steve snider (.,duke energy®adopt 10 cfr 50.69, “risk-informed categorization and treatment of...

Steve Snider Vice President Nuclear Engineering 526 South Church Street, EC-07H Charlotte, NC 28202 980-373-6195 [email protected]

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

( ., DUKE ENERGY®

U.S. Nuclear Regulatory Commission 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


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


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


degradation in reliability of all LSS SSCs would not result in 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 licensee’s 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 licensee’s 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


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.


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.


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.


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/

Uncertainty Discussion

Disposition

1. Wetwell to drywell vacuum breaker cycling Model : Internal Events

In the BNP LERF model, the wetwell to drywell vacuum breaker fail to close (FTC) basic events are quantified based on an estimated number of vacuum breaker cycles assumed to occur during the event. The model currently assumes the vacuum breakers cycle 10 times. It has been postulated that during large LOCAs the vacuum breakers may experience as little as one or no cycles due to the greater pressure in the drywell.

A sensitivity study will be completed for each system categorized to adjust the wetwell to drywell vacuum breaker fail to close basic event to simulate doubling the vacuum breaker cycles and decreasing the number of vacuum breaker cycles to zero to determine the impact of this assumption. These parameters were determined based on engineering judgement to bound the potential range of cycles and provide insights on SSC importance with respect to this parameter. 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.



Index Assumption/


Disposition

2. Modeling of Control Rod Drive (CRD) Support Systems Model : Internal Events

A conservative assumption is made that the CRD system fails without Reactor Building Closed Cooling Water (RBCCW) cooling and is modeled accordingly. Discussions with Operations and Systems Engineering personnel indicate that the CRD pumps may actually have been run with RBCCW inadvertently isolated with no degradation to the CRD pumps, however there is currently no documentation of this.

A sensitivity study will be completed for each system categorized removing the requirement for RBCCW to cool the CRD pump(s). 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.

3. Emergency Buses E1 through E8 dependency on EDG Building HVAC Model : Internal Events

The Brunswick model assumes that the Emergency Busses E1 through E8 located in the Diesel Generator Building Switchgear rooms do not depend on the building’s HVAC system to perform its primary function during a 24-hour mission time based on generic analyses.

The assumption that switchgears E1 through E8 do not depend on the DGB HVAC system is validated and documented by a GOTHIC analysis as discussed in response to RAI-5.a. This independence is built in Brunswick’s PRA model and therefore, this uncertainty has been resolved and will have no impact to the results of the 10 CFR 50.69 categorization process.



Index Assumption/


Disposition

Resolution of this uncertainty is consistent with the guidance in NUREG-1855 Stage F to refine the PRA analysis.

4. Instrument Air Dryer Bypass Valves Model : Internal Events

It is assumed that a failure of the manual isolation valves on either side of the Instrument Air Dryer Bypass Valve (1-SA-PV-5067) would be discovered within three cycles of operation (6 years).

A sensitivity study will be completed for each system categorized to show the impact of potentially longer period of time during which the valves could be failed closed without discovery. A time period of 12 years will be 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.



Index Assumption/


Disposition

5. Staggered vs Non-Staggered Testing for Common Cause Failure (CCF) groups Model : Internal Events

There is a small amount of non-conservative modeling with only applying staggered testing to all CCF groups.

A sensitivity study will be completed for each system categorized to show the impact of all CCF groups being assigned non-staggered testing. This sensitivity study bounds cases for a variety of testing regimes. This study will be completed with those described in NEI 00-04 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



Index Assumption/


Disposition

in the PRA of record prior to no longer completing this sensitivity as a part of system categorization.

6. Common Cause Failures Model : All

For component types where there is no industry common cause data available, generic common cause data is used that is based on the broader population of all component types and common cause failures. This provides some indication of potential common cause faults and may be conservative for some component types and non-conservative for other component types.

The BSEP PRA model is based on industry consensus modeling approaches for its common cause identification and value determination. Therefore, this uncertainty meets the consensus model definition. Additionally, NEI 00-04 Table 5-2 directs a sensitivity study to evaluate common cause failures to their 5th and 95th percentile for all 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 Model : Internal Events

In the LERF model, the basic event modeling containment venting has been simplified to a single probability, and does not use a detailed model of the operator actions and equipment required to perform the action.

The operator actions and associated equipment failures will be added to the model prior to implementation of 50.69. Additionally, any uncertainty from the operator actions to vent containment will be addressed by the NEI 00-04 Table 5-2 sensitivity to evaluate human error basic events to their 5th and 95th percentile for all system categorizations under



Index Assumption/


Disposition

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 Cooling Permissive Model : Internal Events

To satisfy the permissive to allow Residual Heat Removal (RHR) system operation in suppression pool cooling or containment spray mode, either the water level in the reactor vessel must be above the two-thirds core height level or the operator can bypass this requirement by manually positioning the override switch. To simplify the modeling, it is assumed that the water level in the reactor vessel is below the two-thirds core height level and thus the operators must override the two-thirds core height permissive to initiate containment sprays or suppression pool cooling.

The operator actions to start suppression pool cooling or containment sprays reflect operation of these switches as required by procedures. As such, the uncertainty associated with the action will be addressed by the NEI 00-04 Table 5-2 sensitivity to evaluate human error basic events to their 5th and 95th percentile for all system categorizations under 50.69, and the results are presented to the IDP. 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



Index Assumption/


Disposition

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 Related Operator Actions Model: Internal Flood

The BSEP internal flooding model uses only two flood isolation HEPs, XOPER_F25 and XOPER_F60. The XOPER_F25 action is based on minimum time of 25 min needed to respond to a flooding event. The second XOPER-F60 action had been based on a 1E-3 screening value. A detailed analysis for this action has been completed, but the detailed analysis is lower than the value assumed in the model (1E-3).

The operator actions related to flood response are based on industry accepted practices and the value in the model is conservative. Additionally, the uncertainty associated with the action will be addressed by the NEI 00-04 Table 5-2 sensitivity to evaluate human error basic events to their 5th and 95th percentile for all system categorizations under 50.69, and the results are presented to the IDP. 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.

10. Hotwell Makeup Valves’ Impact on

The Component Selection Calculation assumed loss of supports to the hotwell

A sensitivity study will be completed for each system categorized to show the impact of the



Index Assumption/


Disposition

the Condensate Storage Tank Model: Fire

makeup valves (1-CO-LV-1-2 and 2-CO-LV-1- 2) will not impact the long term availability of the Condensate Storage Tank (CST).

hotwell makeup valves on the long-term availability of the CST by failing the hotwell makeup valve such that it drains the CST to 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 Pipe Tunnel Hot Gas Layer Model: Fire

For Electrical Tunnel and Pipe Tunnel, a hot gas layer was assumed not to occur.

A sensitivity study will be completed for each system categorized to show the impact of components in the electrical tunnel and pipe tunnel failing by failing all the components in 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



Index Assumption/


Disposition

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.


Table 2 – Key Assumptions and Uncertainties from the External Flood Hazard Model:

Index Assumption/ Uncertainty Discussion

Disposition

1. Flood Frequency Calculation

500,000 years of storms were produced and hence a key assumption was that this number of years was sufficient for Duke Energy’s purpose in using this data.

Currently, the BSEP External Flood PRA model is utilizing a conservative flood frequency from the IPEEE. This flood frequency is two orders of magnitude 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.


Table 3 – Key Assumptions and Uncertainties from the High Winds Hazard Model:

Index Assumption/


Disposition

1. Missile Fragility A statistical model was developed for the Brunswick missile fragilities instead of using TORMIS directly. The statistical model was based on several TORMIS runs for other plants and used some site and target specific inputs from Brunswick. This approach is assumed to provide reasonable estimates of missile fragilities for Brunswick. The previous TORMIS analyses used to develop the statistical missile model had a similar overall layout, distribution of missile types, and total number of missile to BNP. In addition, the methods followed to collect the site-specific information for all plants were the same. As a result, it was assumed that the missile risk data developed by the statistical model is applicable to Brunswick.

The statistical model described by this uncertainty is used to develop the missile hit probability, which is then used to determine the probability of failure due to missile strike for individual components. To evaluate the impact of this assumption, the probability of failure due to missile strike will be adjusted in a sensitivity study to be completed for each system categorized. Specifically, the probabilities of failure due to missile strike will all be multiplied by 2 and a separate case dividing the probabilities of failure due to missile strike by 2 to explore the impact of varying missile hit probabilities/failures. These 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




Disposition

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 Power (LOOP) event is modeled using the switchyard relay house fragility as a surrogate.

A sensitivity study will be performed for each system categorized to evaluate the impact 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




Disposition

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 perform the BNP high winds human reliability analysis (HRA).

The uncertainty associated with HRA development will be addressed by the NEI 00-04 Table 5-5 sensitivity to evaluate human




Disposition

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.


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 licensee’s 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



Attachment 1

Brunswick 50.69 PRA Implementation Items

U.S. Nuclear Regulatory Commission Page 1 of 1 Serial RA-19-0152 Attachment 1 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 UDescription UResolution

i. The BSEP external flood (XF) model hazard is being updated with more detailed analytical modeling as described in response to RAI 11 in Duke Energy letter dated November 2, 2018. The additional details need a focused scope peer review.

Duke Energy will complete a focused scope peer review of the BSEP External Flood PRA model hazard development prior to implementation of 10 CFR 50.69. Any findings from the focused scope peer review will be resolved and closed per an NRC approved process prior to implementing 10 CFR 50.69.

ii. The BSEP FLEX diesel generator (DG) failure rates will be updated using plant-specific data as described in Attachment 3 of Duke Energy letter dated April 8, 2019.

Duke Energy will update the applicable PRA models with FLEX DG failure rates as described in Attachment 3 of Duke Energy letter dated April 8, 2019 prior to implementing 10 CFR 50.69.

iii. The BSEP LERF model is being updated with additional containment venting modeling as described in response to RAI 4.01 / 17.01 in Duke Energy letter dated April 8, 2019.

The operator actions and associated equipment failures modeling containment venting will be added to the BSEP LERF model as described in response to RAI 4.01 / 17.01 in Duke Energy letter dated 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



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. -2-85- I

INSERT UNIT 1

Amendment Number

Additional Conditions Implementation Date

[NUMBER] 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, 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).

Upon implementation of Amendment No. [XXX].

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 Number

Additional Conditions Implementation Date

[NUMBER] 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, 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).

Upon implementation of Amendment No. [XXX].

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



Attachment 3

Brunswick 50.69 FLEX Diesel Generator Failure Rate Update

U.S. Nuclear Regulatory Commission Page 1 of 1 Serial RA-19-0152 Attachment 3 FLEX Diesel Generator Failure Rates

As described in Duke Energy letter dated February 13, 2019 (Adams Accession No. ML 19044A366), 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.