ML25135A045

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Staff Presentation to ACRS on New and Advanced PRA Topics
ML25135A045
Person / Time
Issue date: 05/21/2021
From: Alissa Neuhausen
NRC/NRR/DANU, NRC/NRR/DRA/APLC, NRC/RES/DRA
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Download: ML25135A045 (1)
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PRA Analysis for New and Advanced Reactors NRR/DRA, NRR/DANU, RES/DRA May 21, 2021

Staff Presentation Agenda

  • Introductory Remarks
  • Absolute vs. Relative Risk Importance Measures
  • PRA Completeness
  • Existing and Planned Guidance
  • Treatment of Uncertainty and Cliff Edge Effects

Introductions

Presenting Staff Supporting Staff Steven Alferink, NRR/DRA Matt Humberstone, RES/DRA Marty Stutzke, NRR/DANU Kevin Coyne, RES/DRA Alissa Neuhausen, NRR/DRA Erick Ball, RES/DRA Jeffery Wood, RES/DRA

Risk Importance Measures Importance measures are commonly used to determine if a basic event is a significant basic event.

ASME/ANS RA-Sa-2009 significant basic event: a basic event that contributes significantly to the computed risks for a specific hazard group. For internal events, this includes any basic event that has an FV importance greater than 0.005 or a RAW importance greater than 2. For hazard groups that are analyzed using methods and assumptions that can be demonstrated to be conservative or bounding, alternative numerical criteria may be more appropriate, and, if used, should be justified.

QU-F6 DOCUMENT the quantitative definition used for significant basic event, significant cutset, and significant accident sequence. If it is other than the definition used in Part 2, JUSTIFY the alternative.

The staff encourages the use of pre-application engagement, including topical reports, for applicants using criteria different than those in the definition of a significant basic event with respect to internal events.

Significant Basic Events Criteria for Significant Basic Events Reference Criteria ASME/ANS RA-Sa-2009 FV > 0.005, RAW > 2 NEI 00-04 FV > 0.005, RAW > 2, RAWCCF > 20 NUMARC 93-01 RRW > 1.005, RAW > 2 BTP 7-19, Revision 9 CDF > 1x10-6

Timeline of Commission Directions 02/12/09 Risk Metrics for New Reactors 10/22/12 SRM to SECY-12-0081 06/30/14 SRM to SECY-13-0137 03/02/11 SRM to SECY-10-0121

SECY-10-0121

  • In 2009, the staff identified potential issues posed by the lower risk estimates of new reactors using the current risk-informed guidance
  • Presented six possible options for addressing the issues related to the application of risk metrics for new reactors
  • Discussed the advantages and disadvantages of each option
  • In SECY-10-0121, the staff requested approval of its recommendation to modify the risk-informed regulatory guidance to
  • recognize the lower risk profiles of new reactors and
  • prevent a significant decrease in the enhanced levels of safety provided by new reactors

SRM-SECY-10-0121

  • The Commission disapproved the staffs recommendation to modify risk guidance for new reactors
  • The Commission reaffirmed that the existing safety goals, safety performance expectations, subsidiary risk goals and associated risk guidance, key principles and quantitative metrics for implementing risk-informed decision-making are sufficient for new plants
  • New reactors with these enhanced margins and safety features should have greater operational flexibility than current reactors

SECY-12-0081

  • The staff proposed three options for consideration to address the different risk metrics used during new reactor licensing and the risk-informed framework for currently operating reactors
  • The staff addressed the ACRS recommendation and stated that an approach involving relative risk was previously considered but was not pursued

SRM-SECY-12-0081

  • The Commission disapproved the staffs recommendation related to the Reactor Oversight Process
  • The Commission directed the staff to give additional consideration to the use of relative risk metrics, or other options, that would provide a more risk-informed approach to the determination of the significance of inspection findings for new reactors
  • The Commission also directed the staff to provide a technical basis if it believes that the use of relative risk metrics is not a viable option for new reactor oversight

SECY-13-0137

  • The staff performed a technical evaluation of the use of relative risk measures for characterizing the significance of inspection findings
  • The staff concluded that, although the relative risk approach has some merit, the shortcomings of the relative risk approach outweigh its benefits
  • Although the staff is not recommending the relative risk approach, the staff will continue to be open to additional ideas as it develops the recommended integrated risk-informed approach with stakeholder input

SRM-SECY-13-0137

  • The Commission disapproved the staffs recommendation to develop an integrated risk-informed approach for evaluating the safety significance of inspection findings for new reactor designs using qualitative measures to supplement the risk evaluations
  • The Commission directed the staff that the SDP should continue to place emphasis on the use of the existing quantitative measures of the change in plant risk for both operating and new reactors

Timeline of Approvals July 2016 NuScale US600 TBD NuScale US460 April 2014 ESBWR February 2025 Holtec SMR-300

Summary of Approvals Criteria for Significant Basic Events Design Baseline CDF Criteria RG 1.174 1x10-5 FV > 0.005, RAW > 2 ESBWR 2x10-8 FV > 0.01, RAW > 5, RAWCCF > 50 NuScale US600 1x10-7 FV > 0.20, CCDF > 3x10-6, CCDFCCF > 1x10-5 NuScale US460 1x10-7 0.20 < FV < 0.90, CCDF > 3x10-6, CCDFCCF > 1x10-5 Holtec SMR-300 1x10-6 FV > 0.02, RAW > 5, RAWCCF > 35 1x10-7 FV > 0.20, RAW > 30, RAWCCF > 60

Topical Report Limitations and Conditions

  • Risk significance determination methodologies are often submitted and approved as topical reports on a case-by-case basis
  • These topical reports typically contain limitations and conditions that:
  • Limit approval to a specific plant design
  • Limit its use to a PRA that the staff finds to be technically acceptable and addresses all modes and all hazards
  • Ensure a risk-informed approach is taken and additional consideration of uncertainties, sensitivities, engineering evaluations and regulations, and maintaining sufficient defense in depth and safety margin will be used to determine a complete list of risk-significant SSCs

Consideration of Criteria for ALWRs

  • The staff recognizes the traditional relative risk criteria are not always appropriate for designs with substantially lower risk profiles
  • Using these criteria could incorrectly identify components as risk significant
  • Approvals of a variety of applications using different thresholds and sliding scale approaches has been efficient and effective
  • The staff has achieved reasonable timelines for approving topical reports based on similarities with past approvals
  • It is not clear that a generic approach would meet all applicant needs
  • The staff has not identified any clear benefit to review timelines
  • The staff is concerned that a generic approach would be too broad or too restrictive, which would negate any advantage of a generic approach
  • Applicants may still submit design-specific methodologies
  • Consideration of absolute risk criteria is consistent with the approach for advanced non-LWR plants

Item Relative Absolute Risk Significant Basic Event A basic event that contributes significantly to baseline risk.

It is defined as any basic event that has an FV [Fussell-Vesely] importance greater than 0.005 or a RAW [risk achievement worth] importance greater than 2 where the importance is normalized against the baseline total integrated risk or risk of a specific combination of source of radioactive material, hazard, and plant operating state.

A basic event that contributes significantly to an absolute risk significance criterion selected for RIDM. It is defined as any basic event that a) contributes at least 1% to any identified absolute risk target; or b) would result in exceeding the criterion if the basic event is assumed to fail with a probability of 1.0.

Risk Significant Event Sequence or Event Sequence Family An event sequence or event sequence family that, when rank-ordered by decreasing frequency, contributes a specified percentage of the baseline risk, or that individually contributes more than a specified percentage of the risk. For this version of the Standard, the aggregate percentage for the set is 95%, and the individual event sequence or event sequence family percentage is 1% of the total integrated risk or risk of a specific combination of source of radioactive material, hazard, and plant operating state.

An event sequence or event sequence family included in a PRA model, defined at the functional or systematic level, that makes a significant contribution to an absolute risk target selected for RIDM. It is defined as any event sequence or event sequence family that contributes at least 1% to any identified absolute risk target.

Risk Significance Definitions in the Non-LWR PRA Standard (ASME/ANS RA-S-1.4-2021)

LMPs Use of Absolute Risk Significance

  • NEI 18-04, Rev. 1, Section 3.3.6, pp. 28-29: The historical approach to evaluating risk importance produced only the relative importance of each basic event because the formulas are normalized against the total calculated risk for the plant, R(base). For advanced non-LWR plants, the frequencies of events involving a release of radioactive material may be very small and those events with releases may involve very small source terms compared with releases from an LWR core damage event. This underlines the importance of using absolute vs. relative risk metrics to establish LBE and SSC risk significance. Hence, it is appropriate to evaluate risk significance not only on a relative basis but also on an absolute basis.
  • ACRS letter, March 19, 2019: For classifying SSCs, We previously addressed these concepts in reports of September 26, 2007, and April 26, 2012, and find the approach logically sound, based on safety importance, and not bound by historical practice.
  • SRM-SECY-19-0117, May 29, 2020
  • Approved
  • The staff should remain open to continuous, critical examination of its thinking regarding approaches and metrics for the licensing of this coming class of advanced reactors.

LMP Identification of Risk-Significant LBEs

  • An AOO, DBE, or BDBE is regarded as risk-significant if the combination of the upper bound (95th percentile) estimates of the frequency and consequence of the LBE are within 1%

of the F-C Target AND the upper bound 30-day TEDE dose at the EAB exceeds 2.5 mrem.

  • Informs the evaluation of defense-in-depth adequacy.

Source: NEI 18-04, Rev. 1, Figure 3-4.

LMP Identification of Risk-Significant SSCs Source: NEI 18-04, Rev. 1, Figure 4-2.

NSRST SSCs: Non-safety-related SSCs that perform risk-significant functions or perform functions that are necessary for defense-in-depth adequacy SR SSCs may (or may not) be risk significant An SSC is regarded as risk-significant if its PRA Safety Function is:

  • Required to keep one or more LBEs inside the F-C Target based on mean frequencies and consequences; or
  • If the total frequency LBEs that involve failure of the SSC PRA Safety Function contributes at least 1% to any of the LMP cumulative risk targets. The LMP cumulative risk targets include:
  • Maintaining the frequency of exceeding 100 mrem to less than 1/plant-year;
  • Meeting the NRC safety goal QHO for individual risk of early fatality; and
  • Meeting the NRC safety goal QHO for individual risk of latent cancer fatality.

PRA Completeness The PRA is to be commensurate with the application for which it is intended and the role the PRA results play in the integrated decisionmaking process.

PRA Completeness for LWRs under Parts 50 and 52

  • SRM-SECY-22-0052, Proposed Rule: Alignment of Licensing Processes and Lessons Learned from New Reactor Licensing
  • Expand the applicability of 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors, to allow design certification applicants, construction permit holders, and combined license holders to risk-inform the categorization of structures, systems, and components.
  • Require operating license applicants for large light-water power reactors to submit a description and analysis of design features for the prevention and mitigation significant severe accidents.

PRA Evolution for LWRs Under Parts 50 and 52 DC: 10 CFR 52.47(a)(27)

ML: 10 CFR 52.157(a)(31)

Essentially Complete Design Except for Site-Specific Elements SDA: 10 CFR 52.137(a)(25)

Essential Complete Design for Major Portions Addressed in SDA Application Except for Site-Specific Elements COL: 10 CFR 52.47(a)(46)

- DC: 10 CFR 52.47(d)(1)

- SDA: 10 CFR 52.47(c)(1)

- ML: 10 CFR 52.47(e)(1)

As-Built Design Operations 10 CFR 50.71(h)(2)

As-Operated Design Construction Permit 10 CFR 50.34(a)(14)

Preliminary Design Operating License 10 CFR 50.34(b)(13)

As-Built Design Level of Detail in the PRA Plant Representation in the PRA Site Has Been Selected Bounding Site Operating Experience Accrues; Periodic PRA Maintenance and Update Early Site Permit

  • Ability to Perform Plant Walkdowns
  • Part 50 OL PRA and Part 52 Fuel Load PRA Have the Same Level of Detail and Degree of Plant Representation
  • Fuel Load and Start-Up Testing
  • Operations As-Operated Design

PRA Completeness for Part 50 LWR CPAs ISG on PRA and Severe Accident Information for CPs The scope and technical acceptability of the CP application PRA depend on the intended use of the information and the level of design maturity.

Consistent with DC/COL-ISG-028, Capability Category I of an NRC endorsed PRA standard is acceptable. DC/COL-ISG-028 is one example of the results of the application process described in the PRA Standard and endorsed in RG 1.200 to determine whether every supporting requirement (SR) is needed for a high-level requirement.

The applicants justification for the scope and level of detail of any PRA or alternative risk evaluation should be consistent with the intended uses of the information from those assessments. The staff should review the applicants plan for assessing any risk contributors not addressed by a PRA or alternative risk evaluation.

PRA Completeness for LMP RG 1.253, Guidance for a Technology-Inclusive Content-of-Application Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors RG 1.253, Section C.3, pp. 12-13 incorporates recommendations provided by the ACRS on December 20, 2023: The applicant should provide justification that the PRA has been performed in such a way that the PRA results are reasonable given the level of maturity of the design, and that the SAR provides sufficient information to support the CP findings. The applicant should also include any necessary commitments to upgrade and maintain the PRA so that its completion status at the OL stage is consistent with its intended uses.

The current state of practice is to use an internal events, at power PRA model as the foundation for representing the response of a facility to perturbations from normal operations. RG 1.253 establishes that the minimum PRA needed for an LMP-based, non-LWR, construction permit application under 10 CFR Part 50 can be an internal events, at-power, reactor PRA logic model. The basis for the staff position in RG 1.253 is that the use of PRA in the LMP methodology is an integral aspect of that approach and is well-defined for the entire design lifecycle.

Existing and Planned Guidance PRA Standard Endorsement Level 1/LERF RG 1.200 rev 4 Level 2 RG 1.200 rev 4 Level 3 No Plans Non-LWR RG 1.247 Advanced LWR RG 1.200 rev 5 LPSD RG 1.200 rev 5 Multi-unit No Plans

  • NRC staff are actively involved in ASME/ANS Joint Committee on Nuclear Risk Management (JCNRM) activities
  • Plans to endorse new Level 1/LERF and Level 2 PRA standards in next revision of RG 1.200

Existing and Planned Guidance

  • NUREG -1855 - Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking
  • 2024 Level 1/LERF PRA standard
  • 2024 Level 2 PRA standard
  • Advanced LWR Considerations
  • Etc.
  • NUREG-2122 - Glossary of Risk-Related Terms in Support of Risk-Informed Decisionmaking
  • Approximately 200 new or revised terms and definitions
  • Discussion updated to be inclusive of non-LWRs
  • Final draft has been prepared for inter-office review
  • Advanced Reactor Risk Metrics
  • RES staff developing white paper
  • Develops risk metric attributes and options
  • Existing Guidance
  • RG 1.253, Rev. 0, Guidance for a Technology-Inclusive Content of Application Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors, March 2024.
  • Planned Guidance
  • DG-1435 (proposed RG 1.247, Rev. 1) - Schedule to be determined.
  • Guidance associated with the proposed final Part 53 rulemaking package
  • DRA-ISG-2024-XXXX, Rev. 0, Content of Risk Assessment and Severe Accident Information in Light-Water Power Reactor Construction Permit Applications - public comments resolved Existing and Planned Guidance

Cliff-Edge Effects Outline

  • Historical Context
  • Some Definitions
  • Examples
  • Mitigation approaches

Cliff-Edge - Some Historical Context Comment on Safety Goal Policy Statement (E.V. Gilby, Gilby Associates, UK, June 16, 1983):

An approach to regulation based upon PRA must lead to emphasis on events beyond design objectives to see if a "cliff edge" situation applies i.e. there must be no significant increase in consequences for relatively small decreases in probability.

In other words studies of margin to failure or margin to a significant degradation of system performance are very important. It is not clear how this emphasis is to be achieved.

Safety Goal Policy Statement (51 FR 30028, August 21, 1986)

The Commission is aware that uncertainties are not caused by use of quantitative methodology in decisionmaking but are merely highlighted through use of the quantification process.

it is necessary that proper attention be given not only to the range of uncertainty surrounding probabilistic estimates, but also to the phenomenology that most influences the uncertainties. For this reason, sensitivity studies should be performed to determine those uncertainties most important to the probabilistic estimates. The results of sensitivity of studies should be displayed showing, for example, the range of variation together with the underlying science or engineering assumptions that dominate this variation.

Cliff-Edge Effect - Some Definitions

flooding risks are of concern due to a cliff-edge effect, in that the safety consequences of a flooding event may increase sharply with a small increase in the flooding level.

  • RG 1.233, Rev 0 (conditionally endorses NEI 18-04, Rev 1) (2020) - cliff-edge effects, involve a dramatic change in plant behavior caused by a small change in a plant parameter
  • NLWR PRA Standard, ASME/ANS RA-S-1.4-2021 (conditionally endorsed by trial use RG 1.247)- an instance of a sudden large variation in a plant conditions in response to a small variation in an input (e.g., change in flood height, grid perturbation based on voltage or frequency exceeding a breaker trip set point)
  • IAEA SSR-2/1, Safety of Nuclear Power Plants: Design (Rev 1, 2016) - in a nuclear power plant, is an instance of severely abnormal plant behaviour caused by an abrupt transition from one plant status to another following a small deviation in a plant parameter, and thus a sudden large variation in plant conditions in response to a small variation in an input.

Cliff-Edge Effects - Some examples Type of Cliff Edge Effect Examples Threshold Effects (margin)

Flood barrier height Protection system setpoints High sensitivity to Parameter Changes Large sensitivity in risk results to small reliability changes (e.g.,

high Birnbaum importance)

Beakaway effects (e.g., chemistry/rapid oxidation)

Phenomena Severe accident behavior Radionuclide transport and release Digital I&C performance Unknowns Hazards screened from further consideration Unknowns

Cliff-Edge Effects - Mitigation (PRA Policy Statement)

(2) PRA and associated analyses (e.g., sensitivity studies, uncertainty analyses, and importance measures) should be used in regulatory matters, where practical within the bounds of the state of-the-art

  • The Policy statements of consideration also address uncertainties The Commission understands that uncertainties exist in any regulatory approach. These uncertainties are derived from knowledge limitations that are not created by PRA, but are often exposed by it.

Cliff-Edge Effects - Mitigation (IAEA)

IAEA Safety Standards series No. SSR-2/1 (2016) - Safety of Nuclear Power Plants:

Design

  • 4.11(b) Application of defence-in-depth - provide assurance that a small deviation in a plant parameter does not lead to a cliff edge effect.
  • 5.21 External hazards - The design of the plant shall provide for an adequate margin to protect items important to safety against levels of external hazards to be considered for design, derived from the hazard evaluation for the site, and to avoid cliff edge effects.
  • 5.73 Safety Analysis - The safety analysis shall provide assurance that uncertainties have been given adequate consideration in the design of the plant and in particular that adequate margins are available to avoid cliff edge effects
  • 5.76(b) Probabilistic approach - situations in which small deviations in plant parameters could give rise to large variations in plant conditions (cliff edge effects) will be prevented IAEA-TECDOC-1791 (2016) - Considerations on the Application of the IAEA Safety Requirements for the Design of Nuclear Power Plants (e.g., Section 8, Cliff Edge Effects and Safety Margins, and Section 9, Design for External Hazards)

Cliff-Effects Mitigation (10 CFR 50.43(e))

10 CFR 50.43(e) (see 72 FR 49494, Aug. 28, 2007)

  • Requirement to demonstrate the performance of new safety features that differ significantly from evolutionary light water reactors or that use simplified, inherent, passive, or other innovative means to accomplish their safety functions
  • Ensure that these new safety features will perform as predicted in the applicants safety analysis report, to provide sufficient data to validate analytical codes, and that the effects of systems interactions are acceptable.
  • May be met with either separate effects or integral system tests; prototype tests; or a combination of tests, analyses, and operating experience.

Cliff-Edge Effects - Mitigation (RG 1.174)

  • RG 1.174 Integrated Decision Making Process - In making a regulatory decision, risk insights (including their associated uncertainties) are integrated with considerations of defense in depth and safety margins. The degree to which the risk insights (and their uncertainties) play a role, and therefore the need for detailed staff review, depends on the application.

Cliff-Edge Effects - Mitigation (RG 1.174)

  • Guidelines are intended to provide assurance that proposed increases in CDF and LERF are small and are consistent with the intent of the Commissions Safety Goal Policy Statement.
  • the boundaries between regions are not definitive. In applying these guidelines, it is particularly important to recognize that the risk metrics calculated using PRA models are a function of the assumptions and approximations made in the development of those models.
  • In the context of integrated decisionmaking, the boundaries between regions are not definitive; the numerical values associated with defining the regions in the figure are to be interpreted as indicative values only.

Source: NEI 18-04, Rev. 1, Figure 3-1.

Cliff Edge Effects - Mitigation (NEI 18-04)

NEI 18-04, Rev 1

  • Event sequences with upper 95th percentile frequencies less than 5x10-7/plant-year are retained in the PRA results and used to confirm there are no cliff edge effects.
  • The integrated decisionmaking process considers if the PRA provides an adequate assessment of cliff-edge effects.

RG 1.233 Clarification - Demarcations are not a hard and fast cutoff - should be considered in conjunction with integrated decisionmaking panel, defense in depth, uncertainties, and an assessment of cliff edge effects.

Cliff Edge Effects Complex and Unknown Cliff Edge Effects Types of effects Complex system interactions Sensitivity and Threshold Effects Phenomena modeling Lack of completeness Examples Combined effects of SSC and human failures Support system interactions Synergistic effects Flood barriers High risk sensitivity to small reliability changes High thermal hydraulic sensitivity Radiological release pathway modeling Rare deterministic health effects Passive system modeling Digital I&C Uknown or unidentified:

Hazards Operating states Phenomena Etc.

Analysis Methods PRA Modeling and Uncertainty Assessments Testing Deterministic use of integrated thermal hydraulic codes External hazard assessment Assess relative risk measures Sensitivity analysis Phenomena Identification and Ranking Deterministic modeling Safety Margin Defense-in-Depth Operating experience, testing Systematic process for identifying risk contributors Comparisons to risk analyses for other plants Completeness uncertainty analysis of unmodeled items

Cliff-Edge Effects - Some concluding thoughts

  • Assessing cliff-edge effects
  • Identify credible hazards/accident sequences with reasonable confidence so the population of unknowns is small contribution to facility risk;
  • For known issues, make defensible determinations to accept or mitigate potential cliff-edge effect based on an integrated risk informed process
  • Mitigation should consider use of diverse means to accomplish safety functions, defense in depth, and adequacy of safety margins
  • Design considerations
  • Key sources of uncertainty should be identified and assessed
  • Design control processes should prevent introduction of new, unanalyzed, potential cliff edge effects
  • Demonstrate performance of new safety features using testing, analysis, and/or operating experience

Cliff-Edge Effects - Some concluding thoughts

  • Uncertainty quantification and sensitivity studies should be used to assess cliff-edge potential (e.g., consider 95th percentile frequencies rather than relying on a numerical cutoff frequency)
  • Numerical guidelines should be assessed using an integrated decision-making process (i.e., risk-informed vice risk-based)
  • Relative risk importance measures can highlight areas of high sensitivity (e.g.,

risk achievement worth or Birnbaum)

  • PRA screening evaluations should consider cliff edge effects to avoid premature elimination of importance sequences (use of inclusive modeling to avoid loss of important risk contributors)
  • PRA model should be as complete as possible (e.g., operating states, hazards, radiological sources, phenomena)
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