ML19060A091

Request for Additional Information Regarding License Amendment Request to Adopt 10 CFR 50.69, Risk Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors

ML19060A091
Person / Time
Site:	Harris
Issue date:	03/18/2019
From:	Barillas M C Plant Licensing Branch II
To:	Hamilton T M Duke Energy Progress
Barillas M C DORL/LPL2-2 301-415-2760
References
EPID L-2018-LLA-0034
Download: ML19060A091 (7)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 18, 2019 Ms. Tanya M. Hamilton Site Vice President Shearon Harris Nuclear Power Plant 5413 Shearon Harris Road M/C HNP01 New Hill, NC 27562-0165

SUBJECT:

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 -REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUEST TO ADOPT 10 CFR 50.69, "RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS AND COMPONENTS FOR NUCLEAR POWER REACTORS" (EPID L-2018-LLA-0034)

Dear Ms. Hamilton:

By application dated February 1, 2018 (Agencywide Documents Access and Management System Accession No. ML 180338768), as supplemented by letter dated October 18, 2018 (ADAMS Accession No. ML 18291A606), Duke Energy Progress, LLC (the licensee) submitted a license amendment request for the Shearon Harris Nuclear Power Plant, Unit 1, requesting to revise the licensing basis to allow for the implementation of the provisions of Title 10 of the Code of Federal Regulations, Part 50, Section 69, "Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors." The U.S. Nuclear Regulatory Commission staff has determined that additional information is needed in order to complete its review. The enclosed request for additional information (RAI) was e-mailed to the licensee in draft form on February 25, 2019, and a clarification call was held on March 7, 2019. An RAI response is due by April 22, 2019. Please note that if a response to this letter is not received by this date, or an acceptable alternate date is not provided in writing, we may deny the application for amendment under the provisions of Title 10 of the Code of Federal Regulations, Part 2, Section 108, "Denial of application for failure to supply information."

T. Hamilton If you have any questions, please contact me at 301-415-2760 or by e-mail to Martha.Barillas@nrc.gov.

Docket No. 50-400

Enclosure:

Request for Additional Information cc: Listserv Sincerely, Martha Barillas, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation DUKE ENERGY PROGRESS, LLC SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 REQUEST FOR ADDITIONAL INFORMATION REGARDING A LICENSE AMENDMENT REQUEST PROPOSING TO ADOPT 10 CFR 50.69, "RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS, AND COMPONENTS FOR NUCLEAR POWER REACTORS" DOCKET NO. 50-400 EPID L-2018-LLA-0034 By letter dated February 1, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 18033B768), as supplemented by letter dated October 18, 2018 (ADAMS Accession No. ML 18291A606), Duke Energy Progress, LLC (Duke Energy, the licensee), submitted a license amendment request (LAR) for Shearon Harris Nuclear Power Plant, Unit 1. The proposed amendment would modify the licensing basis to allow for the implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Section 50.69, "Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors." The provisions of 10 CFR 50.69 allow adjustment of the scope of structures, systems, and components (SSCs) subject to special treatment requirements (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation) based on a method of categorizing SSCs according to their safety significance.

The Nuclear Regulatory Commission (NRC) staff has determined the following request for additional information (RAI) is needed to complete its review. Regulatory Basis Nuclear Energy Institute (NEI) 00-04, Revision 0, "10 CFR 50.69 SSC Categorization Guideline" (ADAMS Accession No. ML052910035), describes a process for determining the safety-significance of SSCs and categorizing them into the four Risk Informed Safety Class categories defined in 10 CFR 50.69. This categorization process is an integrated decisionmaking process that incorporates risk and traditional engineering insights.

NUREG-1855, Revision 1, "Guidelines on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking" (ADAMS Accession No. ML 17062A466), provides guidance on how to treat uncertainties associated with probabilistic risk assessment (PRA) in risk-informed decisionmaking.

Regulatory Guide (RG) 1.200, Revision 2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities" (ADAMS Accession No. ML090410014) describes an acceptable approach for determining whether the quality of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decisionmaking for light-water reactors.

It endorses, with clarifications, the American Society of Mechanical Engineers (ASME)/American Nuclear Society {ANS) PRA Standard ASME/ANS RA-Sa-2009

("ASME/ANS 2009 Standard" or "PRA Standard") (ADAMS Accession No. ML092870592).

Enclosure RAI 5.01 The February 1, 2018, LAR states: The process to categorize each system will be consistent with the guidance in NE/ 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 NE/ 00-04 (i.e., Sections 2 through 12) is integral to providing reasonable confidence" and that "all aspects of NE/ 00-04 must be followed to achieve reasonable confidence in the evaluations required by §50.69(c)(1)(iv)." NEI 00-04 references RG 1.200 as the primary basis for evaluating the technical adequacy of the PRA. RG 1.200 references the ASME/ANS RA-Sa-2009 Standard which requires the identification and documentation of assumptions and 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, Revision 2 defines a key uncertainty as "one that is related to an issue in which there is no consensus approach or model and where the choice of the approach or model is known to have an impact on the risk profile such that it influences a decision being made using the PRA." RG 1.200, Revision 2 defines a key assumption as "one that is made in response to a key source of modeling uncertainty in the knowledge that a different reasonable alternative assumption would produce different results." The term "reasonable alternative" is also defined in RG 1.200, Revision 2. RAI 5 requested the licensee to clarify how key assumptions and (key) uncertainties that could impact the results are identified and included in the evaluation.

In a letter dated October 18, 2018, in the licensee's response to RAI 5, the licensee refers to the integrated risk sensitivity as described in Section 8 of NEI 00-04. For this integrated risk sensitivity study, the unreliability of all low safety significant (LSS) SSCs is increased by a factor of 3 (consistent with NEI 00-04) and the subsequent total risk increase is compared to the RG 1.17 4, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" (ADAMS Accession No. ML 17317A256) acceptable risk increase guidelines.

The licensee stated that this integrated risk sensitivity study, and the subsequent performance monitoring of LSS SSCs, could be used directly to address most of the "in excess of 1000" assumptions and sources of uncertainty instead of identifying and evaluating key assumptions and key uncertainties as described in NUREG-1855, Revision 1. The response also included a table titled "Uncertainties and assumptions not addressed by 10 CFR 50.69 factor of 3 sensitivity/performance monitoring" with 28 entries. The licensee recognized that assumptions and uncertainties that cause SSCs to be excluded from the PRA cannot be addressed by the integrated risk sensitivity.

The entries in the Table are apparently identified and included because they cause SSCs to be excluded.

The dispositions in the Table include dispositions consistent with the NUREG 1855, Revision 1 options of (1) refining the PRA if needed, (2) redefine the application (e.g., add a sensitivity study), or (3) add compensatory measure and monitoring specific to that assumption of uncertainty.

However, the title of the table implies that all the unreported assumptions and uncertainty are evaluated and dispositioned as not being key solely using the factor of 3. Furthermore, most dispositions included in the Table also include the phrase "[a]ny impact of the exclusion of these scenarios on acceptance criteria for 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/high safety significant (HSS) category.

The integrated risk sensitivity study is only performed after all steps in the categorization have been completed and it is not intended to be a change in the risk estimate.

The study simply verifies that the combined impact of any postulated simultaneous degradation in reliability of all LSS SSCs would not result in a significant increases in core damage frequency and large early release frequency.

Therefore, the aggregate risk sensitivity study is intended to capture the uncertainty from relaxation of "special treatment" for candidate LSS SSCs. Other assumptions and uncertainties are related to models and methods used in the PRA and the impact of these assumptions and uncertainties is not considered or included in the integrated risk sensitivity study. NUREG-1855 identifies that one key source of uncertainty is the unknown increase in unreliability associated with the reduced special treatment requirements on LSS SSCs allowed by 10 CFR 50.69. The NU REG states that one acceptable technique to address this specific key source of uncertainty is to increase the unreliability of LSS SSCs by a multiplicative factor in an integrated risk sensitivity study. NEI 00-04 discusses using a factor of 3 to 5 as an acceptable multiplicative factor to address this uncertainty and the licensee selected to use the factor of 3. In contrast, addressing key assumptions and key sources of uncertainty in the PRA might require that SSCs be added to the PRA, might require changes to the model logic, or might require changes in the unreliability (e.g., unreliability increases for unusual uses of SSCs and for consequential failures) greater than the factor of 3 used in the integrated risk sensitivity study. Even for components that are modeled, the integrated risk sensitivity study only addresses the impact of SSCs as they are included in the PRA logic models without addressing any changes to the logic model itself that might be needed to address the key assumption (i.e., because of limitations in scope or level of detail). In addition, the use of the integrated risk sensitivity will result in the licensee identifying potential categorization of a LSS SSC as HSS only if the RG 1.174 risk acceptance guidelines are exceeded.

However, addressing key assumptions and source of uncertainty, can result in a change in categorization even if the RG 1.17 4 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:

a. Clarify which process is used and is meant by the RAI 5 Table title "Uncertainties and Assumptions Not Addressed by 10 CFR 50.69 Factor of 3 Sensitivity/Performance Monitoring" (i.e., which types of uncertainties and assumptions have been addressed by the factor of 3). b. Describe the approach used to identify the assumptions and uncertainties that are used in the base PRA models. c. Describe the approach(es) used to evaluate each assumption and uncertainty to determine whether each assumption and uncertainty is key or not for this application.

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

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

T. Hamilton

SUBJECT:

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 -REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUEST TO ADOPT 10 CFR 50.69, "RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS AND COMPONENTS FOR NUCLEAR POWER REACTORS" (EPID L-2018-LLA-0034)

DATED MARCH 18 2019 DISTRIBUTION:

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