2191 R0 XI.M31: Difference between revisions

Revision as of 19:12, 4 October 2024

XI.M31 REACTOR VESSEL MATERIAL SURVEILLANCE

Program Description

Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix H, requires implementation of a Reactor Vessel Material Surveillance program when the peak neutron fluence at the end of the design life of the vessel exceeds 10¹⁷ n/cm² (E>1MeV). The purpose of the material surveillance program is to monitor the changes in fracture toughness to the ferritic reactor vessel beltline materials. As described in Regulatory Issue Summary 2014-11, beltline materials are those ferritic reactor vessel materials with a projected neutron fluence greater than 10¹⁷ n/cm² (E>1MeV) at the end of the license period (for example, the subsequent period of extended operation), which are evaluated to identify the extent of neutron radiation embrittlement for the material. The surveillance capsules contain reactor vessel material specimens and are located near the inside vessel wall in the beltline region so that the specimens duplicate, as closely as possible, the neutron spectrum, temperature history, and maximum neutron fluence experienced at the reactor vessel’s inner surface. Because of the location of the capsules between the reactor core and the reactor vessel wall, surveillance capsules typically receive neutron fluence exposures that are higher than the inner surface of the reactor vessel. This allows surveillance capsules to be withdrawn and tested prior to the inner surface receiving an equivalent neutron fluence so that the surveillance test results bound the conditions at the end of the subsequent period of extended operation.

The surveillance program must meet the requirements of 10 CFR Part 50, Appendix H. The American Society for Testing Materials (ASTM) standards incorporated by reference in 10 CFR Part 50, Appendix H, include recommended surveillance capsule withdrawal schedules based on plant operation during the original 40-year license term. Therefore, standby capsules or capsules containing reconstituted specimens may need to be incorporated into the Reactor Vessel Material Surveillance program to provide reasonable assurance of appropriate monitoring during the subsequent period of extended operation. Surveillance capsules are designed and located to permit insertion of replacement capsules. If standby capsule(s) will be incorporated into the Reactor Vessel Material Surveillance program for withdrawal and testing to address the subsequent period of extended operation and the capsule(s) has already been withdrawn from the reactor vessel and placed in storage, the surveillance capsule(s) should be reinserted, if necessary, in a location with an appropriate lead factor to ensure that the neutron fluence of the surveillance capsule and the test results will, at a minimum, bound the peak neutron fluence of interest projected to the end of the subsequent period of extended operation.

This program includes withdrawal and testing of at least one surveillance capsule addressing the subsequent period of extended operation, with a neutron fluence of the surveillance capsule between one and two times the peak neutron fluence of interest projected at the end of the subsequent period of extended operation. The peak reactor vessel neutron fluence of interest at the end of the subsequent period of extended operation should address the time-limited aging analyses (TLAAs) described in the following sections of the Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants (SRP-SLR), as applicable: Sections 4.2.2.1.2 (Upper-Shelf Energy), 4.2.3.1.3 (Pressurized Thermal Shock) and 4.2.3.1.4 (Pressure-Temperature Limits) for pressurized water reactors (PWRs); and Sections 4.2.2.1.2 (Upper-Shelf Energy), 4.2.3.1.4 (Pressure Temperature Limits), 4.2.3.1.5 (Elimination of Boiling Water Reactor Circumferential Weld Inspection) and 4.2.3.1.6 (Boiling Water Reactor Axial Welds) for boiling water reactors (BWRs). If a capsule meeting this neutron fluence criterion has not been tested prior to entering the subsequent period of extended operation, then the program includes the withdrawal and testing (or alternatively the retrieval from storage, reinsertion for additional neutron fluence accumulation, if necessary, and testing) of one capsule addressing the subsequent period of extended operation to meet this criterion. If a surveillance capsule was previously identified for withdrawal and testing to address the initial period of extended operation, 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.

An integrated surveillance program (ISP), alternatively, may be considered for a set of reactors that have similar design and operating features, as described in 10 CFR Part 50, Appendix H, Paragraph III.C. The plant-specific implementation of the ISP is consistent with the latest version of the ISP plan that has received approval by the U.S. Nuclear Regulatory Commission (US NRC) for the subsequent period of extended operation.

The objective of this Reactor Vessel Material Surveillance program is to provide sufficient material data and dosimetry to (a) monitor irradiation embrittlement to a neutron fluence level which is greater than the projected peak neutron fluence of interest projected to the end of the subsequent period of extended operation, and (b) provide adequate dosimetry monitoring during the subsequent period of extended operation. If surveillance capsules are not withdrawn during the subsequent period of extended operation, provisions are made to perform dosimetry monitoring. An in-vessel standby capsule, or a standby capsule which has been retrieved from storage and reinserted, when coupled with the use of an US NRC-approved methodology for determining neutron fluence consistent with Regulatory Guide (RG) 1.190, “Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence,” provides an acceptable means of dosimetry monitoring.

The program is a condition monitoring program that measures the increase in Charpy V-notch 30 foot-pound (ft-lb) transition temperature and the drop in the upper-shelf energy (USE) as a function of neutron fluence and irradiation temperature. The data from this surveillance program are used to monitor neutron irradiation embrittlement of the reactor vessel, and are inputs to the neutron embrittlement TLAAs described in Section 4.2 of the SRP-SLR. The Reactor Vessel Material Surveillance program is also used in conjunction with the Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report, AMP X.M2, “Neutron Fluence Monitoring.”

All surveillance capsules, including those previously withdrawn from the reactor vessel, must meet the test procedures and reporting requirements of the applicable ASTM standards referenced in 10 CFR Part 50, Appendix H, to the extent practicable, for the configuration of the specimens in the capsule. Any changes to the surveillance capsule withdrawal schedule, including the incorporation and change of status of standby capsules to capsules scheduled for withdrawal and testing (or alternatively retrieval from storage, reinsertion for additional neutron fluence accumulation, if necessary, and testing) under this program must be approved by the US NRC prior to implementation, in accordance with 10 CFR Part 50, Appendix H, Paragraph III.B.3. Standby capsules placed in storage (e.g., withdrawn from the reactor vessel) are maintained for possible future insertion, and tested specimens are retained in storage for possible reconstitution.

Evaluation and Technical Basis

The Reactor Vessel Material Surveillance program is plant-specific and depends on the composition and availability of the limiting materials, the availability of surveillance capsules, and the projected neutron fluence at the end of the subsequent period of extended operation. In accordance with 10 CFR Part 50, Appendix H, an applicant submits its proposed withdrawal schedule for US NRC approval prior to implementation.

1. Scope of Program: The program addresses neutron embrittlement of all ferritic reactor vessel beltline materials as defined by 10 CFR 50, Appendix G, as the region of the reactor vessel that directly surrounds the effective height of the active core and the adjacent regions of the reactor vessel that are predicted to experience sufficient neutron damage to be considered in the selection of the limiting material with regard to radiation damage. Materials with a projected neutron fluence greater than 10¹⁷ n/cm² (E>1MeV) at the end of the license period (for example, the subsequent period of extended operation), are considered to experience sufficient neutron damage to be included in the beltline. Materials monitored within the licensee’s existing, materials surveillance program typically continue to serve as the basis for the reactor vessel surveillance aging management program (AMP).

For ISPs, the plant-specific implementation of the ISP in this Reactor Vessel Material Surveillance program is maintained consistent with the latest version of the ISP plan that has received approval by the US NRC for the subsequent period of extended operation.

2. Preventive Actions: This program is a surveillance program; no preventive actions are identified.

3. Parameters Monitored or Inspected: The program monitors reduction of fracture toughness of reactor vessel beltline materials due to neutron irradiation embrittlement, through the periodic testing of material specimens at different intervals that have been irradiated in the surveillance capsules that are a part of the program. The program also monitors the long-term operating conditions of the reactor vessel (i.e., vessel beltline operating temperature and neutron fluence, the latter using GALL-SLR AMP X.M2, “Neutron Fluence Monitoring”) that could affect neutron irradiation embrittlement of the reactor vessel.

The program uses two parameters to monitor the effects of neutron irradiation: (a) the increase in the Charpy V-notch 30 ft-lb transition temperature, and (b) the drop in the Charpy V-notch USE. The program uses neutron dosimeters to monitor the neutron fluence of the surveillance capsule and to provide information to benchmark neutron fluence calculations. Low melting point elements or low melting point eutectic alloys may be used as a check on peak specimen irradiation temperature. Results from these temperature monitors are used to ensure that the exposure temperature of the surveillance capsule is consistent with the reactor vessel beltline operating temperature. The Charpy V-notch specimens, neutron dosimeters, and temperature monitors are placed in capsules that are located within the reactor vessel; the capsules are withdrawn periodically to monitor the reduction in fracture toughness due to neutron irradiation.

This program includes withdrawal and testing of at least one capsule addressing the subsequent period of extended operation 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. The peak reactor vessel neutron fluence of interest at the end of the subsequent period of extended operation should address the TLAAs as described in the following sections of the SRP-SLR, as applicable: Sections 4.2.2.1.2 (Upper-Shelf Energy), 4.2.3.1.3 (Pressurized Thermal Shock) and 4.2.3.1.4 (Pressure-Temperature Limits) for PWRs; and Sections 4.2.2.1.2 (Upper-Shelf Energy), 4.2.3.1.4 (Pressure Temperature Limits), 4.2.3.1.5 (Elimination of Boiling Water Reactor Circumferential Weld Inspection) and 4.2.3.1.6 (Boiling Water Reactor Axial Welds) for BWRs. If a capsule meeting this neutron fluence criterion has not been tested prior to entering the subsequent period of extended operation, then the program includes the withdrawal and testing (or alternatively the retrieval from storage, reinsertion for additional neutron fluence accumulation, if necessary, and testing) of one capsule to address the subsequent period of extended operation to meet this criterion. If a surveillance capsule was previously identified for withdrawal and testing to address the initial period of extended operation, 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. Test results are reported consistent with the requirements of 10 CFR Part 50, Appendix H. Because the degree of neutron irradiation embrittlement is a function of the neutron fluence, calculations of the capsule neutron fluence, the reactor vessel wall neutron fluence, and the peak neutron fluence of interest projected to the end of the subsequent period of extended operation are important parts of the program. The methods used to determine both capsule and reactor vessel wall neutron fluence values are consistent with RG 1.190, as described in GALL-SLR AMP X.M2, “Neutron Fluence Monitoring.”

This program uses separate dosimeter capsules or ex-vessel dosimeters to monitor neutron fluence independent of the specimen capsules if there are no surveillance capsules installed in the reactor vessel.

4. Detection of Aging Effects: Reactor vessel materials are monitored by a surveillance program in which surveillance capsules are withdrawn from the reactor vessel and tested consistent with 10 CFR Part 50, Appendix H. The ASTM standards referenced in [Appendix H describe the methods used to monitor irradiation embrittlement (as described in Element 3, above), selection of materials, and the withdrawal schedule for surveillance capsules. Because the withdrawal schedule in Table 1 of ASTM E185-82 is based on plant operation during the original 40-year license term, standby capsules may need to be incorporated into the program as capsules to be tested within a withdrawal schedule that covers the subsequent period of extended operation. Alternatively, this program can propose implementation of in-vessel irradiation of capsule(s) with reconstituted specimens from previously tested capsules and appropriate neutron fluence monitoring.

Alternatively, an ISP for the subsequent period of extended operation may be considered for a set of reactors that have similar design and operating features as described in 10 CFR Part 50, Appendix H, Paragraph III.C. For an ISP, in some cases the plant Reactor Vessel Material Surveillance program may result in no surveillance capsules being irradiated in the plant’s reactor vessel, with the plant relying on data from testing of the ISP capsules from the host plants of the capsules. Additional surveillance capsules may also be needed for the subsequent period of extended operation for an ISP. For ISPs, the plant-specific implementation of the ISP in the Reactor Vessel Material Surveillance program is maintained consistent with the latest version of the ISP plan that has received approval by the US NRC for the subsequent period of extended operation. The plant implements dosimetry monitoring as required by the approved ISP to meet the provision of 10 CFR Part 50, Appendix H, Paragraph III.C.1.b, that each reactor in an ISP has an adequate dosimetry program.

If no in-vessel surveillance capsules are available, an alternative neutron fluence monitoring program uses alternative dosimetry, either from in-vessel capsules or ex-vessel capsules, to monitor neutron fluence during the subsequent period of extended operation. The methods used in this alternative neutron fluence monitoring program are consistent with RG 1.190, including appropriate benchmarking, as described in GALL-SLR Report AMP X.M2, “Neutron Fluence Monitoring.”

If not previously approved, the capsule withdrawal schedule for the Reactor Vessel Material Surveillance program shall be submitted as part of the subsequent license renewal application.

If the reactor vessel exposure conditions (neutron flux, spectrum, irradiation temperature, etc.) are altered, then the basis for the projection of neutron fluence to the end of the subsequent period of extended operation is reviewed and appropriate modifications are made to the Reactor Vessel Material Surveillance program. Any changes to the Reactor Vessel Material Surveillance program must be submitted for US NRC review and approval in accordance with 10 CFR Part 50, Appendix H, prior to implementation.

5. Monitoring and Trending: The program provides data on neutron embrittlement of the reactor vessel materials and neutron fluence data. These data are used to evaluate the TLAAs on neutron irradiation embrittlement (e.g., USE, pressurized thermal shock (PTS), pressure-temperature limits evaluations, etc.) as needed to demonstrate compliance with the requirements of 10 CFR Part 50, Appendix G, and 10 CFR 50.61 or 10 CFR 50.61a for the subsequent period of extended operation, as described in the SRP-SLR, Section 4.2.

The plant-specific surveillance program or ISP has at least one capsule that has attained or will attain neutron fluence between one and two times the peak reactor vessel wall neutron fluence of interest at the end of the subsequent period of extended operation. If a capsule meeting this neutron fluence criterion has not been tested previously, then the program includes withdrawal and testing (or alternatively the retrieval from storage, reinsertion for additional neutron fluence accumulation, if necessary, and testing) of one capsule addressing the subsequent period of extended operation. (If a surveillance capsule was previously identified for withdrawal and testing to address the initial period of extended operation, 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.) The program withdraws, and subsequently tests, the capsule(s) at an outage in which the capsule receives a neutron fluence of between one and one and two times the peak reactor vessel neutron fluence of interest at the end of the subsequent period of extended operation. Test results from this capsule are reported as described in 10 CFR Part 50, Appendix H. If an existing standby capsule that has been previously withdrawn from the reactor vessel is used for testing to meet the neutron fluence criterion for the subsequent period of extended operation and the capsule does not require additional irradiation, then that (formerly standby) capsule is incorporated into the surveillance capsule withdrawal schedule of the Reactor Vessel Material Surveillance program upon receipt of the subsequently renewed license, and reporting of the test results is consistent with 10 CFR Part 50, Appendix H, with the “withdrawal date” of the capsule considered to be no later than the date of the subsequently renewed license. If a plant has ample capsules remaining for future use, all pulled and tested samples placed in storage with reactor vessel neutron fluence less than 37.5 percent of the projected neutron fluence at the end of the subsequent period of extended operation, may be discarded. All pulled and tested samples with a neutron fluence greater than 37.5 percent of the projected reactor vessel neutron fluence at the end of the subsequent period of extended operation and all untested capsules are placed in storage (these specimens and capsules are saved for possible future reconstitution and reinsertion use) unless the applicant has gained US NRC approval to discard the pulled and tested samples or capsules.

If an applicant does not have ample capsules remaining for future use, all withdrawn and tested capsule specimens are placed in storage. These specimens are saved for future reconstitution, in case irradiation embrittlement monitoring by the surveillance program is reestablished. Tested surveillance specimens may be withdrawn from storage and used in research activities (e.g., microstructural examination, mechanical testing, and/or additional irradiation) without US NRC approval if the licensee determines that a sufficient number of specimens will remain.

6. Acceptance Criteria: Although there are no specific acceptance criteria that apply to the surveillance data themselves, the program meets the requirements of 10 CFR Part 50, Appendix H. The reactor vessel embrittlement projections are used to demonstrate compliance with the requirements of 10 CFR Part 50, Appendix G, and 10 CFR 50.61 or 10 CFR 50.61a, and acceptability of other plant-specific analyses, throughout the subsequent period of extended operation, as described in the SRP-SLR, Section 4.2.

7. Corrective Actions: Results that do not meet the acceptance criteria are addressed in the applicant’s corrective action program under those specific portions of the quality assurance (QA) program that are used to meet Criterion XVI, “Corrective Action,” of 10 CFR Part 50, Appendix B. Appendix A of the GALL-SLR Report describes how an applicant may apply its 10 CFR Part 50, Appendix B, QA program to fulfill the corrective actions element of this AMP for both safety-related and nonsafety-related structures and components (SCs) within the scope of this program.

Since the data from this program are used for reactor vessel embrittlement projections to comply with regulations (e.g., 10 CFR Part 50, Appendix G, requirements, and 10 CFR 50.61 or 10 CFR 50.61a limits) through the subsequent period of extended operation, corrective actions would be necessary if these requirements are not satisfied, or if this program fails to meet the requirements of 10 CFR Part 50, Appendix H. If plant operating characteristics exceed the operating restrictions identified previously, such as a lower reactor vessel operating temperature or higher neutron fluence, this program provides reasonable assurance that the impact of actual plant operation characteristics on the extent of reactor vessel embrittlement is evaluated, and the US NRC is notified.

8. Confirmation Process: The confirmation process is addressed through those specific portions of the QA program that are used to meet Criterion XVI, “Corrective Action,” of 10 CFR Part 50, Appendix B. Appendix A of the GALL-SLR Report describes how an applicant may apply its 10 CFR Part 50, Appendix B, QA program to fulfill the confirmation process element of this AMP for both safety-related and nonsafety-related SCs within the scope of this program.

9. Administrative Controls: Administrative controls are addressed through the QA program that is used to meet the requirements of 10 CFR Part 50, Appendix B, associated with managing the effects of aging. Appendix A of the GALL-SLR Report describes how an applicant may apply its 10 CFR Part 50, Appendix B, QA program to fulfill the administrative controls element of this AMP for both safety-related and nonsafety-related SCs within the scope of this program.

10. Operating Experience: The existing reactor vessel material surveillance program provides sufficient material data and dosimetry to (a) monitor irradiation embrittlement at the end of the subsequent period of extended operation, and (b) determine the need for operating restrictions on the inlet temperature, neutron fluence, and neutron flux.

The program is informed and enhanced when necessary through the systematic and ongoing review of both plant-specific and industry operating experience including research and development such that the effectiveness of the AMP is evaluated consistent with the discussion in Appendix B of the GALL-SLR Report.

References

10 CFR Part 50, Appendix B, “Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.” Washington, DC: U.S. Nuclear Regulatory Commission. 2016.

10 CFR Part 50, Appendix G, “Fracture Toughness Requirements.” Washington, DC: U.S. Nuclear Regulatory Commission. 2016.

10 CFR Part 50, Appendix H, “Reactor Vessel Material Surveillance Program Requirements.” Washington, DC: U.S. Nuclear Regulatory Commission. 2016.

10 CFR 50.61, “Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events.” Washington, DC: U.S. Nuclear Regulatory Commission. 2015.

10 CFR 50.61a, “Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events.” Washington, DC: U.S. Nuclear Regulatory Commission. 2015.

ASTM. ASTM E 185-82, “Standard Practice for Conducting Surveillance Tests of Light-Water Cooled Nuclear Power Reactor Vessels.” Philadelphia, Pennsylvania: American Society for Testing Materials. (Versions of ASTM E 185 to be used for the various aspects of the reactor vessel surveillance program are as specified in 10 CFR Part 50, Appendix H). 1982.

_____. ASTM E 185-79, “Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels.” Philadelphia, Pennsylvania: American Society for Testing Materials. 1979.

_____. ASTM E 185-73, “Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels.” Philadelphia, Pennsylvania: American Society for Testing Materials. 1973.

Eason, E.D., G.R. Odette, R.K. Nanstad, and T. Yamamoto. “A Physically Based Correlation of Irradiation-Induced Transition Temperature Shifts for RPV Steels.” ORNL/TM-2006/530. ML081000630. Oak Ridge, Tennessee: Oak Ridge National Laboratory. November 2007.

US NRC. Regulatory Guide 1.99, “Radiation Embrittlement of Reactor Vessel Materials.” Revision 2. Agencywide Documents Access and Management System (ADAMS) Accession No. ML003740284. Washington, DC: U.S. Nuclear Regulatory Commission. May 31, 1988.

_____. Regulatory Guide 1.190, “Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence.” ADAMS Accession No. ML010890301. Washington, DC: U.S. Nuclear Regulatory Commission. March 31, 2001.

_____. Regulatory Issue Summary 2014-11, “Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components.” ADAMS Accession No. ML14149A165. Washington, DC: U.S. Nuclear Regulatory Commission. October 14, 2014.

2191 R0 XI.M31: Difference between revisions

Revision as of 19:12, 4 October 2024

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