RA-19-0152, 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.

ML19099A035
Person / Time
Site:	Brunswick
Issue date:	04/08/2019
From:	Snider S Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA-19-0152
Download: ML19099A035 (33)

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 8, 2019 Serial: RA-19-0152 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. DPR-71 and DPR-62 Docket Nos. 50-325 and 50-324

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 January 10, 2018 (ADAMS Accession No. ML18010A344).

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 November 2, 2018 (ADAMS Accession No. ML18306A523).

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 (SSCs) for Nuclear Power Reactors, dated February 13, 2019 (ADAMS Accession No. ML19044A366)

4. NRC letter, Request for Additional Information Regarding License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components, dated March 7, 2019 (ADAMS Accession No. ML19067A271).

Ladies and Gentlemen:

By letter dated January 10, 2018 (Reference 1), as supplemented by letters dated November 2, 2018 (Reference 2) and February 13, 2019 (Reference 3), Duke Energy Progress, LLC (Duke Energy) submitted a license amendment request (LAR) for Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2. 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

U.S. Nuclear Regulatory Commission Page 2 Serial RA-19-0152 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 7, 2019 (Reference 4), the Nuclear 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 BSEP. Attachment 2 contains proposed markups of the BSEP Renewed Facility Operating License for both Units 1 and 2. The markups supersede those provided in Reference 3. Attachment 3 provides a minor update to the BSEP FLEX diesel generator failure rate that supersedes the failure rate provided in Reference 3 (response to RAI 8-1) .

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 at (980) 373-2062.

I declare under penalty of perjury that the foregoing is true and correct. Executed on April 8, 2019.

Sincerely, Steve Snider Vice President - Nuclear Engineering JLV

Enclosure:

Response to NRC Request for Additional Information cc: Ms. C. Haney, NRC Regional Administrator, Region II Mr. D. J . Galvin, NRC Project Manager, BNP Mr. G. Smith, NRC Sr. Resident Inspector, BNP Mr. W. L. Cox, Ill, Section Chief, N.C. DHSR (Electronic Copy Only)

Chair - North Carolina Utilities Commission (Electronic Copy Only)

U.S. Nuclear Regulatory Commission Page 1 of 22 Serial RA-19-0152 Enclosure Serial: RA-19-0152 Brunswick Steam Electric Plant, Units 1 and 2 Docket Nos. 50-325 and 50-324 / Renewed License Nos. DPR-71 and DPR-62 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 2 of 22 Serial RA-19-0152 Enclosure NRC Request for Additional Information By letter dated January 10, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18010A344), as supplemented by letters dated November 2, 2018 (ADAMS Accession No. ML18306A523) and February 13, 2019 (ADAMS Accession No. ML19044A366), Duke Energy Progress, LLC (the licensee) submitted a license amendment request (LAR) regarding the Brunswick Steam Electric Plant, Units 1 and 2 (Brunswick). 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 identified the need for additional information 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 4.01/17.01:

The January 10, 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

U.S. Nuclear Regulatory Commission Page 3 of 22 Serial RA-19-0152 Enclosure 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 that 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 to identify 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.

RAIs 4 and 17 requested the licensee discuss how key assumptions and (key) uncertainties that could impact the results are identified and included in the evaluation. In the responses to RAIs 4 and 17 in the letter dated November 2, 2018, 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) structures, systems, and components (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 the assumptions and sources of uncertainty instead of identifying and evaluating key assumptions and key uncertainties as described in NUREG-1855. The response to RAI 4 also included a table titled Uncertainties and assumptions not addressed by 10 CFR 50.69 factor of 3 sensitivity/performance monitoring with 32 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 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 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/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

U.S. Nuclear Regulatory Commission Page 4 of 22 Serial RA-19-0152 Enclosure degradation in reliability of all LSS SSCs would not result in a significant increase 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). Further, the licensees proposed approach of changing only the random failure probabilities does not provide any basis or justification for how the proposed approach captures the impact of key modeling assumptions and sources of uncertainty related to external hazard specific failure modes (e.g.

tornado missile impact failure probability). 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 accurately reflect the approved risk-informed process. Based on the above, provide the following information:

RAI- 4.01.a / 17.01.a

a. Clarify which process is used and is meant by the RAI 4 Table title, Uncertainties and Assumptions Not Addressed by 10 CFR 50.69 Factor of 3 Sensitivity/Performance

U.S. Nuclear Regulatory Commission Page 5 of 22 Serial RA-19-0152 Enclosure Monitoring, i.e., which types of uncertainties and assumptions have been addressed by the factor of three.

Duke Energy Response to RAI 4.01.a / 17.01.a:

The following RAI responses in parts b through f supersede the response to RAI 4 and 17 (ADAMS Accession No. ML18306A523). 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 4 is also being superseded by the following response.

Additionally, this response supersedes Attachment 6 of the original LAR and the bulleted list of sensitivities identified in Section 3.2.7 in the original LAR. Duke Energy response to RAI 17-1 is unaffected by these RAI responses.

RAI- 4.01.b / 17.01.b

b. Describe the approach used to identify the assumptions and uncertainties that are used in each of the base PRA models supporting the categorization as stated in the LAR.

The descriptions should be provided separately for internal hazard PRAs (including internal fire) and external hazard PRAs supporting this application.

Duke Energy Response to RAI 4.01.b / 17.01.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.

To identify the assumptions and uncertainties used in the High Winds and External Flood base PRA models supporting the categorization, the PRA documentation was reviewed for plant-specific assumptions and uncertainties. Note that EPRI 1026511 does not provide any generic uncertainties for high wind or external flood PRAs and is therefore not applicable to these two hazards. This identification process is consistent with NUREG-1855 Revision 1 Stage E.

RAI- 4.01.c / 17.01.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.

The descriptions should be provided separately for internal hazard PRAs (including internal fire) and external hazard PRAs supporting this application.

U.S. Nuclear Regulatory Commission Page 6 of 22 Serial RA-19-0152 Enclosure Duke Energy Response to RAI 4.01.c / 17.01.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.

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

RAI- 4.01.d / 17.01.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. The summaries should be provided separately for internal hazard PRAs (including internal fire) and external hazard PRAs supporting this application.

Duke Energy Response to RAI 4.01.d / 17.01.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 cannot be screened using one of the considerations above, then it is retained as key for the application and is presented in part e.

U.S. Nuclear Regulatory Commission Page 7 of 22 Serial RA-19-0152 Enclosure This assessment was applied to all uncertainties and assumptions identified via the methods in part b for the internal hazards and external hazards.

RAI- 4.01.e / 17.01.e

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. The summaries and descriptions should be provided separately for the identified key assumptions and uncertainties related to internal hazard PRAs (including internal fire) and those related to external hazard PRAs supporting this application.

U.S. Nuclear Regulatory Commission Page 8 of 22 Serial RA-19-0152 Enclosure Duke Energy Response to RAI 4.01.e / 17.01.e:

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

Index Assumption/ Disposition Discussion Uncertainty

1. Wetwell to drywell In the BNP LERF model, the wetwell to A sensitivity study will be completed for each vacuum breaker drywell vacuum breaker fail to close system categorized to adjust the wetwell to cycling (FTC) basic events are quantified based drywell vacuum breaker fail to close basic on an estimated number of vacuum event to simulate doubling the vacuum Model : Internal breaker cycles assumed to occur during breaker cycles and decreasing the number of Events the event. The model currently vacuum breaker cycles to zero to determine assumes the vacuum breakers cycle 10 the impact of this assumption. These times. It has been postulated that parameters were determined based on during large LOCAs the vacuum engineering judgement to bound the potential breakers may experience as little as range of cycles and provide insights on SSC one or no cycles due to the greater importance with respect to this parameter.

pressure in the drywell. This sensitivity will be completed with those described in NEI 00-04 Table 5-2 and the results presented to the IDP.

This sensitivity study demonstrates the potential impact on SSC importance with and without the relationship between CRD and RBCCW cooling. As such, SSC importance with respect to the 50.69 application is assessed considering this uncertainty.

Implementation of this sensitivity study is consistent with NEI 00-04 guidance.

U.S. Nuclear Regulatory Commission Page 9 of 22 Serial RA-19-0152 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Index Assumption/ Disposition Discussion Uncertainty

2. Modeling of Control A conservative assumption is made that A sensitivity study will be completed for each Rod Drive (CRD) the CRD system fails without Reactor system categorized removing the requirement Support Systems Building Closed Cooling Water for RBCCW to cool the CRD pump(s). This (RBCCW) cooling and is modeled sensitivity will be completed with those Model : Internal accordingly. Discussions with described in NEI 00-04 Table 5-2 and the Events Operations and Systems Engineering results presented to the IDP.

personnel indicate that the CRD pumps may actually have been run with This sensitivity study demonstrates the RBCCW inadvertently isolated with no potential impact on SSC importance with and degradation to the CRD pumps, without the relationship between CRD and however there is currently no RBCCW cooling. As such, SSC importance documentation of this. with respect to the 50.69 application is assessed considering this uncertainty.

Implementation of this sensitivity study is consistent with NEI 00-04 guidance.

3. Emergency Buses The Brunswick model assumes that the The assumption that switchgears E1 through E1 through E8 Emergency Busses E1 through E8 E8 do not depend on the DGB HVAC system dependency on EDG located in the Diesel Generator Building is validated and documented by a GOTHIC Building HVAC Switchgear rooms do not depend on the analysis as discussed in response to RAI-5.a.

buildings HVAC system to perform its This independence is built in Brunswicks PRA Model : Internal primary function during a 24-hour model and therefore, this uncertainty has been Events mission time based on generic resolved and will have no impact to the results analyses. of the 10 CFR 50.69 categorization process.

U.S. Nuclear Regulatory Commission Page 10 of 22 Serial RA-19-0152 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Index Assumption/ Disposition Discussion Uncertainty Resolution of this uncertainty is consistent with the guidance in NUREG-1855 Stage F to refine the PRA analysis.

4. Instrument Air Dryer It is assumed that a failure of the A sensitivity study will be completed for each Bypass Valves manual isolation valves on either side of system categorized to show the impact of the Instrument Air Dryer Bypass Valve potentially longer period of time during which Model : Internal (1-SA-PV-5067) would be discovered the valves could be failed closed without Events within three cycles of operation (6 discovery. A time period of 12 years will be years). used for the sensitivity based on inspection of system operating history, that shows division swaps, train outages, and maintenance on the bypass valve to occur where the isolation valves would be manipulated. Flow is verified through the bypass line at least once during this period. This sensitivity will be completed with those described in NEI 00-04 Table 5-2 and the results presented to the IDP.

This sensitivity study demonstrates the potential impact on SSC importance based on the instrument air dryer bypass line configuration. As such, SSC importance with respect to the 50.69 application is assessed considering this uncertainty.

Implementation of this sensitivity study is consistent with NEI 00-04 guidance.

U.S. Nuclear Regulatory Commission Page 11 of 22 Serial RA-19-0152 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Index Assumption/ Disposition Discussion Uncertainty

5. Staggered vs Non- There is a small amount of non- A sensitivity study will be completed for each Staggered Testing conservative modeling with only system categorized to show the impact of all for Common Cause applying staggered testing to all CCF CCF groups being assigned non-staggered Failure (CCF) groups groups. testing. This sensitivity study bounds cases for a variety of testing regimes. This study will Model : Internal be completed with those described in NEI 00-Events 04 Table 5-2 and the results presented to the IDP.

This sensitivity shows the impact on SSC importance based on testing techniques for common cause failure groups. As such, SSC importance with respect to the 50.69 application is assessed considering this uncertainty.

Implementation of this sensitivity study is consistent with NEI 00-04 guidance.

Once a review of CCF group testing has been completed and the model updated to reflect those that are on staggered or non-staggered testing, the sensitivity will no longer be completed for each system categorization. If this change is determined to be an upgrade, as defined in the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2), a focused scope peer review will performed and findings resolved and reflected

U.S. Nuclear Regulatory Commission Page 12 of 22 Serial RA-19-0152 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Index Assumption/ Disposition Discussion Uncertainty in the PRA of record prior to no longer completing this sensitivity as a part of system categorization.

6. Common Cause For component types where there is no The BSEP PRA model is based on industry Failures industry common cause data available, consensus modeling approaches for its generic common cause data is used common cause identification and value Model : All that is based on the broader population determination. Therefore, this uncertainty of all component types and common meets the consensus model definition.

cause failures. This provides some indication of potential common cause Additionally, NEI 00-04 Table 5-2 directs a faults and may be conservative for sensitivity study to evaluate common cause some component types and non- failures to their 5th and 95th percentile for all conservative for other component types. system categorizations under 50.69. Results of this sensitivity are presented to the IDP.

The sensitivity study shows the impact on SSC importance considering unknowns regarding common cause failures.

Implementation of this sensitivity study is consistent with NEI 00-04 guidance.

7. Containment Venting In the LERF model, the basic event The operator actions and associated modeling containment venting has been equipment failures will be added to the model Model : Internal simplified to a single probability, and prior to implementation of 50.69. Additionally, Events does not use a detailed model of the any uncertainty from the operator actions to operator actions and equipment vent containment will be addressed by the NEI required to perform the action. 00-04 Table 5-2 sensitivity to evaluate human error basic events to their 5th and 95th percentile for all system categorizations under

U.S. Nuclear Regulatory Commission Page 13 of 22 Serial RA-19-0152 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Index Assumption/ Disposition Discussion Uncertainty 50.69, and the results are presented to the IDP. There is no additional sensitivity required to evaluate this uncertainty.

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.

Implementation of this sensitivity study is consistent with NEI 00-04 guidance.

8. Suppression Pool To satisfy the permissive to allow The operator actions to start suppression pool Cooling Permissive Residual Heat Removal (RHR) system cooling or containment sprays reflect operation in suppression pool cooling or operation of these switches as required by Model : Internal containment spray mode, either the procedures. As such, the uncertainty Events water level in the reactor vessel must associated with the action will be addressed be above the two-thirds core height by the NEI 00-04 Table 5-2 sensitivity to level or the operator can bypass this evaluate human error basic events to their 5th requirement by manually positioning the and 95th percentile for all system override switch. To simplify the categorizations under 50.69, and the results modeling, it is assumed that the water are presented to the IDP. There is no level in the reactor vessel is below the additional sensitivity required to evaluate this two-thirds core height level and thus the uncertainty.

operators must override the two-thirds core height permissive to initiate The sensitivity shows the impact on SSC containment sprays or suppression pool importance in light of unknowns regarding cooling. human error probabilities. As such, SSC

U.S. Nuclear Regulatory Commission Page 14 of 22 Serial RA-19-0152 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Index Assumption/ Disposition Discussion Uncertainty importance with respect to the 50.69 application is assessed in light of this uncertainty.

Implementation of this sensitivity study is consistent with NEI 00-04 guidance.

9. Internal Flood The BSEP internal flooding model uses The operator actions related to flood response Related Operator only two flood isolation HEPs, are based on industry accepted practices and Actions XOPER_F25 and XOPER_F60. The the value in the model is conservative.

XOPER_F25 action is based on Additionally, the uncertainty associated with Model: Internal minimum time of 25 min needed to the action will be addressed by the NEI 00-04 Flood respond to a flooding event. The Table 5-2 sensitivity to evaluate human error second XOPER-F60 action had been basic events to their 5th and 95th percentile based on a 1E-3 screening value. A for all system categorizations under 50.69, detailed analysis for this action has and the results are presented to the IDP.

been completed, but the detailed There is no additional sensitivity required to analysis is lower than the value evaluate this uncertainty.

assumed in the model (1E-3).

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.

Implementation of this sensitivity study is consistent with NEI 00-04 guidance.

10. Hotwell Makeup The Component Selection Calculation A sensitivity study will be completed for each Valves Impact on assumed loss of supports to the hotwell system categorized to show the impact of the

U.S. Nuclear Regulatory Commission Page 15 of 22 Serial RA-19-0152 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Index Assumption/ Disposition Discussion Uncertainty the Condensate makeup valves (1-CO-LV-1-2 and 2- hotwell makeup valves on the long-term Storage Tank CO-LV 2) will not impact the long availability of the CST by failing the hotwell term availability of the Condensate makeup valve such that it drains the CST to Model: Fire Storage Tank (CST). the hotwell during an event rather than the CST water being used for ECCS injection.

This sensitivity will be completed with those described in NEI 00-04 Table 5-3 and the results presented to the IDP.

This sensitivity study demonstrates the potential impact on SSC importance based on the potential for the hotwell makeup valves to impact the CST availability. As such, SSC importance with respect to the 50.69 application is assessed considering this uncertainty.

Implementation of this sensitivity study is consistent with NEI 00-04 guidance.

11. Electrical Tunnel and For Electrical Tunnel and Pipe Tunnel, A sensitivity study will be completed for each Pipe Tunnel Hot Gas a hot gas layer was assumed not to system categorized to show the impact of Layer occur. components in the electrical tunnel and pipe tunnel failing by failing all the components in Model: Fire the tunnels. This sensitivity will be completed with those described in NEI 00-04 Table 5-3 and the results presented to the IDP. On-going Fire testing and analysis may show that cable tray fire prorogation is much less likely

U.S. Nuclear Regulatory Commission Page 16 of 22 Serial RA-19-0152 Enclosure Table 1 - Key Assumptions and Uncertainties from the Internal Hazard Models (Internal Events, Internal Flood, and Fire):

Index Assumption/ Disposition Discussion Uncertainty than currently assumed. If this testing and analysis basis is approved this uncertainty will be reevaluated and the sensitivity may no longer need to be performed; the basis for such will be documented.

This sensitivity study demonstrates the potential impact on SSC importance based on the potential for the hotwell makeup valves to impact the CST availability. As such, SSC importance with respect to the 50.69 application is assessed considering this uncertainty.

Implementation of this sensitivity study is consistent with NEI 00-04 guidance.

U.S. Nuclear Regulatory Commission Page 17 of 22 Serial RA-19-0152 Enclosure Table 2 - Key Assumptions and Uncertainties from the External Flood Hazard Model:

Index Assumption/ Disposition Discussion Uncertainty

1. Flood Frequency 500,000 years of storms were Currently, the BSEP External Flood PRA Calculation produced and hence a key model is utilizing a conservative flood assumption was that this number of frequency from the IPEEE. This flood years was sufficient for Duke frequency is two orders of magnitude Energys purpose in using this data. higher than the flood frequency calculated from the 500,000 year storm simulation.

If the flood frequency determined from the 500,000 year storm is implemented into the BSEP External Flood model, a sensitivity study will be completed for each system categorized that evaluates a flood frequency one order of magnitude in each direction (e.g. if the calculated frequency is 2E-6, then the sensitivity study will evaluate 2E-5 and 2E-7). The results of the sensitivity study will show the impact on SSC importance and the results will be presented to the IDP.

Implementation of this flood frequency is tied to BSEP implementation item i, described in Attachment 1.

Implementation of this sensitivity study is consistent with NEI 00-04 guidance.

U.S. Nuclear Regulatory Commission Page 18 of 22 Serial RA-19-0152 Enclosure Table 3 - Key Assumptions and Uncertainties from the High Winds Hazard Model:

Index Assumption/ Disposition Discussion Uncertainty

1. Missile Fragility A statistical model was developed for the The statistical model described by Brunswick missile fragilities instead of this uncertainty is used to develop using TORMIS directly. The statistical the missile hit probability, which is model was based on several TORMIS then used to determine the runs for other plants and used some site probability of failure due to missile and target specific inputs from Brunswick. strike for individual components.

This approach is assumed to provide To evaluate the impact of this reasonable estimates of missile fragilities assumption, the probability of for Brunswick. failure due to missile strike will be adjusted in a sensitivity study to The previous TORMIS analyses used to be completed for each system develop the statistical missile model had categorized. Specifically, the a similar overall layout, distribution of probabilities of failure due to missile types, and total number of missile missile strike will all be multiplied to BNP. In addition, the methods followed by 2 and a separate case dividing to collect the site-specific information for the probabilities of failure due to all plants were the same. As a result, it missile strike by 2 to explore the was assumed that the missile risk data impact of varying missile hit developed by the statistical model is probabilities/failures. These applicable to Brunswick. factors were determined based on engineering judgement from subject matter experts who have evaluated several high winds models. This sensitivity will be completed with those described in

U.S. Nuclear Regulatory Commission Page 19 of 22 Serial RA-19-0152 Enclosure Table 3 - Key Assumptions and Uncertainties from the High Winds Hazard Model:

Index Assumption/ Disposition Discussion Uncertainty NEI 00-04 Table 5-5 and the results presented to the IDP.

This sensitivity study demonstrates the potential impact on SSC importance if the missile hit probabilities were different. As such, SSC importance with respect to the 50.69 application is assessed considering this uncertainty.

Implementation of this sensitivity study is consistent with NEI 00-04 guidance.

2. HW Induced LOOP A high wind induced Loss of Offsite A sensitivity study will be Power (LOOP) event is modeled using performed for each system the switchyard relay house fragility as a categorized to evaluate the impact surrogate. of the switchyard relay house fragility on component importance. The sensitivity study will multiply the switchyard relay house fragility by 2 and a separate case dividing the switchyard relay house fragility by

2. These factors were determined based on engineering judgement

U.S. Nuclear Regulatory Commission Page 20 of 22 Serial RA-19-0152 Enclosure Table 3 - Key Assumptions and Uncertainties from the High Winds Hazard Model:

Index Assumption/ Disposition Discussion Uncertainty from subject matter experts who have evaluated several high winds models.

This sensitivity study will be completed with those described in NEI 00-04 Table 5-5 and the results presented to the IDP.

This sensitivity study demonstrates the potential impact on SSC importance in response to a wind induced failure of the switchyard relay house (and induced LOOP from the relay house failure). As such, SSC importance with respect to the 50.69 application is assessed considering this uncertainty.

Implementation of this sensitivity study is consistent with NEI 00-04 guidance.

3. Human Reliability An HRA multiplier approach was used to The uncertainty associated with perform the BNP high winds human HRA development will be reliability analysis (HRA). addressed by the NEI 00-04 Table 5-5 sensitivity to evaluate human

U.S. Nuclear Regulatory Commission Page 21 of 22 Serial RA-19-0152 Enclosure Table 3 - Key Assumptions and Uncertainties from the High Winds Hazard Model:

Index Assumption/ Disposition Discussion Uncertainty error basic events to their 5th and 95th percentile for all system categorizations under 50.69 and presented to the IDP. There is no additional sensitivity required to evaluate this uncertainty.

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.

Implementation of this sensitivity study is consistent with NEI 00-04 guidance.

The assumption that all tornado events and straight winds with F1 and greater peak gust winds at the BNP site will automatically induce a LOOP event is no longer an assumption in the BNP HW model.

U.S. Nuclear Regulatory Commission Page 22 of 22 Serial RA-19-0152 Enclosure RAI- 4.01.f / 17.01.f

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. The justification should be provided separately for key assumptions and uncertainties related to internal hazard PRAs (including internal fire) and those related to external hazard PRAs supporting this application.

Duke Energy Response to RAI 4.01.f /17.01.f:

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

Serial: RA-19-0152 Brunswick Steam Electric Plan, Unit Nos. 1 and 2 Docket Nos. 50-325 and 50-324 / Renewed License Nos. DPR-71 and DPR-62 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 Brunswick 50.69 PRA Implementation Items

U.S. Nuclear Regulatory Commission Page 1 of 1 Serial RA-19-0152 The table below identifies the items that are required to be completed prior to implementation of 10 CFR 50.69 at Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2. 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.

Brunswick 50.69 PRA Implementation Items Description U Resolution U

i. The BSEP external flood (XF) model Duke Energy will complete a focused hazard is being updated with more scope peer review of the BSEP External detailed analytical modeling as Flood PRA model hazard development prior to implementation of 10 CFR 50.69.

described in response to RAI 11 in Duke Any findings from the focused scope peer Energy letter dated November 2, 2018. review will be resolved and closed per an The additional details need a focused NRC approved process prior to scope peer review. implementing 10 CFR 50.69.

ii. The BSEP FLEX diesel generator (DG) Duke Energy will update the applicable failure rates will be updated using plant- PRA models with FLEX DG failure rates specific data as described in Attachment as described in Attachment 3 of Duke Energy letter dated April 8, 2019 prior to 3 of Duke Energy letter dated April 8, implementing 10 CFR 50.69.

2019.

iii. The BSEP LERF model is being The operator actions and associated updated with additional containment equipment failures modeling containment venting modeling as described in venting will be added to the BSEP LERF model as described in response to RAI response to RAI 4.01 / 17.01 in Duke 4.01 / 17.01 in Duke Energy letter dated Energy letter dated April 8, 2019. April 8, 2019 prior to implementing 10 CFR 50.69.

Serial: RA-19-0152 Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Docket Nos. 50-325 and 50-324 / Renewed License Nos. DPR-71 and DPR-62 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

Amendment AddWonalCondWons Implementation Number Date 282 During the extended EOG Completion Times Upon implementation of authorized by Amendment No. 282, designated Amendment No. 282.

NLOs will be briefed, each shift, regarding cross-tying 480 V E7 bus to the 480 V E8 bus per 0AOP-36.1, Loss of Any 4kV OR 480V Bus.

282 During the extended EOG Completion Times Upon implementation of authorized by Amendment No. 282, designated Amendment No. 282.

NLOs will be briefed, each shift, regarding starting and tying the SUPP-DG to 4160 V emergency bus E4 per plant procedure 0EOP-01-SBO-08, Supplemental DG Alignment.

282 During the extended EOG Completion Times Upon implementation of authorized by Amendment No. 282, designated Amendment No. 282.

NLOs will be briefed, each shift, regarding load shed procedures and alignment of the FLEX diesel generators.

282 During the extended EOG Completion Times Upon implementation of authorized by Amendment No. 282, a Amendment No. 282.

continuous fire watch shall be established for the Unit 1 Cable Spread Room and for the Balance of Plant busses in the Unit 1 Turbine Building 20 foot elevation.

285 The licensee shall not operate the facility within Upon implementation of the MELLLA+ operating domain with Feedwater Amendment No. 285 Temperature Reduction (FWTR), as defined in the Core Operating Limits Report.

TNSER\ UNIT\

Brunswick Unit 1 App. B-4 Amendment No. 85- I

INSERT UNIT 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 1 License Amendment No. [XXX] dated

[DATE].

Duke Energy will complete the implementation items list in Attachment 1 of Duke letter to NRC dated April 8, 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).

Amendment Number Add~onalCond~ons Implementation Date 310 During the extended EOG Completion Times Upon implementation of authorized by Amendment No. 310, dedicated non- Amendment No. 310.

licensed operators (NLOs) shall be briefed, each shift, regarding cross tying the 4160 V emergency bus E2 to 4160 V emergency bus E4 per plant procedure 0AOP-36.1, Loss of Any 4kV OR 480V Bus.

310 During the extended EOG Completion Times Upon implementation of authorized by Amendment No. 310, dedicated Amendment No. 310.

NLOs will be briefed, each shift, regarding cross-tying 480 V E7 bus to the 480 V E8 bus per 0AOP-36.1, Loss of Any 4kV OR 480V Bus.

310 During the extended EOG Completion Times Upon implementation of authorized by Amendment No. 310, dedicated Amendment No. 310.

NLOs will be briefed, each shift, regarding starting and tying the SUPP-DG to 4160 V emergency bus E4 per plant procedure 0EOP-01-SBO-08, Supplemental DG Alignment.

310 During the extended EOG Completion Times Upon implementation of authorized by Amendment No. 310, designated Amendment No. 310.

NLOs will be briefed, each shift, regarding load shed procedures and alignment of the FLEX diesel generators.

310 During the extended EOG Completion Times Upon implementation of authorized by Amendment No. 310, a continuous Amendment No. 310.

fire watch shall be established for the Unit 2 Cable Spread Room and for the Balance of Plant busses in the Unit 2 Turbine Building 20 foot elevation.

310 During the extended EOG Completion Times Upon implementation of authorized by Amendment No. 310, the FLEX Amendment No. 310.

pump and FLEX Unit 2 hose trailer shall be staged at the south side of the Unit 2 Condensate Storage Tank to support rapid deployment in the event the FLEX pump is needed for Unit 2 inventory control.

313 The licensee shall not operate the facility within the Upon implementation of MELLLA+ operating domain with Feedwater Amendment No. 313.

Temperature Reduction (FWTR), as defined in the Core Operating Limits Report.

Brunswick Unit 2 App. B-4 Amendment No.--a+a-

INSERT UNIT 2 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 April 8, 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).

Serial: RA-19-0152 Brunswick Steam Electric Plan, Unit Nos. 1 and 2 Docket Nos. 50-325 and 50-324 / Renewed License Nos. DPR-71 and DPR-62 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 3 Brunswick 50.69 FLEX Diesel Generator Failure Rate Update

U.S. Nuclear Regulatory Commission Page 1 of 1 Serial RA-19-0152 FLEX Diesel Generator Failure Rates As described in Duke Energy letter dated February 13, 2019 (Adams Accession No. ML19044A366), Duke Energy will update the applicable BNP models with Fails to Start plant-specific failure data prior to implementation of 10 CFR 50.69. Duke Energy will continue to update the FLEX DG failure rate as industry and plant-specific failure data are available, in accordance with the PRA standard and Duke Energy PRA model maintenance procedures.