ML24282A953

Draft Safety Evaluation for Topical Report PWROG 18068 Np, Revision 1, Use of Direct Fracture Toughness for Evaluation of Reactor Pressure Vessel Integrity

ML24282A953
Person / Time
Issue date:	10/18/2024
From:	Walter Kirchner Advisory Committee on Reactor Safeguards
To:	Mirela Gavrilas NRC/EDO
Brown C
References
Download: ML24282A953 (1)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, DC 20555 - 0001 Dr. Mirela Gavrilas Executive Director for Operations U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

DRAFT SAFETY EVALUATION FOR TOPICAL REPORT PWROG-18068-NP, "USE OF DIRECT FRACTURE TOUGHNESS FOR EVALUATION OF REACTOR PRESSURE VESSEL INTEGRITY," REVISION 1

Dear Dr. Gavrilas:

During the 719th meeting of the Advisory Committee on Reactor Safeguards, October 2 through 3, 2024, we completed our review of Pressurized Water Reactor Owners Group (PWROG) topical report PWROG-18068-NP, "Use of Direct Fracture Toughness for Evaluation of Reactor Pressure Vessel Integrity," Revision 1, and the associated safety evaluation (SE). Our Fuels, Materials, & Structures Subcommittee reviewed this matter on September 20, 2024. During these meetings, we had the benefit of discussions with NRC staff and representatives from the PWROG. We also had the benefit of the referenced documents.

CONCLUSIONS AND RECOMMENDATION

1. The use of direct fracture toughness data, proposed in PWROG-18068-NP, Revision 1, represents a significant improvement in the methodology for estimating reactor pressure vessel (RPV) embrittlement.

2. The proposed methodology provides embrittlement predictions more consistent with data and with reduced uncertainty, leading to increased confidence of RPV integrity at anticipated fluences expected during the period of extended operation for subsequent license renewal (SLR).

3. The proposed methodology also leads to more accurate Pressurized Thermal Shock (PTS) evaluation and Pressure-Temperature (P-T) limit curve determination, providing for more flexible operation.

4. The SE should be issued.

BACKGROUND The fracture toughness of the RPV steel provides a key input to calculations that commercial licensees perform to demonstrate the integrity of the vessel during both normal operation and postulated accident conditions (e.g. PTS). Currently, the American Society of Mechanical Engineers (ASME) fracture and arrest toughness (K IC and K IR) correlations, indexed to the October 18, 2024 Dr. M. Gavrilas

reference temperature for nil-ductility transition temperature (RT NDT) of the unirradiated material, describe the toughness of the RPV materials and their variance with temperature.

These correlations were adopted in 1972 as a lower bound representation to a set of 177 toughness values derived from Charpy impact data. These data have been incorporated into the Regulatory Guide (RG) 1.99, Radiation Embrittlement of Reactor Vessel Materials, Revision 2, RPV radiation embrittlement prediction methodology (sum of RT NDT, embrittlement shift prediction, and margin). The use of RT NDT to normalize temperature was designed to account for material heat-to-heat differences in the ductile-brittle transition temperature, thereby collapsing the fracture toughness data onto a single curve. However, this Charpy-based fracture toughness correlation approach is broadly conservative, which propagates forward in the development of PTS limits in Title 10 of the Code of Federal Regulations (10 CFR) 50.61, Fracture toughness requirements for protection against pressurized thermal shock events, as well as the development of P-T limit curves in 10 CFR Part 50, Appendix G, Fracture Toughness Requirements, for operation.

Since the development of the correlation in RG 1.99, Revision 2, the data base for RPV embrittlement has been greatly expanded to 1878 data points. These data have been employed in the development of an updated embrittlement correlation, codified in American Society for Testing and Materials (ASTM) E900-15, Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials. Analysis of this more extensive dataset has shown that the current approach is not the best fracture metric.

Data analysis has shown that the RG 1.99, Revision 2, prediction contains considerable uncertainty. The staff has proposed a path forward to update the RPV embrittlement assessment methodology in SECY-22-0019: Rulemaking Plan for Revision of Embrittlement and Surveillance Requirements for High-Fluence Plants in Long-Term Operation. Alternatives 2 and 3 in SECY-22-0019 would allow the use of E900-15 to estimate the degree of RPV embrittlement. We endorsed the use of Alternative 2 in our April 28, 2022, letter.

In addition to the above issues, the current pressurized water reactor (PWR) fleet initial license period of 40 years is now being extended, or will be extended, with some units now approved for life extension to 80 years. The deviation in the predicted versus measured temperature shift when applying the correlation in RG 1.99, Revision 2, becomes significantly negative (implying reduced margin) at fluences greater than approximately 6 x 1019n/cm2 (E>1 MeV). Several current PWRs are predicted to exceed this value during SLR operation. The methodology proposed in PWROG-18068-NP, Revision 1, makes use of the vastly increased RPV embrittlement database and addresses the issues described above.

DISCUSSION

PWROG-18068-NP, Revision 1, proposes a methodology that justifies the use of direct fracture toughness data to evaluate RPV integrity as an alternative to the requirements of PTS (10 CFR 50.61) and P-T limit curves (10 CFR Part 50, Appendix G). Specifically, the methodology:

a) generates a ductile-brittle transition reference temperature (T0) rather than use of RT NDT, b) adjusts the data for differences between the tested material and the RPV component of interest (master curve approach), c) accounts for test result uncertainty and material variability in the respective RPV component, and d) applies the data using the ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components.

The current approach (RG 1.99, Revision 2) makes use of empirical embrittlement correlations that are based on Charpy data correlated with material toughness. The topical report makes the case for transitioning to the use of a direct toughness approach:

Dr. Transitioning from the current unirradiated reference temperature for nil ductility transition (RT NDT), and the predicted embrittlement shift approach for RPV integrity evaluations to a direct fracture toughness (master curve) approach is expected to benefit RPV operation for license renewal and subsequent license renewal by reducing uncertainties. The available irradiated master curve data show in many cases that substantial conservatism exists []. Thus, application of irradiated master curve data, as an alternative to 10 CFR 50.61 and 10 CFR 50, Appendix G RPV evaluations, is expected to show margin in these analyses.

Establishing a robust fracture toughness basis will ensure public health and safety by reducing uncertainty and enabling a statistical understanding of the actual irradiated RPV fracture toughness.

The approach taken in the proposed topical report uses NRC approved methodologies in ASME Code Section XI, Appendix G, subsection G-2110, Reference Critical Stress Intensity Factor, and the NRC endorsed ASME Code Case N-830, Direct Use of Master Fracture Toughness Curve for Pressure-Retaining Materials of Vessels. The proposed methodology uses the industry consensus ASTM E1921-20, Standard Test Method for Determination of Reference Temperature, T 0, for Ferritic Steels in the Transition Range, and the ASTM E900-15 standard for predicting embrittlement. The approach also ensures uncertainties are properly addressed and appropriately bounding. This methodology represents a significant improvement in estimation of the extent of embrittlement because it makes use of actual toughness data as opposed to empirical correlations. This will allow for both a reduction in and better characterization of uncertainties.

Benefits of this proposed methodology will include improved PTS (10 CFR 50.61) evaluations for license renewal, power uprates, or other operational changes, and extended applicability or improved operating margin of Heat-up/Cool-down (P-T) limit curves (10 CFR Part 50, Appendix G).

We note that this methodology is consistent with and supports the staffs Alternative 2 in SECY-22-0019, which we endorsed in our April 28, 2022, letter report.

SUMMARY

The use of direct fracture toughness data, proposed in PWROG-18068-NP, Revision 1, represents a significant improvement in the methodology for estimating RPV embrittlement. The proposed methodology provides embrittlement predictions more consistent with data and with reduced uncertainty, leading to increased confidence of RPV integrity at anticipated fluences expected during SLR operation. The proposed methodology also leads to more accurate PTS evaluation and P-T limit curve determination, providing for more flexible operation. The SE should be issued.

We are not requesting a formal response from the staff to this letter report.

Sincerely, Walter L. Kirchner Chair Halnon, Gregory signing on behalf of Kirchner, Walter on 10/18/24

Dr. M. Gavrilas

REFERENCES

1. U.S. Nuclear Regulatory Commission, Draft Safety Evaluation for the Pressurized Water Reactor Owners Group Topical Report PWROG18068-NP, Revision 1, Use of Direct Fracture Toughness for Evaluation of RPV Integrity, August 8, 2024 (Agencywide Documents Access and Management System (ADAMS) No. ML24165A212).

2. Pressurized Water Reactor Owners Group Topical Report PWROG-18068-NP, Use of Direct Fracture Toughness for Evaluation of RPV Integrity, Revision 1, July 27, 2021 (ADAMS Accession No. ML21209A933).

3. U.S. Nuclear Regulatory Commission, SECY-22-0019, Rulemaking Plan for the Revision of Embrittlement and Surveillance Requirements for High-Fluence Nuclear Power Plants in Long-Term Operation, March 8, 2022 (ADAMS Accession Package No. ML21314A194).

4. Advisory Committee on Reactor Safeguards, Assessment of the Continued Adequacy of Revision 2 of Regulatory Guide 1.99, November 27, 2019 (ADAMS Accession Package No.

ML19331A231).

5. Advisory Committee on Reactor Safeguards, Rulemaking Plan for the Revision of Embrittlement and Surveillance Requirements for High-Fluence Nuclear Power Plants In Long-Term Operation, April 28, 2022 (ADAMS Accession No. ML22105A325).

6. U.S. Nuclear Regulatory Commission, Regulatory Guide (RG) 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, Revision 20, December 17, 2021 (ADAMS Accession No. ML21181A222).

7. U.S. Nuclear Regulatory Commission, RG 1.99, Radiation Embrittlement of Reactor Vessel Materials, Revision 2, May 31, 1988 (ADAMS Accession No. ML031430205).

8. U.S. Nuclear Regulatory Commission, NUREG1874, Recommended Screening Limits for Pressurized Thermal Shock (PTS), March 1, 2007 (ADAMS Accession No. ML070860156).

9. Ortner S., Styman P., and Long E., The Effects of Flux on the Radiation-Induced Embrittlement of Reactor Pressure Vessel Steels: Review of Current Understanding and Application to High Fluences, Frontiers in Nuclear Engineering, Volume 3, February 2024, pp. 1-20.

10. Erickson, M., Kirk, M., and Stevens, G., EPRI Technical Report 3002016008, Technical Basis for ASME Code Case N-830, Revision 1 (MRP-418): Direct Use of Master Curve Fracture Toughness Curve for Pressure-Retaining Materials of Class 1 Vessels,Section XI, November 19, 2019.

11. ASME Boiler and Pressure Vessel Code,Section XI, Division 1, Code Case N-830, Direct Use of Master Fracture Toughness Curve for Pressure-Retaining Materials of Vessels, September 4, 2014.

12. ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, ASME, July 2023.

Dr. M. Gavrilas

13. ASME Boiler and Pressure Vessel Code,Section XI, Division 1, Appendix G, Fracture Toughness Criteria for Protection Against Failure, ASME, 2017.

14. ASTM E900-15, Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, ASTM International, 2015.

15. ASTM E1921-20, Standard Test Method for Determination of Reference Temperature, T0, for Ferritic Steels in the Transition Range, ASTM International, 2020.

Dr.

SUBJECT:

DRAFT SAFETY EVALUATION FOR TOPICAL REPORT PWROG-18068-NP, REVISION 1, "USE OF DIRECT FRACTURE TOUGHNESS FOR EVALUATION OF REACTOR PRESSURE VESSEL INTEGRITY" Accession No: ML24282A953 Publicly Available (Y/N): Y Sensitive (Y/N): N If Sensitive, which category?

Viewing Rights: NRC Users or ACRS only or See restricted distribution OFFICE ACRS SUNSI Review ACRS ACRS ACRS ACRS NAME CBrown CBrown LBurkhart RKrsek SMoore WKirchner (GHalnon for)

DATE 10/9/24 10/9/24 10/9/24 10/16/24 10/17/24 10/18/24 OFFICIAL RECORD COPYOctober 18, 2024