ML21270A002

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Public Meeting NRC Presentation Reactor Pressure Vessel Embrittlement Monitoring and Prediction in Long Term Operation
ML21270A002
Person / Time
Issue date: 10/18/2021
From: Glenna Lappert, Stewart Schneider
NRC/NMSS/DREFS/RRPB
To:
Schneider, Stewart
References
NRC-2021-0174
Download: ML21270A002 (42)
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Text

Reactor Pressure Vessel

  • All meeting Information (i.e., notice, Embrittlement slides, transcript, and summary) will be Monitoring and publicly available 30 days after this meeting Prediction in Long-Term

Will Begin g.php?MTID=e544840cfd44284243639 Shortly a4ad7a9f8974

- Bridge Line: 800-475-0486 Meeting support contact:

- Passcode: 8425538#

Glenna.Lappert@nrc.gov

Reactor Pressure Vessel Embrittlement Monitoring and Prediction in Long-Term Operation Public Meeting October 18, 2021 ADAMS Accession No. ML21270A002

Agenda

  • Welcome / Introductions / Logistics
  • Public Presentations
  • Discussion and Q&A
  • Closing Remarks, and Adjourn 3

Welcome

Introductions

Welcome

Introductions

  • Meeting visuals are through WebEx.
  • Meeting audio is through the bridgeline.
  • Participants are in listen-only mode until the question and answer period. The operator will open phone lines during this time.
  • This is an Information Meeting with a Question and Answer Session. The purpose of this meeting is for the NRC staff to meet directly with individuals to discuss regulatory and technical issues. Attendees will have an opportunity to ask questions of the NRC staff or give feedback about the issues discussed after all presentations; however, the NRC is not actively soliciting comments towards regulatory decisions at this meeting.
  • This meeting is being transcribed. The transcript will be available through the meeting summary.

5

Meeting Purpose

  • Continue discussion of issues from May 2020 public meeting (ML20168A008)

- Regulatory Guide 1.99 Rev 2 (RG 1.99) and 10 CFR 50.61 embrittlement trend curve

- Appendix H surveillance testing

  • This is a technical discussion; no regulatory decisions will be made at todays meeting
  • NRC staff would like feedback on analysis approach and results 6

Background

Monitoring and Prediction of Embrittlement

  • Embrittlement Trend Curve (ETC) provides estimates of change in fracture toughness (T or RTNDT) as a function of fluence
  • Surveillance capsule testing provides monitoring to ensure ETC predicts plant specific behavior properly
  • Together they are used to determine pressure-temperature (PT) limits for normal operation 40yr 60yr 80yr Heatup & Cooldown Embrittlement (T)

ART 40yr 60yr 80yr Max Pressure per ETC Data 10 CFR 50 App. G RTNDT(u)

Operating Time (years)/Fluence Coolant Temperature ART = Adjusted Reference Temperature 7

Ideal Scenario

  • ETC provides conservative predictions of embrittlement
  • Surveillance data covers all operating periods Potential Uncertainty Sources IF ETC under-predicts measurements IF Limited Surveillance Data is Available 40yr 60yr 80yr 40yr 60yr 80yr

?

Embrittlement (T) Embrittlement (T)

Current Data ART  ? ART ETC  ? ETC

?

Future Data?

RTNDT(u) RTNDT(u)

Operating Time (years)/Fluence Operating Time (years)/Fluence 8

Embrittlement Uncertainty Surveillance data ETC Ideal withdrawal Embrittlement (T)

Embrittlement schedule &

Fluence accurate ETC Capsule test date Expected Operating Time (years)/Fluence uncertainty Operating Time (years)/Fluence Large Uncertainty Delayed withdrawal Embrittlement (T) schedule &  ?

Embrittlement Fluence inaccurate ETC Future data Capsule test date Operating Time (years)/Fluence Large uncertainty Operating Time (years)/Fluence Holistic study needed to understand impact of uncertainty 9

Current Perspective of Potential Issue

  • High confidence that currently operating plants remain safe
  • Recent licensing actions remain valid
  • Insufficient embrittlement monitoring and under predictions of reactor vessel embrittlement will eventually (after about 10 years) impact the staffs confidence in the integrity of the reactor pressure vessel in long-term operation, i.e., both safety margins and performance monitoring may be impacted
  • Further work is needed to determine which plants are impacted by this potential issue 10

Embrittlement Trend Curve

  • May 1988, NRC published RG 1.99, which contained an improved embrittlement trend curve (ETC)

- Fit based on 177 datapoints

- Addressed lower than measured predictions (up to 60°F) of embrittlement in some vessels

  • This ETC was re-evaluated for continued adequacy in 2014 (ML13346A003) and in more detail in 2019 (ML19203A089) 11

Issue - ETC

+180°F Deviates Statistically from mean significant

-180°F DT41J = T41J is a measurement of embrittlement representing the shift in transition temperature from brittle to ductile fracture at an impact toughness of 41J 12

Issue - ETC Scatter greater than RG 1.99 standard deviation Limited data at high fluence 13

Issue - ETC Fluence Function Fluence function begins to flatten at the same fluence level underprediction occurs in Slide 12 14

Surveillance Capsule Delays

  • Appendix H to 10 CFR Part 50 requires periodic monitoring of changes in fracture toughness caused by neutron embrittlement

- ASTM standard (E185-82) allows final capsule fluence to be 2X RPV design fluence - plants change (intended 40-year) design fluence to current license length (e.g., 60 or 80 years)

- ASTM standard (for 40 years) permits holding last capsule without testing

  • Commission finding (Perry decision NRC Administrative Letter 97-04) that staff review of requests to change capsule withdrawal schedules is limited to verification of conformance with the ASTM standard (i.e., not based on technical or safety considerations)

- Capsule withdraw and testing repeatedly delayed in some cases to achieve higher fluence 15

License Renewal

  • Regulations are unchanged; surveillance program addressed in guidance

- Guidance provides flexibility for licensees to demonstrate adequate management of RPV embrittlement due to varying plant-specific circumstances

- Continues reliance on Appendix H program using ASTM E185-82

- GALL Report (NUREG-1801, Rev. 1) for license renewal (40 to 60 years)

- shall have at least one capsule with a projected neutron fluence equal to or exceeding the 60-year peak reactor vessel wall neutron fluence prior to the end of the period of extended operation

- Describes use of reconstituted specimens and use of operating restrictions (neutron flux, spectrum, irradiation temperature, etc.)

- GALL-SLR Report (NUREG-2191) for subsequent license renewal (60 to 80 years)

- withdrawal and testing of at least one capsule . . . with a neutron fluence of the capsule between one and two times the peak neutron fluence of interest at the end of the subsequent period of extended operation - or data from a prior tested capsule

- Specifies - it is not acceptable to redirect or postpone the withdrawal and testing of that capsule to achieve a higher neutron fluence that meets the neutron fluence criterion for the subsequent period of extended operation 16

License Renewal in Practice

- Change is evaluated under current approach of conformance verification

  • (Updated) current licensing basis surveillance program for license renewal/subsequent license renewal is then consistent with the program in GALL/GALL-SLR 17

Issue - Appendix H Performance Monitoring Many licensees have delayed capsules (time and/or fluence),

some recent examples:

Plant Capsule # of times

  1. delayed Turkey Point 5 4 Robinson 5 2 Surry U1 5 2 Surry U2 5 2 North Anna U1 4 2 North Anna U2 4 2 St. Lucie U2 4 1 Point Beach 5 1 Capsule withdrawal schedule changes include Not all plants have delayed delays in both time and/or fluence withdrawal of capsules 18

Potential Impact of Issue 1 3 4 5 (1974) (1985) (2001) (2026) 2 60 Years (1978) (2032) 80 Years (2052)

+180°F

-180°F 19

Potential Impact of Issue 400 Fit with potential additional data 350 300 No correction 75 F 150 F 250 Embrittlement DRTNDT,F 200 150 Data Potential additional data 100 RG1.99 Fit to original data Fit through Plant data 50 Updated fit with additional data 0

1.0E+17 2.0E+19 4.0E+19 6.0E+19 8.0E+19 1.0E+20 1.2E+20 1.4E+20 Fluence, n/cm^2 20

Risk-informed Analysis

  • Considered combined effects of surveillance Defense in depth and embrittlement predictions
  • Leveraged 5 principles of Safety Margins Increase in risk is small Integrated risk-informed decision Decision making Making
  • Targeted sample of plant Change data used, but much meets Performance current Monitoring plant specific information regulations not available 21 21

Analysis Assumptions Plates and Forgings

  • Comparisons based on ASTM E900-15 ETC
  • The NRC staff found that the ASTM E900-15 ETC Welds provided the most accurate characterization of this database*

Analysis Assumptions - Fleet Impact Study

  • A targeted sample of 21 plants
  • Emphasis on high fluence plants, with a few low Cu plants and BWRs to round out
  • Determined changes in adjusted reference temperature resulting from switching ETCs -

embrittlement shift delta (ESD)

  • Results used to benchmark ESD range of risk analysis 23

Results - Fleet Impact Study

  • There is a tendency for material reference temperatures to increase when switching from RG 1.99 to ASTM E900-15.
  • Base materials are more likely to see increases in reference temperatures than weld materials.
  • Only a handful of plant limiting materials will have ESDs > 50 °F, and these tend to be at fluences

~6x1019 n/cm2 .

  • Range of ESDs assumed in risk study bounds fleet impact findings.

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Risk of Failure ESD represents the underprediction of RTNDT Large Uncertainties:

  • Actual plant fluence variations
  • Are these analyses bounding?

Unknown plant-specific considerations

  • How much protection do administrative and other operational limits provide against violating the PT limit?

RG 1.99 Revision 2 Update FAVOR Scoping Study, May 6, 2021, TLR RES/DE/CIB-2020-09, Rev. 1, ML21126A326 25

Through-Wall Crack Frequency Results Transient Type Shallow Flaw 1/4T Flaw Comment BWRs must cooldown on BWR P-T Limit saturation curve, so CPF 1x10-6 for all ESDs CPF 1x10-6 for ESD > 40 °F Cooldowns cooldown on licensed limits not plausible.

BWR Saturation CPF 1x10-6 for all ESDs CPF 1x10-6 for all ESDs Cooldown BWR Leak Test, Additional information is CPF 1x10-6 for Cooldown rate CPF 1x10-6 for all ESDs desired to determine if high ESD > 100 °F 50 °F/hour cooldown rates are possible, BWR Leak Test, or ASME Code action will be CPF 1x10-6 for Cooldown rate CPF 1x10-6 for all ESDs pursued to prohibit.

ESD > 100 °F

> 50 °F/hour Additional information on PWR P-T Limit CPF > 1x10-6 for event frequencies is desired CPF >1x10-6 for ESDs 50 °F Cooldowns ESD 20 °F to confirm TWCF< 1x10-6

/year.

PWR Cooldown, CPF < 1x10-6 for most n/a Actual Transients transients 26

Pressurized Thermal Shock Considerations

- Limits of 270 °F for plates, forgings, and axial weld materials, and 300 °F for circumferential weld materials

  • However, through-wall crack frequency calculated with corrected embrittlement less than 1x10-6 for all cases investigated 27

Safety Margins

  • Uncertainties in risk calculations are high and increasing with time
  • Even though the risk appears low, resolving these issues will help maintain the fundamental safety principles that are the basis of plant design and operation
  • Safety margins, as provided by regulations and current license bases, provide reasonable assurance against brittle fracture 28

Safety Margins Illustration Structural limit Reduced Margin Adequate PT-curve using Margin Pressure, psi RG 1.99 Accurate Operating PT-curve Margin Margin Uncertainty Temp, F Uncertainties increasing due to lack of surveillance, but margin is less due to embrittlement underprediction 29

Performance Monitoring

  • Performance monitoring ensures

- Analysis results remain valid with time

- No unexpected (or unmodelled) adverse safety issue occurs

  • Delaying capsule withdrawal for an extended period with the possibility of no future data represents a lack of performance monitoring 30

Analysis Summary

  • With the current state of knowledge, a generalized analysis suggests the overall risk of brittle fracture is low
  • The uncertainty in these results is high and increases with time

- Plant specific details not considered

  • Under certain conditions, safety margins are impacted and are decreasing as uncertainty increases
  • Delaying capsules at high fluence represents a lack of sufficient performance monitoring
  • Issues are plants with fluences > 6x1019 n/cm2 31

Who is Impacted?

  • Embrittlement Underprediction Percentage of Fleet Surpassing Fluence Levels Percentage of PWRs Surpassing Fluence Levels Year\Fluence 6 x 1019 n/cm2 8 x 1019 n/cm2 6 x 1019 n/cm2 8 x 1019 n/cm2 60 years 6% 0% 9% 0%

80 years 22% 10% 34% 15%

- Plant specific details (e.g., limiting material, etc.) may contribute to which plants are impacted

- More work is needed to determine which plants are impacted

  • Lack of Surveillance Data

- Any plant renewing license that chooses to delay last capsule 32

Staff Goals

  • Currently, regulations are sufficient for reasonable assurance of adequate protection against brittle fracture of vessel
  • Staff wants to ensure continued reasonable assurance in long-term operation

- Provide remedies for the identified issues with RPV surveillance requirements and embrittlement predictions, on a risk-informed, performance basis

  • Do not impact those plants that are not adversely affected by the issues

- Plant-specific surveillance data that covers end of license fluence level

- Projected fluence at end of license < ~3 x 1019 n/cm2 33

Options to Meet Goal

  • Plant-specific action
  • Focused regulatory action
  • Generic communication
  • No action 34

Discussion Topics

  • Is the staffs approach to determine the safety impact of the surveillance and embrittlement issues appropriate?
  • What other options could be considered to address these issues?
  • Are there other potential adverse impacts to plant operations (e.g., unnecessary updates to PT limits) that should be considered?
  • Is now the right time to pursue these issues?

Reminder: The NRC is not actively soliciting comments towards regulatory decisions at this meeting.

35

Summary

  • High confidence that currently operating plants remain safe, and recent licensing actions remain valid
  • Issue will eventually (after about 10 years) impact the staff confidence in the integrity of the reactor pressure vessel in long-term operation, i.e., both safety margins and performance monitoring may be impacted
  • Further work is needed to determine which plants are impacted by this issue
  • Proactively ensure continued reasonable assurance though a risk-informed, performance-based solution

- Staff is considering options - desires focused solution to only those conditions adversely impacted by this issue 36

Public Presentations 37

Discussion 38

Where to Find Information Search for docket ID NRC-2021-0174 39

How did we do?

  • NRC Public Meeting Feedback Form

- Link to the form is available at the public meeting website under the meeting notice

  • Or use this QR code:

40

NRC Contacts David Rudland Senior Technical Lead 301-415-1896 David.Rudland@nrc.gov Stewart Schneider Senior Project Manager 301-415-4123 Stewart.Schneider@nrc.gov 41

Thank You 42

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