ML22272A601
| ML22272A601 | |
| Person / Time | |
|---|---|
| Issue date: | 04/27/2023 |
| From: | Matthew Mcconnell NRC/NRR/DEX |
| To: | |
| Eudy M | |
| Shared Package | |
| ML22060A287 | List: |
| References | |
| DG-1361 RG-1.089, Rev 2 | |
| Download: ML22272A601 (114) | |
Text
May 2021 Response to Public Comments on Draft Regulatory Guide (DG)-1361 Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants.
Proposed Revision 2 of Regulatory Guide (RG) 1.89 On December 17, 2020, and March 18, 2021, the NRC published notices in the Federal Register (85 FR 81958, 86 FR 10133) that Draft Regulatory Guide (DG) 1361 (Proposed Revision 2 of RG 1.89) was available for public comment. The public comment periods ended on February 16 and April 19, 2021, respectively. The NRC received comments from the organizations listed below. The NRC has combined the comments and NRC staff responses in the following table.
Comments were received from the following:
Comment Document 1 James Gleason Huntsville, AL ADAMS Accession No. ML21104A363 Comment Document 2 James Parello Fort Mill, SC, 29708 ADAMS Accession No. ML21022A044 Comment Document 3 Rick Weinacht Curtiss-Wright Nuclear Division Scientech 1360 Whitewater Drive Idaho Falls, ID 83402 ADAMS Accession No. ML21106A271 Comment Document 4 William Horin Nuclear Utility Group on Equipment (NUGEQ)
Qualification - Winston & Strawn LLP 1901 L Street N.W.
Washington, DC, 20036-3506 ADAMS Accession No. ML21041A127 Comment Document 5 William Horin NUGEQ Qualification - Winston & Strawn LLP 1901 L Street N.W.
Washington, DC, 20036-3506 ADAMS Accession No. ML21041A128 Comment Document 6 Robert Konnik Institute of Electrical and Electronics Engineers (IEEE)
ADAMS Accession No. ML21110A055 Comment Document 7 Vincent Bacanskas ADAMS Accession No. ML21050A358 Comment Document 8 Carrie Fosaaen Director, Regulatory Affairs NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 ADAMS Accession No. ML21110A054 Comment Document 9 Carrie Fosaaen Director, Regulatory Affairs NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 ADAMS Accession No. ML21113A276
2 Comment Document 10 Rick Weinacht Curtiss-Wright Nuclear Division Scientech 1360 Whitewater Drive Idaho Falls, ID 83402 ADAMS Accession No. ML21042A003 Comment Document 11 William Horin NUGEQ Qualification - Winston & Strawn LLP 1901 L Street N.W.
Washington, DC, 20036-3506 ADAMS Accession No. ML21050A360 Comment Document 12 William Horin NUGEQ Qualification - Winston & Strawn LLP 1901 L Street N.W.
Washington, DC, 20036-3506 ADAMS Accession No. ML21110A056 Comment Document 13 Jeremy Owen Section Manager Kinetrics 800 Kipling Ave. Unit 2 Toronto, ON, M8Z 5G5 ADAMS Accession No. ML21131A005, ML23055A009 Commenter Section of DG-1361 Specific Comments NRC Resolution Comment Document 1: ML21104A363 James Gleason Section C.1 Comment 1 Section C. 1, a is confusing and does not state explicitly the equipment important to safety that is defined by 10CFR.50.49. It states: 10 CFR 50.49(e)(5) requires, in part, that equipment qualified by test must be preconditioned by natural or artificial (accelerated) aging to its end-of-installed life condition. Therefore, end condition, as defined in Section 3.10 of IEC/IEEE Std. 60780-323, Edition 1, 2016-02, should be considered equivalent to end-of-installed life. Note: Qualified equipment must be capable of performing its design function at the end-of-installed life.
10CFR50.49 Section j (2) states Meets its specified performance requirements when it is subjected to the conditions predicted to be present when it must perform its safety function up to the end of its qualified life.
The staff disagrees with the comment. The comment proposes a change to the DG that would require a change to 10 CFR 50.49(e)(5) because the suggested language does not currently exist in 10 CFR 50.49(e)(5). Rule changes are outside the scope of this RG update, so the staff did not make the proposed change. Furthermore, the staff deleted the regulatory position associated with "end condition" as a result of comment 33. See the staff's response to that comment for additional information.
3 Commenter Section of DG-1361 Specific Comments NRC Resolution Thus, the use of design function at the end-of-installed life is confusing and does not explicitly follow 10CFR50.49. RG 1.89 should refer to end of qualified life and the term end-of-installed life in IEC/IEEE Std. 60780-323, Edition 1, 2016-02, shall mean end of qualified life.
Recommendation Change Section C. 1, a to 10 CFR 50.49(e)(5) in part that important to safety equipment meets its specified performance requirements when it is subjected to the conditions predicted to be present when it must perform its safety function up to the end of its qualified life. The term end-of-installed life in IEC/IEEE Std. 60780-323, Edition 1, 2016-02, shall mean end of qualified life.
James Gleaseon Section C.1 Comment 2 Section C. 1, b is confusing and does not state explicitly the equipment important to safety that is defined by 10CFR.50.49.
Recommendation Replace with: The following description and definition of important to safety should be used instead of the definition in Section 3.12 of IEC/IEEE Std. 60780-323, Edition 1, 2016-02: 10CFR50.49 defines equipment Important to safety in section (b) as follows.
(b) Electric equipment important to safety covered by this section is:
(1) Safety-related electric equipment.
(i) This equipment is that relied upon to remain functional during and following design basis events to ensure (A) The integrity of the reactor coolant pressure boundary; (B) The capability to shut down the reactor and maintain it in a safe shutdown condition; or (C) The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures The staff partially agrees with the comment.
While the staff disagrees with including the suggested text of 10 CFR 50.49(b) in the RG, the staff agrees that its regulations, specifically 10 CFR 50.49 and Appendix A of 10 CFR Part 50, should be cited to clearly describe and define what the NRC considers as equipment important to safety.
Section C.1.b of the DG has been revised to include the following as the description and definition of equipment important to safety:
The following description and definition of equipment important to safety should be used instead of the definition in Section 3.12 of IEC/IEEE Std. 60780-323-2016:
The introduction to 10 CFR Part 50, Appendix A, states that important to safety SSCs are
4 Commenter Section of DG-1361 Specific Comments NRC Resolution comparable to the guidelines in §50.34(a)(1),
§50.67(b)(2), or § 100.11 of this chapter, as applicable.
(ii) Design basis events are defined as conditions of normal operation, including anticipated operational occurrences, design basis accidents, external events, and natural phenomena for which the plant must be designed to ensure functions (b)(1)(i) (A) through (C) of this section.
(2) Non-safety-related electric equipment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of safety functions specified in subparagraphs (b)(1)(i) (A) through (C) of paragraph (b)(1) of this section by the safety-related equipment. (3) Certain post-accident monitoring equipment.
those SSCs that provide reasonable assurance that the facility can be operated without undue risk to public health and safety.
10 CFR 50.49 requires safety-related (Class 1E) electric equipment as defined in 10 CFR 50.49(b)(l) to be environmentally qualified to perform its intended safety functions.
Appendix A to this guide lists typical safety-related equipment and systems. 10 CFR 50.49(b)(2) requires that non-safety-related electric equipment be environmentally qualified if its failure under postulated environmental conditions could prevent satisfactory accomplishment of the safety functions specified in 10 CFR 50.49(b)(1)(i)(A) through (C) by safety-related electric equipment. Appendix B to this guide includes typical examples of non-safety-related electric equipment that may be in scope of 10 CFR 50.49. 10 CFR 50.49(b)(3) requires that certain post-accident monitoring equipment also be environmentally qualified. RG 1.97 includes regulatory guidance for post-accident monitoring equipment.
NOTE: The associated changes are now contained in Section C.1.c of RG 1.89, Rev. 2.
James Gleason Section C.1.c Comment 3 Section C. 1, c is confusing as it states not to use the definition of qualified life in Section 3.12 of IEC/IEEE Std. 60780-323, Edition 1, 2016-02, but does not state explicitly the definition of qualified life to be used and its source.
The staff disagrees with the comment.
Section C.1.c of DG-1361 explicitly states that the following definition of qualified life should be used: period for which an
5 Commenter Section of DG-1361 Specific Comments NRC Resolution There is no definition of qualified life in 10CFR50.49.
- 1. IEEE 323-74 definition of Qualified Life: The period of time for which satisfactory performance can be demonstrated for a specific set of service conditions.
- 2. IEC/IEEE 60780-323 Definition of Qualified Life: period for which an equipment has been demonstrated, through testing, analysis and/or experience, to be capable of functioning within acceptance criteria during specific operating conditions while retaining the ability to perform its safety functions in accident condition or earthquake.
As the IEC/IEEE 60780-323 Definition of Qualified Life expands the definition to include retaining the ability to perform safety functions in accident conditions and earthquakes, it creates backfit and forward fit issues.
Recommendation Use the IEEE 323-74 definition of Qualified Life: The period of time for which satisfactory performance can be demonstrated for a specific set of service conditions.
equipment has been demonstrated, through testing, analysis and/or experience, to be capable of remaining functional during and following design basis events to ensure that the criteria specified in 10 CFR 50.49(b)(1)(i)(A),
(B), and (C) are satisfied.
The staffs clarification of the definition of qualified life in IEC/IEEE 60780-323-2016 is similar to the definition in IEEE 323-1974 except it more clearly explains the term by describing how qualified life is determined (i.e., by testing, analysis, and/or experience) and by citing specific regulatory requirements that must be satisfied in order to establish a qualified life. Additionally, the staffs clarification of the definition in IEC/IEEE 60780-323-2016 does not invalidate the clarification of the definition of qualified life in IEEE Std. 323-1974.
This clarification does not meet the definition of backfitting or forward fitting in MD 8.4. RG 1.89, Rev. 2 is voluntary guidance and represents one acceptable way to satisfy the applicable NRC regulations. The NRC staff is not imposing or expecting specific licensees to implement the guidance in RG 1.89, Rev. 2.
Further, forward fitting occurs only in the context of a licensing action, and the issuance of a RG is not a licensing action.
The staff revised Section C.1.c to remove the specific rule language citation as it was deemed
6 Commenter Section of DG-1361 Specific Comments NRC Resolution unnecessary. NOTE, these changes are reflected in Section C.1.d of RG 1.89, Rev. 2.
James Gleason Section C.1.d Comment 4 Section C. 1, d is confusing since it tries to address the term service life and relates service life, qualified life and shelf life.
There is no use of service life in 10CFR50.49 and its introduction of a new term service life including its relationship to qualified life constitutes a backfit and forward fit.
Recommendation Delete section C.1 d and all discussion of service life.
Add that IEC/IEEE 60780-323 term service life is not endorsed. Please note that that IEC/IEEE 60780-323 proficient use of the term service life and an alternate definition of qualified life may render IEC/IEEE 60780-323 to be not endorsed.
The staff partially agrees with the comment.
The staff agrees to delete C.1.d as service life is not directly associated with equipment qualification and is distinctly different than the term qualified life. Furthermore, service life is addressed elsewhere in the regulations (e.g.,
Appendix B to 10 CFR Part 50).
This clarification does not meet the definition of backfitting or forward fitting in MD 8.4. RG 1.89, Rev. 2 is voluntary guidance and represents one acceptable way to satisfy the applicable NRC regulations. The NRC staff is not imposing or expecting specific licensees to implement the guidance in RG 1.89, Rev. 2.
Further, forward fitting occurs only in the context of a licensing action, and the issuance of a RG is not a licensing action.
The staff disagrees with taking exception to the term service life as defined in IEC/IEEE 60780-323-2016. The staff finds that the period from initial operation to final withdrawal from service of a structure, system or component is an accurate description of the actual service life of a component. Endorsing this term does not imply that any additional requirements need to be met.
7 Commenter Section of DG-1361 Specific Comments NRC Resolution See the staffs disposition of comment 3 for further detail on the staffs endorsement of the term qualified life.
James Gleason Section C.1.e Comment 5 Section C. 1, e is confusing as it notes that The prerequisite for aging electric equipment located in a mild environment is not within the scope of 10 CFR 50.49 and then adds Requirements, including EMC and seismic requirements, shall be specified in the design/purchase specifications.
It is agreed that the prerequisite for aging electric equipment located in a mild environment is not within the scope of 10 CFR 50.49.t There is also no requirement for design/purchase specifications in 10CFR50.49 and there is no requirement for EMC in 10CFR50.49. Thus, introductions for design/purchase specifications requirements and EMC requirements constitutes a backfit and forward fit.
Recommendation Modify Section C. 1, e to the following: In IEC/IEEE 60780-323, the discussion of design/purchase specifications requirements and EMC and seismic requirements are not endorsed The NRC staff partially agrees with the comment. The staff agrees that the aging of electrical equipment in mild environments and electromagnetic compatibility (EMC) and seismic qualification are not requirements under 10 CFR 50.49. However, the staff does not agree with the proposed resolution as electrical equipment in mild environments and EMC and seismic qualification are covered under other applicable regulations listed in the RG. Instead, the staff has elected to edit Section C.1.e to clarify the available guidance for EMC, aging, and seismic as follows
- e.
Paragraph 4 of Section 5.1 of IEC/IEEE Std. 60780 323, Edition 1, 2016 02, notes that Requirements, including EMC
[Electromagnetic Compatibility],
environmental/operational ageing and seismic requirements shall be specified in the design/purchase specifications. Guidance for demonstrating EMC and EMI/RFI qualification is provided in Regulatory Guide 1.180. While not within the scope of 10 CFR 50.49, the requirements for environmental design considerations of equipment located in a mild environment is covered by GDC 4 of Appendix A to 10 CFR Part 50. Guidance for demonstrating seismic qualification is provided in Regulatory Guide 1.100.
8 Commenter Section of DG-1361 Specific Comments NRC Resolution James Gleason Section C.1.f Comment 6 Section C. 1, f is confusing as it states Condition monitoring and associated condition-based qualification methodologies discussed in Section 6.3 of IEC/IEEE Std. 60780-323, Edition 1, 2016-02, represent new approaches for extending or establishing the qualified life of electrical equipment.
If used, these methodologies must ensure that equipment important to safety will perform under the conditions specified in 10 CFR 50.49.
This appears to misstate the purpose and application of condition monitoring.
Section 6.3 of IEC/IEEE Std. 60780-323, Edition 1, 2016-02 States:
Condition monitoring for equipment qualification purposes monitors one or more condition indicators to determine whether equipment remains in a qualified condition.
Condition monitoring is not a new approach for establishing the qualified life of electrical equipment.
The qualified life is established in the regulatory accepted method of aging, including time/temperature effects, radiation and mechanical degradation.
Condition monitoring recognizes that when qualified life is established in the regulatory accepted method, the equipment being qualified is placed into a degraded condition, for which there may be one or more relevant condition indicators of the degraded condition.
The condition indicator shall be measurable, change monotonically with time, be correlated with the safety function performance under DBE 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.
The staff agrees that C.1.f could be presented more clearly to avoid confusion. As such, the staff modified the referenced paragraph to include the following:
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.
As a result of adding the above text, the last sentence in Section C.1.e of DG-1361 was removed. NOTE, this position is now Section C.1.g. of RG 1.89, Rev. 2.
9 Commenter Section of DG-1361 Specific Comments NRC Resolution Therefore, Condition monitoring establishes the degraded condition during the aging part of qualification program.
The regulatory statement: these methodologies must ensure that equipment important to safety will perform under the conditions specified in 10 CFR 50.49, is confusing since condition monitoring is not a qualification method that verifies performance under the conditions specified in 10 CFR 50.49.
The qualification methods that ensure performance under the conditions specified in 10 CFR 50.49 are test, analysis, and test and analysis.
Recommendation Change Section C. 1, f to Condition monitoring recognizes the fact that the aging process in a 10CFR50.49 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 DBE 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.
James Gleason Section C.1.h Comment 7 Section C. 1, h is confusing in: (2) Electric equipment that may be exposed to low-level radiation doses should not generally be considered exempt from radiation qualification testing. Exceptions may be based on qualification by analysis supported by test data or operating experience that verifies that the dose and dose rates will not degrade the operability of the equipment below acceptable values.
This is new and the following RG 1.89 Rev 1 section is missing: (6)
Shielded components need be qualified only to the gamma radiation environment provided it can be demonstrated that the sensitive portions of the component or equipment are not exposed to significant beta radiation dose rates or that the effects of beta radiation, including heating and secondary The staff partially agrees with the comment.
The comment is incorrect in stating that Section C.1.h(2) is new in DG-1361. Section C.1.h(2), is the same as Section C.2.c.(8) of Revision 1 of RG 1.89 (the section was moved to a new location in DG-1361 based on the new RG formatting). Additional information regarding this paragraph can be found in the response to comment 25.
The Division of Operating Reactors (DOR) guidelines indicated that, by using the
10 Commenter Section of DG-1361 Specific Comments NRC Resolution radiation, have no deleterious effects on component performance. If, after considering the appropriate shielding factors, the total beta radiation dose contribution to the equipment or component is calculated to be less than 10%
of the total gamma radiation dose to which the equipment or component has been qualified, the equipment or component is considered qualified for the beta and gamma radiation environment.
The deletion of RG 1.89 Rev 1 section (6) Shielded components, etc, constitutes a forward backfit as it deletes an acceptable process for addressing beta radiation and the addition of Section C. 1, h 2 is an unjustified increase in requirements and therefore a forward fit.
Recommendation Section C. 1, h 2 should be replaced with RG 1.89 Rev 1 section: (6)
Shielded components need be qualified only to the gamma radiation environment provided it can be demonstrated that the sensitive portions of the component or equipment are not exposed to significant beta radiation dose rates or that the effects of beta radiation, including heating and secondary radiation, have no deleterious effects on component performance. If, after considering the appropriate shielding factors, the total beta radiation dose contribution to the equipment or component is calculated to be less than 10% of the total gamma radiation dose to which the equipment or component has been qualified, the equipment or component is considered qualified for the beta and gamma radiation environment..
Technical Information Document (TID)-14844, Calculation of Distance Factors for Power and Test Reactor Sites, source term and conservative assumptions, the beta surface dose would be 1.40 x 108 rad. If this beta dose was conservatively assumed to be 2.0 x 108 rad and shielding factors discussed in the DOR guidelines were applied, and if the beta dose was less than 10% of the total gamma dose to the equipment, then the DOR guidelines indicated that only the gamma dose needed to be considered.
RG 1.89, Revision 1, did not specify that the beta dose should be increased or specify specific shielding factors, with regards to the 10% criteria.
RG 1.183, Revision 0, Appendix I, provided radiological EQ guidance for plants using an alternative source term. While RG 1.183, Revision 0 did provide specific guidance for beta radiation, it did not specify any special criteria for beta radiation similar to the 10%
criteria in the DOR guidelines and RG 1.89, Revision 1. However, while RG 1.183 indicated that it superseded several sections of RG 1.89, Revision 1, for plants using an alternative source term, it did not specifically state that it superseded Section C.2.c(6), which is where the 10% criteria is discussed in RG 1.89, Revision 1.
11 Commenter Section of DG-1361 Specific Comments NRC Resolution The criteria in RG 1.89, Revision 1, Section C.2.c(6) has been re-instated into RG 1.89, Revision 2, with some modifications. If it can be demonstrated that beta radiation has no deleterious effects on equipment, then beta radiation need not be considered. The criterion that if the total beta radiation dose contribution to the equipment or component is calculated to be less than 10% of the total gamma radiation dose to which the equipment or component has been qualified, it is considered qualified for the beta and gamma radiation environment, will also be retained. However, staff notes that 10 CFR 50.49(e)(8) requires that margin be applied to account for unquantified uncertainties considering a portion of the total integrated dose when qualifying equipment may remove some of the margin. Any reduction in margin should be considered in the EQ analysis and applicants and licensees must still ensure that with any reduction in margin, all requirements of 10 CFR 50.49(e) continue to be met. Therefore, Section C.1.h has been revised to include the following:
Shielded components need be qualified only to the gamma radiation environment provided it can be demonstrated that the sensitive portions of the component or equipment are not exposed to significant beta radiation dose rates or that the effects of beta radiation, including heating and secondary radiation, have no deleterious effects on component performance.
If, after considering the appropriate shielding
12 Commenter Section of DG-1361 Specific Comments NRC Resolution factors, the total beta radiation dose contribution to the equipment or component is calculated to be less than 10% of the total gamma radiation dose to which the equipment or component has been qualified, the equipment or component may be considered qualified for the beta and gamma radiation environment, provided that the total integrated dose to equipment remains conservative considering all assumptions made in the analysis, including margin.
These changes do not meet the definition of backfitting or forward fitting in MD 8.4. RG 1.89, Rev. 2 is voluntary guidance and represents one acceptable way to satisfy the applicable NRC regulations. The NRC staff is not imposing or expecting specific licensees to implement the guidance in RG 1.89, Rev. 2.
Further, forward fitting occurs only in the context of a licensing action, and the issuance of a RG is not a licensing action.
NOTE: These changes are reflected in Section C.1.i(2) of RG 1.89, Rev. 2.
James Gleason Section C.1.j Comment 8 Section C. 1, j (1) is confusing as it contains the following:
The 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 sequential application of the plant environment under normal, abnormal, and accident conditions.
The staff partially agrees with the comment.
The staff agrees that C.1.j(1) could be presented more clearly to avoid confusion. As such, the staff has re-written Section C.1.j(1) to state:
Synergistic effects must be considered when these effects are believed to have a significant
13 Commenter Section of DG-1361 Specific Comments NRC Resolution The synergistic effects on materials that are known to have such increased degradation under these conditions should be accounted for when assessing the qualified life.
Section 7.4.1.9.3 Age conditioning of IEC/IEEE Std. 60780-323, Edition 1, 2016-02 contains the discussion of synergistic effects. Historically, synergistic effects in qualification are considered in aging and not accident conditions.
Therefore, the inclusion of accident conditions in determining synergistic effects creates a new requirement and is a forward fit.
The phrase The synergistic effects on materials that are known to have such increased degradation under these conditions should be accounted for when assessing the qualified life is confusing and a new requirement. Synergistic effects, as noted in 10CFR50.49 must be considered when these effects are believed to have a significant effect on equipment performance.
The requirement that synergistic effects on materials need to be accounted for in qualified life is new, as it requires all synergistic effects of materials to be included and inconsistent with the 10CFR50.49 threshold that they must be considered when these effects are believed to have a significant effect on equipment performance.
Recommendation Section C. 1, j (1) modify to: Synergistic effects must be considered when these effects are believed to have a significant effect on equipment performance.
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.
The staff disagrees that synergistic effects need not be considered during accident conditions.
10 CFR 50.49(e)(7) specifies that synergistic effects must be considered when these effects are believed to have a significant effect on equipment performance, without specifying normal operation or accident conditions.
Therefore, synergistic effects that have a significant effect on equipment performance during accident conditions must be considered.
New guidance does not constitute new requirements and, as explained in the response to comment 4, the issuance of new guidance does not constitute forward fitting.
14 Commenter Section of DG-1361 Specific Comments NRC Resolution James Gleason Section C.1.j Comment 9 Section C. 1, j (3) is confusing in that it states: Activation energy values should be based on the testing of the specific compound used in the equipment and on the most relevant material property and property endpoint (i.e., failure mechanism). It is confusing because it constitutes a significant effort to know each compound and to have Arrhenius test data for Activation Energy on every possible compound and failure mechanism.
Additionally, most safety related equipment is made up of many materials, and the basis of aging is to use the lowest activation energy for the assembly when establishing the aging program and qualified life.
Thus, materials that are not the lowest activation energy are aged for more degradation equivalency than the qualified life of the lowest activation energy material.
The NRC studies, such as NUREG/CR-6384 and NUREG/CR-6704 on Arrhenius Theory and its application to environmental qualification have demonstrated the conservatism to establishing qualified life.
Lastly, Activation energy is not a safety function and was never intended to be a quality attribute of a safety related component.
Arrhenius theory and activation energy are intended to place a safety related type test specimen in a reasonable facsimile of the degradation to be seen in service when installed in its application in a nuclear power plant.
Recommendation Section C. 1, j (3) should be deleted.
The staff partially agrees with the comment.
The staff acknowledges that knowing each compound requires a significant effort.
However, the materials within a component need to be known in order to establish and justify environmental qualification.
Activation energy values are an important factor for establishing/determining the qualified life of a component. The clarification provided by the staff in this RG includes meaningful information that should be considered when selecting an activation energy for the purpose of establishing or extending the qualified life of a component. This is especially important given the sensitivity of the activation energy variable in effecting the results of the Arrhenius equation.
The staff agrees that equipment within the scope of this RG can be composed of a variety of materials, and that the basis for aging is to use the activation energy for the most sensitive material within a component. However, the process for determining/selecting activation energies for the various materials within a component should consider the information provided in the proposed RG.
The staff finds that no changes to DG-1361 were necessary as a result of this comment.
James Gleason Section C.1.k Comment 10 The staff disagrees with the comment.
15 Commenter Section of DG-1361 Specific Comments NRC Resolution Section C. 1, k (2) is confusing and unnecessary as it states: Electric equipment located in an area where rapid pressure changes are postulated simultaneously with the most adverse relative humidity should be qualified to demonstrate that the equipment seals and vapor barriers will prevent moisture from penetrating into the equipment to the degree necessary to maintain equipment functionality.
IEC/IEEE Std. 60780-323, Edition 1, 2016-02 identifies interfaces and seals as elements to be identified and maintained as part of qualification and the equipment seals and vapor barriers, when required to ensure the safety function performance must operate properly in the environment in which the equipment is being qualified.
The highlighting of equipment seals and vapor barriers, only where rapid pressure changes are postulated simultaneously, overlooks applications where seals perform safety functions when no pressure variations are requirements.
Recommendation Section C. 1, k (2) should be deleted.
The information included in Section C.1.k(2) of DG-1361, which is from RG 1.89, Rev. 1, Section C.3(b), is supplemental information that is intended to provide further clarity with regard to type testing. Therefore, the staff did not delete Section C.1.k(2).
In areas where there are no pressure changes and only humidity, the staff finds that the IEC/IEEE 60780-323-2016 guidance is adequate.
The staff concludes that no changes to DG-1361 were necessary as a result of this comment.
James Gleason Section C.1.k Comment 11 Section C. 1, k (4) is confusing and unnecessary as it states: Performance characteristics that demonstrate the operability of equipment should be verified before, after, and periodically during testing throughout its range of required operability. Variables indicative of momentary failure that prevent the equipment from performing its safety function (e.g., momentary opening of a relay contact) should be monitored continuously to ensure that momentary failures (if any) have been accounted for during testing. For long-term testing, however, monitoring during periodic intervals may be used if justified.
10CFR50.49 j (2) states that equipment must: Meets its specified performance requirements when it is subjected to the conditions predicted to The staff partially agrees with the comment.
Based on potential confusion with monitoring performance characteristics, the staff modified Section C.1.k(4) to clarify that verifying performance characteristics may need to be performed periodically, as applicable and depending on the equipments safety function.
The staff disagrees that this position represents a forward fit, as explained in the response to comment 4.
16 Commenter Section of DG-1361 Specific Comments NRC Resolution be present when it must perform its safety function up to the end of its qualified life.
The requirement that testing throughout its range of required operability is to be included goes beyond the requirement to demonstrate the safety function and constitutes a forward fit.
Additionally, the phrase: Variables indicative of momentary failure that prevent the equipment from performing its safety function (e.g., momentary opening of a relay contact) should be monitored continuously to ensure that momentary failures (if any) have been accounted for during testing, is excessive and IEC/IEEE Std. 60780-323, Edition 1, 2016-02 7.4.1.6 Monitoring, already requires During testing, both the test environment and the equipments safety function(s) shall be monitored using equipment that provides accuracy and resolution for detecting meaningful changes in the parameters.
Recommendation Section C. 1, k (4): delete James Gleason Section C.1.n Comment 12 Section C. 1, n (1) is confusing and unnecessary as it states:
- 1) A double-transient should be used with equipment that may be vulnerable to thermal binding from different expansion rates of materials during the initial heatup.
Double-transient testing has never been a requirement of 10CFR50.49, DOR Guidelines, RG 1.89, or NUREG-0588.
There has been no requirement to evaluate equipment that may be vulnerable to thermal binding from different expansion rates of materials during the initial heatup.
This constitutes a forward fit.
The staff agrees with the comment. Section C.1.n(1) can be deleted due to it being unnecessary to provide a regulatory position on the use of a double transient profile to address equipment that may be vulnerable to thermal binding from different expansion rates of materials during initial heatup.
A small discussion on use of double-transients during testing has been added to the Background section of the RG for informative purpose only.
17 Commenter Section of DG-1361 Specific Comments NRC Resolution Recommendation Section C. 1, n (1): delete James Gleason Section C.1.n Comment 13 Section C. 1, n (2) is confusing and unnecessary as it states: (2) The use of double transients could help offset tests where the ramp rate (initial temperature rise) of the test is slower than the required profile. This is commonly the result of test chamber and steam supply limitations.
There has been no use of double transients to offset test facility limitations on the initial steam ramp. There is no logical formula for how an offset would be calculated or credited.
This constitutes a forward fit.
Recommendation Section C. 1, n (2): delete The staff agrees that Section C.1.n(2) can be deleted due to it being unnecessary to provide a regulatory position on the use of a double transient profile to offset tests where ramp rate of the test is slower than the required profile.
Section C.1.n(2) has been deleted as a result of this comment.
James Gleason Section C.2.c Comment 14 Section C. 2, c is confusing and unnecessary as it states: An additional stressor to be considered in the qualification of digital systems is smoke exposure from an electrical fire. For smoke exposure, important failure mechanisms are not only long-term effects such as corrosion, but also short-term and perhaps intermittent malfunctions, such as leakage current. Smoke can cause circuit bridging and thus affect the operation of digital equipment.
Because the edge connections and interfaces are typically uncoated, the most likely effect of the smoke is to impede communication and data transfer between subsystems. RG 1.209 provides several references that detail the effects of smoke exposure.
Smoke has never previously been identified to be an environmental parameter or result of a Design Basis Accident. The new requirement to qualify for smoke during a DBA constitutes a forward fit.
Recommendation The staff agrees with the comment with regard to the fact that the regulations address the need for SSCs to remain functional under postulated design basis events, but that smoke exposure is not addressed under 10 CFR 50.49. In addition, guidance for addressing the effects of smoke is covered within Regulatory Guide 1.209, Guidelines for Environmental Qualification of Safety-Related Computer-Based Instrumentation and Control Systems in Nuclear Power Plants.
As a result of this comment, Section C.2.c is being revised to state:
While not a consideration under 10 CFR 50.49, an additional stressor that may need to
18 Commenter Section of DG-1361 Specific Comments NRC Resolution Section C. 2, c, starting at An additional stressor to be considered in the qualification of digital systems is smoke exposure: delete be considered in the qualification of digital systems is smoke exposure from an electrical fire from operational conditions (e.g., fire). For smoke exposure, important failure mechanisms are not only long-term effects such as corrosion, but also short-term and perhaps intermittent malfunctions, such as leakage current. Smoke can cause circuit bridging and thus affect the operation of digital equipment.
Because the edge connections and interfaces are typically uncoated, the most likely effect of the smoke is to impede communication and data transfer between subsystems. RG 1.209 provides several references that detail the effects of smoke exposure.
NOTE: These changes are now reflected in Section C.2.b of RG 1.89, Rev. 2.
This proposed guidance does not meet the definition of forward fitting in MD 8.4. RG 1.89, Rev. 2 is voluntary guidance and represents one acceptable way to satisfy the applicable NRC regulations. The NRC staff is not imposing or expecting specific licensees to implement the guidance in RG 1.89, Rev. 2.
Further, forward fitting occurs only in the context of a licensing action, and the issuance of a RG is not a licensing action.
James Gleason Section C.2.e Comment 15 Section C. 2, e is confusing, contradictory to 10CFR50.49, and unnecessary as it states: Considerations such as the following should be taken into The staff partially agrees with the comment.
Item 1 in Section C.2.e is very general information and while usually true, it may not be true in all situations. Item 2 is obvious and
19 Commenter Section of DG-1361 Specific Comments NRC Resolution account when determining the environment for which the equipment is to be qualified:
(C) equipment outside containment would generally see a less severe environment than equipment inside containment, (2) equipment whose location is shielded from a radiation source would generally receive a smaller radiation dose than equipment at the same distance from the source but exposed to direct radiation, (3) equipment required to initiate protective action would generally be required for a shorter period of time than instrumentation required to operate during and after an accident, and (4) analyses taking into account arrangements of equipment and radiation sources may be necessary to determine whether equipment needed for mitigation of design basis accidents other than LOCA or high-energy line breaks (HELB) could be exposed to a more severe environment than the plant-specific LOCA or HELB environments.
This section has no significance to 10CFR50.49 qualification requirements.
Items (1) and (2) are obvious but are irrelevant since 10CFR50.49 requires equipment in DBA environments to be qualified.
Item (3) discusses equipment performing protective action, but 10CFR50.49 requires equipment be qualified for its safety function.
Item (4) discusses mitigation of design basis accidents instead of 10CFR50.49 Recommendation Section C. 2, e: delete need not be included in the guidance.
Therefore, items 1 and 2 have been deleted.
Item 3 is intended to address differences in the timeframes for which equipment is required to be qualified. Some equipment may have safety functions that are to mitigate accidents and may not be necessary after the initial stages of the accident, while other equipment may need to be qualified to be operable in an extended period in accident conditions. However, this is obvious and need not be stated in the guidance and therefore has been deleted.
Item 4 is intended address that some equipment may have a different bounding accident than other equipment and different radiation sources may be bounding for different equipment. The staff has decided to retain this information but relocated it to the Background section of the guide as this is a more appropriate placement.
Therefore, the entire paragraph C.2.e has been removed from Section C and the following is included in the Background section of RG 1.89, Rev. 2:
When determining the environment for which the equipment is to be qualified, environmental analyses taking into account arrangements of equipment and radiation sources may be necessary to determine whether equipment needed for mitigation of design basis accidents other than loss-of-coolant accidents (LOCA) or
20 Commenter Section of DG-1361 Specific Comments NRC Resolution high-energy line breaks (HELB) could be exposed to a more severe environment than the plant-specific LOCA or HELB environments.
James Gleason Section C.2.f Comment 16 Section C. 2, f is confusing, contradictory to 10 CFR 50.49, and unnecessary as it states:
Electric equipment to be qualified in a nuclear radiation environment should be exposed to radiation, before testing, that simulates the calculated integrated dose (normal and accident) that the equipment must withstand before completion of its intended safety functions.
The requirements in 10 CFR 50.49 are: (4) Radiation. The radiation environment must be based on the type of radiation, the total dose expected during normal operation over the installed life of the equipment, and the radiation environment associated with the most severe design basis accident during or following which the equipment is required to remain functional, including the radiation resulting from recirculating fluids for equipment located near the recirculating lines and including dose-rate effects.
The phrase should be exposed to radiation, before testing, that simulates the calculated integrated dose (normal and accident) that the equipment must withstand before completion of its intended safety functions contradicts 10 CFR 50.49 in that radiation exposure is testing and exposure before testing is confusing.
The phrase withstand before completion of its intended safety functions contradicts 10 CFR 50.49 in that radiation exposure should be the normal does plus the radiation environment associated with the most severe design basis accident during or following which the equipment is required to remain functional.
Recommendation The staff partially agrees with the comment.
The staff agrees that the phrase before testing is confusing, so the staff removed the phrase from Section C.2.f. The staff disagrees that withstand before completion of its intended safety functions contradicts 10 CFR 50.49, as this statement is carried forward from RG 1.89, Rev. 1 and does not represent a new clarification.
NOTE: This position is now Section C.2.c of RG 1.89, Rev.2.
21 Commenter Section of DG-1361 Specific Comments NRC Resolution Section C. 2, f change Electric equipment to be. Intended safety functions.
To: The radiation environment must be based on the type of radiation, the total dose expected during normal operation over the installed life of the equipment, and the radiation environment associated with the most severe design basis accident during or following which the equipment is required to remain functional, including the radiation resulting from recirculating fluids for equipment located near the recirculating lines and including dose-rate effects.
James Gleason Section C.2.f Comment 17 Section C. 2, f is confusing, contradictory to 10 CFR 50.49, and unnecessary as it states: In 10 CFR 100.11, Determination of exclusion area, low population zone, and population center distance (Ref. 36), the NRC provides criteria for evaluating the radiological aspects of the proposed site. A footnote to 10 CFR 100.11 states that the fission product release assumed in these evaluations should be based upon a major accident involving substantial meltdown of the core with subsequent release of appreciable quantities of fission products. The NRC cites Technical Information Document (TID) 14844, Calculation of Distance Factors for Power and Test Reactor Sites (Ref. 37), in 10 CFR Part 100, Reactor Site Criteria, as a source of further guidance on these analyses. Although initially used only for siting evaluations, the TID 14844 source term has been used for design-basis applications, such as EQ of equipment under 10 CFR 50.49. Regulations in 10 CFR 50.67, Accident source term, allows licensees to revise the accident source term used in design-basis radiological consequence analyses.
10CFR50.49 states: (4) Radiation. The radiation environment must be based on the type of radiation, the total dose expected during normal operation over the installed life of the equipment, and the radiation environment associated with the most severe design basis accident during or following which the equipment is required to remain functional, including the radiation resulting The staff partially agrees with the comment.
The staff agrees that Section C.2.f should be revised to remove confusion and to more clearly describe the staffs position that the RG 1.183 accident scenarios may be used in assessing accident EQ radiation doses.
However, the accident scenarios and assumptions developed for the purposes of reactor siting have been used for assessing the total integrated radiation dose for EQ in nearly all currently licensed facilities. In addition, NRC regulatory guides provide one way to meeting the requirements, and the RG revision is not requiring any changes to existing licensees. Applicants and licensees may propose alternative methods to meet NRC requirements.
As a result, Section C.2.f has been rewritten as follows:
Electric equipment to be qualified in a nuclear radiation environment should be exposed to
22 Commenter Section of DG-1361 Specific Comments NRC Resolution from recirculating fluids for equipment located near the recirculating lines and including dose-rate effects.
Since this section uses the radiation from these evaluations should be based upon a major accident involving substantial meltdown of the core with subsequent release of appreciable quantities of fission products and not the radiation environment associated with the most severe design basis accident, it exceeds the requirements of 10CFR50.49 and is a forward fit.
Recommendation Section C. 2, f change Determination of exclusion area,. Radiological consequence analyses.
To: The radiation environment must be based on the type of radiation, the total dose expected during normal operation over the installed life of the equipment, and the radiation environment associated with the most severe design basis accident during or following which the equipment is required to remain functional, including the radiation resulting from recirculating fluids for equipment located near the recirculating lines and including dose-rate effects.
radiation that simulates the calculated integrated dose (normal and accident) that the equipment must withstand before completion of its intended safety functions. Cobalt 60 or cesium 137 would be acceptable gamma radiation sources for EQ.
As required in 10 CFR 50.49(e)(4), the radiation environment must be based on the total dose expected during normal operations over the installed life of the equipment and the radiation environment associated with the most severe design basis accident during or following which the equipment is required to remain functional. In addition, GDC 4 requires that SSCs important to safety shall be designed to accommodate the affects and to be compatible with the environmental conditions, including those associated with postulated accidents. RG 1.183 provides guidance on accident radiological source terms and may be used, as applicable, in combination with Appendix D to this guide for radiation equipment qualification. Alternative source terms and assumptions may be developed for assessing equipment qualification to the radiation environment. Any alternatives will be evaluated on a case-by-case basis.
NOTE: This is now in Section C.2.c of RG 1.89, Rev.2.
This proposed guidance does not meet the definition of forward fitting in MD 8.4. RG
23 Commenter Section of DG-1361 Specific Comments NRC Resolution 1.89, Rev. 2 is voluntary guidance and represents one acceptable way to satisfy the applicable NRC regulations. The NRC staff is not imposing or expecting Specific licensees to implement the guidance in RG 1.89, Rev. 2.
Further, forward fitting occurs only in the context of a licensing action, and the issuance of a RG is not a licensing action.
James Gleason Section C.2.f Comment 18 Section C. 2, f is confusing, contradictory to 10CFR50.49, and unnecessary as it states: RG 1.183 establishes an acceptable alternative source term (AST) and identifies the significant attributes of other ASTs that the NRC staff may find acceptable. For new reactor applications, the safety analysis requirements in 10 CFR 50.34(a)(1) and 10 CFR Part 52 (as applicable) include footnotes describing a fission product release similar to the one in the footnote to 10 CFR 100.11 described above. Although 10 CFR 50.49 does not include a similar footnote, power reactor license applicants have typically considered a core melt accident source term for the 10 CFR 50.49 EQ evaluation consistent with the footnote. Appendix D to this guide includes additional guidance on radiation EQ.
10CFR50.49 states: (4) Radiation. The radiation environment must be based on the type of radiation, the total dose expected during normal operation over the installed life of the equipment, and the radiation environment associated with the most severe design basis accident during or following which the equipment is required to remain functional, including the radiation resulting from recirculating fluids for equipment located near the recirculating lines and including dose-rate effects.
Since this section uses the radiation from a core melt accident source term and not the radiation environment associated with the most severe design basis accident, it exceeds the requirements of 10CFR50.49 and is a forward fit.
See the staffs response to comment 17.
24 Commenter Section of DG-1361 Specific Comments NRC Resolution Recommendation Section C. 2, f change RG 1.183 establishes. Guidance on radiation EQ.
To: The radiation environment must be based on the type of radiation, the total dose expected during normal operation over the installed life of the equipment, and the radiation environment associated with the most severe design basis accident during or following which the equipment is required to remain functional, including the radiation resulting from recirculating fluids for equipment located near the recirculating lines and including dose-rate effects.
Comment Document 2: ML21022A044 James Parello Pg. 8 /
Background
/ 1st Paragraph Comment 19 Editorial: Reference is made in a few places in the main document to the term Class 1E but is not defined until Section C. (STAFF REGULATORY GUIDANCE)
Recommendation Recommend defining Class 1E as safety classification of the electrical equipment and systems per IEC/IEEE 60780-323 on Page 8 / Background /
1st Paragraph.
The staff disagrees with the comment. The staff is neither defining the term Class 1E nor taking exception to this term in this RG. Per footnote 3 in 10 CFR 50.49, the NRC still acknowledges the definition for Class 1E as defined in IEEE Std. 323-1974.
Furthermore, this definition remains the same in IEC/IEEE 60780-323-2016. Therefore, it is unnecessary to define this term in the RG.
However, the staff added the following footnote for clarity to DG-1361, As noted in 10 CFR 50.49, the staff considers Class 1E to be synonymous with the term safety-related.
James Parello Pg. 10 /
Section C.1.b.
Comment 20 Section C.1.b. states it provides a description and definition for the term important to safety. But this is not the case. This section defines the subsections within 10 CFR 50.49 for requirements associated with safety-related and non-safety-related electrical equipment as they apply to See the staffs response comment 2.
25 Commenter Section of DG-1361 Specific Comments NRC Resolution important to safety. The definition for important to safety from 10 CFR 50.49 is actual at the end of Section C.1.c.
The definition in IEC/IEEE 60780-323 Clause 3.12 (equipment important to safety) as it applies to IEEE documents and Class 1E categorization is consistent with 10 CFR 50.49(b)(1)(i) and therefore this first sentence is not needed.
Recommendation Delete the first sentence of Section C.1.b.
James Parello Pg. 10 /
Section C.1.c.
Comment 21 A change in the definition for qualified life is not needed. The use of qualified life is used in conjunction with equipment important to safety.
The proposed definition for qualified life is an applied definition based on equipment important to safety undergoing equipment qualification at the end of its service life. The definition in IEC/IEEE 60780-323 for qualified life is appropriate since it is a global industry standard and should be not referencing requirements from a specific regulatory body. The definition in the standard addresses the period of time demonstrated through the equipment qualification process that the equipment will maintain its ability to perform its designated safety function(s) in an accident condition or a postulated earthquake.
Recommendation Recommend deleting the first paragraph of Section C.1.c and consolidating the remaining information if needed in Section C.1.b See the staffs response to comment 3.
James Parello Pg. 10 /
Section C.1.d.
Comment 22 Equipment service life is the actual period of time the equipment is in service. The definition for service life in IEC/IEEE 60780-323 is the period from initial operation to final withdrawal from service of a structure, system or component. The definition does not imply or infer aging effect outside of service are insignificant. I agree the example of shelf life can impact the qualified life of the equipment but not impact the service life.
See the staffs response to comment 4.
26 Commenter Section of DG-1361 Specific Comments NRC Resolution Recommendation Recommend deleting the presumption that the definition for service life of IEC/IEEE 60780-323 implies that aging effects are insignificant unless the equipment is in service.
James Parello Pg. 11 /
Section C.1.d. / 1st Paragraph Comment 23 Note: The definition for service life provided is the same as in IEC/IEEE 60780-323.
Recommendation Recommend deleting the first paragraph: Therefore, the following definition of service life should be used See the staffs response to comment 4.
James Parello Pg. 11 /
Section C.1.e.
Comment 24 Environmental and operational aging of equipment important to safety to the end of its service life in a mild environment is required by IEC/IEEE 60780-323 if it is determined that the equipment has significant aging mechanisms that impacts the ability of the equipment to perform its safety function(s) prior to Design Basis Events (DBE). In a mild environment a seismic event is a DBE. Examples of equipment aging mechanisms in a mild environment prior to DBE are: wear, vibration, thermal and radiation as a function of time.
Recommendation Recommend deleting Section C.1.e.
The staff disagrees with the comment that Section C.1.e be deleted. Rather, the staff has edited the section as described above in the response to comment 5. The purpose of DG-1361 is to describe an acceptable approach for meeting regulatory requirements for environmental qualification of electrical equipment important to safety and to provide guidance for addressing environmental stressors affecting the long-term reliability of electrical equipment.
Pg. 12 /
Section C.1.h.(2).
Comment 25 This section should be updated constant with Staff Position 2 (Page 16 /
Section 2.c.) for defining a mild radiation environment. The Staff considers a mild radiation environment for electronic equipment to be a total integrated dose less than 10 gray (Gy) (103 rad) and a mild radiation environment for other equipment to be less than 100 Gy (104 rad), to be acceptable.)
Recommendation Recommend the following update to Section C.1.h.(2):
The staff partially agrees with the comment.
The staff agrees that doses less than 103 rad for electronic equipment and 104 rad for other equipment generally does not have a significant impact on equipment performance. This is consistent with the staff position that total integrated doses below these levels may be considered to be located in mild radiation environments.
27 Commenter Section of DG-1361 Specific Comments NRC Resolution Electric equipment that may be exposed to low-level radiation doses (electronic equipment to be a total integrated dose less than 10 gray (Gy)
(103 rad) and other equipment less than 100 Gy (104 rad)) should not generally be considered exempt from radiation qualification testing.
Exceptions for higher doses may be based on qualification by analysis supported by test data or operating experience that verifies that the dose and dose rates will not degrade the operability of the equipment below acceptable values.
The staff also agrees that the RG need not include a statement indicating that low-level radiation should generally not be exempt from qualification testing.
However, while the staff agrees that equipment receiving total integrated doses less than the specified values generally does not require specific qualification testing, the staff does not wish to imply that qualification testing should never be considered to ensure appropriate functionality for these lower total integrated doses. In addition, analyses supported by test data or operating experience are considered acceptable methods used to demonstrate qualification, in accordance with IEC/IEEE Std. 60780-323, and not exemptions from qualification.
As a result, paragraph C.1.h.(2) is unnecessary and has been deleted.
James Parello Pg. 12 /
Section C.1.j.
Comment 26 Information presented regarding aging may be better suited to be with the aging details presently in Clause 7.4.1.9.3 (Age Conditioning).
Recommendation Recommend changing Section 7.3.2 to Section 7.4.1.9.3.
The staff agrees with the comment. The staff changed the reference from Section 7.3.2 to Section 7.4.1.9.3.
James Parello Pg. 13 /
Section C.1.k.(5)
Comment 27 The chemical spray or demineralized water spray during design basis event (DBE) testing needs to be conservatively injected after the peak of the environmental profiles (temperature, pressure). Depending on the nuclear facility, chemical spray or demineralized water spray may be initiated at a The staff agrees with the comment. Section C.1.k(5) has been revised as follows:
Chemical spray or demineralized water spray that is representative of service conditions
28 Commenter Section of DG-1361 Specific Comments NRC Resolution time prior to reaching the peaks of the postulated DBE environmental profile.
During DBE testing if the spray is initiated prior to reaching the peak of the DBE profile then the initial profile ramp and peak may not be met.
Recommendation Recommend the following wording change: (5) Chemical spray or demineralized water spray that is representative of service conditions should be incorporated during simulated event testing after the test chamber reaches the maximum at pressure and temperature conditions that would occur when the spray systems actuate.
should be incorporated during simulated event testing after the test chamber reaches the maximum pressure and temperature conditions that would occur when the spray systems actuate.
James Parello Pg. 14 /
Section C.n.(1)
Comment 28 This section requested Clause 7.4.10 be updated with the following: A double-transient should be used with equipment that may be vulnerable to thermal binding from different expansion rates of materials during the initial heatup. This statement is misleading because the potential for thermal binding of materials with different material expansion rates is also addressed during single-transient DBE testing, thermal aging and thermal cycle testing.
The transient used during equipment qualification testing should be representative of the DBE postulated environment for the nuclear facility as a minimum.
Recommendation Recommend deleting Section C.n.(1)
See the staffs response to comment 12.
James Parello Pg. 14 /
Section C.n.(2)
Comment 29 It is unclear how the use of a double-transient will offset tests where the ramp rate (initial temperature rise) of the test is slower than the required profile. By not meeting the initial ramp you have not demonstrated the equipment can withstand the thermal shock and pressure conditions it will experience when changing from its normal environment through the DBE peak environment.
Recommendation Please include the requirements for a double-transient that are acceptable to the NRC for demonstrating a double-transient DBE can be used to See the staffs response to comment 13.
29 Commenter Section of DG-1361 Specific Comments NRC Resolution conservatively represent the initial ramp of a single-transient DBE that cannot be met.
James Parello Page 16 /
Section 2.c.
Comment 30 This states from RG 1.209 that: An additional stressor to be considered in the qualification of digital systems is smoke exposure from an electrical fire.
Stressors caused by fire and smoke are address in design, construction, installation, and procedural practices (e.g., redundancy, diversity, site location, protective barriers, etc.) for the equipment and the nuclear facility it is to be installed. These potential stressors are addressed by others and not in equipment qualification programs addressed by test, analysis, combined test and analysis, or experience programs documented in IEC/IEEE 60780-323.
10 CFR 50.48 and RG 1.209 are the correct documents to address fire and smoke as it relates to the nuclear facility and the impact it has on electric equipment important to safety (not in RG 1.89).
Recommendation Recommend deleting Section 2.c. starting with An additional stressor to be considered.
See the staffs response to comment 14.
James Parello Page 19 /
References
/ Ref. 9.
And Ref.
- 10.
Comment 31 Editorial: Reference 9 and 10 are out of order has they appear in the main body of the document.
Recommendation Change Reference 9 to Reference 10 and vice-versa.
The staff agrees with the editorial comment, items 9 and 10 have been switched in the Reference list to match the order in which they appear in the main body of the document.
James Parello Page 21 /
References
/ Ref. 36.
Comment 32 Reference 36 should be Chapter 11 of 10 CFR 100 has identified on Page 17 (Section 2.f. / 2nd Paragraph
/ 1st Sentence). The title for Chapter 11 is also missing.
Recommendation The comment is no longer relevant as 10 CFR 100.11 is no longer referenced in RG 1.89, Rev. 2 based on staff resolution of comment
- 17.
30 Commenter Section of DG-1361 Specific Comments NRC Resolution Recommend changing: Chapter 1 to Chapter 11 and adding the following chapter title: Determination of exclusion area, low population zone, and population center distance.
Comment Document 3: ML21106A271 Rick Weinacht of Curtis-Wright Section C.1.a Comment 33 Regulatory Position C.1.a should be deleted. It is not necessary to provide clarification of the definition of end condition in order to use IEC/IEEE 60780/323 to meet 10 CFR 50.49. The proposed clarification treats end condition as a condition of an installed component, when the Standard is using it to define the condition of a test specimen. Moreover, it defines a condition, not a time. End condition, therefore, cannot be equivalent to end-of-installed life. End condition could be said to be the condition of a component at the end of installed life, but the Standard already makes this clear. Finally, the Note at the end of the regulatory position uses the term design function. This is confusing because the term safety function is used elsewhere in DG-1361. If not deleted, Regulatory position C.1.a should at a minimum be reworded to say: end condition, as described in Section 3.10 of IEC/IEEE Std. 60780-323, Edition 1, 2016- 02, should be considered equivalent to end-of-installed life condition.
The staff agrees with the comment. A clarification of the definition of end condition is unnecessary, and Section C.1.a has been deleted.
Rick Weinacht of Curtis-Wright Section C.1.b Comment 34 Regulatory Position C.1.b should be deleted or reworded to simply refer to the 10CFR50.49 definition of important to safety. The regulation is already clear that equipment meeting the 10CFR50.49 definition of important to safety must be qualified. The definition of important to safety in Section 3.12 of IEC/IEEE STD. 60780-323, EDITION 1, 2016-02 does not change the regulatory definition in 10CFR 50.49. Regulatory Position C.1 in Revision 1 of RG 1.89 is much more clearly worded.
See the staffs response to comment 2.
Rick Weinacht of Curtis-Wright Section C.1.c Comment 35 Regulatory Position C.1.c. should be deleted or reworded. Restating text from 10CFR50.49 is entirely unnecessary. It appears that this regulatory position is attempting to provide clarification of the definition of safety function.
See the staffs response to comment 3.
31 Commenter Section of DG-1361 Specific Comments NRC Resolution Defining safety function does not require a change or clarification to the definition of qualified life.
Rick Weinacht of Curtis-Wright Section C.1.d Comment 36 Regulatory Position C.1.d should be deleted. The definition in Section 3.22 of IEC/IEEE 60780- 323 makes no implication of aging effects. The proposed regulatory position attempts to add the period prior the operation phase to the service life. This is contrary to long-standing definitions of service life. The addition of the Note is not necessary. The standard already requires equipment to be tested in its end of life condition, including any adverse aging caused by the pre-operational period.
See the staffs response to comment 4.
Rick Weinacht of Curtis-Wright Section C.1.e Comment 37 Regulatory Position C.1.e should be deleted or reworded. It is not necessary to amend the Standard to limit its scope to match that of 10CFR50.49. If EMC and seismic requirements are not within the scope of 10CFR50.49, then those portions of the Standard are not necessary to be followed to meet the regulation. A much clearer statement of the regulatory position would be:
Portions of IEC/IEEE Std. 60780-323, Edition 1, 2016-02, applicable only to equipment in mild environments, such as the fourth paragraph of Section 5.1, are beyond the scope of 10 CFR 50.49.
See the staffs response to comments 5, 14, and
- 30.
Rick Weinacht of Curtis-Wright Section C.1.f Comment 38 Regulatory Position C.1.f creates regulatory confusion instead of clarification by implying that using the new qualification methodologies may not meet 10CFR50.49. If there is concern that the methodologies may not be acceptable, then the Regulatory Guide should stipulate methods that the NRC considers acceptable. The stated object of the IEEE Standard is to demonstrate and document that equipment can perform safety functions under applicable service conditions. If there is no clarification to the condition monitoring and condition-based qualification methodologies in the Standard, this regulatory position should be deleted.
See the staffs response to comment 6.
Rick Weinacht of Section C.1.j Comment 39 Regulatory Position C.1.j.(3) presents new requirements for justification of activation energies that has rarely been met in past qualification efforts, will The staff disagrees with the comment. The additional guidance provided in Section
32 Commenter Section of DG-1361 Specific Comments NRC Resolution Curtis-Wright increase the time and cost of qualification without a substantial increase in product quality or capability, and fails to recognize the great amount of engineering judgement used and needed to establish a qualified life. The regulatory position should end after the first sentence.
C.1.j.(3) adds clarity for defining, justifying, and documenting activation energy and continues to allow the use of engineering judgement when establishing qualified life.
See the staffs response to comment 9 for additional information on the staffs position on activation energy.
Rick Weinacht of Curtis-Wright Definitions Comment 40 end condition - It is agreed that the definition of end condition, defined in Section 3.10 of IEC/IEEE Standard 6078-323, is synonymous with end-of-installed life condition as described in 10CFR50.49(e)(5). The clarification of Regulatory position C.1.a is not necessary.
See the staffs response to comment 33.
Rick Weinacht of Curtis-Wright Definitions Comment 41 important to safety - Regulatory Position C.1.b does not provide a definition of important to safety. IEC/IEEE STD. 60780-323, EDITION 1, 2016-02 and 10 CFR 50.49 contain definitions of important to safety. The description and definition provided in DG-1361 includes the requirement to environmentally qualify equipment important to safety. The requirement to qualify should not be part of the definition or description. DG-1361 endorses IEC/IEEE STD.
60780-323, EDITION 1, 2016-02 with clarifications as an acceptable approach for meeting environmental qualification regulatory requirements.
Because IEC/IEEE STD. 60780-323, EDITION 1, 2016-02 addresses a larger scope of equipment than 10 CFR 50.49 does, it is expected that their definitions of important to safety would differ. Licensees will refer to 10 CFR 50.49 to determine which equipment important to safety must be environmentally qualified, then refer to IEC/IEEE STD. 60780-323, EDITION 1, 2016-02 to determine if the method used to environmentally qualify equipment is acceptable to the NRC. Regulatory position C.1.b is unnecessary. It could be greatly simplified by stating:
See the staffs response to comment 2.
33 Commenter Section of DG-1361 Specific Comments NRC Resolution 10 CFR 50.49 only requires environmental qualification of electrical equipment installed in harsh environments meeting the 10 CFR 50.49 definition of important to safety as described in 10 CFR 50.49(b).
Rick Weinacht of Curtis-Wright Definitions Comment 42 Qualified Life - A definition is not provided in 10CFR 50.49, 10 CFR 50.2 or the NRC Basic References Glossary. Regulatory position C.1.c of DG-1361 proposes an alternate definition to the one provided in Section 3.20 of IEC/IEEE STD. 60780-323, EDITION 1, 2016-02, contrary to the recommendation in from NRC-sponsored research documented in Brookhaven National Laboratory Technical Report TR-6169-9/97, Supplemental Literature Review on the Environmental Qualification of Safety Related Electric Cables, which states, it is recognized that the qualified life is defined in the applicable IEEE standards, and the current definition should be adhered to. The two definitions are repeated below with the differing portion of the proposed revision of the DG-1361 definition highlighted.
IEC/IEEE STD. 60780-323, EDITION 1, 2016-02 Definition: period for which an equipment has been demonstrated, through testing, analysis and/or experience, to be capable of functioning within acceptance criteria during specific operating conditions while retaining the ability to perform its safety functions in accident condition or earthquake Proposed DG-1361 Definition: period for which an equipment has been demonstrated, through testing, analysis and/or experience, to be capable of remaining functional during and following design basis events to ensure that the criteria specified in 10 CFR 50.49(b)(1)(i)(A), (B) and C are satisfied.
The purpose of proposed wording is not clear. It appears the proposed definition is intended to provide terminology consistent with the regulation.
However, the repeating of the text from 10 CFR 50.49 following the definition is unnecessary. Furthermore, the quoted section of 50.49(b)(1) is defining the equipment that is safety related, not criteria for remaining See the staffs response to comment 35.
34 Commenter Section of DG-1361 Specific Comments NRC Resolution functional. Reference only to 10 CFR 50.49(b)(1) gives the implication that qualified life does not apply to the additional equipment within the scope of 10CFR50.49 as defined in paragraphs 10 CFR 50.49(b)(2) and (b)(3). The definition in the Standard is adequate, consistent with the regulation and essentially identical to the definition in EPRI TR-100844. It is recommended that Regulatory position C.1.c be deleted and the definition in the Standard be accepted without modification or clarification.
Rick Weinacht of Curtis-Wright Definitions Comment 43 Service Life - IEC/IEEE STD. 60780-323, EDITION 1, 2016-02 Section 3.22 provides a definition of service life. EPRI TR-100844 provides a nearly identical definition using the term retirement in place of final withdrawal from service. The term service life is not used in 10 CFR 50.49. There is no need for a clarification of the definition of service life in DG-1361.
Furthermore, contrary to Regulatory position C.1.d, Section 3.22 of IEC/IEEE STD. 60780-323, EDITION 1, 2016-02 makes no implications of aging effects. Regulatory position C.1.d should be deleted.
See the staffs response to comment 4.
Rick Weinacht of Curtis-Wright Comment 44 Significant Aging Mechanism - Section 3.24 of IEC/IEEE STD. 60780-323, EDITION 1, 2016-02 provides the following definition of significant aging mechanism: ageing mechanism that, under normal and abnormal conditions, causes degradation of equipment that progressively and appreciably renders the equipment vulnerable to failure to perform its safety function(s) during the design basis event conditions. Section 7.4.1.9.1 of the JLS clearly states that when significant aging mechanisms are identified, suitable age conditioning shall be included in the type test. This implies if the aging effects assessment determines aging mechanisms are not significant, age conditioning for these aging mechanisms does not need to be included.
10 CFR 50.49(e)(5) requires that consideration must be given to all significant degradation that can have an effect on the functional capability of the equipment. A regulatory position should be provided that acknowledges that aging is only required for significant aging mechanisms as defined in Section 3.24 of IEC/IEEE STD. 60780-323, EDITION 1, 2016-02. During the The staff partially agrees with the comment.
The staff agrees that 10 CFR 50.59(e)(5) requires that consideration must be given to all significant degradation that can have an effect on the functional capability of the equipment.
The staff does not agree that a regulatory position needs to be created since the staff is not taking a position contrary to the definition of significant aging mechanism in IEC/IEEE Std. 60780-323, Edition 1, 2016-2 nor the information the comment references in Section 7.4.1.9.1.
The staff recognizes that it would be useful to include an updated table similar to DOR table C-1 but with 60-80 year information; however,
35 Commenter Section of DG-1361 Specific Comments NRC Resolution EQ Design Bases Assurance Inspections (DBAIs), some inspectors questioned if the wording of IEEE Std. 323-1974, Section 6.3.2(4) specifically stating the conditions under which radiation aging would not need to be included as part of aging, precluded a similar approach for other types of aging, such as thermal aging. Memorandum from H. R. Denton, Director ONRR, to Commissioner Kennedy dated 8/24/1979 comparing the 1971 and 1974 versions of IEEE Std. 323(ADAMS Accession No. 7909210029),
clearly demonstrates that the NRC Staff interpretation has always been that consideration and inclusion of aging effects is only required if the aging effect is significant, stating:
The staff guidelines mentioned above will require that aging be considered, but only for that equipment identified as being susceptible to significant aging effects. Additionally, it would be helpful if thermal and radiation susceptibility data, such as the data in Table C-1 of the DOR Guidelines were provided and updated for service lives of 60 and 80 years. Finally, a regulatory position providing guidelines for adequate justification of exempting pre-aging would be useful.
this information is currently unavailable and would require extensive research. Also, while the staff agrees that guidelines for adequate justification of exempting pre-aging could be useful, the staff is not currently prepared to provide a regulatory position or guidance for what constitutes adequate justification of exempting pre-aging.
No revisions were made to DG-1361 as a result of this comment.
Rick Weinacht of Curtis-Wright Section C Comment 45 DG-1361 Regulatory Position C.1.d states that shelf life can adversely impact qualified life and Regulatory Position C.2.a(3) states the shelf life should be addressed for potential impact on qualified life of equipment that was in utility stock prior to February 22, 1983 and may be used as replacement equipment in lieu of upgrading. Evaluation of shelf lifes impact on qualified life as an environmental qualification requirement is a new regulatory interpretation that needs to be evaluated as a backfit. These new regulatory positions imply that the evaluations should be incorporated into environmental qualification documentation. While in some mostly rare cases, qualified life can be impacted by storage time, it has been long understood that shelf life is a period prior to installation and qualified life begins at installation or operation. The original EQ Reference Manual, EPRI Report TR-100516 (published in 1992), Figure 4.13 shows storage occurring before qualified life and service. This figure is unchanged in Revision 1 of the EQ Reference Manual, EPRI Report 1021067. EPRI Report 1021067, Appendix I The staff agrees with this comment. The staff agrees to delete Sections C.1.d and C.2.a(3) as neither service life nor shelf life is directly associated with 10 CFR 50.49 since the main focus of 10 CFR 50.49 is to establish qualified life of electrical equipment, which can be different from service or shelf life.
Furthermore, service and shelf life are addressed elsewhere in the regulations (e.g.,
Appendix B to 10 CFR Part 50).
36 Commenter Section of DG-1361 Specific Comments NRC Resolution discusses the relationship between shelf life and qualified life and recognizes that nuclear power plant practice assumes shelf life does not impact qualified life. NRC Equipment Qualification Training Manual for Nuclear Regulatory Commission Technical Reviewers and Inspectors (ADAMS Accession Number ML16252A163) Slide 225 recognizes this common practice as well.
EPRI Report 10229259, Plant Engineering: Guidelines for Establishing, Maintaining, and Extending the Shelf Life Capability of Limited Life Items, Revision 1 of NP-6408 (NCIG-13), provides industry guidance on establishing shelf lives. Section 2.1 of EPRI 1022959 notes that one of the underlying assumptions of the guidance is that shelf lives do not impact qualified life except in special circumstances. It has already been recommended that Regulatory Position C.1.d in DG-1361 be deleted. The second sentence of Regulatory Position C.2.a(3) in DG-1361 should also be deleted.
Any discussion of shelf life impacting qualified life in DG-1361 should:
- 1.
recognize the industry guidance in EPRI 1022959
- 2.
clarify that shelf life does not normally impact qualified life, and
- 3.
clarify that shelf life evaluations are not expected to be included in environmental qualification documentation.
Regulatory Position C.2.a.(3) adds a new requirement to address the impact of shelf life of replacement equipment on qualified life. The sentence regarding shelf life should be deleted. Licensees have already has addressed the impact of shelf life on qualified life of the equipment in licensee stock via the licensees Quality Assurance Program and licensees procedures and processes. This additional sentence adds regulatory ambiguity because it neither recognizes the most common relationship between shelf life and qualified life (one does not impact the other), nor does it present any method acceptable to the NRC for addressing this potential impact.
In the presentation Maintaining Qualified Life Equipment and Parts in NPPs from Proceedings of the Workshop on Nuclear Power Plant Aging (NUREG/CP-0036) (1982), Agnihotri poses the question of whether storage
37 Commenter Section of DG-1361 Specific Comments NRC Resolution time should be included in qualified life and recommends the question be answered by the USNRC or through industry research. The EPRI Guidance on shelf life in EPRI 1022959 provides industry research and consensus practices regarding shelf life. Regulatory Position C.2.a.(3) essentially re-asks this forty-year-old question without recognizing the industry research or detailing a practice acceptable to the NRC.
Rick Weinacht of Curtis-Wright Secdtion C.1.f Comment 46 Regulatory Position C.1.f states that condition monitoring and associated condition-based qualification methodologies in section 6.3 of IEC/IEEE 60780-323, Edition 1, 2016-02 must, if used, ensure the equipment will perform under the conditions specified in 10 CFR 50.49. It is unclear why a regulatory position specific to condition monitoring and associated condition-based qualification is necessary. All qualification methodologies must ensure the equipment will perform under the conditions specified in 10 CFR 50.49.
The draft guide states its purpose is to describe an approach to the NRC for meeting EQ regulatory requirements and that it endorses IEC/IEEE 60780-323, Edition 1, 2016-02. The Standard clearly states that condition-based qualification is an adjunct to type testing. Regulatory Position C.1.f provides no clarification, seems to indicate some reluctance in accepting condition monitoring and condition-based qualification as an acceptable practice.
Condition Monitoring was Technical Issue 6.b of the EQ Task Action Plan (ADAMS Accession Number 95050236). The methods described in Section 6.3 of IEC/IEEE 60780-323, Edition 1, 2016-02 are consistent with conclusions of the research conducted under the EQ Task Action Plan and Generic Safety Issue 168 (ADAMS Accession Number ML021360234).
Regulatory Position C.1.f should be deleted.
See the staffs response to comment 6.
Rick Weinacht of Curtis-Wright Section C.1.h Comment 47 Regulatory position C.1.h. should incorporate beta dose reduction methods from Section 4.1.2 of the DOR Guidelines and include the allowance to remove the requirement for additional radiation margin if approved, conservative dose calculations methods are used as stated in NUREG-0588, The staff partially agrees with the comment.
The guidance related to beta radiation has been updated in Section C.1.h as discussed in the response to comment 7. The staff notes that much of the guidance contained in the DOR Guidelines was specific to the TID-14844
38 Commenter Section of DG-1361 Specific Comments NRC Resolution Category I, Section 1.4 and RG 1.89, Revision 1, regulatory Position C.2.c(6).
Specifically, the regulatory guide should confirm:
Beta dose may be reduced by a factor of 10 for the first 30 mils of cable insulation and an additional factor of 10 for the next 40 mils of insulation.
Cables arranged in cable trays inside of containment shall be assumed to be exposed to half the beta dose plus the gamma dose at the containment centerline.
Equipment shall be considered qualified, without any additional radiation margin, if qualified to radiation doses using the methods of Appendix D.
If analysis shows that beta dose to sensitive equipment internals is less than or equal to 10% of the gamma dose, the equipment is considered qualified if qualified to the gamma dose.
The methods for addressing beta radiation are based on sound principles that are irrespective of the regulatory basis for qualification (i.e., DOR/NUREG-0588/10CFR50.49).
Regulatory position C.1.h.(2) should provide more detailed guidance about acceptable radiation exemption analysis (threshold) and pedigree of test data. A table similar to Table C-1 of the DOR Guidelines detailing agreed upon radiation threshold and allowable levels would be very helpful in clearly communicating the amount and pedigree of additional data that is necessary to exempt radiation. At a minimum, Table C-1 of the DOR Guidelines should be validated as an acceptable reference.
source term and assumptions specified in the DOR Guidelines. In addition, much of the DOR Guidelines guidance discussed in the comment was not carried forward into previous versions of RG 1.89 or RG 1.183, Revision 0.
For these reasons, that guidance is not being included in Revision 2 of RG 1.89. However, licensees do not need to change their approach for EQ due to the revision to RG 1.89 and previous guidance can continue to be used by licensees and applicants as a method acceptable to the NRC staff for demonstrating compliance with the regulations when appropriately applied and applicable. As an example, RG 1.183, Rev. 0, indicates that either the TID-14844 or the alternative source term may be used for equipment qualification, even for those plants that use the alternative source term for control room and public dose. However, RG 1.183, Rev. 0, is only applicable within certain parameters, for example, plants with peak burnups up to 62,000 MWD/MTU and plants with 5% enrichment. If a plant were to make a design change that is beyond the bounds of RG 1.183, Revision 0 (such as change to a burnup above 62,000 MWD/MTU), and continue to use guidance in RG 1.183, Revision 0, the licensee needs to consider and address any new or unreviewed issues created and ensure that the proposed implementation of previous guidance is technically justified.
39 Commenter Section of DG-1361 Specific Comments NRC Resolution As noted in the staffs response to comment 44, an update to Table C-1 of the DOR Guidelines would require information that is not currently available, would require extensive research, and is beyond the scope of this RG..
Therefore, Table C-1 of the DOR Guidelines has not been reviewed or updated as part of the RG 1.89 revision.
Rick Weinacht of Curtis-Wright Section C.1.j Comment 48 Regulatory position C.1.j(1) provides statements intended to supplement the guidance of Section 7.3.2 of IEC/IEEE Std. 60780-323, Edition 1, 2016-02.
The supplemental statements in the draft guide are not appropriate solely for Section 7.3.2 as synergistic effects are discussed throughout the Standard.
The basic point of Section 7.3.2 of IEC/IEEE Std. 60780-323, Edition 1, 2016-02, is that the Qualification Plan shall address aging effects. It could be reasonably argued that synergistic effects should be added to Section 7.3.2 as an aging factor that should be considered, but other sections of the Standard adequately cover synergistic effects.
The supplemental statements in the draft guide do not provide regulatory clarity as to what constitutes acceptable accounting for synergistic effects.
Examples:
Revision 1 of RG 1.89 clearly states the known synergistic effects at the time of its publication, namely dose rate effects and sequence effects.
Similar clarity should be provided by the next revision of RG 1.89. The Regulatory Guide should state what the known synergistic effects are, what materials are known to be affected by them, and the acceptable methods for adequately accounting with those affects.
Regulatory position C.2.d adds to the regulatory ambiguity regarding synergistic effects, stating Diffusion-limited oxidation, synergisms, dose-rate effects, and inverse temperature are examples of uncertainties related to aging degradation. The following confusion is created:
o It is unclear what is meant by synergisms in this statement.
See the staffs response to comments 8, 67, 96, 121, 136 and 165 regarding comments on Sections C.1.j(1) and C.2.d of DG-1361.
40 Commenter Section of DG-1361 Specific Comments NRC Resolution O
It is unclear why dose-rate effects are listed separately from synergistic effects.
O While research has shown that diffusion-limited oxidation (DLO) can lead to heterogenous material conditions, this uncertainty has not been shown to impact LOCA performance. The recommendation for Issue A.2 (does DLO impact qualification test results using accelerated aging?) in NUREG/CR-6384, Volume 2, states:
Research has not shown a difference in LOCA performance for cables with and without oxidation diffusion. This issue is not resolved, however, no further research on oxygen diffusion limitation is recommended.
The recommendation for Issue C.4 (is material geometry (slabs vs. cables) important?) in NUREG/CR-6384, Volume 2, states:
past work has not shown any conclusive evidence that these effects would significantly affect qualification results. Therefore, no further studies are recommended in this area.
The Regulatory Guide should clearly state that current qualification practices adequately compensate for any uncertainty created by DLO. If the NRC Staff disagrees that the current qualification practices are sufficient to account for DLO, contrary to its recommendation against further research in this area, the Regulatory Guide should describe acceptable methods for accounting for DLO.
There is no industry consensus that inverse temperature effects have a significant impact on aging degradation. Possible concerns with an inverse temperature effect for some formulations of some compounds is discussed in NUREG/CR-7153, Volume 5, Expanded Materials Degradation Assessment (EMDA), Volume 5: Aging of Cables and Cable Systems. [NUREG/CR-7153 falsely asserts that inverse temperature considerations are summarized in Volume 1 of NUREG/CR-6384, Literature Review of Environmental Qualification of Safety-Related Electric Cables, as an uncertainty in the Arrhenius methodology. The term inverse temperature is not used anywhere in NUREG/CR-6384, Volume 1. The manuscript for NUREG/CR-6384, Volume 1 was completed in October 1995. All but one of the references cited in NUREG/CR-7153 for observation of inverse temperature
41 Commenter Section of DG-1361 Specific Comments NRC Resolution effects were published after that date. The lone reference published prior to October 1995 was published in 1994. This reference is not cited in NUREG/CR-6384, Volume 1.]
It is recommended that the draft guide be revised to summarize research to date and provide acceptable practices for addressing synergistic effects.
Where available, the practices outlined should identify:
o Materials for which synergistic effects have been shown to be significant, and best method for addressing those synergies, o
Materials for which synergistic effect have been shown to be minimal and test methods that are adequate, o
Dose rates that are generally acceptable, and specifically acceptable for certain materials, o
Conservatisms and test condition practices known to eliminate or minimize synergistic effects, such as total test doses above 200 Mrad applied at dose rates less than 1 Mrad/hr, o
Material types shown to have more degradation when exposed to a particular sequential test sequence as compared to another, o
Acceptable or preferred practice when no data is available to indicate if a material is subject to synergistic effects, o
Clear statement of known, significant synergistic effects at the time of regulatory guide development, o
Areas currently being research for possible synergistic effects, but for which there is currently insufficient data to determine if the synergistic effect is significant. Inverse temperature effect is one such area.
Rick Weinacht of Curtis-Wright Section C.1.j Comment 49 DG-1361, Section C.1.j(3) states Activation energy values should be based on the testing of the specific compound used in the equipment and on the most relevant material property and property endpoint (i.e., failure mechanism). It also states, The selected activation energy values should be traceable to a specific test report for which these values were established, including the specific material property for which the activation energy was developed and how that material property is related to the function of the See the staffs response to comment 9.
42 Commenter Section of DG-1361 Specific Comments NRC Resolution material in question. These statements show a lack of recognition of the limited availability of activation energies for specific compounds, material properties and material endpoints, and does not recognize the substantial cost and time required to develop activation energies. These statements represent guidance for definition, justification and documentation of activation energies that goes beyond what is currently required by the regulation, 10 CFR 50.49, and the Standard which the draft guide is attempting to endorse.
Examination of some Unresolved Issues (URIs) issued during the recent round of NRC Inspections under Inspection Procedure 71111.21N will demonstrate the inadequacy of this guidance. URI 05000390, 391/2017007-05 (Watts Bar Inspection Report 2017-007 (ADAMS Accession Number ML17220A153)) and URI 05000395/2018010-06 (VC Summer Inspection Report 2008-010 (ADAMS Accession Number ML18094A162)) both raise issues with the activation energy for electronic components in Barton transmitters. In these two URI cases, an extremely conservative original activation energy of 0.5 eV was assigned by the supplier, Westinghouse. The Westinghouse activation energy basis does not meet the specific compound, material property and material endpoint criteria of DG-1361. The manufacturer, Barton, assigned a higher activation energy of 0.78 eV for the electronic components in later qualification reports. In fact, the activation energy of 0.78 eV has been widely used for electronic components in transmitters of other manufacturers, often citing the same space program report cited in the VC Summer Inspection Report, as well as for other equipment. Although both cited URIs were eventually closed as violations, neither closure resolved the original issue of whether the activation energy used by the licensee was appropriate or adequately justified and documented.
Without additional clarification in a revision to RG 1.89 concerning definition, justification and documentation of activation energy bases, future unresolved issues are likely.
Many additional examples could be discussed, but more examples would only bring additional, unwarranted attention to the qualification significance of thermal qualified lives determined using the Arrhenius methodology and a conservatively selected activation energy based on the best data available at the time.
43 Commenter Section of DG-1361 Specific Comments NRC Resolution The issue can be summarized with two main points:
C.
Activation energy selection and use in qualified life calculations requires the use of a great deal of engineering judgement.
Dr. Sal Carfagno, aging expert and NRC consultant, succinctly makes this point in his paper presented at the 1982 Workshop on Nuclear Power Plant Aging (NUREG/CP-0036). Dr.
Carfagno offers: it becomes all the more obvious that engineering judgment is not only an essential factor in establishing qualified life, it is actually the dominant factor.
NRC Staff has also made similar statements. In the Staffs letter to the Commission dated August 24, 1979 (Accession Number 7909210029) Harold Denton writes:
However, even with its greater detail, IEEE Std. 323-1974 still requires a significant amount of engineering judgement in its implementation especially in the area of aging and margins.
- 2.
Establishing a thermal qualified life has only marginal significance in the safety of a commercial nuclear power plant.
The NRC has concluded that requirements to establish a qualified life and other differences in the 1974 version of IEEE Std. 323 are not safety significant, represent only an incremental improvement and do not warrant backfitting. These conclusions were reached,
- 1.
After release of IEEE 323-1974
- 2.
During the promulgation of the EQ rule in 1983,
- 3.
As part EQ Task Action Plan, Item 3.f begun in 1993 The November 15, 1996 Status Report on the EQ Task Action Plan to the Commission (Accession Number 9611200041) states, At the time of its release, the NRC considered backfitting IEEE 323-1974 to older plants, but recommended against it because the incremental improvements provided by the new standard were not considered safety significant and full implementation of IEEE 323-1974 required further development of other ancillary standards. Public comments and a review by the Advisory
44 Commenter Section of DG-1361 Specific Comments NRC Resolution Committee on Reactor Safeguards (ACRS) did not alter the recommendation concerning backfitting the standard.
In the Staffs letter to the Commission dated August 24, 1979 (Accession Number 7909210029) Harold Denton writes:
The benefit of backfitting either the aging or the margin requirements of the 1974 Standard is a small, unquantifiable increase in the level of assurance that equipment is qualified. Yet the costs in terms of manpower, the testing required to implement these provisions and the possible delay in the staff review effort may be significant.
In the November 15, 1996 Status Report on the EQ Task Action Plan to the Commission, the Staff re-validates the conclusions made in 1979 and after release of IEEE Std. 323-1974:
The staff, therefore, has reasonable assurance that its decision not to backfit older plants to the newest EQ requirements was not flawed and remains valid.
The NRC has repeatedly asserted that requirement to establish a qualified life and pre-age equipment has, at best, an incremental improvement on assurance equipment will perform as required. New requirements on the justification of activation energy are wholly unwarranted and should be deleted.
Rick Weinacht of Curtis-Wright Section C.1.m Comment 50 Regulatory Position C.1.m states: Section 7.4.1.9.3 of IEC/IEEE Std. 60780-323, Edition 1, 2016-02, should be supplemented with the following: For insulating materials, a regression line (IEEE Std. 101, IEEE Guide for the Statistical Analysis of Thermal Life Test Data (Ref. 32)), may be used as a basis for selecting the aging time and temperature. Sample aging times of less than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> should not be used.
This regulatory position is supplementing the current standard with an exact statement from the 1974 version of IEEE 323-1974. In practice, this supplemental statement has very little impact on qualification practices.
However, it represents a failure of the regulatory guidance to embrace the updates to the Standard. It also could be misinterpreted to imply that the Arrhenius methodology and use of regressions lines only applies to insulating materials.
The staff partially agrees with the comment.
The point of carrying forward the statement from IEEE 323-1974 was to provide an option for using statistical analysis of thermal life test data. It was not intended to imply that a regression line alone forms the basis for the aging time and aging temperature.
Nonetheless, the staff modified this statement to note that it applies to organic materials and is not limited to insulating materials.
The staff also clarified the guidance to read as follows: A regression line alone does not form a basis for the aging time and aging
45 Commenter Section of DG-1361 Specific Comments NRC Resolution A regression line alone does not actually form a basis for the aging time and aging temperature. It provides an activation energy proportional to the slope of the regression line that can be used to determine the amount of time at the aging temperature to cause thermal aging equivalent to aging that would occur during the desired service life at the service temperature. The aging time and aging temperature are not a point on the regression line. Once the activation energy is determined, an aging time can be calculated for an assumed aging temperature or an aging temperature can be calculated for an assumed aging time. There are countless aging time and temperature combinations that can be determined from a regression line.
This regulatory position should be deleted, or additional guidance should be provided for materials that are not insulating materials.
temperature. This approach provides an activation energy proportional to the slope of the regression line that can be used to determine the amount of time at the aging temperature to cause thermal aging equivalent to aging that would occur during the desired life at the expected temperature. The aging time and aging temperature are not a point on the regression line. Once the activation energy is determined, an aging time can be calculated for an assumed aging temperature or an aging temperature can be calculated for an assumed aging time.
Rick Weinacht of Curtis-Wright Section C.1.j Comment 51 Regulatory Position C.1.j.(3) states data extrapolations should be minimized by using activation energy values within the temperature range of interest.
And the activation energy should be selected based on the temperature range of the equipment in service. While these statements are consistent with guidance in the relevant IEEE Standards, they fail to provide guidance against which acceptable extrapolation and activation energy selection can be judged.
Terms such as range of interest, applicable temperature range, and good fit are too vague to allow objective agreement if the activation energy is adequately justified. Many Electrical Insulating Material test programs recommend a minimum of 5000 hours0.0579 days <br />1.389 hours <br />0.00827 weeks <br />0.0019 months <br /> at the lowest aging temperature. Even this aging temperature may be well above the service temperature. One could follow the recommendations in the IEEE Standards for development of thermal indices and still be questioned as to whether they met the requirements of the draft guide.
Design Bases Assurance Inspections at Sequoyah, Brunswick, Summer, Watts Bar and St. Lucie all raised issues related to the extrapolation of Arrhenius data. The draft guide offers no guidance that would allow an See the staffs response to comment 9.
The staff partially agrees with the comment.
The staff agrees this information could be useful, however, the staff disagrees with incorporating this information into the revised regulatory guide since specific acceptance criteria for activation energy may not be available because it is material dependent and based on specific failure modes.
46 Commenter Section of DG-1361 Specific Comments NRC Resolution inspector to ascertain if the extrapolations were proper and within regulatory requirements.
Regulatory Position C.1.j.(3) should be reworded to remove ambiguity and recognize which standards, if followed, are adequate for the purposes of defining, justifying, and documenting the basis for activation energy.
Rick Weinacht of Curtis-Wright Section C.1.n Comment 52 Regulatory Position C.1.n discusses the use of double-transient test profiles, explaining the outdated stipulation for using double-transient test profiles as a method for adding margin in the 1974 version of IEEE Std. 323.
Regulatory position C.1.n introduces regulatory ambiguity by implying:
- 1)
A double-transient test profile should be used in some circumstances
- 2)
A double-transient test profile is an adequate method to compensate for test profiles that have ramp rates ramp (initial temperature rise) slower than the required profile.
- 3)
A test profile with a slower ramp rate than the required profile is inadequate for qualification. There is no regulatory basis or research data supporting this regulatory position. It should be deleted.
In fact, NRC-sponsored research has already concluded that single transient DBA testing is acceptable, and double transient DBA testing may be superfluous. As documented in NUREG/CR-6384, Volume 2, Literature Review of Environmental Qualification of Safety-Related Electric Cables.
Use of single transient versus double transient was one of the questions raised as part of the EQ Task Action Plan. The research concluded that this issue was resolved, and no further research was required.
Test laboratories typically meet ramp rate on a best-effort basis. This stipulation is often included in contracts and test plans. Efforts to meet a specified ramp rate as an acceptance criterion often results in significant overshoot and test profile instability. Adequate accident chambers can easily meet peak temperatures typical of most LOCAs in times much less than one minute. Thermal lag analysis shows the difference in equipment temperatures varies little between for test times to peak temperatures of less than one minute compared to a conservatively postulated times on the order of 10 seconds. The regulatory guidance should clearly state that time to reach peak See the staffs response to comments 12 and
- 13.
47 Commenter Section of DG-1361 Specific Comments NRC Resolution temperature, so long as it is less than one minute or so, is of no consequence to qualification test results Rick Weinacht of Curtis-Wright Editorial Page 4 Comment 53 The description of RG 1.40 does not match the Reg Guide number and title.
The description matches the RG 1.140 subject matter. Since RG 1.140 does not provide detail for qualifying equipment, it does not belong in this list. The description of RG 1.40 needs to be revised to match the subject matter of RG 1.40.
The staff agrees with the comment. RG 1.140 is Containment Isolation Provisions for Fluid Systems. RG 1.40 is Qualification of Continuous Duty Safety-Related Motors for Nuclear Power Plants, and endorses IEEE 334-2006, Qualifying Continuous Duty Class 1E Motors for Nuclear Power Generating Stations. Therefore, the original description in the Related Guidance section was incorrect.
The staff has modified the description to read as follows:
RG 1.40, Qualification of Continuous Duty Safety-Related Motors for Nuclear Power Plants, endorses IEEE 334-2006, Qualifying Continuous Duty Class 1E Motors for Nuclear Power Generating Stations, and describes a method that the staff of the NRC considers acceptable to implement regulatory requirements for the qualification of continuous duty safety-related motors for nuclear power plants.
Rick Weinacht of Curtis-Wright Editorial Page 7 Comment 54 The word provides should be changed to provide in the second Background paragraph (Chapter 11 and Appendix A..provide)
The staff agrees with the comment and made the editorial change.
Rick Weinacht of Curtis-Wright Editorial Page 17 Comment 55 Regulatory Position C.2.f is less clear than Regulatory Position C.2.c.(5) of RG 1.89, Revision 1. The new regulatory position mixes in concepts covered in other sections of the Draft Guide. This regulatory position would be clearer if it remained focused on considerations for determining the required See the staffs response to comments 16, 17, and 18.
48 Commenter Section of DG-1361 Specific Comments NRC Resolution radiation dose. Calculational methods should be discussed in a separate regulatory position.
Comment Document 4: ML21041A127 William Horin of NUGEQ General Comment 56 - Summarized 60 Day comment extension and public meeting request Based on this comment and others requesting more time to provide comments on DG-1361, the NRC staff reopened the comment period for an additional 60 days via Federal Register notice, 86 FR 10133 (February 18, 2021).
Based on several public meeting requests related to DG-1361, the NRC staff held a public meeting on May 13, 2021. See ADAMS Accession No. ML21160A276.
No changes to DG-1361 were made as a result of these comments.
Comment Document 5: ML21041A128 William Horin of NUGEQ General Comment 57 - Summarized Information to support comment extension and public meeting request See the staffs response to comment 56.
Comment Document 6: ML21110A055 Robert Konnik of IEEE General Comment 58 We think it should be made clear that the latest revision of IEEE 323 (IEEE/IEC 60780-323) builds on IEEE 323-1974 and equipment qualified to IEEE/IEC 60780-323 would encompass qualification to IEEE 323-1974. In the forward of IEEE 383-2003 it states that Electrical equipment qualified in accordance with either IEEE 323-1974 or IEEE 323-1983 will meet the requirements of IEEE 627-1980 which provide the basic principles for design qualification for all safety systems equipment for use in Nuclear Power Generating Stations. This revision to IEEE 323-1974 was made to clarify its requirements and impose no additional requirements for qualifying Class 1E The staff disagrees with the comment. When RG 1.89, Rev. 2 is issued, both versions of RG 1.89 (which, together, endorse IEEE St. 323-1974 and IEC/IEEE 60780-323-2016 with noted clarifications/exceptions) will provide acceptable methods of meeting the NRCs requirements. Therefore, the comments recommendation isnt necessary. Furthermore, the NRC has not officially endorsed IEEE Std.
49 Commenter Section of DG-1361 Specific Comments NRC Resolution equipment, The 2003 version of IEEE 323 incorporated additional information and clarified several areas that are outlined in the introduction which include the use of IEEE 323 for qualification of equipment in mild environments when desired, design basis event nomenclature, updated test margins, EMI/RFI, and the use of qualified condition. Similarly, in the forward if IEEE/IEC 60780-323, the main technical changes were to harmonize the two documents consider the need to reassess and extend the qualified life. Each revision clarifies and adds information. The white paper by Jim Gleason detailing the major additions and clarifications of IEEE Std 323-2003 Compared to IEEE Std 323-1974 Dated 12/3/07 noted that IEEE Std 323-2003 contains the same qualification methods and process as was contained in IEEE Std 323-1974, but contains additional requirements that have been identified since the development of IEEE Std 323-1974, including lessons learned from NRC research. Similarly, the white paper on IEEE 323-2003 to IEC/IEEE 6078-323: 2016 Changes by IEEE WG SC2.1 Chairman John White and Vice Chairman Robert Konnik dated 5/1/2017, provided information on the more than 40 changes, but as noted, IEC 60780 was based on IEEE 323-1984, so many of the updates were changes in terminology and additions.
Recommendation Add a statement to be clear that equipment qualified to this latest edition of IEEE 323 (IEEE/IEC 60780- 323) encompasses qualification to IEEE 323-1974.
627. The staff may consider endorsing this standard in the future.
No changes to DG-1361 were made as a result of this comment.
Robert Konnik of IEEE Pg. 10 /
Section C.1.a.
Comment 59 Section 3.10 is a general definition of end condition, which is as stated the condition at the end of the aging. It does not imply that this must be the end of the installed life. Equipment may be qualified to a time that is different than what will ultimately be the installed life (continued qualification, condition-based qualification, etc.). It should also be noted that when used in conjunction with condition-based qualification in 7.3.4 In this case, the end condition with margin is the basis of qualification, and the time to reach that See the staffs response to comment 33.
50 Commenter Section of DG-1361 Specific Comments NRC Resolution end condition in service may be more or less than the qualified life established by age conditioning based on the actual service conditions.
Recommendation Recommend that C.1.a be deleted.
Robert Konnik of IEEE Pg. 10 /
Section C.1.d.
Comment 60 Equipment service life is the actual period of time the equipment is in service. The definition for service life in IEC/IEEE 60780-323 is the period from initial operation to final withdrawal from service of a structure, system or component. The definition does not imply or infer aging effect outside of service are insignificant.
If equipment is improperly stored, shelf life can impact the qualified life of the equipment but not impact the service life.
Recommendation Recommend deleting the presumption that the definition for service life of IEC/IEEE 60780-323 implies that aging effects are insignificant unless the equipment is in service.
See the staffs response to comment 4.
Robert Konnik of IEEE Pg. 11 /
Section C.1.e.
Comment 61 Environmental and operational aging of equipment important to safety to the end of its service life in a mild environment is required by IEC/IEEE 60780-323 if it is determined that the equipment has significant aging mechanisms that impacts the ability of the equipment to perform its safety function(s) prior to Design Basis Events (DBE). In a mild environment a seismic event is a DBE. Examples of equipment aging mechanisms in a mild environment prior to DBE are: wear, vibration, thermal and radiation as a function of time.
Recommendation Recommend deleting Section C.1.e.
The staff partially agrees with the comment.
The staff agrees with the comment that environmental stressors like wear, vibration, thermal, and radiation are examples of aging mechanisms in a mild environment and that environmental qualification of electrical equipment in a mild environment is beyond the scope of 10 CFR 50.49, However, the staff disagrees with the recommended deletion of Section C.1.e. Section C.1.e. describes the qualification of equipment in a mild environment. Although the qualification of equipment in a mild environment is not within
51 Commenter Section of DG-1361 Specific Comments NRC Resolution the scope of 10 CFR 50.49, it is within the scope of the RG, which concerns the environmental qualification of certain electric equipment important to safety..
No changes to DG-1361 were made as a result of this comment.
Robert Konnik of IEEE Pg. 12 /
Section C.1.j.
Comment 62 Information presented regarding aging may be better suited to be with the aging details presently in Clause 7.4.1.9.3 (Age Conditioning). Also note that it is not clear that electromagnetic conditions are generally independent of aging and design basis events. For active canceling that may not be the case.
It may generally be the case for passive canceling, but seismic may still effect this. Also note that in 7.4.1.8 of IEEE/IEC 60780-323 it already states that EMI/RFI tests need not use the same sample.
Recommendation Recommend changing Section 7.3.2 to Section 7.4.1.9.3.
See the staffs response to comment 26.
Robert Konnik of IEEE Pg. 12 /
Section C.1.j.(1)
Comment 63 Section 7.4.1.9.3 of IEEE/IEC 60780-323 already states that age conditioning should consider synergistic effects.
Recommendation Recommend changing Section 7.3.2 to Section 7.4.1.9.3.
See the staffs response to comment 26.
Robert Konnik of IEEE Pg. 12 /
Section C.1.j.(2)
Comment 64 Section 7.4.1.9.3 of IEEE/IEC 60780-323 already discusses the maximum temperature during normal operation being used and Arrhenius methodology being acceptable. If there are other methods besides the Arrhenius method (IEEE 98 and 99), it is expected that IEEE will modify an existing standard or develop a new standard that IEEE/IEC 60780-323 can reference.
Recommendation See the staffs response to comment 26.
52 Commenter Section of DG-1361 Specific Comments NRC Resolution Refer to Section 7.4.1.9.3 of IEEE/IEC 60780-323 if this is needed.
Robert Konnik of IEEE Pg. 12 /
Section C.1.j.(3)
Comment 65 Section 7.4.1.9.3 of IEEE/IEC 60780-323 already discusses acceleration of aging and appropriate documentation. Ideally you would want to establish the activation energy using the temperatures that you will operate the equipment in, but this would take at least 60 or 80 years (if this is the expected life), but likely hundreds of years or more. Since this is impracticable, IEEE 98 and 99 were developed to be able to determine the activation energy within a reasonable time.
For some materials, such as cables and splices (IEEE 383), the specific compound must be used as noted but, in some cases, it is not feasible to identify the specific material compound used in the equipment. It is important though to use a conservative activation energy in this case. The industry has used generic materials activation energies as an acceptable method when the exact compounds cannot be determined. The industry has judged the lowest applicable generic published activation energy for materials aging programs for many years as acceptable.
Recommendation Delete Of note, the activation energy should be selected based on the temperature range of the equipment in service to ensure that the equipment remains functional during and following a design-basis event. And replace with Activation energy should be determined using the guidance in IEEE 98 or 99 to ensure that the equipment remains functional during and following a design-basis event.
Remove the wording testing of the specific in second sentence.
Add Testing of the specific material is required by some standards such as IEEE 383. It is recognized that in some cases, it is not practicable to use the specific compound on all parts in a piece of equipment and the use a conservative activation energy may be used if justified.
The staff partially agrees with the comment.
The staff disagrees with deleting the suggested sentence as it provides information that should be considered when determining activation energy values.
The staff modified Section C.1.j(3) to include references to IEEE Std. 98-2016 and IEEE Std.
99-2019 as follows:
IEEE Std. 98-2016, IEEE Standard for the Preparation of Test Procedures for the Thermal Evaluation of Solid Electrical Insulating Materials, (Ref. XX) and IEEE Std. 99-2019, IEEE Recommended Practice for the Preparation of Test Procedures for the Thermal Evaluation of Insulation Systems for Electrical Equipment, (Ref. XX) contains additional technical information and criteria useful for determining activation energy values.
However, the staff is not officially endorsing these IEEE Standards in this RG.
See the staffs response to comments 9 and 161 for additional information pertaining to activation energy guidance included in the RG.
53 Commenter Section of DG-1361 Specific Comments NRC Resolution Robert Konnik of IEEE Page 16 /
Section C.2.c.
Comment 66 This states from RG 1.209 that: An additional stressor to be considered in the qualification of digital systems is smoke exposure from an electrical fire.
Stressors caused by fire and smoke are addressed in design, construction, installation, and procedural practices (e.g., redundancy, diversity, site location, protective barriers, etc.) for the equipment and the nuclear facility it is to be installed. These potential stressors are addressed by others and not in equipment qualification programs addressed by test, analysis, combined test and analysis, or experience programs documented in IEC/IEEE 60780-323.
10 CFR 50.48 and RG 1.209 are the correct documents to address fire and smoke as it relates to the nuclear facility and the impact it has on electric equipment important to safety (not in RG 1.89).
Recommendation Recommend deleting Section 2.c. starting with An additional stressor to be considered.
See the staffs response to comment 14.
Robert Konnik of IEEE Page 16 /
Section C.2.d.
Comment 67 Note that IEEE/IEC 60780-323 section 7.4.1.9.3 discusses the use of Arrhenius aging and the sequence of age conditioning should consider sequential, simultaneous, and synergistic effects in order to achieve the worst state of degradation expected. What are the specific degradation processes that are not amenable to preconditioning that could result in a common cause failure during design basis accidents?
Recommendation It should noted that IEEE/IEC 60780- 323 states that preconditioning for thermal should use the Arrhenius theory as well as the sequence of age conditioning should consider sequential, simultaneous, and synergistic effects in order to achieve the worst state of degradation expected.
Also state the specific degradation processes that are not amenable to preconditioning that could result in a common cause failure during design basis accidents.
The staff agrees with the comment in that the noted statements are relevant. However, it is not necessary to repeat the suggested statement from IEEE/IEC 60780-323-2016 because the proposed RG revision is not superseding the information in Section 7.4.1.9.3 of IEEE/IEC 60780-323-2016. Therefore, the statement remains applicable. In addition, the statement suggested by the comment on specific degradation processes is already incorporated in the Background section of the RG Based on several similar comments, the staff determined that the information in Section C.2.d was less of a staff position and more of general information that would be better suited in the Background section of the revised
54 Commenter Section of DG-1361 Specific Comments NRC Resolution regulatory guide. Therefore, the staff modified the regulatory guide to revise and relocate this information to the Background section.
Robert Konnik of IEEE Page 17 /
Section C.2.e.
Comment 68 We do not know if equipment outside containment would generally experience a less severe environment, but we do know that in some plants more severe environments are outside containment. Is item 4 a new analysis that plants need to perform?
Recommendation Delete item 1 and clarify item 4.
See the staffs response to comment 15.
Robert Konnik of IEEE Page 19 /
References
/ Ref. 9.
And Ref.
- 10.
Comment 69 Editorial: Reference 9 and 10 are out of order has they appear in the main body of the document.
Recommendation Change Reference 9 to Reference 10 and vice-versa.
See the staffs response to comment 31.
Robert Konnik of IEEE Page D-1/Appendix D
Comment 70 Note that IEEE 383 requires testing with normal dose and total integrated dose. Additionally, to perform condition monitoring would need to perform tests without combining normal and accident dose.
Recommendation It should be noted that testing with normal dose is required by some standards and to perform condition monitoring testing would need to be performed without combining normal and accident dose.
The staff partially agrees with the comment.
IEEE 383, as endorsed by RG 1.211. is one acceptable method to comply with 10 CFR 50.49 and similarly, DG-1361, Appendix D, provides one acceptable method for accounting for the radiation environment in environmental qualification. As such, condition-based qualification is one approach that can be used for environmental qualification, and DG-1361, Appendix D does not preclude or prevent the use of condition-based qualification. IAEA NP-T-3.6, Assessing and Managing Cable Ageing in Nuclear Power Plants provides additional information on condition-based qualification.
55 Commenter Section of DG-1361 Specific Comments NRC Resolution The staff agrees that condition-based qualification would require the normal and accident dose be applied separately.
The staff did not make any changes to DG-1361 as a result of the comment.
Comment Document 7: ML21050A358 Vincent Bacanskas Section A Comment 71 Current NRC licensees have gone through a rigorous process of qualification over the past 43 years (1978-2021). Licensees were subject to a review of all qualification documentation in the early 1980s with subsequent staff meetings for corrective actions, follow up inspections on those corrective actions and in the past 5 years, re-inspection to validate the continuation of the program. Many licensees have gone through re-examination of their files for methodology to support extension of their operating licenses. The existing licensing bases are firmly established for operating reactors and reactors currently under construction. Review of the dual logo standard shows no promise of burden reduction on existing licensees, and the wording within DG-1361 quotes existing regulations out of context and creates the potential for backlit/forward fit in many instances. There appears to be NO incentive for an operating reactor to change commitments or licensing bases to incorporate either this revision of the IEEE standard or DG-1361 as written.
Recommendation DG-1361 should thus be revised to indicate that it does NOT apply to existing licensees unless a change in a licensing commitment is made.
The staff does not agree with the comment. RG 1.89, Rev. 2 is applicable to currently operating licensees because the guidance could be useful to them to meet regulations that apply to operating power plants. One such example is provided by the commenter: licensees that seek a change to a commitment concerning the subject matter of the RG. Although the RG applies to currently operating licensees, those licensees are not required to use the guidance in the RG, and the RG does not require licensees to take any actions.
The RG has an exclusion for nuclear power plant licensees that have submitted certifications as required by 10 CFR 50.82(a)(1) or 52.110(a) because those licensees' facilities are no longer operating.
Those licensees do not need to meet the requirements for which RG 1.89 provides implementation guidance, so the guidance would not be useful to those licensees.
This proposed guidance does not meet the definition of backfitting or forward fitting in MD 8.4. RG 1.89, Rev. 2 is voluntary guidance
56 Commenter Section of DG-1361 Specific Comments NRC Resolution and represents one acceptable way to satisfy the applicable NRC regulations. The NRC staff is not imposing or expecting specific licensees to implement the guidance in RG 1.89, Rev. 2.
Further, forward fitting occurs only in the context of a licensing action, and the issuance of a RG is not a licensing action.
No changes to DG-1361 were made as a result of these comments.
Vincent Bacanskas Section A Comment 72 DG-1361 fails to recognize the incorporation of 10CFR50.69 since the publication of 50.49, and where 50.49 is quoted for equipment under its scope in the DG. 50.69 specifically excludes EQ for RISC-3 components and the Commission even states that files such as those required for 50.49 are NOT required for RISC-3 components. (69 FR 68008). EQ (50.49) is identified as a special treatment requirement which is not required to provide reasonable confidence of a components capability to provide a low safety significant design function. While 50.69 was published after 50.49, and 50.49 was not revised to address RISC-3 equipment, it should be identified in the Applicable Regulations section of DG-1361.
Recommendation Footnotes should be added to Staff Regulatory Guidance Section C.1.b indicating that RISC-3 components are not included within this position. This footnote should be repeated whenever the staffs regulatory position includes similar wording. It might be most simply addressed in the Introductory Discussion in the DG so that it is understood before looking at staff guidance in the document.
The NRC staff agrees with the comment that additional discussion of 10 CFR 50.69 would be helpful in the RG. The staff added the following information in the Applicable Regulations section of the DG:
10 CFR 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors, states in part that a holder of a license to operate a light water reactor (LWR) nuclear power plant under 10 CFR Part 50; a holder of a renewed LWR license under 10 CFR Part 54; an applicant for a construction permit or operating license under 10 CFR Part 50; or an applicant for a design approval, a combined license, or manufacturing license under 10 CFR Part 52; may voluntarily comply with the requirements in 10 CFR 50.69 as an alternative to compliance with 10 CFR 50.49 for risk-informed safety class (RISC)-3 and RISC-4 SSCs.
57 Commenter Section of DG-1361 Specific Comments NRC Resolution In the Federal Register (FR) notice (69 FR 68008; November 22, 2004) for the final rule establishing 10 CFR 50.69, the Commission stated that RISC-3 (safety-related low safety significant) and RISC-4 (non-safety-related low safety significant) SSCs will be exempt from the special treatment requirements for qualification methods for environmental conditions and effects and seismic conditions.
Nevertheless, the Commission stated that RISC-3 SSCs continue to be required to be capable of performing their safety-related functions under applicable environmental conditions and effects and seismic conditions, albeit at a lower level of confidence as compared to RISC-1 (safety-related safety significant) SSCs. As specified by the Commission in the FR notice, a licensee implementing 10 CFR 50.69 must consider operating life (aging) and combinations of operating life parameters (synergistic effects) in the design of RISC-3 electrical equipment.
The Commission noted that this is particularly important if the equipment contains materials which are known to be susceptible to significant degradation due to thermal, radiation, and/or wear (cyclic) aging including any known synergistic effects that could impair the ability of the equipment to meet its design-basis function. The Commission direction in the FR notice regarding the capability of RISC-3 SSCs to be able to perform their safety functions under applicable environmental and seismic conditions is clear for licensees who
58 Commenter Section of DG-1361 Specific Comments NRC Resolution have received a license amendment to implement a 10 CFR 50.69 program. With respect to both RISC-3 and RISC-4 SSCs, the Commission decided to remove the RISC-3 and RISC-4 SSCs from detailed, specific requirements that provide the high level of assurance. However, the Commission stated in the FR notice that the functional requirements for these SSCs remain.
In addition, following the guidance in this RG is an acceptable approach for showing that RISC-3 and RISC-4 equipment are capable of performing their functions under applicable environmental conditions and effects including seismic conditions.
Vincent Bacanskas Section A Comment 73 Under Applicable Regulations, the statement related to Criterion 3 of 10 CFR Part 50, Appendix B is written in a matter that could be misleading.
Recommendation These criteria should be listed and described separately so that it does not read that testing is only associated with Criterion III Design Control.
Criterion III lists several methods of design verification and all are acceptable to meet Criterion III.
The staff agrees with the comment.
The NRC revised the RG to remove references to testing and to add further clarity due to this comment as follows:
10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, requires, in part, that the pertinent requirements of this appendix apply to all activities affecting the safety-related functions of structures, systems, and components that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public. These activities include designing, purchasing, fabricating, handling,
59 Commenter Section of DG-1361 Specific Comments NRC Resolution shipping, storing, cleaning, erecting, installing, inspecting, testing, operating, maintaining, repairing, refueling, and modifying.
Vincent Bacanskas Section A Comment 74 Under Related Guidance, the NRC staff states: RG 1.40describes a method that the staff of the NRC considers acceptable to implement regulatory requirements with regard to the design, inspection, and testing of normal atmospheric cleanup systems for controlling releases of airborne This is the first example of a perceived BACKFIT contained within this document. A review of the latest RG 1.40 reveals that there is no language within the Regulatory Guide related to normal atmospheric cleanup systems.
Recommendation It appears that the staff is trying to add equipment qualification requirements to another sub-class of systems that do not perform one of the essential functions outlined in 10CFR50.49. This wording should be removed as it is inconsistent with current regulation.
See the staffs response to comment 53.
Vincent Bacanskas Section C.1 Comment 75 In Regulatory Guidance C.1, the staff throws in guidance related to service life, installed life, and qualified life. The staff guidance should only be to not use those definitions in the dual logo standard. These terms have definitions in IEEE STD 323-1971, IEEE STD 323-1974, DOR guidelines, NUREG-0588, and RG 1.89 R0 and R1 which may NOT be consistent with the current language presented. As said in Comment 1 above, licensing bases for plants with operating licenses are relatively fixed and departure to new definitions may not only confuse inspectors, but potentially represent a threat of backfit if interpreted incorrectly from the initial licensing basis.
See the staffs responses to comments 1, 3, and
- 4.
Vincent Bacanskas Section C.1.f Comment 76 Section C.1.f, the staff states in part: "If used, these methodologies must ensure [emphasis added] that equipment important to safety will perform under the conditions specified in 10CFR50.49.. This appears to be another potential for a BACKFIT. The standards of qualification are, and have been, that we are required to provide reasonable assurance that equipment is See the staffs response to comment 6.
60 Commenter Section of DG-1361 Specific Comments NRC Resolution capable of performing its intended safety function when called upon.
Substituting the wording above [ensure] changes the base requirements.
10CFR50, Appendix A (General Design Criteria) make this abundantly clear in its introduction: Under the provisions of § 50.34, an application for a construction permit must include the principal design criteria for a proposed facility. Under the provisions of 10 CFR 52.47, 52.79, 52.137, and 52.157, an application for a design certification, combined license, design approval, or manufacturing license, respectively, must include the principal design criteria for a proposed facility. The principal design criteria establish the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety; that is, structures, systems, and components that provide reasonable assurance [emphasis added]
that the facility can be operated without undue risk to the health and safety of the public.
Vincent Bacanskas Sectdion C.1.h Comment 77 Regulatory Guidance C.1.h discusses establishing the radiation qualification dose but clearly does not include the additional guidance provided in RG 1.89 R1. As this is the only document that it appears in, deleting the guidance would represent a backfit. Specifically, RG 1.89, Regulatory Position C.1.c(6) states: Shielded components need be qualified only to the gamma radiation environment provided that it can be demonstrated that the sensitive portions of the component or equipment are not exposed to significant beta radiation dose rates including heating and secondary radiation, have no deleterious effects on component performance. If, after considering appropriate shielding factors, the total beta radiation dose contribution to which the equipment or component is calculated to be less than 10% of the total gamma radiation dose to which the equipment has been qualified, the equipment or component is considered qualified for the beta and radiation environment. Considering the number of times that this provision was used to justify TID doses, removing this consideration would be a significant backfit for existing licensees. Also, RES performed research related to the actual radiation types, etc., that would be seen while exposed to a Cobalt-60 source. This was See the staffs response to comment 7.
61 Commenter Section of DG-1361 Specific Comments NRC Resolution published in NUREG/CR-5231, Cobalt 60 Simulation of LOCA Radiation Effects.
Recommendation Perhaps review of this document would provide the appropriate insights to be included in the draft RG.
Vincent Bacanskas Section C.1.j Comment 78 Staff Regulatory Guidance C.1.j(1) - 10CFR50.49(e)(7) states: Synergistic effects must be considered when these effects are believed to have a significant effect on equipment performance. NUREG-0588, Part 1 refers one to NUREG/CR-0276 and NUREG/CR-0401 for synergistic effects.
NUREG/CR-0276 says that there were no cable failures during this particular research program and that no significant functional or mechanical synergisms exist (Test Summary Sec. 2.2.4). NUREG/CR-0401 contains essentially the same paragraph from NUREG/CR-0276 and includes a section on ethylene propylene insulation with PVC jacket from Savannah River where they indicate degradation occurred and they expect synergistic effects contributed.
The entire focus on this seems without value as little to no additional research was published on this. Furthermore, to state in RG 1.89 R1 that the test sequence of IEEE 323-1974 shall be followed; then there is a staff expectation that radiation followed by thermal aging is the preferred sequence. This once again, raises regulatory confusion. In essence, the entire staff position on synergism appears to have little scientific basis. While Sandia did document that some materials degraded at different rates with varying radiation dose rates, using the Merriam-Webster online dictionary, this is not a synergistic effect. The presence of this dose rate phenomena is clearly proven and must be a consideration in radiation aging.
Recommendation It would benefit the industry and the NRC staff to remove the position with regard to synergism, state as was included in NUREG-0588 that a simple literature search is sufficient and instead provide a regulatory position on dose rate effects as part of DG-1361. This would be of great benefit to EQ The staff partially agrees with the comment. 10 CFR 50.49(e)(7) states that synergistic effects must be considered when these effects are believed to have a significant effect on equipment performance. The guidance is for situations where there are known synergisms that could affect equipment qualification.
Given the existing guidance, as noted in the comment, and absent any new information, the staff decided the best approach was to carry forward the staff position on synergistic effects from RG 1.89, Rev. 1.
See the staffs response to comment 8 for additional information on the staffs position on synergistic effects.
62 Commenter Section of DG-1361 Specific Comments NRC Resolution programs everywhere and there is nothing in the present research that even implies safety could be deleteriously affected by synergistic effects in EQ testing.
Vincent Bacanskas Section C.1.j Comment 79 Section C.1.j(3) discusses the use of an Activation Energy and imposes specific requirements that are absent from any Regulatory Document.
Specifically, Regulatory Guide 1.89, R1 states in C.5.(c): The aging acceleration rate and activation energies used during qualification testing and the basis upon which the rate and activation energy were established should be defined, justified, and documented. The section in DG-1361 not only adds additional requirements [Backfit] but in some instances is technically incorrect.
Many of the currently used activation energy values were developed by insulation system material research laboratories (non-nuclear) in the 1960s and 1970s. AIEE Guide for the Statistical Analysis of Test Data was published in 1968 and remained in effect until the publication of IEEE 101-1972 IEEE Guide for the Statistical Analysis of Thermal Life Test Data. This IEEE document was published by the IEEE Standards Coordinating Committee on Thermal Rating as a NON-NUCLEAR standard. IEEE 101, while referenced in IEEE 323-1974, was never endorsed by the NRC. IEEE 101-1972 states in part: Procedures for estimating the thermal life of electrical insulation systems and materials call for life tests at several temperatures, USUALLY WELL ABOVE THE EXPECTED NORMAL OPERATING TEMPERATURE. [emphasis added] By the selection of high temperatures for the tests, life of the insulation samples will be terminated, according to some selected failure criterion or criteria, within relatively short times-typically one week to one year. The additional criteria added in the DG contradicts the standards used for the very testing it wishes to backfit.
The paragraphs in IEEE 101-1972 go further to describe the appropriate methods to develop an activation energy. Furthermore, many activation energies were provided to licensees by manufacturers or equipment qualification test laboratories in full qualification reports or studies. The bases See the staffs response to comment 9 regarding Activation energy.
This proposed guidance does not meet the definition of backfitting in MD 8.4. RG 1.89, Rev. 2 is voluntary guidance and represents one acceptable way to satisfy the applicable NRC regulations. The NRC staff is not imposing or expecting specific licensees to implement the guidance in RG 1.89, Rev. 2.
63 Commenter Section of DG-1361 Specific Comments NRC Resolution documents for the activation energies are referenced in the equipment qualification reports. These test reports were furnished to the licensee (in most cases) under an approved Appendix B program and the wording by the staff is attempting to transfer that responsibility to the licensees. Please also refer to 48 FR 2732 regarding the need for a central file and the appropriateness of test laboratory files.
Vincent Bacanskas Section C.1 Comment 80 While the DG provides a statement in the Staff Regulatory Guidance section on applicability to equipment located in a Mild Environment, there should be a more definitive statement as position C.1 that 50.49 explicitly excludes equipment in a Mild Environment and paragraphs associated with this equipment are excluded from this DG.
The staff agrees with the comment that the RG should have a more definitive statement regarding the fact that 10 CFR 50.49 does not require environmental qualification for electrical equipment located in a mild environment. As a result of this comment, the staff added the following in the Background section of the RG:
Requirements for dynamic and seismic qualification of electric equipment important to safety; protection of electric equipment important to safety against other natural phenomena and external events; and environmental qualification of electric equipment important to safety located in a mild environment are not included within the scope of 10 CFR 50.49.
Comment Document 8: ML21110A054 Carrie Fosaaen of NuScale No page number Generic Comment Comment 81 This proposed revision exceeds the scope of the original stated purpose of RG 1.89, which is to ensure equipment important to safety remains functional during and following design basis accidents. Expanding the scope of this RG could lead, as suggested by comments below, to unintended consequences such as back-fitting and forward-fitting implications.
The staff disagrees with the comment. The scope of Revision 2 to RG 1.89 is consistent with the purpose of 10 CFR 50.49. Any additional information incorporated into this RG provides guidance based on almost 40 years of experience since the issuance of the previous version of this RG.
64 Commenter Section of DG-1361 Specific Comments NRC Resolution Recommendation Align RG with the scope and intent of 10 CFR 50.49, and clarify as needed to ensure the revision is within the scope of the stated purpose of 10 CFR 50.49.
Resolutions for specific instances are identified in subsequent comments.
This proposed guidance does not meet the definition of backfitting or forward fitting in MD 8.4. RG 1.89, Rev. 2 is voluntary guidance and represents one acceptable way to satisfy the applicable NRC regulations. The NRC staff is not imposing or expecting specific licensees to implement the guidance in RG 1.89, Rev. 2.
Further, forward fitting occurs only in the context of a licensing action, and the issuance of a RG is not a licensing action.
Carrie Fosaaen of NuScale Pages 1-2, applicable regulations Comment 82 The applicable regulations list is confusing and potentially misleading.
Separating into Part 50 and Part 52 creates the impression the requirements for different applicants/licensees are different. Note that design approval applicants (SDA and DC) need only provide an equipment list per 10 CFR 50.49, while license applicants (OL, COL, and ML) must also describe the EQ program and license holders must establish the program. Furthermore, the introduction to the Part 52 regulations makes reference to design control measures, such as testing which is not supported by a regulation in Part 52.
Recommendation Consolidate the applicable regulations into a single list and clarify applicability for various types of applicants.
The staff partially agrees with the comment.
While some of the criteria apply to both Part 50 and 52 applicants and licensees, the Applicable Regulations sections is meant to delineate the applicable regulations without specifying what requirements each type of applicant or licensee (construction permit, operating license, combined license (COL), manufacturing license, standard design approval, design certification) must meet. Therefore, the staff is maintaining the separate lists of applicable Part 50 regulations and Part 52 regulations.
The staff agrees that 10 CFR 52.47, 52.79 &
52.137 do not discuss design control measures.
The staff revised DG-1361 to remove and that design control measures, such as testing, be used to check the adequacy of the design in the introduction to Part 52 in the Applicable Regulations section.
65 Commenter Section of DG-1361 Specific Comments NRC Resolution Carrie Fosaaen of NuScale Page 2, 3rd bullet from bottom Comment 83 Incorrect citation. Reads in part,... For a manufacturing license as defined in 10 CFR 52.157, only electric equipment defined in 10 CFR 50.49(b) which is within the scope of the manufactured reactor must be included in the EQ program. A manufacturing license is not defined by 10 CFR 52.157, and that reference is unnecessary. 10 CFR 52.157 does require an ML applicant to provide a list of electric equipment important to safety and would be appropriate to address in a separate bullet.
Recommendation Revise to state: For a manufacturing license, only electric equipment defined in 10 CFR 50.49(b) which is within the scope of the manufactured reactor must be included in the EQ program.
The staff partially agrees with the comment.
The staff agrees that it did not correctly state the language related to 10 CFR 52.157 in the DG. The staff does not agree to add the recommended language suggested in the comment. However, the staff revised the Applicable Regulations section to clarify the requirement within 10 CFR 52.157 for an applicant for a manufacturing license to provide a list of electric equipment important to safety that is required by 10 CFR 50.49(d).
Carrie Fosaaen of NuScale Page 4, 3rd bullet from top Comment 84 RG 1.180 is not relevant to environmental qualification under 10 CFR 50.49.
Keep the focus of this RG to satisfying the requirements of 10 CFR 50.49 by removing unrelated RGs that are beyond the scope of 50.49.
Recommendation Remove reference to RG 1.180 and other RGs that are not related to compliance with 10 CFR 50.49.
The NRC partially agrees with the comment.
The NRC agrees that the Regulatory Analysis for DG-1361 referred to providing guidance for demonstrating compliance with 10 CFR 50.49.
However, the Purpose section of DG-1361 stated that the DG describes an approach that is acceptable to the NRC staff to meet regulatory requirements for environmental qualification of certain electric equipment important to safety.
This includes guidance for qualification of equipment in a mild environment even though that is not within the scope of 10 CFR 50.49, as explained in the response to comments 5 and
- 24.
66 Commenter Section of DG-1361 Specific Comments NRC Resolution Carrie Fosaaen of NuScale Page 7, 2nd paragraph from bottom Comment 85 The DG states:
Chapter 11 and Appendix A to the Electric Power Research Institutes (EPRIs) Plant Support Engineering: Nuclear Power Plant Equipment Qualification Reference Manual, Revision 1, issued September 2010 (Ref.
26), provides a detailed regulatory history of electrical and mechanical equipment qualification. While the agency has not officially endorsed this EPRI document, the NRC staff has reviewed Chapter 11 and Appendix A and found that it reflects an accurate representation of the regulatory history of electrical and mechanical equipment qualification.
EPRIs EQ Reference Manual is widely used in the industry. NRCs endorsement of that document would support regulatory efficiency and clarity.
Recommendation Endorse either in part or whole the EPRI EQ Reference Manual (Reference 26). Specific endorsements should include the criteria for determining significant aging mechanisms as found in the EPRI Reference Manual, Page 4-3 and 4-4, first formalized in IEEE 627-1980.
The staff agrees that EPRIs Plant Support Engineering: Nuclear Power Plant Equipment Qualification Reference Manual, Revision 1, is widely used by the nuclear industry.
However, the staff is not prepared to endorse this document beyond what was identified in the DG.
No changes to DG-1361 were made as a result of this comment.
Carrie Fosaaen of NuScale Page 7, 2nd paragraph from bottom Comment 86 In the Background section, the proposed RG states, For the purposes of this guide, the primary objective of qualification is to demonstrate that equipment important to safety can perform its safety function(s) without experiencing common-cause failures before, during, and after applicable design-basis events.
Note that 10 CFR 50.49 is not associated with preventing common cause failures before a design basis accident, and it does not address the environmental conditions of events other than design basis accidents.
Although all design basis events are relevant to the scope of electric equipment addressed (see 10 CFR 50.49(b)(1)), the qualification program required by 10 CFR 50.49 is specific to precluding environmentally-induced The staff agrees with the comment. The staff revised the Background section to remove before and added clarification on design-basis events as defined in 10 CFR 50.49(b)(1)(ii).
67 Commenter Section of DG-1361 Specific Comments NRC Resolution common cause failures during or following exposure to harsh environmental conditions that result from a design basis accident (see 10 CFR 50.49(d)(1) and (e)).
Recommendation Revise to state... common-cause failures during and following applicable design-basis accidents.
Carrie Fosaaen of NuScale Page 8, first paragraph, last sentence Comment 87 In the Background section, the proposed revision to the RG makes the following statement: The qualification specifications in IEC/IEEE 60780-323, Edition 1, 2016-02, when met, demonstrate and document the ability of equipment to perform safety function(s) under applicable service conditions, including design-basis events, reducing the risk of common-cause equipment failure. This statement implies that the NRC is increasing the scope of 10 CFR 50.49 and RG 1.89 to more than design basis accidents, to envelope other applicable service conditions.
While aging is required as part of the EQ program, 10 CFR 50.49 is intended to prevent environmentally-induced common cause failures of electrical equipment important to safety during and following design basis accidents.
Note that 10 CFR 50.49(d) and (e) require qualification to parameters for DBAs and not other DBEs.
Recommendation Revise to state... perform safety function(s) during and following design basis accidents by reducing the risk of common-cause equipment failure.
The staff partially agrees with the comment.
The staff agrees that one of the main purposes of 10 CFR 50.49 is to prevent environmentally induced common cause failures of electrical equipment important to safety. The staff also agrees that 10 CFR 50.49(d) and (e) require qualification to address design basis accidents ;
however, the staff disagrees that it is expanding the scope of 10 CFR 50.49 because 10 CFR 50.49(b) states that electrical equipment important to safety covered by 10 CFR 50.49 is safety-related equipment relied on to remain functional during and following design basis events to ensure: (1) the integrity of the reactor coolant pressure boundary; (2) the capability to shut down the reactor and maintain it in a safe shutdown condition; or (3) the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guidelines in § 50.34(a)(1), § 50.67(b)(2), or § 100.11.
No changes were made to DG-1361 as a result of this comment.
68 Commenter Section of DG-1361 Specific Comments NRC Resolution Carrie Fosaaen of NuScale Comment 88 The proposed revision to the RG states, 10 CFR 50.49 requires safety-related electric equipment (Class 1E) as defined in 10 CFR 50.49(b)(l) to be qualified to perform its intended safety functions.
While correct, this description and other aspects of the proposed revision do not recognize 10 CFR 50.69, which excludes Risk-Informed Safety Class (RISC)-3 components from EQ 10 CFR 50.49 (RISC-3 structures, systems and components (SSCs) means safety-related SSCs that perform low safety significant functions).
Recommendation Revise RG 1.89 to include provisions for licensees conforming to 10 CFR 50.69.
See the staffs response to comment 72.
Carrie Fosaaen of NuScale Page 10, paragraph 1.d, and page 11 carryover Comment 89 Reads in part,... however, the period before the operational phase of the SSC (i.e., shelf life) could also adversely impact the qualified life.
This language implies that the proposed revision to RG 1.89 is invoking shelf life requirements as part of the qualified life determination.
Shelf life is not required per 10 CFR 50.49(e)(5). The commonly applied industry standard is that qualified life starts once the equipment is operational, and does not consider shelf life. Shelf life is managed in accordance with 10 CFR 50 Appendix B, Criterion 13.
Recommendation Revise statements to indicate that shelf life is not to be considered as part of compliance with 10 CFR 50.49.
See the staffs response to comment 4.
Carrie Fosaaen of NuScale Page 11, paragraph e Comment 90 This paragraph recognizes that mild environments are beyond the scope of 10 CFR 50.49, but the proposed provision includes EMC and seismic in the list See the staffs response to comment 24
69 Commenter Section of DG-1361 Specific Comments NRC Resolution of requirements that must be considered. EMC and seismic are not in the scope of 10 CFR 50.49.
Recommendation Revise statements to indicate that EMC and seismic requirements are not required for compliance with 10 CFR 50.49 and are addressed per other regulations.
Carrie Fosaaen of NuScale Page 11, Paragraph g Comment 91 Paragraph g. lists methods acceptable to the NRC staff for calculating and establishing containment pressure and temperature envelopes. Subparagraph (2) states For pressurized water reactors (PWRs) with a dry containment, LOCA or MSLB containment environment should be calculated using CONTEMPT-LT or equivalent industry codes. Identifying a specific code and its equivalent is unduly restrictive. Similar statements are in subparagraphs (3) and (4). Revise the guidance on acceptable codes and replace with statement technology-neutral discussion such as: Containment pressure and temperature environment should be calculated using codes which are consistent with the licensees design and licensing basis.
Recommendation Containment response methodologies are reviewed and accepted by the NRC as part of the application.
The staff partially agrees with the comment.
The staff agrees with the proposal to incorporate the following statement in the RG:
Containment pressure and temperature environment should be calculated using codes that are consistent with the licensees design and licensing basis.
The staff revised DG-1361 to add the above statement as a footnote to each instance where the phrase equivalent industry codes is used in Section C.1.g.
The staff also revised the first bullet under Section C.1.g (NOTE: Section C.1.h in RG 1.89, Rev. 2) to note that Typical methods for calculating mass and energy release rates for LOCAs and MSLBs are referenced in Appendix C to the RG. The staff also revised the title of Appendix C to note that the methods listed are Typical.
The staff agrees with adding the last recommended statement, with minor edits as follows: Containment response methodologies require review and approval by the NRC.
70 Commenter Section of DG-1361 Specific Comments NRC Resolution Carrie Fosaaen of NuScale Page 12, paragraph i Comment 92 Remove this paragraph. Electromagnetic conditions are generally independent of aging and design-basis events. Therefore, qualification can be established on a different sample than the sample subjected to aging and design-basis events. As stated in comment 9, EMC is not in the scope of 10 CFR 50.49. And it is not needed since the dual logo standard already contains this clarification in the note in Section 7.4.1.8.c.
Recommendation Delete this paragraph The staff agrees with the comment. The clarification is not needed, and Section C.2.i from DG-1361 has been deleted. Further, Note 1 within the IEC/IEEE 60780 standard clause 7.4.1.8.c already includes a sentence which states: For convenience, EMI/RFI susceptibility testing and operational test under extreme conditions may be performed on a separate test specimen.
Carrie Fosaaen of NuScale Page 12, paragraph j.(3)
Comment 93 The guidance related to activation energy is restrictive and could result in requiring new tests to derive activation energy values in lieu of use of conservatively-established activation energy values that are justified as being appropriate for the application. If followed as written, this restriction could lead to unnecessary effort to evaluate activation energies without a benefit to safety.
Additionally, the paragraph states, The selected activation energy values should be traceable to a specific test report for which these values were established. Material properties that have historically been used in industry do not all have traceable test reports, because some come from research papers from labs and other facilities.
Recommendation Revise paragraph to allow use of conservatively-established activation energy values that have been found to be acceptable, and to clarify need for traceable test reports.
The staff disagrees with the comment. The intent of the guidance on activation energy was not to require new tests. The staff recognizes that specific data from qualification test reports may be stored elsewhere (e.g., at a vendor facility, in a centralized file, etc.). The point of this particular guidance is that the activation energy values need to be defined, justified, and documented as specified in 10 CFR 50.49 and that the specific material properties need to be considered to determine an accurate activation energy value.
See the staffs response to comment 9 for additional information on the staffs position on activation energy as it pertains to environmental qualification of electrical equipment.
71 Commenter Section of DG-1361 Specific Comments NRC Resolution Carrie Fosaaen of NuScale Page 16, paragraph b Comment 94 The guidance related to commercial grade dedication is out of place in RG 1.89. Supply chain-related activities are beyond the scope of 10 CFR 50.49.
Recommendation Remove guidance related to commercial grade dedication.
The staff agrees with the comment.
The NRC revised the RG to remove references to RG 1.164 due to this comment and removed Section C.2.b from DG-1361.
Carrie Fosaaen of NuScale Page 16, paragraph c Comment 95 This paragraph introduces requirements that are beyond the scope of 10 CFR 50.49. The scope of 50.49 is qualification for electric equipment in a harsh environment, not mild. A mild environment is defined as an environment that would at no time be significantly more severe than the environment that would occur during normal plant operation, including anticipated operational occurrences.
Therefore, by defining a mild radiation environment using a fixed threshold without respect to the normal operating environment of electronic equipment, this paragraph c. conflicts with the terms of 10 CFR 50.49(c) and broadens the scope of EQ. Furthermore, smoke from an electrical fire is not a condition during or following a design basis accident, and therefore not within the scope of the EQ program required by 10 CFR 50.49.
Recommendation Delete paragraph c. as it conflicts with and exceeds the requirements of 10 CFR 50.49.
The staff partially agrees with the comment.
The comment is correct regarding the definition of a mild environment in 10 CFR 50.49(c). However, this does not mean that equipment located in an environment that would at no time be significantly more severe than the environment that would occur during normal plant operation, including anticipated operational occurrences, need not be qualified.
GDC 4 requires that equipment important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents. RG 1.89 uses the term mild radiation environment to define a radiation environment that is below the total integrated dose that would normally result in potential degradation to materials. This definition was included in Revision 3 of NUREG-0800, Section 3.11 and was documented in NUREG-1503, Final SER ABWR, Chapter 3, Design of Structures, Components, Equipment, and Systems, and NUREG-1793, Final SER AP1000, Chapter 3, Design of Structures, Components, Equipment, and Systems. RG 1.89 uses the dose criteria
72 Commenter Section of DG-1361 Specific Comments NRC Resolution to specify when the radiation is such that environmental degradation may be a concern.
The staff did not revise DG-1361 regarding aspects of radiation qualification. However, the DG was revised concerning the need to address stressors (like smoke) that are not covered within 10 CFR 50.49 requirements.
The document now makes it clear that qualification for possible exposure to smoke is not covered within 10 CFR 50.49, and that guidance for addressing additional stressors like smoke are contained within Regulatory Guide 1.209. Section C.2.c was modified to include a pointer to Regulatory Guide 1.209 for guidance in addressing the effects of smoke.
NOTE, this position is now Section C.2.b. of RG 1.89, Rev. 2.
See the staffs responses to comments 14 and 25 for additional information on qualification of equipment in a mild environment.
Carrie Fosaaen of NuScale Page 16, paragraph d Comment 96 The draft guide introduces a new term, inverse temperature, without explaining what it is or how to comply. Inverse Temperature Effects is first published in NUREG/CR-7153, Volume 5. This publication was focused on cables and cable systems only. Therefore, it is not established whether inverse temperature effects does or does not impact other broad types of elastomers. There is no existing NRC guidance on how to address inverse temperature effects. If Staff intend to address inverse temperature effects, it should be fully and transparently considered by Staff in an appropriate regulatory action. The mention of inverse temperature effects in DG-1361 implies additional requirements outside the normal regulatory process.
The staff partially agrees with the comment.
NUREG/CR 7153, Expanded Materials Degradation Assessment (EMDA): Volume 5:
Aging of Cables and Cable Systems, Section 3.3, Inverse Temperature, states that The observed inverse temperature effect, where polymer degradation occurs more rapidly for constant dose rates as the combined environment temperature is lowered, represents an example in which material aging and
73 Commenter Section of DG-1361 Specific Comments NRC Resolution Additionally, Item (2) of paragraph d. states,... concurrent radiation and thermal aging or sequential aging, as well as the order of radiation and thermal aging, based on which produces the worst-case degradation; and...
The phrase worst-case is ambiguous and inconsistent with requirements of 10 CFR 50.49(e)(5) for aging to end-of-installed-life condition. This requirement was not present in RG 1.89 Rev 1. This statement may create unintended back-fit/forward-fit consequences.
Recommendation Remove discussion about inverse temperature effects.
Remove item (2).
lifetime prediction cannot be represented adequately by conventional approaches, such as the Arrhenius methodology. The inverse temperature effect is applicable to certain XLPO and EPR insulation materials. The staff added the following definition on inverse temperature to the Background section of RG 1.89, Rev. 2:
Inverse temperature or reverse temperature effect is where polymer degradation occurs more rapidly for constant dose rates as the combined environment temperature is lowered.
NUREG/CR 7153, Expanded Materials Degradation Assessment (EMDA): Volume 5:
Aging of Cables and Cable Systems, Section 3.3, Inverse Temperature provides additional information.
Based on several similar comments, the staff determined that the information in Section C.2.d was less of a staff position and more of general information that would be better suited in the Background section of the revised regulatory guide. Therefore, the staff modified the regulatory guide to revise and relocate this information to the Background section.
The staff is not imposing additional requirements in the Background section. The information is provided for consideration based on experience and research (EMDA and references listed in it).
74 Commenter Section of DG-1361 Specific Comments NRC Resolution Also, in response to this comment, the staff added the following text to the Background section of DG-1361:
Diffusion-limited oxidation, synergisms, dose-rate effects, and inverse temperature effects are examples of such effects.
Experience suggests that consideration should be given, for example, to the following:
The staff also replaced the worst-case degradation with more severe degradation in the Background section in response to this comment.
Carrie Fosaaen of NuScale Page 20, item 26 Comment 97 Incorrect citation. Reads in part... Electric Power Research Institute, (EPRI)
Nuclear Energy Institute (NEI) EPRI/NEI Report No. 1021067.. This is an EPRI report only.
Recommendation Delete NEI from report title The staff agrees with the editorial comment and removed the incorrect reference to the Nuclear Energy Institute in association with Report No. 1021067 in the list of References in DG-1361.
Carrie Fosaaen of NuScale Appendix A
Comment 98 The list of typical safety-related electric equipment or systems is typical only for operating power reactors (large, non-passive LWRs). Additionally for such designs, containment combustible gas control is no longer required to be safety-related function (pursuant to revised 10 CFR 50.44). Emergency systems to achieve safe shutdown is also unclear, because beyond those systems already listed, it would appear to only include the system for residual heat removal.
Recommendation Clarify list is not applicable for passive designs and other new technologies, and revise combustible gas control and safe shutdown bullets.
The staff partially agrees with the comment.
Footnote 9, Appendix A, of DG-1361 states:
In Title 10 of the Code of Federal Regulations (10 CFR) section 50.49(b)(1), the NRC identifies safety-related electric equipment as a subset of electric equipment important to safety and defines it as the equipment that is relied upon to remain functional during and following design-basis events to ensure (1) the integrity of the reactor coolant pressure boundary, (2) the capability to shut down the reactor and
75 Commenter Section of DG-1361 Specific Comments NRC Resolution maintain it in a safe-shutdown condition, or (3) the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guidelines in 10 CFR Part 100, Reactor Site Criteria.
Based on the footnote 9, any equipment or systems not under the category of 10 CFR 50.49(b)(1) is non-safety related. Non-safety related equipment or systems do not belong to the list in Appendix A.
For some designs, combustible gas control does not have a safety related function. Based on footnote 9 in Appendix A, combustible gas control systems do not belong to the list in Appendix A.
Emergency systems to achieve safe shutdown are design specific. They can cover all systems to achieve safe shutdown, not just residual heat removal systems. This term corresponds to 10 CFR 50.49(b)(1).
The list in Appendix A, Emergency systems to achieve safe shutdown, is general for a purpose to include all designs (large, non-passive, passive, small module, or advanced).
Based on this comment, the staff revised Appendix A to state For large light-water reactors, the following are typical safety-related electric equipment or systems. Some
76 Commenter Section of DG-1361 Specific Comments NRC Resolution items on this list may not be applicable to passive designs, small modular designs, or advanced reactors. However, emergency systems to achieve safe shutdown could be safety-related electric systems for any design.
The staff also revised Appendix A to delete the bullet emergency systems to achieve safe shutdown since it is ambiguous and includes systems listed above.
Carrie Fosaaen of NuScale Appendix B, Page B-1 Comment 99 This introduction to this appendix provides a confusing description for its regulatory basis, suggesting the example equipment is explicitly within the scope of 10 CFR 50.49. Rather, these are examples of non-safety-related electric equipment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of the specified safety functions, pursuant to 10 CFR 50.49(b)(2).
Additionally, this version of the proposed RG removed the following language from the previous version:
Associated circuits, as defined in Regulatory Guide 1.75, "Physical Independence of Electric Systems," need only be qualified to ensure that they will not fail under postulated environmental conditions in a manner that could prevent satisfactory accomplishment of safety functions by safety-related equipment.
This statement accurately describes the purpose and acceptance criterion for the qualification of the non-safety-related electric equipment within the scope of 10 CFR 50.49(b)(2).
Recommendation The staff partially agrees with the comment.
As Appendix B to DG-1361 is carried forward from the previous RG revision, the staff disagrees that the regulatory basis description is inaccurate or confusing. Nevertheless, the staff updated the appendix to include the information suggested by the comment from the previous revision of RG 1.89. This information was unintentionally omitted.
The following paragraph was added to Appendix B:
Associated circuits, as defined in Regulatory Guide 1.75, Physical Independence of Electric Systems, need only be qualified to ensure that they will not fail under postulated environmental conditions in a manner that could prevent satisfactory accomplishment of safety functions by safety-related equipment.
77 Commenter Section of DG-1361 Specific Comments NRC Resolution Clarify the basis and intent of Appendix B. Re-insert deleted provision in RG 1.89 Rev 2.
Carrie Fosaaen of NuScale Appendix C, Page C-1 Comment 100 The list of acceptable methods for calculating mass and energy releases could be misconstrued as limiting for applicants.
Mass and energy release methodologies are reviewed as part of plant licensing. This Appendix should be generically applicable to all reactor designs. For example, NuScales use of NRELAP5 code was found acceptable by the staff. If a list of currently acceptable methods is to be included, it should be clarified as only examples of methods the Staff have previously evaluated for existing designs.
Also, there is no mention of other design basis accidents that may require methods for calculating mass and energy release.
Recommendation Include a generic position in Appendix C that mass and energy releases are developed using a methodology that is consistent with the licensing basis of the plant. Clarify that the list of existing methods are examples for existing designs and not intended to restrict future methods.
See the staffs response to comment 91.
In addition, the staff revised DG-1361 to include a footnote to Appendix C that states:
Mass and energy releases are developed using a methodology that is consistent with the licensing basis of the plant. The listed methods are examples for existing designs.
Carrie Fosaaen of NuScale Appendix D, Generic Comment Comment 101 Appendix D is dedicated to RG 1.183 which does not apply to all facilities.
The DG makes no mention of source term for non-AST source term for EQ and is silent on the guidance in RG 1.195.
Recommendation Revise Appendix D to address both existing and new designs.
The staff partially agrees with the comment.
The staff intended to make it clear that other source term methodologies may be used, as appropriate, for the radiation accident EQ dose.
This includes RG 1.195. Appendix D has been revised to make it clearer that other approved source term methodologies may be used.
The staff does not intend to reference all acceptable accident source term methodologies and guidance documents that have been used or
78 Commenter Section of DG-1361 Specific Comments NRC Resolution may be used in the future for evaluating accident EQ doses.
In addition, the description of RG 1.183 in the Related Guidance section has been updated, as follows:
RG 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors (Ref. 13), provides guidance to licensees of operating power reactors on acceptable applications of alternative source terms (ASTs); the scope, nature, and documentation of associated analyses and evaluations; consideration of impacts on analyzed risk; and content of submittals related to the use of ASTs in radiological consequence analyses at operating power reactors. RG 1.183 can be used in radiological accident analysis and provides acceptable accident source term methodologies that may be used for EQ, as applicable.
Therefore, for those applicants and licensees that RG 1.183 is applicable, RG 1.183 is referenced in this guide to describe acceptable source term methodologies to be used for EQ.
However, RG 1.183 is not the only approved methodology for accident source terms and additional source term methodologies may be approved in the future. While other accident source term methodologies are not specifically referenced in this guide, approved accident source term methodologies for EQ may continue to be used (provided that they remain
79 Commenter Section of DG-1361 Specific Comments NRC Resolution applicable) and new methodologies may be considered by the staff. The source term methodologies used must be applicable to the specific applicant or licensee and adequate to address EQ requirements.
Carrie Fosaaen of NuScale Appendix D, Page D-1, paragraph D-2.1 Comment 102 This discussion refers to the survivability period. Equipment survivability has a defined meaning with respect to beyond design basis events (see 10 CFR 50.44 and RG 1.7), so the term may introduce confusion within the context of 10 CFR 50.49 compliance.
Recommendation Revise to use a different term, such as post-accident operating time or mission time.
The staff agrees with the comment.
The second paragraph in Section D-2.1 has been revised to remove the term survivability period and the sentence is reworded as, The period of exposure should be consistent with the design basis event qualification for the EQ equipment being evaluated. In addition, the staff removed the third sentence, which defined survivability period.
Carrie Fosaaen of NuScale General Comment 103 Appendix E was eliminated from this proposed version of RG 1.89. Appendix E contains valuable qualification documentation requirements and categories for equipment. See RG 1.89 Rev 1, Page 1.89-17.
Recommendation:
Restore Appendix E to RG 1.89 The staff agrees with the comment. The staff will include Appendix E from RG 1.89, Rev. 1 in RG 1.89, Rev. 2 since it provides useful information on qualification documentation for equipment.
In addition, the staff edited the guidance to include the following: Section 8 of IEC/IEEE Std. 60780-323, Edition 1, 2016-02, provides guidance on documentation. Additional documentation guidance can be found in Appendix E of this Regulatory Guide. NOTE:
this change is now in Section C.1.n of RG 1.89, Rev. 2.
Carrie Fosaaen of NuScale General Comment 104 Neither the joint logo nor the proposed RG revision addresses that, if a licensee or entity has previously met provisions of IEEE 323-1974, whether the joint logo and RG would accept the previous testing per IEEE 323-1974 See the staffs response to comment 58.
80 Commenter Section of DG-1361 Specific Comments NRC Resolution as being equivalent from an environmental qualification perspective. Without these endorsements it would be incumbent upon an entity to reconcile the differences each time.
This reconciliation would add burden and cost to not only the entities using the IEEE 323-1974 versions, but also to the staff during inspections and other activities. It might even require additional testing, which may lead to forward-fit implications. Further, if a component is replaced with a component that is tested to the joint logo standard and an IEEE 323-1974 applicant wants to use it, it is uncertain if this would be allowed without reconciliations.
Recommendation Clarify such that the joint logo and RG apply to both scenarios: RG 1.89 to endorse the use of IEEE 323-1974 as acceptable to meeting the joint logo, and RG 1.89 to endorse the joint logo as meeting the requirements for IEEE 323-1974.
Carrie Fosaaen of NuScale General Comment 105 A design specific review standard was issued to NuScale for the DCA application. This DSRS 3.11 included additional guidance that was not in RG 1.89.
Recommendation Conduct a reconciliation between the DSRS 3.11 and RG 1.89 to ensure the RG encompasses updated requirements.
The staff partially agrees with this comment.
The staff reviewed the NuScale design specific review standard (DSRS) 3.11 (ADAMS Accession No. ML15355A455) to ensure generic aspects were addressed in the proposed revision to RG 1.89. The NuScale DSRS 3.11 is applicable only to the NuScale design, and all facets of the DSRS 3.11 may not apply to all designs.
Nonetheless, the staff revised RG 1.89 to add In SRM-SECY-05-0197, Review of Operational Programs in a Combined License Application and General Emergency Planning Inspections, Tests, Analyses, and Acceptance Criteria, the NRC staff describes operational programs for new nuclear power plants as
81 Commenter Section of DG-1361 Specific Comments NRC Resolution programs that are required by regulation, are reviewed by the NRC staff for acceptability with the results documented in the safety evaluation report (SER), and will be verified for implementation by NRC inspectors. For example, SECY-05-0197 specifies the EQ program as an operational program.
Furthermore, the NRCs 10 CFR Part 52 regulations already require consideration of risk and as such the staff did not find it necessary to include additional information on risk in RG 1.89, Rev. 2. The NuScale DSRS 3.11 considers the application of risk insights.
Carrie Fosaaen of NuScale General Comment 106 10 CFR 50.49 footnote 4 refers to RG 1.97 Rev. 2. In BTP 7-10 the staff provided guidance for RG 1.97 Rev. 3 and 4 as it relates to intent of 10 CFR 50.49.
Recommendation Revise BTP 7-10 to update 1.97 revision and provide staff interpretation for meeting 10 CFR 50.49. Include correct revision of RG 1.97, which is Rev 5.
The staff recognizes that there is a reference to a previous revision of RG 1.97 in the footnote for 10 CFR 50.49. A separate work task is underway to revise and update NUREG-0800, Chapter 7 in its entirety, which includes, among other things, BTP 7-10. Included in this activity is the task to review all references and ensure the most current applicable reference is addressed in the updated Standard Review Plan. Revision 5 of Reg Guide 1.97 incorporated the 2016 version of IEEE-497, which included criteria for addressing Type F variables, which are used for monitoring beyond design basis events with resulting fuel damage. The approach taken to address beyond design basis event instrumentation reliability does not require the qualification process described within 10 CFR 50.49 requirements. Instead, such devices must be
82 Commenter Section of DG-1361 Specific Comments NRC Resolution demonstrated to be available and reliable to support their intended functions when needed through design, analysis, and testing.
The staff notes that updating the reference to RG 1.97 Rev. 2 in 10 CFR 50.49 is outside the scope and intent of DG-1361. All applicants and licensees are encouraged to use the latest revisions of RGs for the most up to date guidance from the NRC, if appropriate for their particular needs or as required by their individual licensing basis.
No changes to DG-1361 were made as a result of this comment.
Carrie Fosaaen of NuScale General Comment 107 10 CFR Part 52 licensees must address ITAAC. The DG does not speak to ITAAC-related EQ.
Recommendation Revise RG 1.89 to reference RG 1.215 for EQ-related ITAAC closures.
The staff agrees with the comment.
The DG does not address ITAAC related to EQ, and Part 52 licensees must address ITAAC to obtain a 10 CFR 52.103(g) finding and load fuel.
The staff revised the Applicable Regulations section to add 10 CFR 52.99, Inspection during construction; ITAAC schedules and notifications; NRC notices, and 10 CFR 52.97, Issuance of combined licenses. As required by 10 CFR 52.99, licensees must notify the NRC that the prescribed inspections, tests, and analyses have been performed and that the prescribed acceptance criteria are met for each ITAAC included in their COL. In addition, as required by 10 CFR 52.97,COLs
83 Commenter Section of DG-1361 Specific Comments NRC Resolution must contain ITAAC that are necessary and sufficient to provide reasonable assurance that the facility has been constructed and will be operated in accordance with the license; the Atomic Energy Act of 1954, as amended; and NRC rules and regulations. For those facilities that are required to have an EQ program, the aforementioned regulations could necessitate EQ-related ITAACs.
In addition, the staff added RG 1.215, Guidance for ITAAC Closure Under 10 CFR Part 52, to the Related Guidance section. RG 1.215 provides guidance on documenting the completion of ITAAC for the implementation of 10 CFR 52.99 and is applicable to EQ-related ITAAC.
Comment Document 9: ML21113A276 Carrie Fosaaen of NuScale General Comment-108 - Summarized 60 Day comment extension and public meeting request See the staffs response to comment 56.
Comment Document 10: ML21042A003 Rick Weinacht of Curtiss-Wright Comment-109 - Summarized 60 Day comment extension and public meeting request See the staffs response to comment 56.
Comment Document 11: ML21050A360 William Horin of NUGEQ General Comment 110 - Summarized This Comment Document was superseded by Comment Document 12 below (see ML ML23086C099).
See the staffs responses to Comment Document 12 below.
Comment Document 12: ML21110A056
84 Commenter Section of DG-1361 Specific Comments NRC Resolution William Horin of NUGEQ General Comment 111 Consistency with the stated objective in Section 2 of the Regulatory Analysis for DG-1361, which states:
The objective of this regulatory action is to access the need to revise NRC guidance and provide applicants with an updated method to demonstrate compliance with 10 CFR 50.49, Environmental qualifications of electric equipment important to safety for nuclear power plants.
Keeping RG 1.89 specific to an acceptable method of meeting 10 CFR 50.49 also results in consistency with RG 1.209, Guidelines for Environmental Qualification of Safety-Related Computer-Based Instrumentation and Control Systems.
There are several examples of where DG-1361 goes beyond the scope of providing another acceptable method of complying with 10 CFR 50.49. For specific examples, see Comments 1.2 through 1.5.
The staff disagrees with the comment. RG 1.89, Rev. 2 provides a means of satisfying 10 CFR 50.49 and other regulations associated with the environmental qualification of equipment as stated in the Applicable Regulations section of the RG. The staff added the following sentence in the Purpose section of RG 1.89, Rev. 2 to clarify that the RG provides guidance for satisfying design criteria identified in Appendix A to 10 CFR Part 50, such as GDC 4: This RG also provides guidance for addressing environmental stressors affecting the long-term reliability of electric equipment. See the staffs responses to comments 112 and 114 for more information on this addition.
The staff included references to other RGs to inform the reader of other resources available when qualifying equipment. Some of the RGs are referenced because they are helpful to address regulatory requirements related to environmental qualification. The staff revised the Related Guidance section to state, The following documents facilitate qualification under other requirements, include additional information for qualifying specific equipment, or provide an additional level of detail for qualifying equipment.
William Horin of NUGEQ Section C.1.e / p11 Comment 112 Any guidance related to the content of design / qualification / procurement specifications should not be construed as being limited to environmental qualification under 50.49.
The staff partially agrees with the comment to the extent that the implementation of the guidance for addressing qualification criteria that are not within the scope of 10 CFR 50.49
85 Commenter Section of DG-1361 Specific Comments NRC Resolution The intent of the clarification is also confusing since the same rational, used for excluding aging of mild environment equipment, is true with respect to EMC and seismic requirements (which are also not within the scope of 50.49).
The staff position in C.1.e makes a clarification related to Section 5.1 of IEC/IEEE Std. 60780-323 by removing the prerequisite for aging for electric equipment located in mild environment since this equipment is not within the scope of 50.49. This clarification appears to be inappropriate since design and procurement specifications include requirements related to equipment qualification, which are not limited to environmental qualification requirement related to 50.49 compliance.
Recommendation Proposed Change: Delete Position C.1.e is not complete. But rather than deleting Section C.1.e as suggested by the commenter, the staff has modified its position as follows::
This RG also provides guidance for addressing environmental stressors affecting the long-term reliability of electric equipment.
Section C.1.e has also been modified to clarify that aging for mild environment equipment is not considered under 10 CFR 50.49, but rather under General Design Criterion 4 of 10 CFR 50 Appendix A.
See the staffs response to comment 5 for additional information.
William Horin of NUGEQ Section C.2.c / p16 Comment 113 There is no need to include or address smoke as fire is not an event that is covered by environmental qualification under 50.49. Smoke effects are adequately addressed in RG 1.209 as well as Appendix R.
The staff position in C.2.c brings up smoke exposure from a fire as an additional stressor to be considered in the qualification of digital systems.
Smoke exposure from a fire is not a condition addressed or required by 10CFR50.49 and results from an event other than a design basis accident.
Recommendation Proposed Change: Remove discussion in C.2.c regarding the consideration of smoke in the qualification of digital systems.
See the staffs response to comment 14.
William Horin of NUGEQ Section C.1.i / p11 Comment 114 This is another example of where the DG is providing guidance that is outside the scope of 10CFR50.49. EMI/RFI or electrical power surges are not identified in 10CFR50.49. As indicated in the C.1.i, EMC and electrical power surges are independent from DBEs and DBAs.
The staff agrees with the comment that guidance on EMC stressors does not fall under 10 CFR 50.49. Section C.1.e was revised to provide clarification of this, as well as to state that guidance for addressing EMC and
86 Commenter Section of DG-1361 Specific Comments NRC Resolution There is currently sufficient guidance to address EMC in RG 1.180, Guidelines for Evaluating Electromagnetic and Radio Frequency Interference in Safety-Related Instrumentation and Control Systems.
The supplemental guidance in C.1.i is clarifying that the testing for EMC can be separate from EQ testing. This clarification is not necessary since the dual logo standard already reflects this (See Note 1 in Section 7.4.1.8.c on page 22 of the dual logo standard).
Recommendation Proposed Change: Remove text.
EMI/RFI may be found in Regulatory Guide 1.180. In addition to endorsing IEC/IEEE 60780-323-2016 with clarifications, the staff intended to include supplemental guidance for addressing environmental stressors of all types.
The Purpose section has been modified to include the following statement:
This RG also provides guidance for addressing environmental stressors affecting the long-term reliability of electric equipment.
William Horin of NUGEQ Section B,
Background
/ p7 Comment 115 The rewording of the primary objective of qualification represents an example of expanding its scope beyond an acceptable method to meet 10CFR50.49.
Several parts of DG-1361 use the term design-basis events in place of design-basis accidents. The use of design-basis events is inconsistent with, and represents an expansion from, the design basis accident parameters specified in 10CFR50.49 (d)(1), (d)(2), (d)(3) and I that need to be considered when establishing environmental qualification.
10CFR50.49 is not associated with the prevention of environmentally induced common-cause failures prior to a design basis event.
The terminology of.before, during and after. is also inconsistent with Regulatory Position C.1.c. It also reflects a change from Section B of RG 1.89 R1 as well as Section B of RG 1.209. The wording in the Background section of DG-1361 is misleading and could be interpreted as expanding beyond the scope of 10CFR50.49. Section B of DG-1361 reflects a change from Part B of RG 1.89 R1 which reads; For the purpose of this guide, qualification is a verification of design limited to demonstrating that the electric equipment is capable of performing its safety function under significant environmental stresses resulting from design basis accidents in order to avoid common cause failures. (Emphasis added)
The staff partially agrees with the comment.
The staff disagrees that the proposed RG represents an expansion of scope because 10 CFR 50.49(b) states that electric equipment important to safety covered by 10 CFR 50.49 is safety-related equipment relied on to remain functional during and following design basis events to ensure: (1) the integrity of the reactor coolant pressure boundary; (2) the capability to shut down the reactor and maintain it in a safe shutdown condition; or (3) the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guidelines in § 50.34(a)(1), § 50.67(b)(2), or § 100.11.
The staff agrees that 10 CFR 50.49(d) and (e) require qualification to address design basis accidents.
87 Commenter Section of DG-1361 Specific Comments NRC Resolution The revised wording in DG-1361 currently reads; For the purposes of this guide, the primary objective of qualification is to demonstrate that equipment important to safety can perform its safety function(s) without experiencing common-cause failures before, during and after applicable design-basis events. (Emphasis added)
The comments on before, during, and after are addressed in the staff responses to comments 86 and 87. No other changes to the RG were made due to this comment.
William Horin of NUGEQ Section A, Related Guidance /
p4 Comment 116 IEEE Std. 323-1974 as modified by RG 1.89R1 also provides an equally acceptable approach for establishing EQ of electrical equipment in accordance with 10CFR50.49.
The wording of provide the fundamental approach should be reworded to clarify that this is one acceptable approach, but not the only acceptable standard to use in establishing EQ in accordance with 10 CFR 50.49.
Recommendation Proposed Rewording: Change provide the fundamental approach to provides one acceptable approach so that it is clear that qualification to standards such as IEEE 323-74 (as endorsed by RG 1.89 R1) or any applicable daughter standards remains an acceptable approach.
The staff partially agrees with the comment.
The staff removed the sentence that contained provides the fundamental approach from the RG in acknowledgement that there could be alternative approaches to establishing EQ of electrical equipment. However, no additional suggested changes were made as a result of this comment.
William Horin of NUGEQ Section C.2.b / p16 Comment 117 The guidance in Section C.2.b appears to be broadening the scope of RG 1.89 to overlap with RG 1.164. The first sentence in C.2.b indicates that no significant changes in form, fit or function should have occurred since the performance of the original qualification testing. The staff position that there has been no significant change to the item being procured since its original qualification seems related to maintaining test report applicability for like-for-like replacements. The staff position then goes on to state that since visual examinations or material-type verifications alone may not be sufficient to determine whether significant changes have not occurred, a combination of material testing along with partial requalification testing of the components may be necessary. As worded, this seems to infer that some level of requalification is warranted even for like-for-like replacements in order to establish the basis for test report applicability. In effect, this appears to be See response to comment 94
88 Commenter Section of DG-1361 Specific Comments NRC Resolution treating the dedication of a like-for-like replacement items in a manner that more closely resembles dedication for equivalent or alternative replacements. Both of these procurement scenarios (like-for-like and equivalent) are currently addressed in Section B.3.2 of EPRI Report 3002002982, which is endorsed by RG 1.164.
This is an example of a new or expanded regulatory position that appears to be inconsistent with RG 1.164.
William Horin of NUGEQ Comment 118 Comments 2.2 - 2.18 are provided in response to the specific request in the FRN for DG-1361 to identify any concerns related to backfitting or forward fitting.
Section IV of FRN / Vol. 85, No. 243 / December 17, 2020.
DG-1361 contains multiple examples of regulatory positions that differ from RG 1.89, R1 positions on the same topic (in the direction of being more restrictive), and even some examples that simply differ from the language of 10 CFR 50.49. In either instance, NRC processes and procedures dictate an issue-by-issue evaluation of such differences for backfit or forward fit implications.
Section D, Implementation of DG-1361 does not identify or address any new or changed staff positions. Our review of DG-1361 has identified multiple examples of new or revised staff positions are described in the following comments.
The staff disagrees with the comment.
This proposed guidance does not meet the definition of backfitting or forward fitting in MD 8.4. RG 1.89, Rev. 2 is voluntary guidance and represents one acceptable way to satisfy the applicable NRC regulations. The NRC staff is not imposing or expecting specific licensees to implement the guidance in RG 1.89, Rev. 2.
Further, forward fitting occurs only in the context of a licensing action, and the issuance of a RG is not a licensing action.
No changes were made to DG-1361 as a result of this comment.
William Horin of NUGEQ General Comment 119 The RG in its introduction should clearly repeat 10CFR50.49 (k):
(k) Applicants for and holders of operating licenses are not required to requalify electric equipment important to safety in accordance with the provisions of this section if the Commission has previously required qualification of that equipment in accordance with "Guidelines for Evaluating Environmental Qualification of Class 1E Electrical Equipment in Operating Reactors," November 1979 (DOR Guidelines), or NUREG-0588 (For Comment version), "Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment."
The NRC staff partially agrees with the comment. The staff agrees that the NRC has previously found existing licensees' programs to be in compliance with 10 CFR 50.49, existing licensees are not required to requalify electric equipment important to safety under 10 CFR 50.49(k), and DG-1361 presents positions different from positions in RG 1.89, Rev. 1 that could have backfitting implications if used to interpret current requirements in 10 CFR 50.49
89 Commenter Section of DG-1361 Specific Comments NRC Resolution The NRC staff has previously found the existing Licensees programs in compliance with the regulation, yet this DG presents many potential backfit positions if used to interpret current requirements as opposed to clearly limiting guidance to new reactors.
in a way that would require existing licensees to requalify electric equipment important to safety.
However, as explained in Section D, Implementation, of DG-1361, the staff does not intend to use the guidance in DG-1361 to support NRC staff actions in a manner that would constitute backfitting or forward fitting, and if a licensee believes that the NRC is using this guidance in a manner that constitutes backfitting or forward fitting, then the licensee can file a backfit or forward fit appeal.
The guidance in RG 1.89, Rev. 2 is not limited to new reactors. Existing reactor licensees can use this guidance if they choose to do so. The staff also finds it unnecessary to repeat entire portions of 10 CFR 50.49 in the proposed RG.
No changes to DG-1361 were made as a result of this comment.
William Horin of NUGEQ General Comment 120 To clarify what would be needed for a licensee, who is committed to IEEE Std 323-1974, to accept or install a component that has been qualified in accordance with the dual logo standard.
Recommendation Please add a statement that clarifies whether equipment qualification testing to the standard endorsed under this RG meets the requirements of IEEE Std 323-1974 and satisfies the requirements for compliance with a prior license basis. Also see comment 3.2.
See the staffs response to comment 58.
90 Commenter Section of DG-1361 Specific Comments NRC Resolution William Horin of NUGEQ Section C.2.d (2) /
p16 Comment 121 10 CFR 50.49(e)(5) does not require the aging in a test program to produce a worst-case state of degradation. As noted in Comment 2.12, the staffs research into simultaneous vs. sequential test sequence exposures did not identify a significant effect on performance.
This is a new regulatory position that is not included in RG 1.89 R1. The existing regulatory requirement in 10CFR50.49(e)(5) is for accelerated aging during a qualification test program to simulate an end-of-installed life condition. The expectation to have an accelerated aging program to produce the worst-case or most-severe degradation is a new regulatory position that extends beyond 10 CFR 50.49.
Recommendation Proposed Change: Remove C.2.d (2).
Also See Comment 2.19.
See the staffs responses to comments 67 and 96 for additional details.
William Horin of NUGEQ Section C.1.c / p10 Comment 122 The change in definition for qualified life should clearly reflect that the need to establish a qualified life is specifically limited to equipment subject to harsh environment qualification under 50.49. As noted in RG 1.209, the need to establish a qualified life for mild environment equipment does not apply.
The proposed change in the definition of qualified life and the use of design basis events would infer qualified life extends to equipment that is relied upon for events not addressed by 50.49 EQ programs.
Also see Comment 2.6 The staff partially agrees with the comment.
See the staffs response to comment 3 for details on the staffs proposed definition for qualified life.
The staff added a regulatory position in the RG to clarify that the staff recognizes that environmental qualification of electric equipment located in a mild environment is beyond the scope of 10 CFR 50.49. The staff added the following to Section C.1.a:
Guidance endorsed in this RG could also be used to satisfy GDC 4 requirements for the design of structures, systems, and components important to safety to accommodate the effects of and to be compatible with the environmental
91 Commenter Section of DG-1361 Specific Comments NRC Resolution conditions associated with normal operation, maintenance, testing, and postulated accidents, including LOCAs.
William Horin of NUGEQ Section C.1.c / p10-11 Comment 123 By being specific to 50.49 (b)(1), the proposed definition of qualified life could be misinterpreted as exempting EQ equipment that is qualified based on 50.49 (b)(2) or (b)(3) functions from having a qualified life.
The definition of qualified life, which is specific to 50.49 should be specific to design basis accidents, since design basis events includes scenarios which are excluded pI50.49 (c).
Since the qualified life does not (in and by itself) ensure the performance of a safety function under harsh DBA conditions, it is suggested that the basis be tI to 50.49(e)(5) in lieu of 50.49(b)(1).
Suggest design basis events be changed to design basis accidents to be consistent with the requirements in 50.49 (d) and (e). Also See comment 2.5.
The staff partially agrees with the comment.
The staff revised DG-1361 to eliminate the potential confusion with the definition of qualified life to include a reference to 10 CFR 50.49(b) instead of limiting to 10 CFR 50.49(b)(1).
See the staffs responses to comments 3, 21, 80, and 86 for details on the staffs resolution to other aspects of this comment.
William Horin of NUGEQ Section C.1.d / p10-11 Comment 124 The proposed wording in the note inappropriately blends service life and qualified life. These definitions have distinct and separate meanings with qualified life being specific to 50.49 qualification. The need to consider degradation prior to installation in establishing qualified life is a new regulatory position used on 50.49(e)(5) which states; Equipment qualified by test must be preconditioned by natural or artificial (accelerated) aging to its end-of-installed life condition. Consideration must be given to all significant types of degradation which can have an effect on the functional capability of the equipment. Application of this regulatory interpretation of 50.49 is incorrect from a regulatory perspective, whether considered in current licensing bases or future licensing bases relying on 50.49.
There is no apparent need to supplement the definition of service life in the dual logo standard. The addition of the note comingles the definition of service life and qualified life. The added note is also inconsistent with the following unmodified definition of service life; period from initial operation See the staffs response to comment 4.
92 Commenter Section of DG-1361 Specific Comments NRC Resolution to final withdrawal from service of a structure, system or component, as it excludes time prior to initial operation (e.g., storage time).
Recommendation This note should be removed.
Also See Comments 2.8 and 2.9.
William Horin of NUGEQ Section C.1.d / p10-11 Comment 125 As indicated in EPRI 1022959, shelf-life programs are fundamentally based on ensuring that qualified life is not reduced by the length of time in storage.
The control of aging during storage is addressed for applicable equipment through a licensees shelf-life program (10 CFR 50 Appendix B, Criterion XIII). A properly executed shelf life and storage program effectively prevents significant aging effects from occurring. This approach is also consistent with the NRCs Equipment Qualification Training Manual for Nuclear Regulatory Commission Technical Reviewers and Inspectors (see training slide 225; Accession No. ML16252A163).
This change is essentially introducing a new regulatory position that significant aging occurs during storage regardless of the shelf life or storage conditions that must be addressed under the provisions of Criterion XIII of 10 CFR 50 Appendix B. Also See Comments 2.7 and 2.9.
See the staffs response to comment 4.
William Horin of NUGEQ Section C.1.d / 10-p11 Comment 126 This appears to be a new regulatory position or expectation since 50.49(e)(5) only requires the consideration of significant aging mechanisms. According to EPRI Report 1021067 (EQ Reference Manual (Reference 26 of DG-1361)),
Section I.4, Nuclear plant practice has been to assume that the shelf storage life of a component does not affect its in-service qualified life. That is, when installed, the item is like new. This assumption is reasonable provided proper storage conditions are used and conservative shelf-life limitations are specified.
This wording is inconsistent with current industry practice where qualified life and shelf life are typically treated separately. The degree of degradation of a properly packaged and stored item that is subject to a shelf-life program See the staffs response to comment 4.
93 Commenter Section of DG-1361 Specific Comments NRC Resolution is not significant compared to the inherent level of uncertainty in defining a qualified life that is based on the Arrhenius methodology.
Also See Comments 2.7 and 2.8.
William Horin of NUGEQ Section C.1.h (1) /
p12 Comment 127 To make it clear that this is not indicating any change from the radiation sources that are identified as being significant in the current licensing basis, which can vary. Consistency with RG 1.183 and 1.195. This statement should be clarified as meaning all significant radiation sources that are considered as part of the licensing basis.
Recommendation Proposed Change: The radiation qualification should factor in doses from all significant radiation sources at the equipment location.
The staff partially agrees with the comment.
The staff agrees that only radiation sources that are significant to the total integrated dose to the qualified equipment need to be considered and that initial qualification should be based on the significant sources expected to affect the equipment. However, the staff notes that if during operation the qualified equipment is exposed to a new significant radiation source that wasnt part of the initial licensing basis for that equipment, then the dose from the new source and the impact to the original analysis for the qualified equipment should be considered (i.e., a new source that wasnt part of the initial licensing basis for the equipment should not be ignored).
The sentence in section C.1.h.(1) of DG-1361 that states, The radiation qualification should factor in doses from all potential radiation sources at the equipment location, has been revised to The radiation qualification should factor in doses from all radiation sources that significantly impact the total integrated dose to the equipment. NOTE: This sentence is now in Section C.1.i.(1) of RG 1.89, Rev. 2.
William Horin of NUGEQ Section C.1.h (2) /
p12 Comment 128 Keeping the guidance in RG 1.89 specific to an acceptable method of complying with 10CFR50.49.
See the staffs response to comment 25.
94 Commenter Section of DG-1361 Specific Comments NRC Resolution Consistency with the guidance in RG 1.89 R1 Position C.2.c (8).
Recommendation This statement should be clarified that this is referring to low-level radiation doses that exceed the radiation harsh or radiation damage thresholds (i.e., the lowest dose that induces an observable change in physical properties of a material).
Also see Comment 2.18.
William Horin of NUGEQ Section C.1.j (1) /
p12 Comment 129 The staffs research results from NUREG/CR-0275, NUREG/CR-4301, and NUREG/CR-4091 did not identify a significant effect on performance based on simultaneous vs. sequential test sequences. Consistency with 50.49(e)(7).
10CFR50.49(e)(5) does not require the aging in a test program to produce a worst-case state of degradation. The position differs from the guidance in C.5.a of RG 1.89R1, which indicates the need to account for synergistic effects that have been identified prior to the initiation of qualification test program.
The guidance in C.1.j (1) brings up the difference between simultaneous vs.
sequential testing as a synergistic effect. NRC research presented in NUREG/CR-0275, NUREG/CR-4301 and NUREG/CR-4091 addresses simultaneous vs. sequential test sequences. This research did not identify a significant effect on performance. Per 50.49(e)(7), the need to address synergistic effects is conditional upon the synergism having a significant effect on equipment performance.
This statement also appears to be addressing dose rate effects and test sequence effects. Additional information on the effect of dose rate and test sequence effects is provided in NUREG/CR-2127 and NUREG/CR-3629.
Recommendation Proposed Change: Eliminate discussion on simultaneous vs. sequential aging and revise or reword for consistency with previous staff position in Section C.5.a of RG 1.89 R1.
The staff agrees with the comment. C.1.j(1) has been rewritten to be presented more clearly. See the staffs response to comment 8 for additional information.
95 Commenter Section of DG-1361 Specific Comments NRC Resolution William Horin of NUGEQ Section C.1.j.3 /
p12 Section C.1.j.3 /
p13 Comment 130 The proposed wording represents a backfit for current plant licensing bases, would constitute a forward fit for current plants taking actions that would appropriately be managed premised on the existing licensing basis or under the specific regulatory terms of 10 CFR 50.49, related to the selection and justification for selecting an activation energy value for a material.
This information is generally unavailable to the licensee and where it does exist, it is often considered proprietary by many manufacturers.
This new staff position:
a) Could be interpreted as meaning that the basis for the activation energy must include data within the temperature range that the equipment or component is exposed to during normal operation. This would directly conflict with IEEE Std. 101 b) Doesnt address or provide options for justifying an activation energy (either by the manufacturer, vendor, or licensee) when activation energy data for the specific formulation is not available or is only available for a generic material family.
c) Doesnt reflect the position in IP 71111.21N regarding validation of information in EQ Reports (e.g., Activation Energy) from approved 10CFR50 Appendix B suppliers.
A significant portion of the activation energies selected by manufacturers is derived from materials databases, academic research, and testing performed for or by other organizations, which the vendors may consider proprietary or otherwise retained in their record system.
This is a new staff position that would impose very specific expectations for activation energies that are not requirements as set forth in 10 CFR 50.49, and therefore applicable to any plant to which 10 CFR 50.49 is applied, and are changes to current guidance in Regulatory Guide 1.89, Rev. 1, Section C.5.c as has been applied to currently licensed plants [and potentially to replacement equipment in those currently licensed plants.].
The staff partially agrees with the comment.
The staff understands that certain data may not be immediately available to licensees since some information may be maintained by a vendor/manufacturer. For this reason, the staff has modified the RG to include reference to, Select Topics Regarding the Environmental Qualification Process, of Inspection Procedure 71111 Attachment 21N, Design Bases Assurance Inspection (Programs), (ADAMS Accession No. ML19036A556) dated February 5, 2019, which provides additional clarification on select environmental qualification topics, including activation energy.
The staff disagrees with the comment that the information in the RG, including the above reference conflicts with the intent of IEEE Std. 101. The RG, and Attachment 2 of IP71111.21N note that For organic materials, a regression line (IEEE Std. 101, IEEE Guide for the Statistical Analysis of Thermal Life Test Data (Ref. 32)), may be used as a basis for selecting the aging time and temperature.
Sample aging times of less than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> should not be used.
The staff disagrees that this position represents a backfit, forward fit, or NRC staff expectations. This guidance does not meet the definition of backfitting or forward fitting in MD 8.4. RG 1.89, Rev. 2 is voluntary guidance and represents one acceptable way to satisfy
96 Commenter Section of DG-1361 Specific Comments NRC Resolution The staff position that activation energies should be based on the data from the specific compound is new. While this may be the optimum approach, the guidance needs to account for situations where the specific formulation is proprietary or not currently available.
The staff position that activation energy should be selected based on the temperature range of the equipment in service is also new. Activation energies that are based on isothermal testing and analysis per IEEE 101 are not based on data points within the range of normal operating temperatures when the equipment is in service. These data points are higher and extrapolated down to the equipments service temperature.
The expectation of the activation energy being traceable to a specific test report is also a new regulatory position compared to C.5.c of RG 1.89 R1.
the applicable NRC regulations. The NRC staff is not imposing or expecting specific licensees to implement the guidance in RG 1.89, Rev. 2.
Further, forward fitting occurs only in the context of a licensing action, and the issuance of a RG is not a licensing action.
See the staffs response to comments 9 and 50 for additional information.
William Horin of NUGEQ Section C.1.f / p11 Comment 131 Consistency with Management Directive MD 6.6. The use of the term must ensure should be clarified since it could be interpreted as going beyond providing reasonable assurance that certain important to safety electrical equipment is not susceptible to environmentally induced common cause failures. Proposed Wording: If used, these methodologies should continue to provide reasonable assurance that equipment important to safety will perform as required under design basis accident conditions. The use of the term must ensure should be clarified since it could be interpreted as going beyond providing reasonable assurance that certain important to safety electrical equipment is not susceptible to environmentally induced common cause failures.
Recommendation Proposed Wording: If used, these methodologies should continue to provide reasonable assurance that equipment important to safety will perform as required under design basis accident conditions.
The sentence that used the term must ensure has been removed from the guidance. See the staffs response to comment 6 for additional information.
97 Commenter Section of DG-1361 Specific Comments NRC Resolution William Horin of NUGEQ Section C.1.l / p13-14 Comment 132 The environmental parameters discussed in RG 1.183 is limited to dose. As discussed in Comment 2.16, the source term used for EQ has margins that are inherently conservative.
The proposed wording is not consistent with the staffs resolution of the Nuclear Power Engineering Committee (NPEC) comment number 73 to NUREG-0588 that clarifies the intent of margin requirements in Section 6.3.1.5 of IEEE Std. 323-1974.
The reference to RG 1.183 is confusing as well as a change from the reference to Position C.4 in RG 1.89 R1. The current staff position in C.4 of RG 1.89 R1 covers margin and is not specific or limited to the application of quantified margin for dose. This staff position is effectively invoking AST or RG 1.183 on Part 50 plants.
Recommendation Proposed Change: Reword for clarity and remove wording related to margins being applied to the environmental parameters discussed in RG 1.183. Also see Comment 2.16 The staff partially agrees with the comment.
The reference to RG 1.183 in Section C.1.l was made in error. The second sentence in C.1.l has been removed.
The staff does not intend to require the use of RG 1.183 or any other source term methodology as part of the update to RG 1.89.
Nor does the staff intend to require any changes to the source term methodology used in currently licensed plants. The source term methodology used must be applicable to the specific plant and adequate to meet applicable requirements.
William Horin of NUGEQ Comment 133 NUREG-0588 (1.4 (1)) states that additional radiation margin identified in Section 6.3.1.5 of IEEE Std. 323-1974 are not required if the required accident radiation dose is developed in accordance with the methodology in Appendix D of 0588. This position recognizes the inherent conservatism in the methodology used to define the integrated accident doses used for EQ.
Since these conservatisms are quantifiable, this approach would satisfy 50.49(e)(8). The source term for EQ (based on earlier EQ guidance, RG 1.31, RG 1.42, RG 1.1833 or RG 1.1954) are significantly more severe than the allowable level of fuel failure under design basis accident conditions. The level of conservatism in the assumptions used to define the accident dose should be more than sufficient to eliminate the need to arbitrarily add an additional 10%. This is further supported by NUREG/CR-53135 which The staff disagrees with the comment. While the source term methodology used for many components may contain significant conservatisms, for others it may not (e.g.,
components that receive a maximum accident dose from an accident other than the maximum hypothetical accident LOCA). In addition, EQ analysis vary based on the specific plant and licensee. Also, RG 1.89 has recommended certain potential non-conservatisms, with the inclusion of the 10% margin added to the accident dose calculations. For example, beta dose may not need to be considered if it is less than 10% of the gamma dose (see the response
98 Commenter Section of DG-1361 Specific Comments NRC Resolution indicates in Section VII.6.8 that core melt in-containment radiation conditions have yet to be calculated to this accuracy (e.g., within a factor of 2).
Recommendation The suggested margin for accident radiation dose in 7.4.1.7 of the dual logo standard should not be required when the integrated accident dose is developed consistent with RG 1.183 or RG 1.195 (or earlier analysis performed per Appendix D of NUREG-0588, Appendix D of RG 1.89 R1, RG 1.3, or 1.4).
to comment 7). Therefore, the staff does not find that the 10% margin should be removed.
As a result, no changes to DG-1361 were made as a result of this comment.
See the staffs response to comment 101 for additional details.
William Horin of NUGEQ Section C.2.a(3) /
p15 Comment 134 Compared to RG 1.89 R1, this new guidance appears to narrow the focus on consideration of shelf life for DOR or NUREG-0588 qualified equipment that has been in stock prior to February 22, 1983. Sound reasons to the contrary does not appear to the appropriate location since it is redundant with C.1.d.
Recommendation Proposed Change: Remove the proposed additional wording related to shelf life being addressed with respect to potential impact on qualified life. Also see comments related to C.1.d, such as 2.7, 2.8 and 2.9.
See the staffs responses to comment 4.
William Horin of NUGEQ Section C.2.c / p16 Comment 135 Consistent with the staffs response to Comment 37 to NUREG-0588 and Table C-1 of the DOR Guidelines, there should be some recognition or allowance for equipment to be classified as being in a mild radiation only environment when it can be demonstrated that the total integrated dose for which the equipment is being qualified is below the lowest radiation damage threshold for any of the items that are relied upon for the equipment to perform the credited important to safety function(s).
The addition of radiation thresholds changes the definition of mild as provided in 50.49. The proposed revision to RG 1.89 should recognize that the distinction between mild and harsh radiation environments is directly related to elevated stressors under DBA conditions that could result in environmentally induced common cause failures. Equipment items made of The staff disagrees with the comment. The guidance provided, distinguishing mild and harsh radiation environments, is consistent with previous staff positions documented in NUREG-0800, NUREG-1793, and NUREG-1503.
In addition, demonstrating that the total integrated dose that the equipment is exposed to is less than the damage threshold for the equipment is part of demonstrating that the equipment is appropriately qualified. The staff does not intend to identify different thresholds for a harsh radiation environment for different
99 Commenter Section of DG-1361 Specific Comments NRC Resolution materials with radiation damage thresholds above the total integrated dose levels are not subject to radiation induced common cause failures.
Recommendation Proposed Change: Add the following to the end of the statement in C.2.c:
Total integrated dose requirements that are above these thresholds may also be considered a mild radiation environment when it can be demonstrated that the radiation damage threshold for the equipment is higher than the required total integrated dose to which the equipment is being qualified to. Also see Comment 3.3.
types of equipment other than those specified in Section C.2. (This is now C.2.b)
See the staffs response to comment 17 for additional information on total integrated dose as it pertains to the environmental qualification of electrical equipment.
No changes were made to DG-1361 as a result of this comment.
William Horin of NUGEQ Section C.2.d (2) /
p16 Comment 136 As noted in Comment 2.4, the imposition of an aging sequence that produces the worst-case degradation goes beyond the regulatory requirement in 10 CFR 50.49(e)(5) to precondition the test specimen(s) to their end-of-installed-life condition.
Regulatory Staff position C.2.d is largely based on wording from Section B, Discussion section of RG 1.89, Rev 1. However, the wording in C.2.d.(2) is a new staff position. The inclusion of inverse temperature effects has been added as an example of uncertainties with respect to the ability of an accelerated aging program being able to simulate an end-of-installed life condition.
Also see Comment 2.4.
See the staffs response to comments 48, 67, 96, and 121.
William Horin of NUGEQ General Comment 137 It would be helpful to end users of the revised RG if the guidance was structured in a way that clearly differentiated between the guidance for Part 52 plants vs. Part 50 plants. Having clear and concise delineation of requirements that apply would be consistent with NRC Management Directive MD 6.6.
The following comments address the need for DG-1361 to clearly differentiate between Part 50 and Part 52 guidance. There are several The staff partially agrees with the comment.
While the staff agrees that it would be helpful to be able to delineate between each type of applicant/licensee, some of the criteria apply to both Part 50 and 52 applicants and licensees.
The Applicable Regulations section lists the applicable regulations without specifying the criteria each type of applicant/licensee (construction permit, operating license,
100 Commenter Section of DG-1361 Specific Comments NRC Resolution examples in DG-1361 where the distinction between specific guidance relative to Part 50 or Part 52 plants is not clearly delineated. See following comments.
combined license, manufacturing license, standard design approval, design certification) must meet.
See the staffs response to comment 82 for additional information on the staffs position on delineating between the requirements for Part 50 and Part 52 facilities.
William Horin of NUGEQ General Comment 138 Consistency with Management Directive MD 6.6.
The relevant guidance that would be applicable to Part 50 plants should be clearly identified for situations such as when an existing plant decides to install a new or replacement item that has been type tested to the dual logo standard. For existing plants, this should be able to be done using CLB/CDB for environmental conditions. For example, equipment that is type tested to the dual logo standard could be qualified without having to adopt or apply AST for the purpose of environmental qualification. For existing plants, it should be clarified that if there are any specific aspects of type testing to the dual logo standard that would need to be reconciled or addressed to ensure continued compliance to IEEE Std. 323-1974.
The staff agrees that the current license/design basis remains applicable. This RG, as with the previous version, provides one acceptable means of complying with the regulations.
No changes were made to DG-1361 as a result of this comment.
William Horin of NUGEQ Section C.2.c / p16 Comment 139 Consistency with Management Directive MD 6.6.
These NUREGs are used as the reference for accepting a mild radiation environment for electronic equipment as 103 rad and 104 rad for other equipment. As worded, it is not clear if this position is specific to just Part 52 plants. Additional discussion should be provided to clarify why regulatory positions in SERs for Part 52 plants is applicable to Part 50 plants.
Note that this citation may be impacted depending on the resolution of Comment 2.18.
The staff disagrees with the comment. The RG update is applicable to any nuclear power plant that references it. Existing facilities are not required to conform to the updated RG unless they chose to. Note that RG 1.89, Revision 1 provided the value of 104 rads for non-electronic components and indicated that electronic components may have a lower threshold. Therefore, the inclusion of the 103 rad threshold for electronic equipment adds a threshold when there was none specified in RG
101 Commenter Section of DG-1361 Specific Comments NRC Resolution 1.89, Revision 1. No changes were made to DG-1361 as a result of this comment.
See the staffs response to comment 25 for additional information.
William Horin of NUGEQ Applicable Guidance /
p4 and Appendix D / D-2 Comment 140 The text should be revised or clarified. The proposed revision specifically focuses on AST, which would be appropriate for a Part 52 plant, but not necessarily applicable to EQ for a Part 50 plant. The Draft Guide is not clear whether Technical Information Document (TID) 14844, Calculation of Distance Factors for Power and Test Reactor Sites remains a valid methodology for calculating source terms. The proposed revision to RG 1.89 does not recognize or reference RG 1.195, Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors or the acceptability for Part 50 plants to use source terms developed in accordance with NUREG-0588, Appendix D of RG 1.89 R1, or RGs 1.3 or 1.4 which are based on TID-14844.
The proposed change to the guidance and Appendix D, Qualification in the Radiation Environment regarding the use of alternate source term (AST) is an example where the DG is primarily focused on Part 52 plants.
There is no need to repeat the information in Appendix D, Section D-2, that is already contained in Appendix I to RG 1.183.
Recommendation Proposed Change: Appendix D should simply refer to Appendix I of RG 1.183 and expand upon the continued acceptability of source terms based on TID-14844 for EQ purposes, even if the plant has adopted AST for other radiological analysis. Appendix D-2 should also cover the resolution of GSI-187, which concluded that licensees may continue to use TID-14844 for EQ even if they adopt AST (See ML011210348). Also See Comment 5.2.
The staff disagrees with the comment.
This revision to RG 1.89 focuses on AST (and therefore, the guidance provided in RG 1.183) for radiological accident guidance because that is the most recent radiological source term guidance. Other approved source term methodologies and EQ methodologies, including those provided in TID-14844 and RG 1.89, Revision 1, continue to be acceptable to licensees currently using them, provided that design changes affecting the source term are not being made. Additional justification is needed for use of TID-14844 or other previously approved methodologies for fuel enrichment greater than 5% and peak burnup greater than 62,000 MWD/MTU.
In addition, Section D-2.2 of Appendix D of DG-1361 indicates that approved alternative assumptions may be used. This is intended to specify that the source terms and source term methodologies in RG 1.183 are not the only acceptable source terms or methodologies that can be used.
102 Commenter Section of DG-1361 Specific Comments NRC Resolution Appendix I has been removed from the next proposed revision of RG 1.183 (see DG-1389 (ADAMS Accession No. ML21204A065)).
See the staffs response to comment 101 for additional details.
William Horin of NUGEQ References
/ p19-20 Comment 141 As noted in Section 3.11 of currently issued DSRS, these documents contain guidance acceptable to the staff for environmental design and qualification of computer-specific requirements that should be used in conjunction with NUREG-0588 and RG 1.89, as appropriate, for evaluating computer specific requirements.
DG 1361 doesnt reference or point to RG 1.152 or IEEE Std 7-4.3.2. At a minimum, inclusion of these references this would seem appropriate for Part 52 plants.
The staff agrees with the comment. As noted in the Design Specific Review Standard (DSRS) for the NuScale Small Modular Reactor design, there are applicable environmental criteria associated with the use of computer-based equipment in safety related applications. The staff augmented the Related Guidance section in RG 1.89, Rev. 2 to include appropriate documents outlining criteria for evaluation of the performance of digital devices, including embedded digital devices, under the environmental conditions expected.
Specifically, the staff added RG 1.152, which endorses IEEE Std. 7-4.3.2-2003.
William Horin of NUGEQ Section C.2.f / p17 Comment 142 While Part 52 plants will utilize AST, Part 50 plants continue to use a radiological source term based on TID-14844 based on the resolution of GSI-187 (which dealt with the potential impact of postulated cesium concentration on Equipment Qualification). GSI-187 was closed out based on the conclusion by the staff that there was no clear basis for backfitting the requirement to modify the design basis for equipment qualification to adopt AST. (See ML011210348).
This section should also address or cover the guidance in RG 1.195 in addition to RG 1.183. RG 1.3 and RG 1.4 were withdrawn based on the guidance in these documents being updated and incorporated into RG 1.183 and 1.195.
See the staffs response to comments 101 and 140.
103 Commenter Section of DG-1361 Specific Comments NRC Resolution In addition to the guidance regarding AST and RG 1.183, this section should also address the continued acceptability of the TID-14844 source term for defining accident doses for EQ, even for plants which have adopted AST.
Suggest inclusion of RG 1.195 for plants where EQ is not based on AST.
Also see comments 3.4 and 5.2.
William Horin of NUGEQ General Comment 143 The proposed revision to RG 1.89 does not address or cover some of the potential burden reduction areas that the staff previously indicated would be considered in the next revision to RG 1.89. The staffs prior indication to consider these in a future revision of RG 1.89 was not captured or identified in the staffs periodic review that concluded in 2018 that a revision to this regulatory guide was warranted.
For example:
- 1) Graded Qualification Methods Based on Severity of Accident Environment NRC response: We are considering whether to clarify the option to qualify a component specific to the environment through a revision to Regulatory Guide (RG) 1.89.
- 2) Alternative Qualification Methods for Equipment Exposed to Radiation-Only Harsh Conditions NRC response: We are considering whether to clarify the option to address EQ for radiation only environments through a revision to RG 1.89.
Reference Accession No. ML040510309 The NUGEQ believes that these examples could be achievable under existing regulatory direction with license amendments (e.g., application of 10 CFR 50.59, or application of existing guidance).
The Group would like to see such positions referenced in DG-1361, perhaps in an Appendix.
The staff partially agrees with the comment.
The staff reviewed the February 20, 2004, letter from the NRC to William A. Horin (ADAMS Accession No. ML040510309), in response to the comment. DG-1361 does not prohibit alternative qualification methods from being used nor does it prevent qualifying for select harsh environmental conditions. For example, if equipment is only exposed to a radiation harsh environment and temperature, pressure, etc. do not cause significant aging degradation of equipment as result of the environmental conditions, then only radiation would have to be considered when qualifying the equipment. The staff finds that the proposed RG and IEC/IEEE 60780-323-2016, contain adequate guidance on qualifying equipment subject to a radiation environment.
With regard to alternative qualification methods, applicants or licensees may use alternative approaches, if appropriately justified, and consistent with current regulatory practice and applicable NRC requirements.
104 Commenter Section of DG-1361 Specific Comments NRC Resolution Based on this, the staff determined that no additional information needed to be incorporated into the revision to RG 1.89.
William Horin of NUGEQ Section C.1.g (1) /
p11 App C / C-1 Comment 144 The guidance in Appendix C doesnt cover Part 52 plants.
The cited methods in Appendix C are unchanged from RG 1.89 R1 and only cover B&W, CE, Westinghouse and GE designs that were licensed under Part
- 50. It would be more appropriate to make Appendix C more generic by linking the methodology for mass and energy release to be consistent with the methodology used to define the containment response to design basis accidents for the safety analysis or consistent with the methodology used for HELB analysis. This approach would result in the appendix being applicable to both Part 50 and Part 52 plants.
Recommendation Proposed Change: C.1.g should be reworded in a more generic manner that avoids the use of specific codes since these are established and approved as part of the plants design and licensing basis.
The staff partially agrees with the comment.
The staff revised the guidance to include a footnote to Appendix C in RG 1.89, Rev. 2.
that states:
Mass and energy releases are developed using a methodology that is consistent with the licensing basis of the plant. The listed methods are examples for existing designs.
See the staffs responses to comments 90 and 99 for additional details.
William Horin of NUGEQ Applicable Guidance /
p4 and Appendix D / D-2 Comment 145 For Part 50 plants, GSI-187 (which deals with the potential impact of postulated cesium concentration on Equipment Qualification) was closed out based on the conclusion by the staff that there was no clear basis for backfitting the requirement to modify the design basis for equipment qualification to adopt AST (See ML011210348).
Consistent with the resolution of GSI-187, it is common industry practice for operating plants to continue to use a source term based on TID-14844 for establishing EQ even if the plant has otherwise adopted AST.
Since DG-1361 covers both Part 50 and Part 52 plants, it should continue to address radiological source terms based on TID-14844 as well as AST. The guidance in Section D-2 focuses exclusively on AST. The methodology and See the staffs response to comments 101 and 140.
105 Commenter Section of DG-1361 Specific Comments NRC Resolution sample calculation for EQ radiation dose using a non-AST source term has been removed from Appendix D.
Recommendation Proposed Change: DG-1361 should include a reference to RG 1.195 along with changes needed to satisfy 10CFR50.49 (if any). The proposed revision to RG 1.89 should also reflect the conclusion from the resolution of GSI-187 that licensees can continue to use TID-14844 for EQ even if they adopt AST.
Also see Comment 3.4.
William Horin of NUGEQ Appendix A
Comment 146 Consistency between regulatory guidance documents.
Appendix A is used to provide examples of typical safety-related electrical equipment (Class 1E) or systems. Suggest that the additional clarification or examples from Chapter 3.11 of NUREG-0800 (SRP) be considered for inclusion.
Recommendation Proposed Change: Add additional clarification consistent with Footnote 6:
6 (A) Electrical equipment that are essential for shutting down the reactor and maintaining it in a safe shutdown condition, containment isolation, reactor core cooling, and containment and reactor heat removal, or otherwise are essential in preventing significant release of radioactive material to the environment following a design basis accident; (B) Electrical equipment that initiates the above functions automatically; (C) Electrical equipment that is used by the operators to initiate the above functions manually; (D) Electrical equipment whose failure can prevent the satisfactory accomplishment of one or more of the above safety functions; (E) Other electrical equipment important to safety, as described in 10 CFR 50.49(b)(1) and (2); (F) Certain post-accident monitoring equipment, as described in 10 CFR 50.49(b)(3) and RG 1.97; and (G) Protection and safety systems as described in 10 CFR 50.55a(h) and RG 1.209.
The staff agrees with the comment. The staff inserted the suggested information as a footnote in Appendix A of RG 1.89, Rev. 2.
106 Commenter Section of DG-1361 Specific Comments NRC Resolution William Horin of NUGEQ Appendix B
/ B-1 Comment 147 Since the DG will still apply to Part 50 plants, the guidance should be retained. Ensuring associated circuits of non-safety-related equipment will not fail and prevent satisfactory accomplishment of safety functions by safety-related equipment should remain cited as a typical example of non-safety-related equipment being addressed by 50.49.
The discussion on associated circuits has been removed. This is relevant information pertaining to the identification or exemption of equipment being classified as subject to EQ per 50.49(b)(2).
Recommendation Proposed Change: Reinstate first paragraph of Appendix B from RG 1.89 R1.
See the staffs response to comment 99.
William Horin of NUGEQ App D, Section D-1
/ D-1 Comment 148 The period of exposure may be limited to a qualified life that is less than the plant license.
Recommendation Proposed Change: The period of exposure for a normal operational dose is generally the duration of the plant license; however, the period of exposure may be limited to the qualified life of the equipment.
The staff partially agrees with the comment.
The staff deleted the sentence of concern as it was deemed unnecessary. The important aspect of environmental qualification is that environmental parameters such as radiation are addressed for the entire duration for which the equipment is installed in the plant.
William Horin of NUGEQ Missing App E Comment 149 The guidance in Appendix E specifically defines EQ categories that are not part of the documentation requirements in the dual logo standard. The EQ categories in Appendix E, Sections 3.a thru 3.d are still relevant and in use by both Part 50 and Part 52 licensees.
The current guidance in Appendix E of RG 1.89 R1 remains relevant to Part 50 and Part 52 plants. Appendix E should remain part of RG 1.89 since it includes specific information related to qualification documentation.
See the staffs response to comment 103.
107 Commenter Section of DG-1361 Specific Comments NRC Resolution Recommendation Suggested Change: Reinstate and retain Appendix E.
William Horin of NUGEQ Related Guidance /
p3 Comment 150 10CFR50.69 is relevant since it can eliminate safety-related electrical equipment from the scope of the EQ program if it is classified as RISC-3.
Consistency with existing regulation that can change the scope of equipment subject to 10 CFR 50.49.
The proposed revision to RG 1.89 should recognize and reflect EQ program scope and implementation basis in accordance with risk-informed rule 10CFR50.69 by acknowledging the nexus between EQ program scope and that rule in the Regulatory Guidance section.
Recommendation Proposed Change: Add or address 10CFR50.69.
Also see Comment 6.4.
See the staffs response to comment 72.
William Horin of NUGEQ Related Guidance /
p3 Section C.1 Comment 151 Section 6.2 of IEC/IEEE 60780-323 is specific to reassessing qualified life, but there is no specific reference to the existing guidance in X.E1 related to the reanalysis of TLAA. Consistency with existing regulatory guidance.
Staff has provided acceptable methods related to reassessing qualified life in NUREG-1801 & NUREG-2191.Section X.E1 contains specific guidance related to the reanalysis of an EQ aging evaluations or TLAA.
Recommendation Proposed change - Add a staff position in Section C.1 that covers Section 6.2 of the dual logo standard by providing a reference to the existing guidance in Section X.E1 of NUREG-1801 & NUREG-2191.
The staff agrees with the comment. The staff added a staff position in Section C.1.f of RG 1.89, Rev. 2, to supplement Section 6.2 of IEC/IEEE Std. 60780-323, Edition 1, 2016-02, with the following:
X.E1, Environmental Qualification of Electric Equipment, of NUREG-2191, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR)
Report, and X.E1, Environmental Qualification (EQ) of Electric Components, of NUREG-1801, Generic Aging Lessons Learned Report, note that under 10 CFR 54.21(c)(1)(iii), plant EQ programs, which implement the requirements of 10 CFR 50.49 (as further defined and clarified by the DOR
108 Commenter Section of DG-1361 Specific Comments NRC Resolution Guidelines, NUREG-0588, and Regulatory Guide 1.89), are viewed as aging management programs for license renewal and subsequent license renewal. Reanalysis of an aging evaluation to extend the qualification of components under 10 CFR 50.49(e) is performed on a routine basis as part of an EQ program. Reanalysis evaluates the original attributes, assumptions and conservatisms for environmental conditions, and other factors of an aging evaluation to demonstrate that equipment qualified life can be extended.
Important attributes for the reanalysis of an aging evaluation include analytical methods, data collection and reduction methods, underlying assumptions, acceptance criteria, and corrective actions. These attributes are discussed further in Section X.E1 of NUREG-2191 and NUREG-1801.
William Horin of NUGEQ Reason for Revision /
p7 Comment 152 Consistency between RG 1.89 and RG 1.209.
Note that the NRC has endorsed the use of IEEE 323-2003 for qualifying computer-based I&C equipment in mild environment applications in RG 1.209.
Recommendation Proposed Change: Reword to clarify that the NRC did not officially endorse these standards for qualification of electrical equipment to harsh design basis accident conditions.
The staff agrees with the comment. A clarification has been included to identify that guidance for ensuring that digital and computer-based devices in mild environments may be found in RG 1.209. Digital devices proposed for installation in harsh environments and used in SSCs for achieving design basis functions should meet the requirements of 10 CFR 50.49.
109 Commenter Section of DG-1361 Specific Comments NRC Resolution William Horin of NUGEQ Section C.1.b / p10 Comment 153 Consistency with 10CFR50.69. To clarify that not all safety-related electrical equipment that must function during or following exposure to harsh accident conditions is subject to 50.49 if they are classified as RISC-3 under 10CFR50.69.
This statement does not reflect or recognize that electrical equipment important to safety that is classified as RISC-3 under 10CFR50.69 is not subject to 10CFR50.49. DG-1361 should clarify that under 10 CFR 50.69, safety-related equipment that perform low risk significant functions (i.e.,
RISC-3) are not subject to 10 CFR 50.49 requirements. Also See comment 6.1 Recommendation Proposed Change: 10 CFR 50.49 requires safety-related electric equipment (Class 1E) as defined in 10 CFR 50.49(b)(1) to be qualified to perform intended safety functions, unless classified as RISC-3 under 10 CFR 50.69.
The staff disagrees with the comment and with incorporating the specific suggested language.
However, the staff revised the guidance to include information on 10 CFR 50.69. See the staffs response to comment 72 for additional details.
William Horin of NUGEQ Section C.2.f / p17 Comment 154 This statement is confusing, particularly the interjection of before testing when discussing various elements of a test sequence.
Recommendation Suggest changing before testing to before the DBA simulation to remove any confusion regarding what testing the radiation test needs to be done prior to. It would also be appropriate to retain the guidance from C.2.c of RG 1.89 R1.
The staff agrees with the comment that the statement could lead to confusion.
See the response to comments to 16 and 17.
William Horin of NUGEQ Applicable Regulations
/p2 Comment 155 The reference to 10 CFR 52.157, Contents of applications; technical information in final safety analysis report does not provide a definition of a manufacturing license.
The citation to 10 CFR 52.157 appears to be a typo because the Commission findings necessary for issuance of a manufacturing license are set forth in See the staffs response to comment 83.
110 Commenter Section of DG-1361 Specific Comments NRC Resolution 10 CFR 52.167. This may be intended to reference 10 CFR 52.167 for the definition of a manufacturing license.
William Horin of NUGEQ References
/ p 20 Comment 156 The citation to EPRI Report 1021067 should not include NEI.
Recommendation Suggested citation: Electric Power Research Institute (EPRI), Nuclear Power Plant Equipment Qualification Reference Manual, Revision 1. EPRI, Palo Alto, CA: 2010. 1021067 See the staffs response to comment 97.
William Horin of NUGEQ Appendix C Footnote 10
/ C-1 Comment 157 Correct the embedded hyperlink address to remove the word and at the end.
The embedded hyperlink to Doc Collections includes the word and as part of the address.
The staff agrees with the comment and has fixed the link.
Comment Document 13: ML21131A005 Jeremy Owen of Kinetrics C.1.a Comment 158 In the context of IEC/IEEE 60780-323 end condition means the condition of the equipment after completion of the aging treatment. End of installed life may end up being different than the qualified life, for instance if condition-based qualification is applied. As such this paragraph is confusing and does not provide clarity for the terms referenced.
Recommendation Recommend deleting this paragraph or providing further clarification between end of installed life vs qualified life.
The 4th line states:
Note: Qualified equipment must be capable of performing its design function at the end-of-installed life. To be specific, suggest adding the word qualified to read: Note: Qualified equipment must be capable of performing its design function at the end-of-installed qualified life.
See the staffs responses to comment 33.
111 Commenter Section of DG-1361 Specific Comments NRC Resolution Jeremy Owen of Kinetrics C.1.b Comment 159 This section is intended to provide clarity about the phrase important to safety. The last paragraph does that but the rest of the section, specifically the second paragraph, do not provide useful information and make the section more confusing than it needs to be.
Recommendation Recommend removing the second paragraph.
See the staffs response to comment 2.
Jeremy Owen of Kinetrics C.1.d Comment 160 The term service life in IEC/IEEE 60780- 323 does not imply anything about aging effects outside of the time the equipment was in service. While improper control of shelf life can affect qualified life, it does not relate to service life.
Recommendation Recommend removing this section as it introduces confusion between service life and qualified life in relation to the impact of improper control of shelf life.
See the staffs response to comment 4.
Jeremy Owen of Kinetrics C.1.j(3)
Comment 161 While it is preferred to use the activation for the actual compound being tested, it is not always practical, and the accepted industry approach has been to use available conservative values. More often than not, activation energies for a specific compound are not available.
Recommendation Given the importance of activation energies for qualified life of equipment, as much guidance as possible should be given on selection of activation energies.
It is recommended to refer to IEEE 98, 99 and/or UL Std 746B for the determination of activation energies. The section should also indicate that while the activation energy for the specific material being considered is The staff partially agrees with the comment.
The staff is not implying that the lowest activation energy for a material must be used.
To clarify this, RG 1.89, Rev. 2 includes the following: The selected activation energy should be representative of the most limiting material in a component/sub-component when determining qualified life.
The staff modified the guidance to include references to IEEE Std. 98-2016 and IEEE Std. 99-2019 as discussed in the response to comment 65.
112 Commenter Section of DG-1361 Specific Comments NRC Resolution sometimes required, such as in IEEE 383, the conservative approach is also acceptable if properly justified.
Selecting the lowest activation energy from a group available for a specific failure parameter may be too conservative. Guidance could be given that the activation energy for the material that has the closest UL temperature index to the material being evaluated should be selected rather than the lowest activation energy in the group.
The staff did not have adequate time to review the referenced UL standard to consider it for incorporation into the RG.
Jeremy Owen of Kinetrics C.1.l Comment 162 This section endorses the margins presented in Section 7.4.1.7 of IEC/IEEE Std. 60780-323 are acceptable. The margins presented are only applicable to the accident conditions. Is there any guidance for margins applied to normal conditions?
Recommendation Provide guidance for margins to use for normal conditions.
The staff partially agrees with the comment.
While there is no guidance on how to incorporate margin under normal conditions, conservatisms are generally applied, which inherently add margin. The staff finds that the margins added to accident conditions is adequate to satisfy the requirements for the environmental qualification of electric equipment. Therefore, the staff does not find it necessary to modify DG-1363.
See the staffs response to comment 132 for changes to C.1.l.
Jeremy Owen of Kinetrics C.1.n Comment 163 It is not clear how a double peak should be used for equipment vulnerable to thermal binding or when there are limitations of the steam supply during testing.
Recommendation Recommend to providing additional guidance or specific examples.
See the staffs response to comment 12. The portion of Section C.1.n that comment 163 is associated with has been deleted.
Jeremy Owen of Kinetrics C.2.c Comment 164 This section refers to additional stressors such as smoke exposure. This type of aging mechanism is not part of the scope of IEC/IEEE 60780-323 See the staffs response to comment 14.
113 Commenter Section of DG-1361 Specific Comments NRC Resolution Recommendation Recommend this comment should be removed. Alternatively, it should be made clearer that while this aging mechanism is addressed in other documents such as RG 1.209.
Jeremy Owen of Kinetrics C.2.d(1)
Comment 165 This section refers to preconditioning of test samples employing the Arrhenius methodology. It is not clear as to what aspect of preconditioning this statement refers to and what the reader should consider. IEC/IEEE 60780-323 clearly describes the use of the Arrhenius methodology.
Recommendation This statement should be clarified to indicate what it alludes to, otherwise it should either refer to the discussion in IEC/IEEE 60780- 323 on that topic or be removed. Item #2 is also discussed in IEC/IEEE 60780-323, so this statement does not provide any additional clarity.
The staff agrees that Section 7.4.1.9.3 of IEC/IEEE 60780-323-2016 discusses the use of Arrhenius aging and the sequence of age conditioning considers sequential, simultaneous, and synergistic effects.
The intent of Section C.2.d in DG-1361 was to address uncertainties with regard to the processes and environmental factors that could result in such degradation. This Section specifically stated, Experience suggests that consideration should be given to address these uncertainties.
See the staffs responses to comments 67, 95, 121, and 136 for additional information.
Jeremy Owen of Kinetrics C.2.e(4)
Comment 166 The 7th line bullet # 4 states: analyses taking into account arrangements of equipment and radiation sources may be necessary to determine whether equipment needed for mitigation of design basis accidents other than LOCA or high-energy line breaks (HELB) could be exposed to a more severe environment than the plant specific LOCA or HELB environments.
Recommendation This could be clarified. In order to be clear, suggest defining the other DBAs that are more severe than the plant specific LOCA or HELB.
The staff disagrees with the comment. While the LOCA or HELB is the most severe accident for most equipment, the design basis accident that is most severe to a particular piece of equipment is dependent on the plant and the location of the equipment. For example, if the equipment is located near the main steam piping, the limiting design basis accident for that equipment may be a main steam line break or steam generator tube rupture.
Since the limiting accident is specific to the plant and the location of the equipment,
114 Commenter Section of DG-1361 Specific Comments NRC Resolution identifying a specific accident or providing a full list of potential accidents is not possible or reasonable.
See the staffs response to comment 15 for additional information.