ML24232A072

From kanterella
Jump to navigation Jump to search
NRC Staff Slides for PWROG Rmc Meeting August 2024
ML24232A072
Person / Time
Issue date: 08/21/2024
From: Meena Khanna
NRC/NRR/DRA
To:
References
Download: ML24232A072 (40)
v • d • e


Text

NRC PRESENTATION PWROG RMC MEETING AUGUST 21, 2024 MEENA KHANNA, ACTING DIRECTOR DIVISION OF RISK ASSESSMENT OFFICE OF NUCLEAR REACTOR REGULATION

AGENDA

  • NRC Perspectives: Data Analysis and Jensen Hughes White Paper - John Lane
  • Group for Risk Evaluation and Assessment Tools Review Status - Antonios Zoulis 2

EDO AND NRR LEADERSHIP CHANGES

  • Mike King - Special Assistant to the EDO for ADVANCE Act Implementation
  • Greg Bowman - Deputy Office Director for Advanced Reactors
  • Mike Franovich - Acting Deputy Office Director for Reactor Programs (Licensing, Oversight, and Administration)
  • Jeremy Bowen - Division Director, NRR, Division of Advance Reactors and Non-power Production and Utilization Facilities
  • Michelle Sampson - Division Director, NRR, Division of New and Renewed Licenses 3

ADVANCE ACT Clear sign from Congress and President that the NRC must find more efficient ways to meet its important safety and security mission.

Provisions include:

Update to mission statement - consideration of efficiency and the benefit of nuclear energy technology to society Leveraging Risk-Informed Decisionmaking and process efficiencies Budgeting flexibilities Develop strategies (e.g., micro-reactors, fusion technology)

International cooperation Reporting requirements to Congress More 4

RISK INITIATIVES

  • Generic Environment Impact Statement Rule Issued August 6, 2024, 89 FR 64166
  • Reactor Accident Analysis Modernization (RAAM)
  • EPRI Technical Report on Alternate 50.69 Methodology
  • TSTF - 603, "Revise the Risk Informed Completion Time Backstop"
  • NRC and Industry Risk Initiatives/Alignment on Priorities
  • RIDM 2.0 Internal Workshops/Initiatives and Public Risk Forum 5

CODE CASE N-752 SUBMITTALS AND PROGRESS MEENA KHANNA

CODE CASE N-752 - BACKGROUND

  • Risk-Informed categorization and treatment for repair/replacement activities in Class 2 and 3 systems for passive components and pressure retaining portions of components
  • Process uses the consequence portion of the EPRI RI-ISI methodology supplemented with additional considerations to define final high and low safety significance
  • Similar to methodology approved for passive categorization in 50.69 LARs 7

CODE CASE N-752 - BACKGROUND, CONTD Cross-disciplinary personnel required includes:

PRA plant operations system design safety or accident analysis Frequency for Feedback and Process Adjustment required at minimum once every two refueling outages Defined treatment requirements address:

design control

procurement & installation

configuration control

corrective action 8

CODE CASE N-752 - SUBMITTALS & REVIEW STATUS PLANTS STATUS Arkansas Nuclear One, Units 1 & 2 Approved Oconee Units 1, 2 & 3 Approved Grand Gulf Unit 1 Approved Riverbend Unit 1 Approved Waterford Unit 3 Approved St Lucie Units 1 & 2 Approved Seabrook Approved Point Beach Units 1 & 2 Approved Hatch Units 1 & 2 Under Review Farley Units 1 & 2 Under Review Vogtle Units 1 & 2 Under Review 9

Q&A 10

ELIJAH DICKSON RISK-INFORMING CONTROL ROOM DESIGN CRITERIA

  • Increased Enrichment (IE) Rulemaking On September 8, 2023, the FRN (88 FR 61986) was published requesting comments (due November 22, 2023) on the Increased Enrichment Rulemaking Regulatory Basis.

On November 6, 2023, the FRN (88 FR 76143) was published, extending the public comment period from the originally November 22, 2023, due date to January 22, 2024.

Three alternatives for control room design criterion of 10 CFR 50.67 and GDC-19 are under consideration.

Within the Regulatory Basis document, FRN (88 FR 61986), staff recommended pursuing rulemaking to amend the control room design criteria and update the current regulatory guidance accordingly with revised assumptions and models and continue to maintain appropriate and prudent safety margins.

RISK-INFORMING CONTROL ROOM (CR) DESIGN CRITERIA 12

RISK-INFORMING CR DESIGN CRITERIA, CONT'D

  • IE Rulemaking, Contd
  • Leveraging Commission-directed PRA policies Severe Reactor Accident policy statement (50 FR 32138; August 8, 1985)

Safety Goals for the Operations of Nuclear Power Plants; Policy Statement (51 FR 30028; August 21, 1986 PRA Policy Statement (60 FR 42622, August 16, 1995)

Commission SRM-SECY-98-144, Staff RequirementsSECY-98-144White Paper on Risk-Informed and Performance-Based Regulations, (ML003753601)

  • Assessing advancements in radiation protection recommendations and standards, as well as PRA methodologies and technology, has propelled the field of nuclear risk assessment to a higher level of maturity.

13

RISK-INFORMING CR DESIGN CRITERIA, CONTD Risk-Informed and Performance-Based Control Room Design Criteria Framework for RG 1.183 Rev. 2.1 Policy: SRM-SECY-98-144, Staff RequirementsSECY-98-144 White Paper on Risk-Informed and Performance-Based Regulation, dated March 1, 1999, defines the terms and Commission expectations for risk-informed and performance-based regulation (ML003753601).

Approach: Leverage IE RM proposal to amend the control room design criteria numerical value to a higher, but still safe performance criteria, but allow flexibilities for higher values with a consideration of the plant-specific risk profile or risk information (ML23032A504).

Purpose:

Enables a performance-based evaluation using traditional deterministic radiological consequence analysis methods within defined risk informed boundaries.

Intention: Provide flexibility when determining how to meet an established acceptance criterion in a way that encourages and rewards safety of the facility.

Source: Information Sharing Meeting - 6/12/24 - ML24162A245, Public Draft Guide Excerpt File - ML24162A323 1 - White Paper: Method for Graded Risk-Informed Performance-Based Control Room Design Criteria Framework, to be released with DG-1425 (RG 1.183 Rev. 2) 14

RISK-INFORMING CR DESIGN CRITERIA, CONTD

  • Framework:
  • Leverages in part, a facilities safe design and operations to justify a higher control room design criterion with a lower plant-specific risk profile using modern probabilistic risk assessment methods.

Criteria Range: regulatory-based and national-and international organization recommendations for radiation exposures under emergency conditions range from 10 to 25 rem TEDE.1 Risk-Metrics: RG 1.174 CDF and LERF Criteria, IPE and IPEEE, and EAB and LPZ.2 Source: Public Workshop (3 of 3) - M24066A177 1 - White Paper: Brock, T, et. al. Control Room Design Criteria and Radiological Health Effects. (ML23027A059) 2 - Public Workshop (3 of 3), Update to RG 1.183, Revision 1 - Alternative Radiological Source Terms, (M24066A177) 15

Q&A 16

ENHANCING OVERSIGHT OF PRA CONFIGURATION CONTROL (PCC)

LUNDY PRESSLEY

Issued Reports on Tabletops and Conducted public meetings with Industry Stakeholders Mar/Jun 2023 Developed Working Group recommendations for near-term path forward and Obtained NRC Mgt.

Alignment on Proposal Feb/Jun 2023 PCC Project Milestones Issued Memo to Establish Cross-Regional Review Panels and SDP Guidance Jun 2024 Started OpESS Gather Data June 2024 Ongoing Public meetings on PCC SDP and gathered feedback from Industry Stakeholders Mar/Apr and Jun 2024 Conducted Regional KM Session on OpESS Jan 2024 Issue near-term oversight guidance (OpESS)

Jan 2024 Revise OpESS as necessary/

if needed 2026 Develop long-term PCC oversight process 2026 / 2027 18

KEY MESSAGES Operating Experience Smart Sample (OpESS) is a balanced and performance-based approach that provides maximum flexibility for regional implementation The OpESS (ML23255A006) is now available for general use.

OpESS KM session was provided to inspectors (Jan 2024), and NRR is available to provide "Just-in-Time" training/support on OpESS to regional inspectors as needed.

Cross-regional review panel guidance (ML24081A131) and PCC specific Examples of Minor determination (ML24152A029) were issued and will ensure consistent dispositioning of issues across the Regional Offices.

OpESS has been utilized at multiple licensees.

19

SIGNIFICANCE DETERMINATION PROCESS (SDP) KEY POINTS

  • Be riskSMART framework was utilized to determine a near-term path forward.
  • Cross-regional review panel charter provides screening guidance and refers to and utilizes existing processes.
  • Reliable consistent SDP outcomes.
  • Maintain regulatory predictability.

20

GRADED APPROACH TO OpESS IMPLEMENTATION

  • OpESS applicability based upon Risk-Informed Programs (RIP)
  • IF FPTI AND (50.69 + SFCP) THEN OpESS = LOW NOTE: Or other extenuating circumstances.

21

OpESS LESSONS LEARNED

  • Minimal resources required (NRC & Licensee) to complete the OpESS.*
  • Inspectors and Licensee have provided positive feedback on the interactions during implementation of OpESS.*
  • Cross Regional Review Panel has been exercised.
  • Some remaining opportunities in 2024 for OpESS implementation.
  • NOTE: Very limited sample size.

22

Q&A 23

NRC PERSPECTIVES: DATA ANALYSIS &

JENSEN HUGHES' PAPER JOHN C LANE

JENSEN HUGHES WHITE PAPER KEY POINTS

  • J-H White Paper Observations:
  • A number of data variables increased a tremendous amount over the 2007 study
  • Average run hours per demand for standby EDP is only 1.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
  • FTR > 1 hr failure rate estimate is too high
  • J-H White Paper Approach:
  • Re-bin the events such that those with slightly more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> into the runs for one hour bin and reevaluate the FTR<1H rate
  • Take the average of the NUREG/CR-6926 (1/15) ratio and the EDG data derived ratio (1/2.8) for a ratio of 1/8.9
  • Multiply the FTR<1H rate with 1/8.9 to determine the FTR>1H rate 25

FTR1H AND FTR>1H CONSIDERATIONS

  • NUREG/CR-6928 splits FTR into FTR1H and FTR>1H based on observations that there were significant differences between rates for FTR1H and FTR>1H
  • The ratio of FTR>1H to FTR1H rates ranges from 0.45 to 0.006" (Section 5.5 of NUREG/CR-6928)
  • The process used for FTR1H and FTR>1H categories is approximate
  • EPIX/IRIS failure records rarely indicate how long a component was operating when it experienced an FTR event
  • Section 5.4 and Appendix A.1.2 of NUREG/CR-6928 describes the process (also used in the 2010, 2015, and 2020 updates) 26

EARLY VS LATE COMPONENT FAILURE RATES OVER TIME 27

  • Use fundamentally the same process in data collection, characterization, and analysis
  • NUREG/CR-6928 and its updates reflect different periods of the industry operating experience
  • NUREG/CR-6928: 1998-2002
  • 2010 Update: 1998-2010
  • 2015 Update: 1998-2015
  • 2020 Update: 2006-2020
  • Section 6 of the 2020 update report (INL/EXT-21-65055) compares the 2020 and 2015 results
  • Of about 300 UR templates, there are 20 templates that have a 50% or more increase from the 2015 update values, 60 templates that have a 50% or more decrease from the 2015 update values NRC OpE Data Analysis Consistency 28
  • Enhancements over time
  • The fixed data period from 1997 has been changed to a rolling 15-year period in response to the industry comments
  • Revisited LOOP recovery data in response to the comments and new information from Jensen Hughes
  • Uncertainties remain in the data
  • Data source the licensee reporting in INPO IRIS database
  • Data coding licensee IRIS training and INL coders
  • For standby equipment and FTR events
  • Demands and run hours in IRIS are mostly estimated (Section 16.5 of INPO 19-002, Rev. 1)
  • While an FTR > 1 hr test could have run a few more hours before it failed; it could also have run only for 2 hrs in the 8-hr test NRC OpE Data Analysis Consistency, CONTD 29

FTR ISSUE BECAME PROMINENT

  • During the 2020 update, we also discovered potential FTR1H and FTR>1H issue since more templates now had larger FTR>1H rate than the FTR1H rate
  • 7 of 18 templates have larger FTR1H rate (MDP, EDG, FAN, CTF, HTG, ACH, and MDC)
  • But now 8 templates have larger FTR>1H rate (TDP, TDP-AFW, TDP-HCI-RCI, EDP, EDP-AFW, PDF, CTG, and EDC)
  • 3 other templates have the same rate (HPCS EDG, SBO EDG, and CHL)
  • The established convention was retained for the 2020 update to avoid a perceived one-time drastic change to the risk profile 30

ISSUES WITH J-H PROPOSED CHANGES

  • Arbitrarily taking the average of two denominators -NUREG/CR-6928 (1/15) ratio and the EDG derived ratio (1/2.8) to arrive at 1/8.9
  • Does not appear to have a basis
  • NUREG/CR-6928 with 1998-2002 data, has wide range of ratios of FTR>1H to FTR1H rates = 0.45 (1/2.2) to 0.006 (1/167)
  • This large variation from 0.45 to 0.006 should be an indicator not to use a simplified approach for general application
  • The 1998-2002 data does not reflect the current performance
  • Components now demonstrate FTR>1 H to FTR1 H greater than 1 31

ISSUES WITH J-H PROPOSED CHANGES, CONT'D

  • The J-H approach was applied to a limited number of the templates
  • The recommended failure rates for FTR>1H in J-H Table 4 appear to use incorrect FTR1H values from the 2020 INL update:
  • Table 4 shows 7.13E-4 for TDP-FTR>1H, which corresponds to a value of 6.35E-3 for FTR1H (6.35E-3
  • 1/8.9 = 7.13E-4)
  • However, the 2020 update value for FTR1H is 2.56E-3 (6.35E-3 is the 2020 update value for FTR>1H)
  • Same error appears for other templates as well, e.g., TDP-HCI-RCI-FTR>1H 32

PATH FORWARD

  • NRC will continue to examine alternative approaches to treat FTR1H and FTR>1H (e.g., combining them into a single failure mode of FTR) in the next parameter update using CY2025 data
  • Will now have two consecutive updates to study the issue
  • The approach would apply to standby equipment FTR templates, other than EDG whose FTLR data is collected in the IRIS database 33

DISCUSSION AND Q&A 34

GROUP FOR RISK EVALUATION AND ASSESSMENT TOOLS REVIEW (GREATR) STATUS ANTONIOS M. ZOULIS

GROUP FOR RISK EVALUATION AND ASSESSMENT TOOLS REVIEW (GREATR)

  • Public meeting held on July 17, 2024 (ML24185A244) to discuss completed updates and propose draft future changes:
  • Enhanced Realism for Offsite/Onsite Power Recovery Probabilities
  • Improved Clarity & Consistency on Initiating Event Assessments
  • Update to the Risk Assessment of Operational Events Handbook, Volume 1, Internal Events
  • Risk Assessment of Operational Events Handbook (RASP) changes will be available in draft during the September 25, 2024, Reactor Oversight Process Bimonthly meeting.
  • Interest shown during public meeting for additional public interaction on proposed model changes for common-cause failures. Tentatively scheduled for October 2024.

36

DISCUSSION AND Q&A 37

BACKGROUND SLIDES

IE RULEMAKING AND DG PROJECT SCHEDULE Milestone Tentative Completion Public Workshops* and Information Meetings**

January 2024 - June 2024 DG Internal Review begins June 2024 DG Internal Review completed October 2024 Pre-Decisional DG Publicly Available to Support ACRS Briefings Fall 2024 ACRS Briefings (Staff to respond as needed)

Fall 2024 - Spring 2025 Increased Enrichment Proposed Rule package to Commission (will include DG referenced in SECY paper)

June 2025 ***

  • The term Workshop means a Comment-Gathering Meeting as described in the NRCs policy statement on public meetings (see 86 FR 14964)
    • The term Information Meeting means NRC will inform attendees and allow questions (see 86 FR 14964)
      • Aligned with IE RM schedule 39

ACRONYMS ASME = American Society of Mechanical Engineers BWROG = Boiling Water Reactor Owner's Group CDF = Core Damage Frequency CETI - Comprehensive Engineering Team Inspection CFR = Code of Federal Regulations DG = Diesel Generator EDG = Emergency Diesel Generator EPRI = Electric Power Research Institute FTR = Failure-to-Run failure rate FTLR: Failure-to-Load-Run FPTI = Fire Protection Team Inspection FRN = Federal Register Notice GREATR = Group for Risk Evaluation and Assessment Tools Review IE = Increased Enrichment KM = Knowledge Management LAR = License Amendment Request LERF = Large Early Release Frequency LCO = Limiting Condition for Operation NFPA = National Fire Protection Association NRC = Nuclear Regulatory Commission NRR = Office of Nuclear Reactor Regulation OpESS = Operating Experience Smart Sample PCC = PRA Configuration Control PRA = Probabilistic Risk Assessment PWROG = Pressurized Water Reactor Owner's Group RASP = Risk Assessment of Operational Events Handbook RG = Regulatory Guide RICT = Risk-Informed Completion Time RIDM = Risk-Informed Decisionmaking RI-ISI = Risk-Informed Inservice Inspection RIP = Risk-Informed Programs SDP = Significance Determination Process SECY = Commission Papers SFCP = Surveillance Frequency Control Program SRM = Staff Requirements Memoranda TR = Technical Report TSTF = Technical Specifications Task Force 40

Retrieved from "https://kanterella.com/w/index.php?title=ML24232A072&oldid=3751195"