ML24066A177

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RG 1-183 Public Workshop (3 of 3) - NRC Slides - ML24066A177
ML24066A177
Person / Time
Issue date: 03/08/2024
From: David Garmon-Candelaria
NRC/NRR/DRA/ARCB
To:
References
RG-1.183, Rev 1
Download: ML24066A177 (41)
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Text

Public Workshop (3 of 3)

Update to RG 1.183, Revision 1 - Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors March 8, 2024 ADAMS Accession No. ML24066A177

Introductions

2 Please ensure to identify yourself before speaking throughout the meeting.

NRC Opening Remarks Michael Franovich, Director Division of Risk Assessment Office of Nuclear Reactor Regulation 3

Workshop Agenda Workshop Format

- Breaks: Approximately every 1 - 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />

- Discussion: Encouraged throughout the workshop Project Schedule Generalized Framework for Developing a Graded, Risk-Informed Method for the Control Room Design Criteria Non-LOCA Release Fraction Calculations External Presentations (as applicable) 4

Workshop Format 5

86 FR 14964 - NRC Policy on Enhancing Participation in NRC Public Meetings Comment-Gathering Meeting - Purpose is for the NRC to obtain feedback on regulatory issues and NRC actions - in this case for an update to RG 1.183 Focused on allowing participants opportunity to provide opinions, perspectives and feedback NRC staff will take notes during meeting and use these notes to inform decision-making Meeting summaries will include high-level summaries of notes NRC staff has taken 5

RG 1.183 Update Stakeholder Involvement

  • Public Workshops and Information Meetings

- 3 Workshops

- 2 Information Meetings

  • Contact Project Lead
  • Meeting Feedback
  • ACRS Meeting - approximately November 2024
  • Draft Guide Comment Period - approximately June 2025

- Relationship with Proposed Increased Enrichment Rulemaking

  • Comment period on Proposed Increased Enrichment Rulemaking Regulatory Basis ended on January 22, 2024 6

Project Schedule 7

Milestone Tentative Completion Public Workshops* and Information Meetings**

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

November 2024 Increased Enrichment Proposed Rule package to Commission (will include DG referenced in SECY paper)

December 2024 DG Available for Public Comment via Proposed IAW Increased Enrichment Rulemaking SECY June 2025 (Estimated)

  • 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)

Generalized Framework for Developing a Graded Risk-informed Method for the Control Room Design Criteria Elijah Dickson, Ph.D.

Sr. Reliability and Risk Analyst Division of Risk Assessment Office of Nuclear Reactor Regulation

Overview BACKGROUND METHOD ANALYSIS METHODOLOGY ANALYSIS AND RESULTS 9

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Background===

  • Consider Commission-directed PRA-related policies, which provide direction on certain changes to the development and implementation of its regulations using risk-informed, and ultimately performance based, approaches.

1985 Severe Reactor Accident policy statement (50 FR 32138; August 8, 1985)

PRA Policy Statement (60 FR 42622, August 16, 1995)

SRM-SECY-98-144, Staff RequirementsSECY-98-144White Paper on Risk-Informed and Performance-Based Regulations (ADAMS Accession No. ML003753601), March 1, 1999 SECY-19-0036, Application of the Single Failure Criterion to NuScale Power LLC's Inadvertent Actuation Block Valves, July 2, 2019 NRC memorandum to the Executive Director for Operations, Implementing Commission Direction on Applying Risk-Informed Principles in Regulatory Decision Making, (ADAMS Accession No. ML19319C832), November 18, 2019

  • Present a generalized method to develop a framework for a graded risk-informed, performance-based acceptance criteria.

Purpose would be to enable a clear deterministic evaluation using traditional deterministic radiological consequence analyses within defined risk-informed boundaries.

  • Assess the method through several examples using various risk-and deterministic metrics commonly used in nuclear risk analyses.

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Method

  • Method: systematic mapping of a predetermined range of acceptable dose-based control room design values onto a specified risk-metric.
  • Range of acceptable design values would be defined based on an assessment of regulatory precedence and organizations responsible for making radiation protection recommendations.
  • Risk-metrics would be defined based on an assessment of those commonly used in nuclear risk analysis.
  • In practice, a lower plant-specific risk metric would justify a higher control room design criterion.

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Analysis: Risk-and Deterministic Metrics Used in Nuclear Safety Analyses Survey and assess several risk-and deterministic metrics applied in nuclear safety analyses.

Deterministic Metrics:

  • Exclusion Area Boundary
  • Low Population Zone
  • Develop data sets to define boundary conditions and binning:
  • EAB and LPZ datasets extracted from licensee's approved analysis of record for their initial adoption of 10 CFR 50.67.

Risk Metrics:

  • Core Damage Frequency
  • Large Early Release
  • Develop data sets to define boundary conditions and binning:
  • 1980s CDF dataset extracted from the Individual Plant Examinations (IPEs) results.

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METHODOLOGY ANALYSIS AND RESULTS

  • Developed three examples of a graded control room design criterion with a TEDE criteria range being arbitrarily selected for illustration purpose only.
  • Example 1: Risk-informed model based on RG 1.174 CDF criteria
  • Example 2: Risk-informed model based on IPE CDF data
  • Example 3: Deterministic model based on EAB or LPZ data
  • Assess each example to understand how the framework would apply to the operating reactor fleets plant-specific information and contemporary understanding of radiological health effects and radiation protection recommendations.

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Example 1: Risk-informed model based on RG 1.174 that leverages facility-specific risk information (TEDE criteria range is arbitrarily selected for illustration only)

Risk-Metric Range: Regulatory Guide 1.174 CDF Criteria within three-bins.

Criteria Range (Arbitrary): 5 to 10 rem TEDE Scientific Underpinning: not necessarily consistent with HPS PS010-4, ICRP Pub. 109, NCRP Pub. 180 14

Example 1: Analysis and Results (TEDE criteria range is arbitrarily selected for illustration only)

Data: Develop modern CDF datasets from the NRC SPAR models and various risk-informed applications representing contemporary risk profiles form the 2020s.

Filter 2020s CDFs into each bin.

~ 57% of the facilities have plant-specific CDF results lower than 1.E-5 and would corresponding to a control room design value of 10 rem TEDE.

~ 43% of the facilities have plant specific CDF values between 1E-4 and 1E-5 and would correspond to a control room design criteria value of 7.5 rem TEDE.

Table 2: Example 1 Analysis Results of Binning CDF Data into RG 1.174 Histogram Figure 2: Example 1 Analysis Results of Binning CDF Data into RG 1.174 Histogram CDF Data Binned Percent of Facilities Binned CDF < 1.E-5 55 59%

1E-4 CDF 1E-5 39 41%

CDF > 1E-4 (or no PRA) 0 0%

RG 1.174 CDF Ranges Analysis of CDF Data Binning 15

Example 2: Analysis and Results (TEDE criteria range is arbitrarily selected for illustration only)

Data: Develop Contemporary CDF datasets from the NRC SPAR models and various risk-informed applications to assess contemporary risk profiles form the 2020s.

Filter 2020s CDFs into each bin.

Compare 1980s and 2020 CDF data Improving risk profiles would allow for flexible control room design criteria

~ 93% of the fleet would have a control room design criteria of 25 rem TEDE.

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Example 2: Risk-informed model based on IPE CDF data (TEDE criteria range is arbitrarily selected for illustration only)

Risk-Metric Range: A CDF dataset developed from IPEs results representing understood risk-profiles from the 1980s. IPE data was quartiled to define lower and upper boundaries for each of the five bins.

Criteria Range (Arbitrary): 5 to 25 rem TEDE Scientific Underpinning: EPA PAG Manual, Part 100 Site Criteria 17

Example 3: Deterministic model based on EAB or LPZ data (TEDE criteria range is arbitrarily selected for illustration only)

Each bin boundary is defined as multiples of 5, from zero to

25. Each bin is correlated to a corresponding value 25 rem EAB siting value (can also use LPZ value, next slide)

Criteria Range (Arbitrary): 5 to 25 rem TEDE Scientific Underpinning: EPA PAG Manual, Part 100 Site Criteria 18

Example 3: Analysis and Results (TEDE criteria range is arbitrarily selected for illustration only)

Data: Developed from reactor fleets analysis of record results for the EAB and LPZ MHA-LOCA results from approved license amendment requests to adopt 10 CFR 50.67.

Filter Facility-specific EAB and LPZ data into each bin.

EAB criteria is more limiting that the LPZ criteria.

~ 42% of the facilities would obtain a control room design criteria of 25 rem with EAB model

~ 82% of the facilities would obtain a control room design criteria of 25 rem with LPZ model 19

Supporting / Background 20

DBAs vs. Actual Event Sequences The DBAs are not intended to be actual event sequences but, rather, are intended to be surrogates to enable a deterministic evaluation of the performance of plant engineered safety features, such as the control room habitability systems. (see the following three slides for illustration)

These analyses are intentionally conservative in order to address uncertainties in accident progression, fission product transport, and atmospheric dispersion, and to provide desirable defense-in-depth.

An actual accident sequence may not progress as modeled by a DBA (e.g., may involve multiple failures), resulting in a greater challenge to the control room systems. If the challenge exceeds the design basis, radiation exposures in the control room may exceed those envisioned in the design bases of the control room habitability features. In such an event, the facility radiation protection and emergency response programs implement measures to minimize radiation dose. (see the slide for accident management guidelines)

The NRCs comprehensive radiation protection and emergency response regulatory framework and the conservatism in the agencys deterministic radiological consequence analysis assumptions ensures reasonable assurance that adequate protective measures can and would be taken in an actual radiological emergency.

The range of considered acceptance criteria are well-established values based on the probability and severity of the accident being assessed.

Sources of Health Effects Information:

RG 8.29, Instruction Concerning Risks from Occupational Radiation Exposure.

Information Assistance Request NRR-2022-019, Control Room Design Criteria and Radiological Health Effects, U.S. Nuclear Regulatory Commission, June 2023 (ML23027A059).

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Containment Leakage Pathway Model Containment (Drywell)

Reactor Building Control Room Environment SGTS Filter CREVS Filter Unfiltered In leakage Containment Leakage, La Containmen t Sprays EAB LPZ 22

Engineered Safety Feature Pathway Model Suppression Pool Reactor Building Control Room Environment SGTS Filter CREVS Filter Unfiltered In leakage ESF Leakage EAB LPZ 23

Base-Case Nodalization:

RG 1.183 Rev. 0 Main Steamline Isolation Valve Leakage Reactor Vessel Control Room Environment CREVS Filter Unfiltered In leakage MSL A (with broken recirculation line)

Volume 1 Failed Open MSIVs Turbine Stop Valves Seismic 1/QA Cat. I Aerosol Depositio n

EAB LPZ MSL A Volume 2 MSL B, C, D Vol 1 Containment Containmen t Sprays Deterministi c pipe break Reactor Building 24

Control Room Design Criterion of 10 CFR 50.67 and GDC-19: Typical Role of Accident Management Guidelines General Operating Procedures Alarm Response Procedures Abnormal Operating Procedures Emergency Operating Procedures Severe Accident Management Guideline Normal Critical safety functions threatened Critical safety functions lost or degraded Core degraded and beyond FLEX Support Guidelines Severe Accident Management Guideline Normal Operation s

Off-normal /

Abnormal Events Accidents Severe Accidents Accident Managemen t Regime Supporting Procedures 25

Non-LOCA Gap Release Fractions for RG 1.183 Rev. 2 March 8, 2024 Workshop Chris Van Wert, Senior Technical Advisor Joseph Messina, Reactor Systems Engineer Division of Safety Systems Office of Nuclear Reactor Regulation 26

Overview

  • Purpose
  • Background
  • Assumptions
  • Draft PWR Gap Release Fractions
  • Draft BWR Gap Release Fractions
  • Enrichment Sensitivity Studies 27

Purpose

  • The NRC is seeking feedback on the assumptions used to generate the draft non-LOCA gap release fractions presented, particularly on the draft power history curves 28

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Background===

  • A gap release fractions are the amount of a fission product that will be located in the fuel-cladding gap of a fuel rod relative to the amount produced and therefore be released if rod failure occurs
  • The Non-LOCA release fractions in RG 1.183 Rev. 1 are applicable if operation remains below the power history provided in the RG.
  • The power history curve extends to 68 GWd/MTU (rod-average)
  • Appendix I provides an analytical procedure to calculate release fractions for other power histories, fuel designs, etc.

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Background===

  • The ANS-5.4 (2011) standard is used to calculate the volatile fission product release fractions
  • Details and derivation of the ANS-5.4 standard is provided in NUREG/CR-7003
  • Database used to develop ANS-5.4 included IE and HBU fuel 30 Figure 1 from NUREG/CR-7003, Background and Derivation of ANS-5.4 Standard Fission Product Release Model

Assumptions

  • Axial peaking factors:

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Draft PWR Gap Release Fractions Group Rev 1 Fraction Draft Rev 2 Fraction (17x17)

I-131 0.07 0.04 I-132 0.07 0.04 Kr-85 0.40 0.34 Other Noble Gases 0.06 0.03 Other Halogens 0.04 0.02 Alkali Metals 0.20 0.17 32

Draft BWR Gap Release Fractions 33 Group Rev 1 Fraction Draft Rev 2 Fraction I-131 0.03 0.01 I-132 0.03 0.01 Kr-85 0.32 0.26 Other Noble Gases 0.03 0.01 Other Halogens 0.02 0.01 Alkali Metals 0.16 0.13

Enrichment Sensitivity Study

  • Non-LOCA releases were found to increase with increasing enrichment 34 Enrichment (wt%)

Total FGR

(%)

I-132 BE Release Frac Kr-85m BE Release Frac Kr-85 BE Release Frac 4

21.09 4.62E-03 1.71E-03 2.20E-01 5

20.53 5.42E-03 2.01E-03 2.42E-01 6

23.23 6.08E-03 2.25E-03 2.54E-01 7

25.31 6.54E-03 2.42E-03 2.68E-01 8

27.38 6.90E-03 2.56E-03 2.75E-01

Trends

  • Long-lived radionuclides tend to peak late in life
  • Short-lived radionuclides tend to peak at the knee of the power history curve 35 0

0.001 0.002 0.003 0.004 0.005 0.006 0.007 0.008 0

20 40 60 80 Best Estimate Release Fraction Rod Average Burnup (GWd/MTU)

I-132 0

0.05 0.1 0.15 0.2 0.25 0.3 0

20 40 60 80 Best Estimate Release Fraction Rod-Average Burnup (GWd/MTU)

Kr-85

Comments/Questions?

36

External Presentations 37

Break 38

Discussion 39

RG 1.183 Update Stakeholder Involvement

- Reminder -

  • Public Workshops and Information Meetings

- 3 Workshops

- 2 Information Meetings

  • Contact Project Lead
  • Meeting Feedback
  • ACRS Meeting - approximately November 2024
  • Draft Guide Comment Period - approximately June 2025

- Relationship with Proposed Increased Enrichment Rulemaking

  • Comment period on Proposed Increased Enrichment Rulemaking Regulatory Basis ends on January 22, 2024 40

Contacts David Garmon, Project Lead/

Technical Contact david.garmon@nrc.gov 301-415-3512 Chris Van Wert, Technical Contact christopher.vanwert@nrc.gov 301-415-3217 https://www.nrc.gov/pmns/mtg?do=details&Code=20231392 How did we do?

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