ML24138A064
| ML24138A064 | |
| Person / Time | |
|---|---|
| Issue date: | 05/17/2024 |
| From: | Matthew Mcconnell NRC/NRR/DEX/ELTB |
| To: | |
| References | |
| Download: ML24138A064 (13) | |
Text
Regulatory Guide 1.89, Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants, Revision 2 Matthew McConnell Senior Electrical Engineer U.S. Nuclear Regulatory Commission 1
2 U.S. Environmental Qualification (EQ) Requirements Formal EQ requirements for electrical equipment located in harsh areas are stated within 10 CFR Part 50.49, Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants, which was issued January 21, 1983.
EQ principles are embodied in General Design Criteria 1, 2, and 4 of Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50 and in Criterion III, Design Control, Criterion XI, Test Control, and Criterion XVII, Quality Assurance Records, of Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, to 10 CFR Part 50.
Nuclear power reactors are committed to differing NRC EQ requirements.
3 Environmental Qualification History in U.S. (cont.)
Nuclear power reactors are licensed to the following requirements:
Division of Operating Reactor (DOR) Guidelines, Guidelines for Evaluating Qualification of Class 1E Electrical Equipment in Operating Reactors, November 3, 1979.
NUREG-0588, Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment, Revision 1 (for Comment Version) Category II.
NUREG-0588 Category I as emphasized in Regulatory Guide (RG) 1.89, Qualification of Class 1E Equipment for Light-Water-Cooled Nuclear Power Plants," which endorses Institute of Electrical and Electronic Engineers (IEEE)
Standard (Std.) 323-1974, "IEEE Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations, Revision 1.
4 Environmental Qualification History in U.S. (cont.)
The applicability of the DOR Guidelines, or NUREG-0588 Category I or II, relates to the date of issuance of the construction permit for the nuclear power reactors.
Nuclear power reactors with construction permits issued prior to 1971 had to meet DOR Guidelines.
In cases where the DOR guidelines do not provide sufficient detail but NUREG-0588 Category II does, NUREG-0588 is expected to be used.
Nuclear power reactors with construction permits between 1971 and mid-1974 had to meet NUREG-0588 Category II requirements IEEE Std. 323-1971, IEEE Trial-Use Standard: General Guide for Qualifying Class I Electric Equipment for Nuclear Power Generating Stations.
Did not include guidance on aging, margins, or documentation Nuclear power reactors with construction permits after mid-1974 had to meet NUREG-0588, Category I requirements as emphasized in RG 1.89, Revision 1, which was issued June 1984, endorses IEEE Std. 323-1974, subject to supplementary provisions.
5 Environmental Qualification History in U.S. (cont.)
In accordance with 10 CFR 50.49(l), replacement electric equipment installed after the issuance of 10 CFR 50.49 (February 22, 1983) must be qualified according to the provisions of 10 CFR 50.49 unless there are sound reasons to the contrary.
A list of sound reasons is provided in RG 1.89 Same in both revisions.
A more detailed regulatory history can be found in Chapter 11 and Appendix A to the Electric Power Research Institute (EPRI)
Report No. 1021067, Plant Support Engineering: Nuclear Power Plant Equipment Qualification Reference Manual, Revision 1, issued September 2010.
6 RG 1.89 Revision 2 Issued April 2023 Draft Guide (DG) -1361 issued December 2020 Approximately 160 public comments Endorses, with clarifications, the English portion of the dual logo International Electrotechnical Commission (IEC)/IEEE Std.
60780-323, Nuclear FacilitiesElectrical Equipment Important to SafetyQualification, Edition 1, 2016-02 Reflects almost 40 years of experience gained in implementing regulatory requirements and industry research and testing related to EQ.
Contains information specific for EQ for both older plants and newer reactors licensed under 10 CFR Parts 50 and 52.
7 RG 1.89 Revision 2 (cont.)
Numerous terms clarified based on differences between U.S.
NRC and International definitions.
Clarified that mild environment is not within the scope of 10 CFR 50.49 (GDC 4 of Appendix A to 10 CFR Part 50)
Equipment important to safety defined by 10 CFR Part 50, Appendix A and 10 CFR 50.49 Qualified life to be consistent with 10 CFR 50.49(b)
Includes references to documents that facilitate qualification under other requirements, include additional information for qualifying specific equipment, or provide an additional level of detail for qualifying equipment For example, RG 1.63, Electric Penetration Assemblies in Containment Structures for Nuclear Power Plants
8 RG 1.89 Revision 2 (cont.)
Removed the double peak test profile and clarifies that it is not required but may be used to increase the severity of the design-basis event test.
As emphasized in IEC/IEEE Std. 60780 323, Edition 1, 2016-02, it is essential that the test profile envelopes the plant-specific design bases.
Added reference to aspects of Generic Aging Lessons Learned for Subsequent License Renewal (NUREG-2191) and License Renewal (NUREG-1801) for reassessing qualified life.
9 RG 1.89 Revision 2 (cont.)
Condition monitoring Clarified that condition monitoring recognizes the fact that the aging process in a 10 CFR 50.49 test method qualification program can be an acceptable process of determining end of qualified life, if it is proven during a qualification by test program to be a condition indicator that must be measurable, change monotonically with time, be correlated with the safety function performance under design-basis event conditions, be linked to the functional degradation of the qualified equipment, and have a consistent trend from unaged through the limit of the qualified pre-accident condition.
10 RG 1.89 Revision 2 (cont.)
Synergistic effects Clarified that synergistic effects must be considered when these effects are believed to have a significant effect on equipment performance.
A synergistic effect is the result of the combined environmental effects of the plant conditions such as radiation, humidity, and temperature that could result in greater degradation of equipment in relation to individual application of the plant environmental effects under normal, abnormal, and accident conditions.
If synergistic effects have been identified prior to the initiation of qualification, they should be accounted for in the qualification program.
Synergistic effects known at this time are dose rate effects and effects resulting from the different sequence of applying radiation and (elevated) temperature.
11 RG 1.89 Revision 2 (cont.)
Clarified that the expected operating temperature of the equipment under service conditions should be accounted for in thermal aging.
The Arrhenius equation is considered an acceptable method of addressing accelerated thermal aging within the limitation of state-of-the-art technology.
The use of other aging methods should be justified, and the staff will evaluate it on a case-by-case basis.
Clarified that the aging acceleration rate and activation used during qualification testing and the basis upon which the rate and activation energy were established should be defined, justified, and documented.
Selected activation energy should be representative of the most limiting material in a component/sub-component when determining qualified life.
12 RG 1.89 Revision 2 (cont.)
Periodic surveillance and testing programs Acceptable to account for uncertainties about age-related degradation that could affect the functional capability of equipment.
Results of such programs will be acceptable as ongoing qualification to modify the designated life (or qualified life) of equipment and should be incorporated into the maintenance and refurbishment/replacement schedules.
Radiation considerations for metal oxide semiconductor devices.
Radiation qualification for electronic components may have a lower exposure threshold.
Update to the guidance for performing radiological assessments (Annex D).
Formerly included in RG 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors.
13 QUESTIONS?