementation

[NUMBER]

of 10 CFR 50.69.

Duke Energy will complete the implementation items list in Attachment 1 of Duke Energy letter to the NRC dated April 23, 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).