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
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 Comment Document 2 Comment Document 3 James Gleason James Parello Rick Weinacht Huntsville, AL Fort Mill, SC, 29708 Curtiss-Wright Nuclear Division ADAMS Accession No. ML21104A363 ADAMS Accession No. ML21022A044 Scientech 1360 Whitewater Drive Idaho Falls, ID 83402 ADAMS Accession No. ML21106A271 Comment Document 4 Comment Document 5 Comment Document 6 William Horin William Horin Robert Konnik Nuclear Utility Group on Equipment NUGEQ Institute of Electrical and Electronics Engineers (IEEE)
(NUGEQ) Qualification - Winston & Strawn LLP ADAMS Accession No. ML21110A055 Qualification - Winston & Strawn LLP 1901 L Street N.W.
1901 L Street N.W. Washington, DC, 20036-3506 Washington, DC, 20036-3506 ADAMS Accession No. ML21041A128 ADAMS Accession No. ML21041A127 Comment Document 7 Comment Document 8 Comment Document 9 Vincent Bacanskas Carrie Fosaaen Carrie Fosaaen ADAMS Accession No. ML21050A358 Director, Regulatory Affairs Director, Regulatory Affairs NuScale Power, LLC NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon Oregon 97330 97330 ADAMS Accession No. ML21110A054 ADAMS Accession No. ML21113A276 May 2021
Comment Document 10 Comment Document 11 Comment Document 12 Rick Weinacht William Horin William Horin Curtiss-Wright Nuclear Division NUGEQ NUGEQ Scientech Qualification - Winston & Strawn LLP Qualification - Winston & Strawn LLP 1360 Whitewater Drive 1901 L Street N.W. 1901 L Street N.W.
Idaho Falls, ID 83402 Washington, DC, 20036-3506 Washington, DC, 20036-3506 ADAMS Accession No. ML21042A003 ADAMS Accession No. ML21050A360 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 Specific Comments NRC Resolution DG-1361 Comment Document 1: ML21104A363 James Section C.1 Comment 1 The staff disagrees with the comment. The Gleason Section C. 1, a is confusing and does not state explicitly the equipment comment proposes a change to the DG that important to safety that is defined by 10CFR.50.49. It states: 10 CFR would require a change to 10 CFR 50.49(e)(5) 50.49(e)(5) requires, in part, that equipment qualified by test must be because the suggested language does not preconditioned by natural or artificial (accelerated) aging to its end-of- currently exist in 10 CFR 50.49(e)(5). Rule installed life condition. Therefore, end condition, as defined in Section 3.10 changes are outside the scope of this RG of IEC/IEEE Std. 60780-323, Edition 1, 2016-02, should be considered update, so the staff did not make the proposed equivalent to end-of-installed life. Note: Qualified equipment must be change. Furthermore, the staff deleted the capable of performing its design function at the end-of-installed life. regulatory position associated with "end condition" as a result of comment 33. See the staff's response to that comment for additional 10CFR50.49 Section j (2) states Meets its specified performance information.
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.
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Commenter Section of Specific Comments NRC Resolution DG-1361 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 Section C.1 Comment 2 The staff partially agrees with the comment.
Gleaseon Section C. 1, b is confusing and does not state explicitly the equipment While the staff disagrees with including the important to safety that is defined by 10CFR.50.49. suggested text of 10 CFR 50.49(b) in the RG, the staff agrees that its regulations, specifically Recommendation 10 CFR 50.49 and Appendix A of 10 CFR Part Replace with: The following description and definition of important to 50, should be cited to clearly describe and safety should be used instead of the definition in Section 3.12 of IEC/IEEE define what the NRC considers as equipment Std. 60780-323, Edition 1, 2016-02: 10CFR50.49 defines equipment important to safety.
Important to safety in section (b) as follows.
(b) Electric equipment important to safety covered by this Section C.1.b of the DG has been revised to section is: include the following as the description and (1) Safety-related electric equipment. definition of equipment important to safety:
(i) This equipment is that relied upon to remain functional during and following design basis events to ensure The following description and definition of (A) The integrity of the reactor coolant pressure equipment important to safety should be used boundary; instead of the definition in Section 3.12 of (B) The capability to shut down the reactor and maintain IEC/IEEE Std. 60780-323-2016:
it in a safe shutdown condition; or (C) The capability to prevent or mitigate the consequences The introduction to 10 CFR Part 50, Appendix of accidents that could result in potential offsite exposures A, states that important to safety SSCs are 3
Commenter Section of Specific Comments NRC Resolution DG-1361 comparable to the guidelines in §50.34(a)(1), those SSCs that provide reasonable assurance
§50.67(b)(2), or § 100.11 of this chapter, as applicable. that the facility can be operated without undue (ii) Design basis events are defined as conditions of risk to public health and safety.
normal operation, including anticipated operational occurrences, design basis accidents, external events, and 10 CFR 50.49 requires safety-related (Class natural phenomena for which the plant must be designed 1E) electric equipment as defined in 10 CFR to ensure functions (b)(1)(i) (A) through (C) of this section. 50.49(b)(l) to be environmentally qualified to perform its intended safety functions.
(2) Non-safety-related electric equipment whose failure Appendix A to this guide lists typical safety-under postulated environmental conditions could prevent related equipment and systems. 10 CFR satisfactory accomplishment of safety functions specified 50.49(b)(2) requires that non-safety-related in subparagraphs (b)(1)(i) (A) through (C) of paragraph (b)(1) electric equipment be environmentally of this section by the safety-related equipment. (3) Certain qualified if its failure under postulated post-accident monitoring equipment. 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 Section Comment 3 The staff disagrees with the comment.
Gleason C.1.c 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 Section C.1.c of DG-1361 explicitly states that not state explicitly the definition of qualified life to be used and its source. the following definition of qualified life should be used: period for which an 4
Commenter Section of Specific Comments NRC Resolution DG-1361 There is no definition of qualified life in 10CFR50.49. equipment has been demonstrated, through testing, analysis and/or experience, to be
- 1. IEEE 323-74 definition of Qualified Life: The period of time for which capable of remaining functional during and satisfactory following design basis events to ensure that the performance can be demonstrated for a specific set of service conditions. criteria specified in 10 CFR 50.49(b)(1)(i)(A),
- 2. IEC/IEEE 60780-323 Definition of Qualified Life: period for which an (B), and (C) are satisfied.
equipment has been demonstrated, through testing, analysis and/or experience, to be capable of functioning within acceptance criteria during The staffs clarification of the definition of specific operating conditions while retaining the ability to perform its safety qualified life in IEC/IEEE 60780-323-2016 functions in accident condition or earthquake. is similar to the definition in IEEE 323-1974 except it more clearly explains the term by As the IEC/IEEE 60780-323 Definition of Qualified Life expands the describing how qualified life is determined definition to include retaining the ability to perform safety functions in (i.e., by testing, analysis, and/or experience) accident conditions and earthquakes, it creates backfit and forward fit issues. and by citing specific regulatory requirements that must be satisfied in order to establish a Recommendation qualified life. Additionally, the staffs Use the IEEE 323-74 definition of Qualified Life: The period of time for clarification of the definition in IEC/IEEE which satisfactory performance can be demonstrated for a specific set of 60780-323-2016 does not invalidate the service conditions. 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 5
Commenter Section of Specific Comments NRC Resolution DG-1361 unnecessary. NOTE, these changes are reflected in Section C.1.d of RG 1.89, Rev. 2.
James Section Comment 4 The staff partially agrees with the comment.
Gleason C.1.d Section C. 1, d is confusing since it tries to address the term service life and relates service life, qualified life and shelf life. The staff agrees to delete C.1.d as service life is not directly associated with equipment There is no use of service life in 10CFR50.49 and its introduction of a new qualification and is distinctly different than the term service life including its term qualified life. Furthermore, service life relationship to qualified life constitutes a backfit and forward fit. is addressed elsewhere in the regulations (e.g.,
Appendix B to 10 CFR Part 50).
Recommendation Delete section C.1 d and all discussion of service life. This clarification does not meet the definition Add that IEC/IEEE 60780-323 term service life is not endorsed. Please note of backfitting or forward fitting in MD 8.4. RG that that IEC/IEEE 60780-323 proficient use of the term service life and an 1.89, Rev. 2 is voluntary guidance and alternate definition of qualified life may render IEC/IEEE represents one acceptable way to satisfy the 60780-323 to be not endorsed. 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.
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Commenter Section of Specific Comments NRC Resolution DG-1361 See the staffs disposition of comment 3 for further detail on the staffs endorsement of the term qualified life.
James Section Comment 5 The NRC staff partially agrees with the Gleason C.1.e Section C. 1, e is confusing as it notes that The prerequisite for aging electric comment. The staff agrees that the aging of equipment located in a mild environment is not within the scope of 10 CFR electrical equipment in mild environments and 50.49 and then adds Requirements, including EMC and seismic electromagnetic compatibility (EMC) and requirements, shall be specified in the design/purchase seismic qualification are not requirements specifications. under 10 CFR 50.49. However, the staff does not agree with the proposed resolution as It is agreed that the prerequisite for aging electric equipment located in a mild electrical equipment in mild environments and environment is not within the scope of 10 CFR 50.49.t EMC and seismic qualification are covered under other applicable regulations listed in the There is also no requirement for design/purchase specifications in RG. Instead, the staff has elected to edit 10CFR50.49 and there is no requirement for EMC in 10CFR50.49. Thus, Section C.1.e to clarify the available guidance introductions for design/purchase for EMC, aging, and seismic as follows specifications requirements and EMC requirements constitutes a backfit and forward fit. e. Paragraph 4 of Section 5.1 of IEC/IEEE Std. 60780 323, Edition 1, 2016 02, notes that Requirements, including EMC Recommendation [Electromagnetic Compatibility],
Modify Section C. 1, e to the following: In IEC/IEEE 60780-323, the environmental/operational ageing and seismic discussion of design/purchase specifications requirements and EMC and requirements shall be specified in the seismic requirements are not endorsed 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.
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Commenter Section of Specific Comments NRC Resolution DG-1361 James Section Comment 6 The staff agrees that C.1.f could be presented Gleason C.1.f Section C. 1, f is confusing as it states Condition monitoring and associated more clearly to avoid confusion. As such, the condition-based qualification methodologies discussed in Section 6.3 of staff modified the referenced paragraph to IEC/IEEE Std. 60780-323, include the following:
Edition 1, 2016-02, represent new approaches for extending or establishing the qualified life of electrical equipment. Condition monitoring recognizes the fact that If used, these methodologies must ensure that equipment important to safety the aging process in a 10 CFR 50.49 test will perform under the conditions specified in 10 CFR 50.49. method qualification program can be an acceptable process of determining end of This appears to misstate the purpose and application of condition monitoring. qualified life, if it is proven during a qualification by test program to be a condition Section 6.3 of IEC/IEEE Std. 60780-323, Edition 1, 2016-02 States: indicator that must be measurable, change Condition monitoring for equipment qualification purposes monitors one or monotonically with time, be correlated with the more condition indicators to determine whether equipment remains in a safety function performance under design-basis qualified condition. event conditions, be linked to the functional degradation of the qualified equipment, and Condition monitoring is not a new approach for establishing the qualified life have a consistent trend from unaged through of electrical equipment. the limit of the qualified pre-accident condition.
The qualified life is established in the regulatory accepted method of aging, including time/temperature effects, radiation and mechanical degradation. As a result of adding the above text, the last sentence in Section C.1.e of DG-1361 was Condition monitoring recognizes that when qualified life is established in the removed. NOTE, this position is now Section regulatory accepted method, the equipment being qualified is placed into a C.1.g. of RG 1.89, Rev. 2.
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.
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Commenter Section of Specific Comments NRC Resolution DG-1361 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 Section Comment 7 The staff partially agrees with the comment.
Gleason C.1.h 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 The comment is incorrect in stating that radiation qualification testing. Exceptions may be based on qualification by Section C.1.h(2) is new in DG-1361. Section analysis supported by test data or operating experience that verifies that the C.1.h(2), is the same as Section C.2.c.(8) of dose and dose rates will not degrade the operability of the equipment below Revision 1 of RG 1.89 (the section was moved acceptable values. to a new location in DG-1361 based on the new RG formatting). Additional information This is new and the following RG 1.89 Rev 1 section is missing: (6) regarding this paragraph can be found in the Shielded components need be qualified only to the gamma radiation response to comment 25.
environment provided it can be demonstrated that the sensitive portions of the component or equipment are not exposed to significant beta radiation dose The Division of Operating Reactors (DOR) rates or that the effects of beta radiation, including heating and secondary guidelines indicated that, by using the 9
Commenter Section of Specific Comments NRC Resolution DG-1361 radiation, have no deleterious effects on component performance. If, after Technical Information Document (TID)-14844, considering the appropriate shielding factors, the total beta radiation dose Calculation of Distance Factors for Power and contribution to the equipment or component is calculated to be less than 10% Test Reactor Sites, source term and of the total gamma radiation dose to which the equipment or component has conservative assumptions, the beta surface dose been qualified, the equipment or component is considered qualified for the would be 1.40 x 108 rad. If this beta dose was beta and gamma radiation environment. conservatively assumed to be 2.0 x 108 rad and shielding factors discussed in the DOR The deletion of RG 1.89 Rev 1 section (6) Shielded components, etc, guidelines were applied, and if the beta dose constitutes a forward backfit as it deletes an acceptable process for addressing was less than 10% of the total gamma dose to beta radiation and the addition of Section C. 1, h 2 is an unjustified increase the equipment, then the DOR guidelines in requirements and therefore a forward fit. indicated that only the gamma dose needed to be considered.
Recommendation Section C. 1, h 2 should be replaced with RG 1.89 Rev 1 section: (6) RG 1.89, Revision 1, did not specify that the Shielded components need be qualified only to the gamma radiation beta dose should be increased or specify environment provided it can be specific shielding factors, with regards to the demonstrated that the sensitive portions of the component or equipment are 10% criteria.
not exposed to significant beta radiation dose rates or that the effects of beta radiation, including heating and secondary radiation, have no deleterious RG 1.183, Revision 0, Appendix I, provided effects on component performance. If, after considering the appropriate radiological EQ guidance for plants using an shielding factors, the total beta radiation dose contribution to the equipment alternative source term. While RG 1.183, or component is calculated to be less than 10% of the total gamma radiation Revision 0 did provide specific guidance for dose to which the equipment or component has been qualified, the equipment beta radiation, it did not specify any special or component is considered qualified for the beta and gamma radiation criteria for beta radiation similar to the 10%
environment.. 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.
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Commenter Section of Specific Comments NRC Resolution DG-1361 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 11
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Section Comment 8 The staff partially agrees with the comment.
Gleason C.1.j Section C. 1, j (1) is confusing as it contains the following: The staff agrees that C.1.j(1) could be The synergistic effect is the result of the combined environmental effects of presented more clearly to avoid confusion. As the plant conditions such as radiation, humidity, and temperature that could such, the staff has re-written Section C.1.j(1) to result in greater degradation of equipment in relation to sequential application state:
of the plant environment under normal, abnormal, and accident conditions.
Synergistic effects must be considered when these effects are believed to have a significant 12
Commenter Section of Specific Comments NRC Resolution DG-1361 The synergistic effects on materials that are known to have such increased effect on equipment performance. A degradation under these conditions should be accounted for when assessing synergistic effect is the result of the combined the qualified life. environmental effects of the plant conditions such as radiation, humidity, and temperature Section 7.4.1.9.3 Age conditioning of IEC/IEEE Std. 60780-323, Edition 1, that could result in greater degradation of 2016-02 contains the discussion of synergistic effects. Historically, equipment in relation to individual application synergistic effects in qualification are considered in aging and not accident of the plant environmental effects under conditions. normal, abnormal, and accident conditions.
Therefore, the inclusion of accident conditions in determining synergistic If synergistic effects have been identified prior effects creates a new requirement and is a forward fit. to the initiation of qualification, they should be accounted for in the qualification program.
The phrase The synergistic effects on materials that are known to have such Synergistic effects known at this time are dose increased degradation under these conditions should be accounted for when rate effects and effects resulting from the assessing the qualified life is confusing and a new requirement. Synergistic different sequence of applying radiation and effects, as noted in 10CFR50.49 must be considered when these effects are (elevated) temperature.
believed to have a significant effect on equipment performance.
The staff disagrees that synergistic effects need The requirement that synergistic effects on materials need to be accounted for not be considered during accident conditions.
in qualified life is new, as it requires all synergistic effects of materials to be 10 CFR 50.49(e)(7) specifies that synergistic included and inconsistent with the 10CFR50.49 threshold that they must be effects must be considered when these effects considered when these effects are believed to have a significant effect on are believed to have a significant effect on equipment performance. equipment performance, without specifying normal operation or accident conditions.
Recommendation Therefore, synergistic effects that have a Section C. 1, j (1) modify to: Synergistic effects must be considered when significant effect on equipment performance these effects are believed to have a significant effect on equipment during accident conditions must be considered.
performance.
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.
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Commenter Section of Specific Comments NRC Resolution DG-1361 James Section Comment 9 The staff partially agrees with the comment.
Gleason C.1.j Section C. 1, j (3) is confusing in that it states: Activation energy values The staff acknowledges that knowing each should be based on the testing of the specific compound used in the compound requires a significant effort.
equipment and on the most relevant material property and property endpoint However, the materials within a component (i.e., failure mechanism). It is confusing because it constitutes a significant need to be known in order to establish and effort to know each compound and to have Arrhenius test data for Activation justify environmental qualification.
Energy on every possible compound and failure mechanism.
Activation energy values are an important Additionally, most safety related equipment is made up of many materials, factor for establishing/determining the and the basis of aging is to use the lowest activation energy for the assembly qualified life of a component. The clarification when establishing the aging program and qualified life. provided by the staff in this RG includes meaningful information that should be Thus, materials that are not the lowest activation energy are aged for more considered when selecting an activation energy degradation equivalency than the qualified life of the lowest activation energy for the purpose of establishing or extending the material. qualified life of a component. This is especially important given the sensitivity of the activation The NRC studies, such as NUREG/CR-6384 and NUREG/CR-6704 on energy variable in effecting the results of the Arrhenius Theory and its application to environmental qualification have Arrhenius equation.
demonstrated the conservatism to establishing qualified life.
The staff agrees that equipment within the Lastly, Activation energy is not a safety function and was never intended to scope of this RG can be composed of a variety be a quality attribute of a safety related component. of materials, and that the basis for aging is to use the activation energy for the most sensitive Arrhenius theory and activation energy are intended to place a safety related material within a component. However, the type test specimen in a reasonable facsimile of the degradation to be seen in process for determining/selecting activation service when installed in its application in a nuclear power plant. energies for the various materials within a component should consider the information Recommendation provided in the proposed RG.
Section C. 1, j (3) should be deleted.
The staff finds that no changes to DG-1361 were necessary as a result of this comment.
James Section Comment 10 The staff disagrees with the comment.
Gleason C.1.k 14
Commenter Section of Specific Comments NRC Resolution DG-1361 Section C. 1, k (2) is confusing and unnecessary as it states: Electric The information included in Section C.1.k(2) equipment located in an area where rapid pressure changes are postulated of DG-1361, which is from RG 1.89, Rev. 1, simultaneously with the most adverse relative humidity should be qualified to Section C.3(b), is supplemental information demonstrate that the equipment seals and vapor barriers will prevent moisture that is intended to provide further clarity with from penetrating into the equipment to the degree necessary to maintain regard to type testing. Therefore, the staff did equipment functionality. not delete Section C.1.k(2).
IEC/IEEE Std. 60780-323, Edition 1, 2016-02 identifies interfaces and seals In areas where there are no pressure changes as elements to be identified and maintained as part of qualification and the and only humidity, the staff finds that the equipment seals and vapor barriers, when required to ensure the safety IEC/IEEE 60780-323-2016 guidance is function performance must operate properly in the environment in which the adequate.
equipment is being qualified.
The staff concludes that no changes to DG-The highlighting of equipment seals and vapor barriers, only where rapid 1361 were necessary as a result of this pressure changes are postulated simultaneously, overlooks applications where comment.
seals perform safety functions when no pressure variations are requirements.
Recommendation Section C. 1, k (2) should be deleted.
James Section Comment 11 The staff partially agrees with the comment.
Gleason C.1.k Section C. 1, k (4) is confusing and unnecessary as it states: Performance Based on potential confusion with monitoring characteristics that demonstrate the operability of equipment should be performance characteristics, the staff modified verified before, after, and periodically during testing throughout its range of Section C.1.k(4) to clarify that verifying required operability. Variables indicative of momentary failure that prevent performance characteristics may need to be the equipment from performing its safety function (e.g., momentary opening performed periodically, as applicable and of a relay contact) should be monitored continuously to ensure that depending on the equipments safety function.
momentary failures (if any) have been accounted for during testing. For long-term testing, however, monitoring during periodic intervals may be used if The staff disagrees that this position represents justified. a forward fit, as explained in the response to comment 4.
10CFR50.49 j (2) states that equipment must: Meets its specified performance requirements when it is subjected to the conditions predicted to 15
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Section Comment 12 The staff agrees with the comment. Section Gleason C.1.n Section C. 1, n (1) is confusing and unnecessary as it states: C.1.n(1) can be deleted due to it being
- 1) A double-transient should be used with equipment that may be vulnerable unnecessary to provide a regulatory position on to thermal binding from different expansion rates of materials during the the use of a double transient profile to address initial heatup. equipment that may be vulnerable to thermal binding from different expansion rates of Double-transient testing has never been a requirement of 10CFR50.49, DOR materials during initial heatup.
Guidelines, RG 1.89, or NUREG-0588.
A small discussion on use of double-transients There has been no requirement to evaluate equipment that may be vulnerable during testing has been added to the to thermal binding from different expansion rates of materials during the Background section of the RG for informative initial heatup. purpose only.
This constitutes a forward fit.
16
Commenter Section of Specific Comments NRC Resolution DG-1361 Recommendation Section C. 1, n (1): delete James Section Comment 13 The staff agrees that Section C.1.n(2) can be Gleason C.1.n Section C. 1, n (2) is confusing and unnecessary as it states: (2) The use of deleted due to it being unnecessary to provide a double transients could help offset tests where the ramp rate (initial regulatory position on the use of a double temperature rise) of the test is slower than the required profile. This is transient profile to offset tests where ramp rate commonly the result of test chamber and steam supply limitations. of the test is slower than the required profile.
There has been no use of double transients to offset test facility limitations on Section C.1.n(2) has been deleted as a result of the initial steam ramp. There is no logical formula for how an offset would be this comment.
calculated or credited.
This constitutes a forward fit.
Recommendation Section C. 1, n (2): delete James Section Comment 14 The staff agrees with the comment with regard Gleason C.2.c Section C. 2, c is confusing and unnecessary as it states: An additional to the fact that the regulations address the need stressor to be considered in the qualification of digital systems is smoke for SSCs to remain functional under postulated exposure from an electrical fire. For smoke exposure, important failure design basis events, but that smoke exposure is mechanisms are not only long-term effects such as corrosion, but also short- not addressed under 10 CFR 50.49. In term and perhaps intermittent malfunctions, such as leakage current. Smoke addition, guidance for addressing the effects of can cause circuit bridging and thus affect the operation of digital equipment. smoke is covered within Regulatory Guide Because the edge connections and interfaces are typically uncoated, the most 1.209, Guidelines for Environmental likely effect of the smoke is to impede communication and data transfer Qualification of Safety-Related Computer-between subsystems. RG 1.209 provides several references that detail the Based Instrumentation and Control Systems in effects of smoke exposure. Nuclear Power Plants.
Smoke has never previously been identified to be an environmental parameter As a result of this comment, Section C.2.c is or result of a Design Basis Accident. The new requirement to qualify for being revised to state:
smoke during a DBA constitutes a forward fit.
While not a consideration under 10 CFR Recommendation 50.49, an additional stressor that may need to 17
Commenter Section of Specific Comments NRC Resolution DG-1361 Section C. 2, c, starting at An additional stressor to be considered in the be considered in the qualification of digital qualification of digital systems is systems is smoke exposure from an electrical smoke exposure: delete 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 Section Comment 15 The staff partially agrees with the comment.
Gleason C.2.e Section C. 2, e is confusing, contradictory to 10CFR50.49, and unnecessary Item 1 in Section C.2.e is very general as it states: Considerations such as the following should be taken into information and while usually true, it may not be true in all situations. Item 2 is obvious and 18
Commenter Section of Specific Comments NRC Resolution DG-1361 account when determining the environment for which the equipment is to be need not be included in the guidance.
qualified: Therefore, items 1 and 2 have been deleted.
(C) equipment outside containment would generally see a less severe Item 3 is intended to address differences in the environment than equipment inside containment, (2) equipment timeframes for which equipment is required to whose location is shielded from a radiation source would generally be qualified. Some equipment may have safety receive a smaller radiation dose than equipment at the same distance functions that are to mitigate accidents and from the source but exposed to direct radiation, (3) equipment may not be necessary after the initial stages of required to initiate protective action would generally be required for a the accident, while other equipment may need shorter period of time than instrumentation required to operate during to be qualified to be operable in an extended and after an accident, and (4) analyses taking into account period in accident conditions. However, this is arrangements of equipment and radiation sources may be necessary to obvious and need not be stated in the guidance determine whether equipment needed for mitigation of design basis and therefore has been deleted.
accidents other than LOCA or high-energy line breaks (HELB) could be exposed to a more severe environment than the plant- specific Item 4 is intended address that some equipment LOCA or HELB environments. may have a different bounding accident than other equipment and different radiation sources This section has no significance to 10CFR50.49 qualification requirements. may be bounding for different equipment. The staff has decided to retain this information but Items (1) and (2) are obvious but are irrelevant since 10CFR50.49 requires relocated it to the Background section of the equipment in DBA environments to be qualified. guide as this is a more appropriate placement.
Item (3) discusses equipment performing protective action, but Therefore, the entire paragraph C.2.e has been 10CFR50.49 requires equipment be qualified for its safety function. removed from Section C and the following is included in the Background section of RG Item (4) discusses mitigation of design basis accidents instead of 10CFR50.49 1.89, Rev. 2:
Recommendation When determining the environment for which Section C. 2, e: delete 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 19
Commenter Section of Specific Comments NRC Resolution DG-1361 high-energy line breaks (HELB) could be exposed to a more severe environment than the plant-specific LOCA or HELB environments.
James Section Comment 16 The staff partially agrees with the comment.
Gleason C.2.f Section C. 2, f is confusing, contradictory to 10 CFR 50.49, and unnecessary The staff agrees that the phrase before testing as it states: is confusing, so the staff removed the phrase from Section C.2.f. The staff disagrees that Electric equipment to be qualified in a nuclear radiation environment should withstand before completion of its intended be exposed to radiation, before testing, that simulates the calculated integrated safety functions contradicts 10 CFR 50.49, as dose (normal and accident) that the equipment must withstand before this statement is carried forward from RG 1.89, completion of its intended safety functions. Rev. 1 and does not represent a new clarification.
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 NOTE: This position is now Section C.2.c of during normal operation over the installed life of the equipment, and the RG 1.89, Rev.2.
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 20
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Section Comment 17 The staff partially agrees with the comment.
Gleason C.2.f Section C. 2, f is confusing, contradictory to 10 CFR 50.49, and unnecessary The staff agrees that Section C.2.f should be as it states: In 10 CFR 100.11, Determination of exclusion area, low revised to remove confusion and to more population zone, and population center distance (Ref. 36), the NRC provides clearly describe the staffs position that the RG criteria for evaluating the radiological aspects of the proposed site. A footnote 1.183 accident scenarios may be used in to 10 CFR 100.11 states that the fission product release assumed in these assessing accident EQ radiation doses.
evaluations should be based upon a major accident involving substantial However, the accident scenarios and meltdown of the core with subsequent release of appreciable quantities of assumptions developed for the purposes of fission products. The NRC cites Technical Information Document (TID) reactor siting have been used for assessing the 14844, Calculation of Distance Factors for Power and Test Reactor Sites total integrated radiation dose for EQ in nearly (Ref. 37), in 10 CFR Part 100, Reactor Site Criteria, as a source of further all currently licensed facilities. In addition, guidance on these analyses. Although initially used only for siting NRC regulatory guides provide one way to evaluations, the TID 14844 source term has been used for design- basis meeting the requirements, and the RG revision applications, such as EQ of equipment under 10 CFR 50.49. Regulations in is not requiring any changes to existing 10 CFR 50.67, Accident source term, allows licensees to revise the accident licensees. Applicants and licensees may source term used in design-basis radiological consequence analyses. propose alternative methods to meet NRC requirements.
10CFR50.49 states: (4) Radiation. The radiation environment must be based on the type of radiation, the total dose expected during normal operation over As a result, Section C.2.f has been rewritten as the installed life of the equipment, and the radiation environment associated follows:
with the most severe design basis accident during or following which the equipment is required to remain functional, including the radiation resulting Electric equipment to be qualified in a nuclear radiation environment should be exposed to 21
Commenter Section of Specific Comments NRC Resolution DG-1361 from recirculating fluids for equipment located near the recirculating lines radiation that simulates the calculated and including dose-rate effects. integrated dose (normal and accident) that the equipment must withstand before completion Since this section uses the radiation from these evaluations should be based of its intended safety functions. Cobalt 60 or upon a major accident involving substantial meltdown of the core with cesium 137 would be acceptable gamma subsequent release of appreciable quantities of fission products and not the radiation sources for EQ.
radiation environment associated with the most severe design basis accident, it exceeds the requirements of 10CFR50.49 and is a forward fit. As required in 10 CFR 50.49(e)(4), the radiation environment must be based on the Recommendation total dose expected during normal operations Section C. 2, f change Determination of exclusion area, . Radiological over the installed life of the equipment and the consequence analyses. radiation environment associated with the most severe design basis accident during or To: The radiation environment must be based on the type of radiation, the following which the equipment is required to total dose expected during normal operation over the installed life of the remain functional. In addition, GDC 4 requires equipment, and the radiation environment associated with the most severe that SSCs important to safety shall be designed design basis accident during or following which the equipment is required to to accommodate the affects and to be remain functional, including the radiation resulting from recirculating fluids compatible with the environmental conditions, for equipment located near the recirculating lines and including dose-rate including those associated with postulated effects. 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 22
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Section Comment 18 See the staffs response to comment 17.
Gleason C.2.f 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.
23
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Pg. 8 / Comment 19 The staff disagrees with the comment. The Parello Background Editorial: Reference is made in a few places in the main document to the term staff is neither defining the term Class 1E nor
/ 1st Class 1E but is not defined until Section C. (STAFF REGULATORY taking exception to this term in this RG. Per Paragraph GUIDANCE) footnote 3 in 10 CFR 50.49, the NRC still acknowledges the definition for Class 1E as Recommendation defined in IEEE Std. 323-1974.
Recommend defining Class 1E as safety classification of the electrical equipment and systems per IEC/IEEE60780-323 on Page 8 / Background / Furthermore, this definition remains the same 1st Paragraph. 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 Pg. 10 / Comment 20 See the staffs response comment 2.
Parello Section Section C.1.b. states it provides a description and definition for the term C.1.b. 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 24
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Pg. 10 / Comment 21 See the staffs response to comment 3.
Parello Section A change in the definition for qualified life is not needed. The use of C.1.c. 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 James Pg. 10 / Comment 22 See the staffs response to comment 4.
Parello Section Equipment service life is the actual period of time the equipment is in C.1.d. 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.
25
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Pg. 11 / Comment 23 See the staffs response to comment 4.
Parello Section Note: The definition for service life provided is the same as in IEC/IEEE C.1.d. / 1st 60780-323.
Paragraph Recommendation Recommend deleting the first paragraph: Therefore, the following definition of service life should be used James Pg. 11 / Comment 24 The staff disagrees with the comment that Parello Section Environmental and operational aging of equipment important to safety to the Section C.1.e be deleted. Rather, the staff has C.1.e. end of its service life in a mild environment is required by IEC/IEEE 60780- edited the section as described above in the 323 if it is determined that the equipment has significant aging mechanisms response to comment 5. The purpose of DG-that impacts the ability of the equipment to perform its safety function(s) prior 1361 is to describe an acceptable approach for to Design Basis Events (DBE). In a mild environment a seismic event is a meeting regulatory requirements for DBE. Examples of equipment aging mechanisms in a mild environment prior environmental qualification of electrical to DBE are: wear, vibration, thermal and radiation as a function of time. equipment important to safety and to provide guidance for addressing environmental Recommendation stressors affecting the long-term reliability of Recommend deleting Section C.1.e. electrical equipment.
Pg. 12 / The staff partially agrees with the comment.
Comment 25 Section The staff agrees that doses less than 103 rad for C.1.h.(2). This section should be updated constant with Staff Position 2 (Page 16 / electronic equipment and 104 rad for other Section 2.c.) for defining a mild radiation environment. The Staff equipment generally does not have a significant considers a mild radiation environment for electronic equipment to be a impact on equipment performance. This is total integrated dose less than 10 gray (Gy) (103 rad) and a mild consistent with the staff position that total radiation environment for other equipment to be less than 100 Gy (104 integrated doses below these levels may be rad), to be acceptable.) considered to be located in mild radiation Recommendation environments.
Recommend the following update to Section C.1.h.(2):
26
Commenter Section of Specific Comments NRC Resolution DG-1361 The staff also agrees that the RG need not Electric equipment that may be exposed to low-level radiation doses include a statement indicating that low-level (electronic equipment to be a totalintegrated dose less than 10 gray (Gy) radiation should generally not be exempt from (103 rad) and other equipmentless than 100 Gy (104 rad)) should not qualification testing.
generally be considered exempt from radiation qualification testing.
Exceptions for higher doses may be based on qualification by analysis However, while the staff agrees that equipment supported by test data or operating experience that verifies that the dose receiving total integrated doses less than the and dose rates will not degrade the operability of the equipment below specified values generally does not require acceptable values. 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 Pg. 12 / Comment 26 The staff agrees with the comment. The staff Parello Section Information presented regarding aging may be better suited to be with the changed the reference from Section 7.3.2 to C.1.j. aging details presently in Clause 7.4.1.9.3 (Age Conditioning). Section 7.4.1.9.3.
Recommendation Recommend changing Section 7.3.2 to Section 7.4.1.9.3.
James Pg. 13 / Comment 27 The staff agrees with the comment. Section Parello Section The chemical spray or demineralized water spray during design basis event C.1.k(5) has been revised as follows:
C.1.k.(5) (DBE) testing needs to be conservatively injected after the peak of the environmental profiles (temperature, pressure). Depending on the nuclear Chemical spray or demineralized water spray facility, chemical spray or demineralized water spray may be initiated at a that is representative of service conditions 27
Commenter Section of Specific Comments NRC Resolution DG-1361 time prior to reaching the peaks of the postulated DBE environmental profile. should be incorporated during simulated event During DBE testing if the spray is initiated prior to reaching the peak of the testing after the test chamber reaches the DBE profile then the initial profile ramp and peak may not be met. maximum pressure and temperature conditions that would occur when the spray systems Recommendation actuate.
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 testchamber reaches the maximum at pressure and temperature conditionsthat would occur when the spray systems actuate.
James Pg. 14 / Comment 28 See the staffs response to comment 12.
Parello Section This section requested Clause 7.4.10 be updated with the following: A C.n.(1) 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)
James Pg. 14 / Comment 29 See the staffs response to comment 13.
Parello Section It is unclear how the use of a double-transient will offset tests where the ramp C.n.(2) 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 28
Commenter Section of Specific Comments NRC Resolution DG-1361 conservatively represent the initial ramp of a single-transient DBE that cannot be met.
James Page 16 / Comment 30 See the staffs response to comment 14.
Parello Section 2.c. 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.
James Page 19 / Comment 31 The staff agrees with the editorial comment, Parello References Editorial: Reference 9 and 10 are out of order has they appear in the main items 9 and 10 have been switched in the
/ Ref. 9. body of the document. Reference list to match the order in which they And Ref. appear in the main body of the document.
- 10. Recommendation Change Reference 9 to Reference 10 and vice-versa.
James Page 21 / Comment 32 The comment is no longer relevant as 10 CFR Parello References Reference 36 should be Chapter 11 of 10 CFR 100 has identified on Page 17 100.11 is no longer referenced in RG 1.89,
/ Ref. 36. (Section 2.f. / 2nd Paragraph Rev. 2 based on staff resolution of comment
/ 1st Sentence). The title for Chapter 11 is also missing. 17.
Recommendation 29
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Section Comment 33 The staff agrees with the comment. A Weinacht of C.1.a Regulatory Position C.1.a should be deleted. It is not necessary to provide clarification of the definition of end Curtis- clarification of the definition of end condition in order to use IEC/IEEE condition is unnecessary, and Section C.1.a Wright 60780/323 to meet 10 CFR 50.49. The proposed clarification treats end has been deleted.
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.
Rick Section Comment 34 See the staffs response to comment 2.
Weinacht of C.1.b Regulatory Position C.1.b should be deleted or reworded to simply refer to Curtis- the 10CFR50.49 definition of important to safety. The regulation is already Wright 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.
Rick Section Comment 35 See the staffs response to comment 3.
Weinacht of C.1.c Regulatory Position C.1.c. should be deleted or reworded. Restating text from Curtis- 10CFR50.49 is entirely unnecessary. It appears that this regulatory position is Wright attempting to provide clarification of the definition of safety function.
30
Commenter Section of Specific Comments NRC Resolution DG-1361 Defining safety function does not require a change or clarification to the definition of qualified life.
Rick Section Comment 36 See the staffs response to comment 4.
Weinacht of C.1.d Regulatory Position C.1.d should be deleted. The definition in Section 3.22 of Curtis- IEC/IEEE 60780- 323 makes no implication of aging effects. The proposed Wright 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.
Rick Section Comment 37 See the staffs response to comments 5, 14, and Weinacht of C.1.e Regulatory Position C.1.e should be deleted or reworded. It is not necessary 30.
Curtis- to amend the Standard to limit its scope to match that of 10CFR50.49. If Wright 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.
Rick Section Comment 38 See the staffs response to comment 6.
Weinacht of C.1.f Regulatory Position C.1.f creates regulatory confusion instead of clarification Curtis- by implying that using the new qualification methodologies may not meet Wright 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.
Rick Section Comment 39 Weinacht of C.1.j Regulatory Position C.1.j.(3) presents new requirements for justification of The staff disagrees with the comment. The activation energies that has rarely been met in past qualification efforts, will additional guidance provided in Section 31
Commenter Section of Specific Comments NRC Resolution DG-1361 Curtis- increase the time and cost of qualification without a substantial increase in C.1.j.(3) adds clarity for defining, justifying, Wright product quality or capability, and fails to recognize the great amount of and documenting activation energy and engineering judgement used and needed to establish a qualified life. The continues to allow the use of engineering regulatory position should end after the first sentence. judgement when establishing qualified life.
See the staffs response to comment 9 for additional information on the staffs position on activation energy.
Rick Definitions Comment 40 See the staffs response to comment 33.
Weinacht of end condition - It is agreed that the definition of end condition, defined in Curtis- Section 3.10 of IEC/IEEE Standard 6078-323, is synonymous with end-of-Wright installed life condition as described in 10CFR50.49(e)(5). The clarification of Regulatory position C.1.a is not necessary.
Rick Definitions Comment 41 See the staffs response to comment 2.
Weinacht of important to safety - Regulatory Position C.1.b does not provide a definition Curtis- of important to safety. IEC/IEEE STD. 60780-323, EDITION 1, 2016-02 and Wright 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:
32
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Definitions Comment 42 See the staffs response to comment 35.
Weinacht of Qualified Life - A definition is not provided in 10CFR 50.49, 10 CFR 50.2 or Curtis- the NRC Basic References Glossary. Regulatory position C.1.c of DG-1361 Wright 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 33
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Definitions Comment 43 See the staffs response to comment 4.
Weinacht of Service Life - IEC/IEEE STD. 60780-323, EDITION 1, 2016-02 Section 3.22 Curtis- provides a definition of service life. EPRI TR-100844 provides a nearly Wright 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.
Rick Comment 44 The staff partially agrees with the comment.
Weinacht of Significant Aging Mechanism - Section 3.24 of IEC/IEEE STD. 60780-323, The staff agrees that 10 CFR 50.59(e)(5)
Curtis- EDITION 1, 2016-02 provides the following definition of significant aging requires that consideration must be given to all Wright mechanism: ageing mechanism that, under normal and abnormal conditions, significant degradation that can have an effect causes degradation of equipment that progressively and appreciably renders on the functional capability of the equipment.
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 The staff does not agree that a regulatory that when significant aging mechanisms are identified, suitable age position needs to be created since the staff is conditioning shall be included in the type test. This implies if the aging not taking a position contrary to the definition effects assessment determines aging mechanisms are not significant, age of significant aging mechanism in IEC/IEEE conditioning for these aging mechanisms does not need to be included. Std. 60780-323, Edition 1, 2016-2 nor the information the comment references in Section 10 CFR 50.49(e)(5) requires that consideration must be given to all 7.4.1.9.1.
significant degradation that can have an effect on the functional capability of the equipment. A regulatory position should be provided that acknowledges The staff recognizes that it would be useful to that aging is only required for significant aging mechanisms as defined in include an updated table similar to DOR table Section 3.24 of IEC/IEEE STD. 60780-323, EDITION 1, 2016-02. During the C-1 but with 60-80 year information; however, 34
Commenter Section of Specific Comments NRC Resolution DG-1361 EQ Design Bases Assurance Inspections (DBAIs), some inspectors this information is currently unavailable and questioned if the wording of IEEE Std. 323-1974, Section 6.3.2(4) would require extensive research. Also, while specifically stating the conditions under which radiation aging would not need the staff agrees that guidelines for adequate to be included as part of aging, precluded a similar approach for other types justification of exempting pre-aging could be of aging, such as thermal aging. Memorandum from H. R. Denton, Director useful, the staff is not currently prepared to ONRR, to Commissioner Kennedy dated 8/24/1979 comparing the 1971 and provide a regulatory position or guidance for 1974 versions of IEEE Std. 323 (ADAMS Accession No. 7909210029), what constitutes adequate justification of clearly demonstrates that the NRC Staff interpretation has always been that exempting pre-aging.
consideration and inclusion of aging effects is only required if the aging effect is significant, stating: No revisions were made to DG-1361 as a result The staff guidelines mentioned above will require that aging be considered, of this comment.
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.
Rick Section C Comment 45 The staff agrees with this comment. The staff Weinacht of DG-1361 Regulatory Position C.1.d states that shelf life can adversely impact agrees to delete Sections C.1.d and C.2.a(3) as Curtis- qualified life and Regulatory Position C.2.a(3) states the shelf life should be neither service life nor shelf life is directly Wright addressed for potential impact on qualified life of equipment that was in associated with 10 CFR 50.49 since the main utility stock prior to February 22, 1983 and may be used as replacement focus of 10 CFR 50.49 is to establish qualified equipment in lieu of upgrading. Evaluation of shelf lifes impact on qualified life of electrical equipment, which can be life as an environmental qualification requirement is a new regulatory different from service or shelf life.
interpretation that needs to be evaluated as a backfit. These new regulatory Furthermore, service and shelf life are positions imply that the evaluations should be incorporated into addressed elsewhere in the regulations (e.g.,
environmental qualification documentation. While in some mostly rare cases, Appendix B to 10 CFR Part 50).
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 35
Commenter Section of Specific Comments NRC Resolution DG-1361 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 36
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Secdtion Comment 46 See the staffs response to comment 6.
Weinacht of C.1.f Regulatory Position C.1.f states that condition monitoring and associated Curtis- condition-based qualification methodologies in section 6.3 of IEC/IEEE Wright 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.
Rick Section Comment 47 The staff partially agrees with the comment.
Weinacht of C.1.h Regulatory position C.1.h. should incorporate beta dose reduction methods The guidance related to beta radiation has been Curtis- from Section 4.1.2 of the DOR Guidelines and include the allowance to updated in Section C.1.h as discussed in the Wright remove the requirement for additional radiation margin if approved, response to comment 7. The staff notes that conservative dose calculations methods are used as stated in NUREG-0588, much of the guidance contained in the DOR Guidelines was specific to the TID-14844 37
Commenter Section of Specific Comments NRC Resolution DG-1361 Category I, Section 1.4 and RG 1.89, Revision 1, regulatory Position C.2.c(6). source term and assumptions specified in the Specifically, the regulatory guide should confirm: DOR Guidelines. In addition, much of the
- Beta dose may be reduced by a factor of 10 for the first 30 mils of DOR Guidelines guidance discussed in the cable insulation and an additional factor of 10 for the next 40 mils of comment was not carried forward into previous insulation. versions of RG 1.89 or RG 1.183, Revision 0.
- Cables arranged in cable trays inside of containment shall be assumed For these reasons, that guidance is not being to be exposed to half the beta dose plus the gamma dose at the containment included in Revision 2 of RG 1.89. However, centerline. licensees do not need to change their approach
- Equipment shall be considered qualified, without any additional for EQ due to the revision to RG 1.89 and radiation margin, if qualified to radiation doses using the methods of previous guidance can continue to be used by Appendix D. licensees and applicants as a method acceptable
- If analysis shows that beta dose to sensitive equipment internals is to the NRC staff for demonstrating compliance less than or equal to 10% of the gamma dose, the equipment is considered with the regulations when appropriately qualified if qualified to the gamma dose. applied and applicable. As an example, RG 1.183, Rev. 0, indicates that either the TID-The methods for addressing beta radiation are based on sound principles that 14844 or the alternative source term may be are irrespective of the regulatory basis for qualification (i.e., DOR/NUREG- used for equipment qualification, even for 0588/10CFR50.49). those plants that use the alternative source term for control room and public dose. However, Regulatory position C.1.h.(2) should provide more detailed guidance about RG 1.183, Rev. 0, is only applicable within acceptable radiation exemption analysis (threshold) and pedigree of test certain parameters, for example, plants with data. A table similar to Table C-1 of the DOR Guidelines detailing agreed peak burnups up to 62,000 MWD/MTU and upon radiation threshold and allowable levels would be very helpful in clearly plants with 5% enrichment. If a plant were to communicating the amount and pedigree of additional data that is necessary make a design change that is beyond the to exempt radiation. At a minimum, Table C-1 of the DOR Guidelines should bounds of RG 1.183, Revision 0 (such as be validated as an acceptable reference. 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.
38
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Section Comment 48 See the staffs response to comments 8, 67, 96, Weinacht of C.1.j Regulatory position C.1.j(1) provides statements intended to supplement the 121, 136 and 165 regarding comments on Curtis- guidance of Section 7.3.2 of IEC/IEEE Std. 60780-323, Edition 1, 2016-02. Sections C.1.j(1) and C.2.d of DG-1361.
Wright 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.
39
Commenter Section of Specific Comments NRC Resolution DG-1361 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 40
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Section Comment 49 See the staffs response to comment 9.
Weinacht of C.1.j DG-1361, Section C.1.j(3) states Activation energy values should be based Curtis- on the testing of the specific compound used in the equipment and on the Wright 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 41
Commenter Section of Specific Comments NRC Resolution DG-1361 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.
42
Commenter Section of Specific Comments NRC Resolution DG-1361 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 43
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Section Comment 50 The staff partially agrees with the comment.
Weinacht of C.1.m Regulatory Position C.1.m states: Section 7.4.1.9.3 of IEC/IEEE Std. 60780- The point of carrying forward the statement Curtis- 323, Edition 1, 2016-02, should be supplemented with the following: For from IEEE 323-1974 was to provide an option Wright insulating materials, a regression line (IEEE Std. 101, IEEE Guide for the for using statistical analysis of thermal life test Statistical Analysis of Thermal Life Test Data (Ref. 32)), may be used as a data. It was not intended to imply that a basis for selecting the aging time and temperature. Sample aging times of less regression line alone forms the basis for the than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> should not be used. aging time and aging temperature.
This regulatory position is supplementing the current standard with an exact Nonetheless, the staff modified this statement statement from the 1974 version of IEEE 323-1974. In practice, this to note that it applies to organic materials and supplemental statement has very little impact on qualification practices. is not limited to insulating materials.
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 The staff also clarified the guidance to read as Arrhenius methodology and use of regressions lines only applies to insulating follows: A regression line alone does not form materials. a basis for the aging time and aging 44
Commenter Section of Specific Comments NRC Resolution DG-1361 A regression line alone does not actually form a basis for the aging time and temperature. This approach provides an aging temperature. It provides an activation energy proportional to the slope activation energy proportional to the slope of of the regression line that can be used to determine the amount of time at the the regression line that can be used to aging temperature to cause thermal aging equivalent to aging that would determine the amount of time at the aging occur during the desired service life at the service temperature. The aging temperature to cause thermal aging equivalent time and aging temperature are not a point on the regression line. Once the to aging that would occur during the desired activation energy is determined, an aging time can be calculated for an life at the expected temperature. The aging assumed aging temperature or an aging temperature can be calculated for an time and aging temperature are not a point on assumed aging time. There are countless aging time and temperature the regression line. Once the activation energy combinations that can be determined from a regression line. is determined, an aging time can be calculated This regulatory position should be deleted, or additional guidance should be for an assumed aging temperature or an aging provided for materials that are not insulating materials. temperature can be calculated for an assumed aging time.
Rick Section Comment 51 See the staffs response to comment 9.
Weinacht of C.1.j Regulatory Position C.1.j.(3) states data extrapolations should be minimized Curtis- by using activation energy values within the temperature range of interest. The staff partially agrees with the comment.
Wright And the activation energy should be selected based on the temperature range The staff agrees this information could be of the equipment in service. While these statements are consistent with useful, however, the staff disagrees with guidance in the relevant IEEE Standards, they fail to provide guidance against incorporating this information into the revised which acceptable extrapolation and activation energy selection can be judged. regulatory guide since specific acceptance Terms such as range of interest, applicable temperature range, and good criteria for activation energy may not be fit are too vague to allow objective agreement if the activation energy is available because it is material dependent and adequately justified. Many Electrical Insulating Material test programs based on specific failure modes.
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 45
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Section Comment 52 See the staffs response to comments 12 and Weinacht of C.1.n Regulatory Position C.1.n discusses the use of double-transient test profiles, 13.
Curtis- explaining the outdated stipulation for using double-transient test profiles as a Wright 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 46
Commenter Section of Specific Comments NRC Resolution DG-1361 temperature, so long as it is less than one minute or so, is of no consequence to qualification test results Rick Editorial Comment 53 The staff agrees with the comment. RG 1.140 Weinacht of Page 4 The description of RG 1.40 does not match the Reg Guide number and title. is Containment Isolation Provisions for Fluid Curtis- The description matches the RG 1.140 subject matter. Since RG 1.140 does Systems. RG 1.40 is Qualification of Wright not provide detail for qualifying equipment, it does not belong in this list. The Continuous Duty Safety-Related Motors for description of RG 1.40 needs to be revised to match the subject matter of RG Nuclear Power Plants, and endorses IEEE 1.40. 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 Editorial Comment 54 The staff agrees with the comment and made Weinacht of Page 7 The word provides should be changed to provide in the second the editorial change.
Curtis- Background paragraph (Chapter 11 and Appendix A..provide)
Wright Rick Editorial Comment 55 See the staffs response to comments 16, 17, Weinacht of Page 17 Regulatory Position C.2.f is less clear than Regulatory Position C.2.c.(5) of and 18.
Curtis- RG 1.89, Revision 1. The new regulatory position mixes in concepts covered Wright in other sections of the Draft Guide. This regulatory position would be clearer if it remained focused on considerations for determining the required 47
Commenter Section of Specific Comments NRC Resolution DG-1361 radiation dose. Calculational methods should be discussed in a separate regulatory position.
Comment Document 4: ML21041A127 William General Comment 56 - Summarized Based on this comment and others requesting Horin of 60 Day comment extension and public meeting request more time to provide comments on DG-1361, NUGEQ 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 General Comment 57 - Summarized See the staffs response to comment 56.
Horin of Information to support comment extension and public meeting request NUGEQ Comment Document 6: ML21110A055 Robert General Comment 58 The staff disagrees with the comment. When Konnik of We think it should be made clear that the latest revision of IEEE 323 RG 1.89, Rev. 2 is issued, both versions of RG IEEE (IEEE/IEC 60780-323) builds on IEEE 323-1974 and equipment qualified to 1.89 (which, together, endorse IEEE St. 323-IEEE/IEC 60780-323 would encompass qualification to IEEE 323-1974. In 1974 and IEC/IEEE 60780-323-2016 with the forward of IEEE 383-2003 it states that Electrical equipment qualified in noted clarifications/exceptions) will provide accordance with either IEEE 323-1974 or IEEE 323-1983 will meet the acceptable methods of meeting the NRCs requirements of IEEE 627-1980 which provide the basic principles for design requirements. Therefore, the comments qualification for all safety systems equipment for use in Nuclear Power recommendation isnt necessary. Furthermore, Generating Stations. This revision to IEEE 323-1974 was made to clarify its the NRC has not officially endorsed IEEE Std.
requirements and impose no additional requirements for qualifying Class 1E 48
Commenter Section of Specific Comments NRC Resolution DG-1361 equipment, The 2003 version of IEEE 323 incorporated additional 627. The staff may consider endorsing this information and clarified several areas that are outlined in the introduction standard in the future.
which include the use of IEEE 323 for qualification of equipment in mild environments when desired, design basis event nomenclature, updated test No changes to DG-1361 were made as a result margins, EMI/RFI, and the use of qualified condition. Similarly, in the of this comment.
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.
Robert Pg. 10 / Comment 59 See the staffs response to comment 33.
Konnik of Section Section 3.10 is a general definition of end condition, which is as stated the IEEE C.1.a. 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 installedlife (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 49
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Pg. 10 / Comment 60 See the staffs response to comment 4.
Konnik of Section Equipment service life is the actual period of time the equipment is in IEEE C.1.d. 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.
Robert Pg. 11 / Comment 61 The staff partially agrees with the comment.
Konnik of Section Environmental and operational aging of equipment important to safety to the IEEE C.1.e. end of its service life in a mildenvironment is required by IEC/IEEE 60780- The staff agrees with the comment that 323 if it is determined that the equipment has significant aging mechanisms environmental stressors like wear, vibration, that impacts the ability of the equipment to perform its safety function(s) thermal, and radiation are examples of aging prior to Design Basis Events (DBE). In a mild environment a seismic event mechanisms in a mild environment and that is a DBE. Examples of equipment aging mechanisms in a mild environment environmental qualification of electrical prior to DBE are: wear, vibration,thermal and radiation as a function of time. equipment in a mild environment is beyond the scope of 10 CFR 50.49, However, the staff Recommendation disagrees with the recommended deletion of Recommend deleting Section C.1.e. 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 50
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Pg. 12 / Comment 62 See the staffs response to comment 26.
Konnik of Section Information presented regarding aging may be better suited to be with the IEEE C.1.j. 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.
Robert Pg. 12 / Comment 63 See the staffs response to comment 26.
Konnik of Section Section 7.4.1.9.3 of IEEE/IEC 60780-323 already states that age conditioning IEEE C.1.j.(1) should consider synergistic effects.
Recommendation Recommend changing Section 7.3.2 to Section 7.4.1.9.3.
Robert Pg. 12 / Comment 64 See the staffs response to comment 26.
Konnik of Section Section 7.4.1.9.3 of IEEE/IEC 60780-323 already discusses the maximum IEEE C.1.j.(2) 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 51
Commenter Section of Specific Comments NRC Resolution DG-1361 Refer to Section 7.4.1.9.3 of IEEE/IEC 60780-323 if this is needed.
Robert Pg. 12 / Comment 65 The staff partially agrees with the comment.
Konnik of Section Section 7.4.1.9.3 of IEEE/IEC 60780-323 already discusses acceleration of The staff disagrees with deleting the suggested IEEE C.1.j.(3) aging and appropriate documentation. Ideally you would want to establish the sentence as it provides information that should activation energy using the temperatures that you will operate the equipment be considered when determining activation in, but this would take at least 60 or 80 years (if this is the expected life), but energy values.
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 The staff modified Section C.1.j(3) to include reasonable time. references to IEEE Std. 98-2016 and IEEE Std.
99-2019 as follows:
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 IEEE Std. 98-2016, IEEE Standard for the identify the specific material compound used in the equipment. It is important Preparation of Test Procedures for the Thermal though to use a conservative activation energy in this case. The industry has Evaluation of Solid Electrical Insulating used generic materials activation energies as an acceptable method when the Materials, (Ref. XX) and IEEE Std. 99-2019, exact compounds cannot be determined. The industry has judged the lowest IEEE Recommended Practice for the applicable generic published activation energy for materials aging programs Preparation of Test Procedures for the Thermal for many years as acceptable. Evaluation of Insulation Systems for Electrical Equipment, (Ref. XX) contains additional Recommendation technical information and criteria useful for Delete Of note, the activation energy should be selected based on the determining activation energy values.
temperature range of the equipment in service to ensure that the equipment However, the staff is not officially endorsing remains functional during and following a design-basis event. And replace these IEEE Standards in this RG.
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 See the staffs response to comments 9 and 161 design- basis event. for additional information pertaining to activation energy guidance included in the RG.
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.
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Commenter Section of Specific Comments NRC Resolution DG-1361 Robert Page 16 / Comment 66 See the staffs response to comment 14.
Konnik of Section This states from RG 1.209 that: An additional stressor to be considered in IEEE C.2.c. 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.
Robert Page 16 / Comment 67 The staff agrees with the comment in that the Konnik of Section Note that IEEE/IEC 60780-323 section 7.4.1.9.3 discusses the use of noted statements are relevant. However, it is IEEE C.2.d. Arrhenius aging and the sequence of age conditioning should consider not necessary to repeat the suggested statement sequential, simultaneous, and synergistic effects in order to achieve the worst from IEEE/IEC 60780-323-2016 because the state of degradation expected. What are the specific degradation processes proposed RG revision is not superseding the that are not amenable to preconditioning that could result in a common cause information in Section 7.4.1.9.3 of IEEE/IEC failure during design basis accidents? 60780-323-2016. Therefore, the statement remains applicable. In addition, the statement Recommendation suggested by the comment on specific It should noted that IEEE/IEC 60780- 323 states that preconditioning for degradation processes is already incorporated thermal should use the Arrhenius theory as well as the sequence of age in the Background section of the RG conditioning should consider sequential, simultaneous, and synergistic effects in order to achieve the worst state of degradation expected. Based on several similar comments, the staff determined that the information in Section Also state the specific degradation processes that are not amenable to C.2.d was less of a staff position and more of preconditioning that could result in a common cause failure during design general information that would be better suited basis accidents. in the Background section of the revised 53
Commenter Section of Specific Comments NRC Resolution DG-1361 regulatory guide. Therefore, the staff modified the regulatory guide to revise and relocate this information to the Background section.
Robert Page 17 / Comment 68 See the staffs response to comment 15.
Konnik of Section We do not know if equipment outside containment would generally IEEE C.2.e. 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.
Robert Page 19 / Comment 69 See the staffs response to comment 31.
Konnik of References Editorial: Reference 9 and 10 are out of order has they appear in the main IEEE / Ref. 9. body of the document.
And Ref.
- 10. Recommendation Change Reference 9 to Reference 10 and vice-versa.
Robert Page D- Comment 70 The staff partially agrees with the comment.
Konnik of 1/Appendix Note that IEEE 383 requires testing with normal dose and total integrated IEEE D dose. Additionally, to perform condition monitoring would need to perform IEEE 383, as endorsed by RG 1.211. is one tests without combining normal and accident dose. acceptable method to comply with 10 CFR 50.49 and similarly, DG-1361, Appendix D, Recommendation provides one acceptable method for accounting It should be noted that testing with normal dose is required by some standards for the radiation environment in environmental and to perform condition monitoring testing would need to be performed qualification. As such, condition-based without combining normal and accident dose. 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.
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Commenter Section of Specific Comments NRC Resolution DG-1361 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 Section A Comment 71 The staff does not agree with the comment. RG Bacanskas Current NRC licensees have gone through a rigorous process of qualification 1.89, Rev. 2 is applicable to currently operating over the past 43 years (1978-2021). Licensees were subject to a review of all licensees because the guidance could be useful qualification documentation in the early 1980s with subsequent staff to them to meet regulations that apply to meetings for corrective actions, follow up inspections on those corrective operating power plants. One such example is actions and in the past 5 years, re- inspection to validate the continuation of provided by the commenter: licensees that seek the program. Many licensees have gone through re-examination of their files a change to a commitment concerning the for methodology to support extension of their operating licenses. The existing subject matter of the RG. Although the RG licensing bases are firmly established for operating reactors and reactors applies to currently operating licensees, those currently under construction. Review of the dual logo standard shows no licensees are not required to use the guidance promise of burden reduction on existing licensees, and the wording within in the RG, and the RG does not require DG-1361 quotes existing regulations out of context and creates the potential licensees to take any actions.
for backlit/forward fit in many instances. There appears to be NO incentive for an operating reactor to change commitments or licensing bases to The RG has an exclusion for nuclear power incorporate either this revision of the IEEE standard or DG-1361 as written. plant licensees that have submitted certifications as required by 10 CFR Recommendation 50.82(a)(1) or 52.110(a) because those DG-1361 should thus be revised to indicate that it does NOT apply to existing licensees' facilities are no longer operating.
licensees unless a change in a licensing commitment is made. 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 55
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Section A Comment 72 The NRC staff agrees with the comment that Bacanskas DG-1361 fails to recognize the incorporation of 10CFR50.69 since the additional discussion of 10 CFR 50.69 would publication of 50.49, and where 50.49 is quoted for equipment under its scope be helpful in the RG. The staff added the in the DG. 50.69 specifically excludes EQ for RISC-3 components and the following information in the Applicable Commission even states that files such as those required for 50.49 are NOT Regulations section of the DG:
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 10 CFR 50.69, Risk-informed categorization confidence of a components capability to provide a low safety significant and treatment of structures, systems and design function. While 50.69 was published after 50.49, and 50.49 was not components for nuclear power reactors, states revised to address RISC-3 equipment, it should be identified in the Applicable in part that a holder of a license to operate a Regulations section of DG-1361. light water reactor (LWR) nuclear power plant under 10 CFR Part 50; a holder of a renewed Recommendation LWR license under 10 CFR Part 54; an Footnotes should be added to Staff Regulatory Guidance Section C.1.b applicant for a construction permit or operating indicating that RISC-3 components are not included within this position. This license under 10 CFR Part 50; or an applicant footnote should be repeated whenever the staffs regulatory position includes for a design approval, a combined license, or similar wording. It might be most simply addressed in the Introductory manufacturing license under 10 CFR Part 52; Discussion in the DG so that it is understood before looking at staff guidance may voluntarily comply with the requirements in the document. 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.
56
Commenter Section of Specific Comments NRC Resolution DG-1361 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 57
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Section A Comment 73 The staff agrees with the comment.
Bacanskas 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. The NRC revised the RG to remove references to testing and to add further clarity due to this Recommendation comment as follows:
These criteria should be listed and described separately so that it does not read that testing is only associated with Criterion III Design Control. 10 CFR Part 50, Appendix B, Quality Criterion III lists several methods of design verification and all are acceptable Assurance Criteria for Nuclear Power Plants to meet Criterion III. 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, 58
Commenter Section of Specific Comments NRC Resolution DG-1361 shipping, storing, cleaning, erecting, installing, inspecting, testing, operating, maintaining, repairing, refueling, and modifying.
Vincent Section A Comment 74 See the staffs response to comment 53.
Bacanskas 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.
Vincent Section C.1 Comment 75 See the staffs responses to comments 1, 3, and Bacanskas In Regulatory Guidance C.1, the staff throws in guidance related to service 4.
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.
Vincent Section Comment 76 See the staffs response to comment 6.
Bacanskas C.1.f 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 59
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Sectdion Comment 77 See the staffs response to comment 7.
Bacanskas C.1.h 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 60
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Section Comment 78 The staff partially agrees with the comment. 10 Bacanskas C.1.j Staff Regulatory Guidance C.1.j(1) - 10CFR50.49(e)(7) states: Synergistic CFR 50.49(e)(7) states that synergistic effects effects must be considered when these effects are believed to have a must be considered when these effects are significant effect on equipment performance. NUREG-0588, Part 1 refers believed to have a significant effect on one to NUREG/CR-0276 and NUREG/CR-0401 for synergistic effects. equipment performance. The guidance is for NUREG/CR-0276 says that there were no cable failures during this particular situations where there are known synergisms research program and that no significant functional or mechanical synergisms that could affect equipment qualification.
exist (Test Summary Sec. 2.2.4). NUREG/CR-0401 contains essentially the Given the existing guidance, as noted in the same paragraph from NUREG/CR-0276 and includes a section on ethylene comment, and absent any new information, the propylene insulation with PVC jacket from Savannah River where they staff decided the best approach was to carry indicate degradation occurred and they expect synergistic effects contributed. forward the staff position on synergistic effects The entire focus on this seems without value as little to no additional research from RG 1.89, Rev. 1.
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 See the staffs response to comment 8 for expectation that radiation followed by thermal aging is the preferred additional information on the staffs position sequence. This once again, raises regulatory confusion. In essence, the entire on synergistic effects.
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 61
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Section Comment 79 See the staffs response to comment 9 Bacanskas C.1.j Section C.1.j(3) discusses the use of an Activation Energy and imposes regarding Activation energy.
specific requirements that are absent from any Regulatory Document.
Specifically, Regulatory Guide 1.89, R1 states in C.5.(c): The aging This proposed guidance does not meet the acceleration rate and activation energies used during qualification testing and definition of backfitting in MD 8.4. RG 1.89, the basis upon which the rate and activation energy were established should Rev. 2 is voluntary guidance and represents be defined, justified, and documented. The section in DG-1361 not only adds one acceptable way to satisfy the applicable additional requirements [Backfit] but in some instances is technically NRC regulations. The NRC staff is not incorrect. imposing or expecting specific licensees to implement the guidance in RG 1.89, Rev. 2.
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 62
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Section C.1 Comment 80 The staff agrees with the comment that the RG Bacanskas While the DG provides a statement in the Staff Regulatory Guidance section should have a more definitive statement on applicability to equipment located in a Mild Environment, there should be regarding the fact that 10 CFR 50.49 does not a more definitive statement as position C.1 that 50.49 explicitly excludes require environmental qualification for equipment in a Mild Environment and paragraphs associated with this electrical equipment located in a mild equipment are excluded from this DG. 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 No page Comment 81 The staff disagrees with the comment. The Fosaaen of number This proposed revision exceeds the scope of the original stated purpose of RG scope of Revision 2 to RG 1.89 is consistent NuScale Generic 1.89, which is to ensure equipment important to safety remains functional with the purpose of 10 CFR 50.49. Any Comment during and following design basis accidents. Expanding the scope of this RG additional information incorporated into this could lead, as suggested by comments below, to unintended consequences RG provides guidance based on almost 40 such as back- fitting and forward-fitting implications. years of experience since the issuance of the previous version of this RG.
63
Commenter Section of Specific Comments NRC Resolution DG-1361 Recommendation Align RG with the scope and intent of 10 CFR 50.49, and clarify as needed to This proposed guidance does not meet the ensure the revision is within the scope of the stated purpose of 10 CFR 50.49. definition of backfitting or forward fitting in Resolutions for specific instances are identified in subsequent comments. 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 Pages 1-2, Comment 82 The staff partially agrees with the comment.
Fosaaen of applicable The applicable regulations list is confusing and potentially misleading.
NuScale regulations Separating into Part 50 and Part 52 creates the impression the requirements While some of the criteria apply to both Part 50 for different applicants/licensees are different. Note that design approval and 52 applicants and licensees, the Applicable applicants (SDA and DC) need only provide an equipment list per 10 CFR Regulations sections is meant to delineate the 50.49, while license applicants (OL, COL, and ML) must also describe the applicable regulations without specifying what EQ program and license holders must establish the program. Furthermore, the requirements each type of applicant or licensee introduction to the Part 52 regulations makes reference to design control (construction permit, operating license, measures, such as testing which is not supported by a regulation in Part 52. combined license (COL), manufacturing license, standard design approval, design Recommendation certification) must meet. Therefore, the staff is Consolidate the applicable regulations into a single list and clarify maintaining the separate lists of applicable Part applicability for various types of applicants. 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.
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Commenter Section of Specific Comments NRC Resolution DG-1361 Carrie Page 2, 3rd Comment 83 The staff partially agrees with the comment.
Fosaaen of bullet from Incorrect citation. Reads in part, . . . For a manufacturing license as defined NuScale bottom in 10 CFR 52.157, only electric equipment defined in 10 CFR 50.49(b) which The staff agrees that it did not correctly state is within the scope of the manufactured reactor must be included in the EQ the language related to 10 CFR 52.157 in the program. A manufacturing license is not defined by 10 CFR 52.157, and DG . The staff does not agree to add the that reference is unnecessary. 10 CFR 52.157 does require an ML applicant to recommended language suggested in the provide a list of electric equipment important to safety and would be comment. However, the staff revised the appropriate to address in a separate bullet. Applicable Regulations section to clarify the requirement within 10 CFR 52.157 for an Recommendation applicant for a manufacturing license to Revise to state: For a manufacturing license, only electric equipment defined provide a list of electric equipment important in 10 CFR 50.49(b) which is within the scope of the manufactured reactor to safety that is required by 10 CFR 50.49(d).
must be included in the EQ program.
Carrie Page 4, 3rd Comment 84 The NRC partially agrees with the comment.
Fosaaen of bullet from RG 1.180 is not relevant to environmental qualification under 10 CFR 50.49. The NRC agrees that the Regulatory Analysis NuScale top Keep the focus of this RG to satisfying the requirements of 10 CFR 50.49 by for DG-1361 referred to providing guidance for removing unrelated RGs that are beyond the scope of 50.49. demonstrating compliance with 10 CFR 50.49.
However, the Purpose section of DG-1361 Recommendation stated that the DG describes an approach that is Remove reference to RG 1.180 and other RGs that are not related to acceptable to the NRC staff to meet regulatory compliance with 10 CFR 50.49. 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.
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Commenter Section of Specific Comments NRC Resolution DG-1361 Carrie Page 7, 2nd Comment 85 The staff agrees that EPRIs Plant Support Fosaaen of paragraph The DG states: Engineering: Nuclear Power Plant Equipment NuScale from Chapter 11 and Appendix A to the Electric Power Research Institutes Qualification Reference Manual, Revision 1, bottom (EPRIs) Plant Support Engineering: Nuclear Power Plant Equipment is widely used by the nuclear industry.
Qualification Reference Manual, Revision 1, issued September 2010 (Ref. However, the staff is not prepared to endorse 26), provides a detailed regulatory history of electrical and mechanical this document beyond what was identified in equipment qualification. While the agency has not officially endorsed this the DG.
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 No changes to DG-1361 were made as a result electrical and mechanical equipment qualification. of this comment.
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.
Carrie Page 7, 2nd Comment 86 The staff agrees with the comment. The staff Fosaaen of paragraph In the Background section, the proposed RG states, For the purposes of revised the Background section to remove NuScale from this guide, the primary objective of qualification is to demonstrate that before and added clarification on design-bottom equipment important to safety can perform its safety function(s) without basis events as defined in 10 CFR experiencing common-cause failures before, during, and after applicable 50.49(b)(1)(ii).
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 66
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Page 8, first Comment 87 The staff partially agrees with the comment.
Fosaaen of paragraph, In the Background section, the proposed revision to the RG makes the NuScale last following statement: The qualification specifications in IEC/IEEE 60780- The staff agrees that one of the main purposes sentence 323, Edition 1, 2016-02, when met, demonstrate and document the ability of of 10 CFR 50.49 is to prevent environmentally equipment to perform safety function(s) under applicable service conditions, induced common cause failures of electrical including design-basis events, reducing the risk of common-cause equipment equipment important to safety. The staff also failure. This statement implies that the NRC is increasing the scope of 10 agrees that 10 CFR 50.49(d) and (e) require CFR 50.49 and RG 1.89 to more than design basis accidents, to envelope qualification to address design basis accidents ;
other applicable service conditions. however, the staff disagrees that it is expanding the scope of 10 CFR 50.49 because 10 CFR While aging is required as part of the EQ program, 10 CFR 50.49 is intended 50.49(b) states that electrical equipment to prevent environmentally-induced common cause failures of electrical important to safety covered by 10 CFR 50.49 is equipment important to safety during and following design basis accidents. safety-related equipment relied on to remain Note that 10 CFR 50.49(d) and (e) require qualification to parameters for functional during and following design basis DBAs and not other DBEs. events to ensure: (1) the integrity of the reactor coolant pressure boundary; (2) the capability to Recommendation shut down the reactor and maintain it in a safe Revise to state . . . perform safety function(s) during and following design shutdown condition; or (3) the capability to basis accidents by reducing the risk of common- cause equipment failure. 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.
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Commenter Section of Specific Comments NRC Resolution DG-1361 Carrie Comment 88 See the staffs response to comment 72.
Fosaaen of The proposed revision to the RG states, 10 CFR 50.49 requires safety-related NuScale 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.
Carrie Page 10, Comment 89 See the staffs response to comment 4.
Fosaaen of paragraph Reads in part, . . . however, the period before the operational phase of the NuScale 1.d, and SSC (i.e., shelf life) could also adversely impact the qualified life.
page 11 carryover 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.
Carrie Page 11, Comment 90 See the staffs response to comment 24 Fosaaen of paragraph e This paragraph recognizes that mild environments are beyond the scope of 10 NuScale CFR 50.49, but the proposed provision includes EMC and seismic in the list 68
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Page 11, Comment 91 The staff partially agrees with the comment.
Fosaaen of Paragraph g Paragraph g. lists methods acceptable to the NRC staff for calculating and The staff agrees with the proposal to NuScale establishing containment pressure and temperature envelopes. Subparagraph incorporate the following statement in the RG:
(2) states For pressurized water reactors (PWRs) with a dry containment, Containment pressure and temperature LOCA or MSLB containment environment should be calculated using environment should be calculated using codes CONTEMPT-LT or equivalent industry codes. Identifying a specific code that are consistent with the licensees design and its equivalent is unduly restrictive. Similar statements are in and licensing basis.
subparagraphs (3) and (4). Revise the guidance on acceptable codes and replace with statement technology- neutral discussion such as: Containment The staff revised DG-1361 to add the above pressure and temperature environment should be calculated using codes statement as a footnote to each instance where which are consistent with the licensees design and licensing basis. the phrase equivalent industry codes is used in Section C.1.g.
Recommendation Containment response methodologies are reviewed and The staff also revised the first bullet under accepted by the NRC as part of the application. 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.
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Commenter Section of Specific Comments NRC Resolution DG-1361 Carrie Page 12, Comment 92 The staff agrees with the comment. The Fosaaen of paragraph i Remove this paragraph. Electromagnetic conditions are generally clarification is not needed, and Section C.2.i NuScale independent of aging and design-basis events. Therefore, qualification can be from DG-1361 has been deleted. Further, Note established on a different sample than the sample subjected to aging and 1 within the IEC/IEEE 60780 standard clause design-basis events. As stated in comment 9, EMC is not in the scope of 10 7.4.1.8.c already includes a sentence which CFR 50.49. And it is not needed since the dual logo standard already contains states: For convenience, EMI/RFI this clarification in the note in Section 7.4.1.8.c. susceptibility testing and operational test under extreme conditions may be performed on a Recommendation separate test specimen.
Delete this paragraph Carrie Page 12, Comment 93 The staff disagrees with the comment. The Fosaaen of paragraph The guidance related to activation energy is restrictive and could result in intent of the guidance on activation energy was NuScale j.(3) requiring new tests to derive activation energy values in lieu of use of not to require new tests. The staff recognizes conservatively-established activation energy values that are justified as being that specific data from qualification test reports appropriate for the application. If followed as written, this restriction could may be stored elsewhere (e.g., at a vendor lead to unnecessary effort to evaluate activation energies without a benefit to facility, in a centralized file, etc.). The point of safety. this particular guidance is that the activation energy values need to be defined, justified, and Additionally, the paragraph states, The selected activation energy values documented as specified in 10 CFR 50.49 and should be traceable to a specific test report for which these values were that the specific material properties need to be established. Material properties that have historically been used in industry considered to determine an accurate activation do not all have traceable test reports, because some come from research energy value.
papers from labs and other facilities.
See the staffs response to comment 9 for Recommendation additional information on the staffs position Revise paragraph to allow use of conservatively-established activation energy on activation energy as it pertains to values that have been found to be acceptable, and to clarify need for traceable environmental qualification of electrical test reports. equipment.
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Commenter Section of Specific Comments NRC Resolution DG-1361 Carrie Page 16, Comment 94 The staff agrees with the comment.
Fosaaen of paragraph b The guidance related to commercial grade dedication is out of place in RG NuScale 1.89. Supply chain-related activities are beyond the scope of 10 CFR 50.49. The NRC revised the RG to remove references to RG 1.164 due to this comment and removed Recommendation Section C.2.b from DG-1361.
Remove guidance related to commercial grade dedication.
Carrie Page 16, Comment 95 The staff partially agrees with the comment.
Fosaaen of paragraph c This paragraph introduces requirements that are beyond the scope of 10 CFR The comment is correct regarding the NuScale 50.49. The scope of 50.49 is qualification for electric equipment in a harsh definition of a mild environment in 10 CFR environment, not mild. A mild environment is defined as an environment that 50.49(c). However, this does not mean that would at no time be significantly more severe than the environment that equipment located in an environment that would occur during normal plant operation, including anticipated operational would at no time be significantly more severe occurrences. than the environment that would occur during normal plant operation, including anticipated Therefore, by defining a mild radiation environment using a fixed threshold operational occurrences, need not be qualified.
without respect to the normal operating environment of electronic equipment, GDC 4 requires that equipment important to this paragraph c. conflicts with the terms of 10 CFR 50.49(c) and broadens safety shall be designed to accommodate the the scope of EQ. Furthermore, smoke from an electrical fire is not a condition effects of and to be compatible with the during or following a design basis accident, and therefore not within the environmental conditions associated with scope of the EQ program required by 10 CFR 50.49. normal operation, maintenance, testing, and postulated accidents. RG 1.89 uses the term Recommendation mild radiation environment to define a Delete paragraph c. as it conflicts with and exceeds the requirements of 10 radiation environment that is below the total CFR 50.49. 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 71
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Page 16, Comment 96 The staff partially agrees with the comment.
Fosaaen of paragraph d The draft guide introduces a new term, inverse temperature, without NuScale explaining what it is or how to comply. Inverse Temperature Effects is first NUREG/CR 7153, Expanded Materials published in NUREG/CR-7153, Volume 5. This publication was focused on Degradation Assessment (EMDA): Volume 5:
cables and cable systems only. Therefore, it is not established whether Aging of Cables and Cable Systems, Section inverse temperature effects does or does not impact other broad types of 3.3, Inverse Temperature, states that The elastomers. There is no existing NRC guidance on how to address inverse observed inverse temperature effect, where temperature effects. If Staff intend to address inverse temperature effects, it polymer degradation occurs more rapidly for should be fully and transparently considered by Staff in an appropriate constant dose rates as the combined regulatory action. The mention of inverse temperature effects in DG-1361 environment temperature is lowered, represents implies additional requirements outside the normal regulatory process. an example in which material aging and 72
Commenter Section of Specific Comments NRC Resolution DG-1361 lifetime prediction cannot be represented Additionally, Item (2) of paragraph d. states,. . . concurrent radiation and adequately by conventional approaches, such thermal aging or sequential aging, as well as the order of radiation and as the Arrhenius methodology. The inverse thermal aging, based on which produces the worst- case degradation; and . . . temperature effect is applicable to certain The phrase worst-case is ambiguous and inconsistent with requirements of XLPO and EPR insulation materials. The staff 10 CFR 50.49(e)(5) for aging to end-of-installed-life condition. This added the following definition on inverse requirement was not present in RG 1.89 Rev 1. This statement may create temperature to the Background section of RG unintended back-fit/forward-fit consequences. 1.89, Rev. 2:
Inverse temperature or reverse temperature Recommendation effect is where polymer degradation occurs Remove discussion about inverse temperature effects. more rapidly for constant dose rates as the Remove item (2). 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).
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Commenter Section of Specific Comments NRC Resolution DG-1361 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 Page 20, Comment 97 The staff agrees with the editorial comment Fosaaen of item 26 Incorrect citation. Reads in part. . . Electric Power Research Institute, (EPRI) and removed the incorrect reference to the NuScale Nuclear Energy Institute (NEI) EPRI/NEI Report No. 1021067. . This is an Nuclear Energy Institute in association with EPRI report only. Report No. 1021067 in the list of References in DG-1361.
Recommendation Delete NEI from report title Carrie Appendix Comment 98 The staff partially agrees with the comment.
Fosaaen of A The list of typical safety-related electric equipment or systems is typical NuScale only for operating power reactors (large, non-passive LWRs). Additionally Footnote 9, Appendix A, of DG-1361 states:
for such designs, containment combustible gas control is no longer required to In Title 10 of the Code of Federal Regulations be safety-related function (pursuant to revised 10 CFR 50.44). Emergency (10 CFR) section 50.49(b)(1), the NRC systems to achieve safe shutdown is also unclear, because beyond those identifies safety-related electric equipment as a systems already listed, it would appear to only include the system for residual subset of electric equipment important to safety heat removal. and defines it as the equipment that is relied upon to remain functional during and following Recommendation design-basis events to ensure (1) the integrity Clarify list is not applicable for passive designs and other new technologies, of the reactor coolant pressure boundary, (2) and revise combustible gas control and safe shutdown bullets. the capability to shut down the reactor and 74
Commenter Section of Specific Comments NRC Resolution DG-1361 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 75
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Appendix Comment 99 The staff partially agrees with the comment.
Fosaaen of B, Page B-1 This introduction to this appendix provides a confusing description for its As Appendix B to DG-1361 is carried forward NuScale regulatory basis, suggesting the example equipment is explicitly within the from the previous RG revision, the staff scope of 10 CFR 50.49. Rather, these are examples of non-safety-related disagrees that the regulatory basis description electric equipment whose failure under postulated environmental conditions is inaccurate or confusing. Nevertheless, the could prevent satisfactory accomplishment of the specified safety functions, staff updated the appendix to include the pursuant to 10 CFR 50.49(b)(2). information suggested by the comment from the previous revision of RG 1.89. This Additionally, this version of the proposed RG removed the following information was unintentionally omitted.
language from the previous version:
The following paragraph was added to Associated circuits, as defined in Regulatory Guide 1.75, "Physical Appendix B:
Independence of Electric Systems," need only be qualified to ensure that they will not fail under postulated environmental conditions in a manner that could Associated circuits, as defined in Regulatory prevent satisfactory accomplishment of safety functions by safety-related Guide 1.75, Physical Independence of Electric equipment. Systems, need only be qualified to ensure that they will not fail under postulated This statement accurately describes the purpose and acceptance criterion for environmental conditions in a manner that the qualification of the non-safety-related electric equipment within the scope could prevent satisfactory accomplishment of of 10 CFR 50.49(b)(2). safety functions by safety-related equipment.
Recommendation 76
Commenter Section of Specific Comments NRC Resolution DG-1361 Clarify the basis and intent of Appendix B. Re-insert deleted provision in RG 1.89 Rev 2.
Carrie Appendix Comment 100 See the staffs response to comment 91.
Fosaaen of C, Page C-1 The list of acceptable methods for calculating mass and energy releases could NuScale be misconstrued as limiting for applicants. In addition, the staff revised DG-1361 to include a footnote to Appendix C that states:
Mass and energy release methodologies are reviewed as part of plant licensing. This Appendix should be generically applicable to all reactor Mass and energy releases are developed using designs. For example, NuScales use of NRELAP5 code was found a methodology that is consistent with the acceptable by the staff. If a list of currently acceptable methods is to be licensing basis of the plant. The listed methods included, it should be clarified as only examples of methods the Staff have are examples for existing designs.
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.
Carrie Appendix Comment 101 The staff partially agrees with the comment.
Fosaaen of D, Generic Appendix D is dedicated to RG 1.183 which does not apply to all facilities.
NuScale Comment The DG makes no mention of source term for non-AST source term for EQ The staff intended to make it clear that other and is silent on the guidance in RG 1.195. source term methodologies may be used, as appropriate, for the radiation accident EQ dose.
Recommendation This includes RG 1.195. Appendix D has been Revise Appendix D to address both existing and new designs. 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 77
Commenter Section of Specific Comments NRC Resolution DG-1361 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 78
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Appendix Comment 102 The staff agrees with the comment.
Fosaaen of D, Page D- This discussion refers to the survivability period. Equipment survivability NuScale 1, has a defined meaning with respect to beyond design basis events (see 10 The second paragraph in Section D-2.1 has paragraph CFR 50.44 and RG 1.7), so the term may introduce confusion within the been revised to remove the term survivability D-2.1 context of 10 CFR 50.49 compliance. period and the sentence is reworded as, The period of exposure should be consistent with Recommendation the design basis event qualification for the EQ Revise to use a different term, such as post-accident operating time or mission equipment being evaluated. In addition, the time. staff removed the third sentence, which defined survivability period.
Carrie General Comment 103 The staff agrees with the comment. The staff Fosaaen of Appendix E was eliminated from this proposed version of RG 1.89. Appendix will include Appendix E from RG 1.89, Rev. 1 NuScale E contains valuable qualification documentation requirements and categories in RG 1.89, Rev. 2 since it provides useful for equipment. See RG 1.89 Rev 1, Page 1.89-17. information on qualification documentation for equipment.
Recommendation:
Restore Appendix E to RG 1.89 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 General Comment 104 See the staffs response to comment 58.
Fosaaen of Neither the joint logo nor the proposed RG revision addresses that, if a NuScale 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 79
Commenter Section of Specific Comments NRC Resolution DG-1361 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 General Comment 105 The staff partially agrees with this comment.
Fosaaen of A design specific review standard was issued to NuScale for the DCA The staff reviewed the NuScale design specific NuScale application. This DSRS 3.11 included additional guidance that was not in RG review standard (DSRS) 3.11 (ADAMS 1.89. Accession No. ML15355A455) to ensure generic aspects were addressed in the proposed Recommendation revision to RG 1.89. The NuScale DSRS 3.11 Conduct a reconciliation between the DSRS 3.11 and RG 1.89 to ensure the is applicable only to the NuScale design, and RG encompasses updated requirements. 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 80
Commenter Section of Specific Comments NRC Resolution DG-1361 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 General Comment 106 The staff recognizes that there is a reference to Fosaaen of 10 CFR 50.49 footnote 4 refers to RG 1.97 Rev. 2. In BTP 7-10 the staff a previous revision of RG 1.97 in the footnote NuScale provided guidance for RG 1.97 Rev. 3 and 4 as it relates to intent of 10 CFR for 10 CFR 50.49. A separate work task is 50.49. underway to revise and update NUREG-0800, Chapter 7 in its entirety, which includes, Recommendation among other things, BTP 7-10. Included in Revise BTP 7-10 to update 1.97 revision and provide staff interpretation for this activity is the task to review all references meeting 10 CFR 50.49. Include correct revision of RG 1.97, which is Rev 5. 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 81
Commenter Section of Specific Comments NRC Resolution DG-1361 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 General Comment 107 The staff agrees with the comment.
Fosaaen of 10 CFR Part 52 licensees must address ITAAC. The DG does not speak to NuScale ITAAC-related EQ. The DG does not address ITAAC related to EQ, and Part 52 licensees must address ITAAC Recommendation to obtain a 10 CFR 52.103(g) finding and load Revise RG 1.89 to reference RG 1.215 for EQ-related ITAAC closures. 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 82
Commenter Section of Specific Comments NRC Resolution DG-1361 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 General Comment-108 - Summarized See the staffs response to comment 56.
Fosaaen of 60 Day comment extension and public meeting request NuScale Comment Document 10: ML21042A003 Rick Comment-109 - Summarized See the staffs response to comment 56.
Weinacht of 60 Day comment extension and public meeting request Curtiss-Wright Comment Document 11: ML21050A360 William General Comment 110 - Summarized See the staffs responses to Comment Horin of This Comment Document was superseded by Comment Document 12 below Document 12 below.
NUGEQ (see ML ML23086C099).
Comment Document 12: ML21110A056 83
Commenter Section of Specific Comments NRC Resolution DG-1361 William General Comment 111 The staff disagrees with the comment. RG Horin of Consistency with the stated objective in Section 2 of the Regulatory Analysis 1.89, Rev. 2 provides a means of satisfying 10 NUGEQ for DG-1361, which states: CFR 50.49 and other regulations associated The objective of this regulatory action is to access the need to revise NRC with the environmental qualification of guidance and provide applicants with an updated method to demonstrate equipment as stated in the Applicable compliance with 10 CFR 50.49, Environmental qualifications of electric Regulations section of the RG. The staff added equipment important to safety for nuclear power plants. the following sentence in the Purpose section Keeping RG 1.89 specific to an acceptable method of meeting 10 CFR 50.49 of RG 1.89, Rev. 2 to clarify that the RG also results in consistency with RG 1.209, Guidelines for Environmental provides guidance for satisfying design criteria Qualification of Safety-Related Computer-Based Instrumentation and Control identified in Appendix A to 10 CFR Part 50, Systems. such as GDC 4: This RG also provides guidance for addressing environmental There are several examples of where DG-1361 goes beyond the scope of stressors affecting the long-term reliability of providing another acceptable method of complying with 10 CFR 50.49. For electric equipment. See the staffs responses specific examples, see Comments 1.2 through 1.5. 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 Section Comment 112 The staff partially agrees with the comment to Horin of C.1.e / p11 Any guidance related to the content of design / qualification / procurement the extent that the implementation of the NUGEQ specifications should not be construed as being limited to environmental guidance for addressing qualification criteria qualification under 50.49. that are not within the scope of 10 CFR 50.49 84
Commenter Section of Specific Comments NRC Resolution DG-1361 The intent of the clarification is also confusing since the same rational, used is not complete. But rather than deleting for excluding aging of mild environment equipment, is true with respect to Section C.1.e as suggested by the commenter, EMC and seismic requirements (which are also not within the scope of the staff has modified its position as follows::
50.49).
The staff position in C.1.e makes a clarification related to Section 5.1 of This RG also provides guidance for IEC/IEEE Std. 60780-323 by removing the prerequisite for aging for electric addressing environmental stressors affecting equipment located in mild environment since this equipment is not within the the long-term reliability of electric equipment.
scope of 50.49. This clarification appears to be inappropriate since design and procurement specifications include requirements related to equipment Section C.1.e has also been modified to clarify qualification, which are not limited to environmental qualification that aging for mild environment equipment is requirement related to 50.49 compliance. not considered under 10 CFR 50.49, but rather under General Design Criterion 4 of 10 CFR 50 Recommendation Appendix A.
Proposed Change: Delete Position C.1.e See the staffs response to comment 5 for additional information.
William Section Comment 113 See the staffs response to comment 14.
Horin of C.2.c / p16 There is no need to include or address smoke as fire is not an event that is NUGEQ 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.
William Section Comment 114 The staff agrees with the comment that Horin of C.1.i / p11 This is another example of where the DG is providing guidance that is outside guidance on EMC stressors does not fall under NUGEQ the scope of 10CFR50.49. EMI/RFI or electrical power surges are not 10 CFR 50.49. Section C.1.e was revised to identified in 10CFR50.49. As indicated in the C.1.i, EMC and electrical provide clarification of this, as well as to state power surges are independent from DBEs and DBAs. that guidance for addressing EMC and 85
Commenter Section of Specific Comments NRC Resolution DG-1361 There is currently sufficient guidance to address EMC in RG 1.180, EMI/RFI may be found in Regulatory Guide Guidelines for Evaluating Electromagnetic and Radio Frequency 1.180. In addition to endorsing IEC/IEEE Interference in Safety-Related Instrumentation and Control Systems. 60780-323-2016 with clarifications, the staff The supplemental guidance in C.1.i is clarifying that the testing for EMC can intended to include supplemental guidance for be separate from EQ testing. This clarification is not necessary since the dual addressing environmental stressors of all types.
logo standard already reflects this (See Note 1 in Section 7.4.1.8.c on page 22 of the dual logo standard). The Purpose section has been modified to include the following statement:
Recommendation Proposed Change: Remove text. This RG also provides guidance for addressing environmental stressors affecting the long-term reliability of electric equipment.
William Section B, Comment 115 The staff partially agrees with the comment.
Horin of Background The rewording of the primary objective of qualification represents an example NUGEQ / p7 of expanding its scope beyond an acceptable method to meet 10CFR50.49. The staff disagrees that the proposed RG Several parts of DG-1361 use the term design-basis events in place of represents an expansion of scope because 10 design-basis accidents. The use of design-basis events is inconsistent with, CFR 50.49(b) states that electric equipment and represents an expansion from, the design basis accident parameters important to safety covered by 10 CFR 50.49 is specified in 10CFR50.49 (d)(1), (d)(2), (d)(3) and I that need to be considered safety-related equipment relied on to remain when establishing environmental qualification. functional during and following design basis 10CFR50.49 is not associated with the prevention of environmentally induced events to ensure: (1) the integrity of the reactor common-cause failures prior to a design basis event. coolant pressure boundary; (2) the capability to The terminology of .before, during and after. is also inconsistent with shut down the reactor and maintain it in a safe Regulatory Position C.1.c. It also reflects a change from Section B of RG shutdown condition; or (3) the capability to 1.89 R1 as well as Section B of RG 1.209. The wording in the Background prevent or mitigate the consequences of section of DG-1361 is misleading and could be interpreted as expanding accidents that could result in potential offsite beyond the scope of 10CFR50.49. Section B of DG-1361 reflects a change exposures comparable to the guidelines in § from Part B of RG 1.89 R1 which reads; 50.34(a)(1), § 50.67(b)(2), or § 100.11.
For the purpose of this guide, qualification is a verification of design limited to demonstrating that the electric equipment is capable of performing The staff agrees that 10 CFR 50.49(d) and (e) its safety function under significant environmental stresses resulting from require qualification to address design basis design basis accidents in order to avoid common cause failures. (Emphasis accidents.
added) 86
Commenter Section of Specific Comments NRC Resolution DG-1361 The comments on before, during, and after The revised wording in DG-1361 currently reads; are addressed in the staff responses to For the purposes of this guide, the primary objective of qualification is to comments 86 and 87. No other changes to the demonstrate that equipment important to safety can perform its safety RG were made due to this comment.
function(s) without experiencing common-cause failures before, during and after applicable design-basis events. (Emphasis added)
William Section A, Comment 116 The staff partially agrees with the comment.
Horin of Related IEEE Std. 323-1974 as modified by RG 1.89R1 also provides an equally The staff removed the sentence that contained NUGEQ Guidance / acceptable approach for establishing EQ of electrical equipment in provides the fundamental approach from the p4 accordance with 10CFR50.49. RG in acknowledgement that there could be The wording of provide the fundamental approach should be reworded to alternative approaches to establishing EQ of clarify that this is one acceptable approach, but not the only acceptable electrical equipment. However, no additional standard to use in establishing EQ in accordance with 10 CFR 50.49. suggested changes were made as a result of this comment.
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.
William Section Comment 117 See response to comment 94 Horin of C.2.b / p16 The guidance in Section C.2.b appears to be broadening the scope of RG 1.89 NUGEQ 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 87
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Comment 118 The staff disagrees with the comment.
Horin of Comments 2.2 - 2.18 are provided in response to the specific request in the NUGEQ FRN for DG-1361 to identify any concerns related to backfitting or forward This proposed guidance does not meet the fitting. definition of backfitting or forward fitting in Section IV of FRN / Vol. 85, No. 243 / December 17, 2020. MD 8.4. RG 1.89, Rev. 2 is voluntary guidance DG-1361 contains multiple examples of regulatory positions that differ from and represents one acceptable way to satisfy RG 1.89, R1 positions on the same topic (in the direction of being more the applicable NRC regulations. The NRC staff restrictive), and even some examples that simply differ from the language of is not imposing or expecting specific licensees 10 CFR 50.49. In either instance, NRC processes and procedures dictate an to implement the guidance in RG 1.89, Rev. 2.
issue-by-issue evaluation of such differences for backfit or forward fit Further, forward fitting occurs only in the implications. context of a licensing action, and the issuance Section D, Implementation of DG-1361 does not identify or address any of a RG is not a licensing action.
new or changed staff positions. Our review of DG-1361 has identified multiple examples of new or revised staff positions are described in the No changes were made to DG-1361 as a result following comments. of this comment.
William General Comment 119 The NRC staff partially agrees with the Horin of The RG in its introduction should clearly repeat 10CFR50.49 (k): comment. The staff agrees that the NRC has NUGEQ (k) Applicants for and holders of operating licenses are not required to previously found existing licensees' programs requalify electric equipment important to safety in accordance with the to be in compliance with 10 CFR 50.49, provisions of this section if the Commission has previously required existing licensees are not required to requalify qualification of that equipment in accordance with "Guidelines for Evaluating electric equipment important to safety under 10 Environmental Qualification of Class 1E Electrical Equipment in Operating CFR 50.49(k), and DG-1361 presents positions Reactors," November 1979 (DOR Guidelines), or NUREG-0588 (For different from positions in RG 1.89, Rev. 1 that Comment version), "Interim Staff Position on Environmental Qualification of could have backfitting implications if used to Safety-Related Electrical Equipment." interpret current requirements in 10 CFR 50.49 88
Commenter Section of Specific Comments NRC Resolution DG-1361 in a way that would require existing licensees The NRC staff has previously found the existing Licensees programs in to requalify electric equipment important to compliance with the regulation, yet this DG presents many potential backfit safety.
positions if used to interpret current requirements as opposed to clearly limiting guidance to new reactors. 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 General Comment 120 See the staffs response to comment 58.
Horin of To clarify what would be needed for a licensee, who is committed to IEEE NUGEQ 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.
89
Commenter Section of Specific Comments NRC Resolution DG-1361 William Section Comment 121 See the staffs responses to comments 67 and Horin of C.2.d (2) / 10 CFR 50.49(e)(5) does not require the aging in a test program to produce a 96 for additional details.
NUGEQ p16 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.
William Section Comment 122 The staff partially agrees with the comment.
Horin of C.1.c / p10 The change in definition for qualified life should clearly reflect that the See the staffs response to comment 3 for NUGEQ need to establish a qualified life is specifically limited to equipment subject to details on the staffs proposed definition for harsh environment qualification under 50.49. As noted in RG 1.209, the need qualified life.
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 The staff added a regulatory position in the RG design basis events would infer qualified life extends to equipment that is to clarify that the staff recognizes that relied upon for events not addressed by 50.49 EQ programs. environmental qualification of electric Also see Comment 2.6 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 90
Commenter Section of Specific Comments NRC Resolution DG-1361 conditions associated with normal operation, maintenance, testing, and postulated accidents, including LOCAs.
William Section Comment 123 The staff partially agrees with the comment.
Horin of C.1.c / p10- By being specific to 50.49 (b)(1), the proposed definition of qualified life The staff revised DG-1361 to eliminate the NUGEQ 11 could be misinterpreted as exempting EQ equipment that is qualified based on potential confusion with the definition of 50.49 (b)(2) or (b)(3) functions from having a qualified life. qualified life to include a reference to 10 CFR The definition of qualified life, which is specific to 50.49 should be specific 50.49(b) instead of limiting to 10 CFR to design basis accidents, since design basis events includes scenarios which 50.49(b)(1).
are excluded pI50.49 (c).
Since the qualified life does not (in and by itself) ensure the performance of a See the staffs responses to comments 3, 21, safety function under harsh DBA conditions, it is suggested that the basis be 80, and 86 for details on the staffs resolution tI to 50.49(e)(5) in lieu of 50.49(b)(1). to other aspects of this comment.
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.
William Section Comment 124 See the staffs response to comment 4.
Horin of C.1.d / p10- The proposed wording in the note inappropriately blends service life and NUGEQ 11 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 91
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Section Comment 125 See the staffs response to comment 4.
Horin of C.1.d / p10- As indicated in EPRI 1022959, shelf-life programs are fundamentally based NUGEQ 11 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.
William Section Comment 126 See the staffs response to comment 4.
Horin of C.1.d / 10- This appears to be a new regulatory position or expectation since 50.49(e)(5)
NUGEQ p11 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 92
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Section Comment 127 The staff partially agrees with the comment.
Horin of C.1.h (1) / To make it clear that this is not indicating any change from the radiation NUGEQ p12 sources that are identified as being significant in the current licensing basis, The staff agrees that only radiation sources that which can vary. Consistency with RG 1.183 and 1.195. This statement should are significant to the total integrated dose to the be clarified as meaning all significant radiation sources that are considered as qualified equipment need to be considered and part of the licensing basis. that initial qualification should be based on the significant sources expected to affect the Recommendation equipment. However, the staff notes that if Proposed Change: The radiation qualification should factor in doses from all during operation the qualified equipment is significant radiation sources at the equipment location. 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 Section Comment 128 See the staffs response to comment 25.
Horin of C.1.h (2) / Keeping the guidance in RG 1.89 specific to an acceptable method of NUGEQ p12 complying with 10CFR50.49.
93
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Section Comment 129 The staff agrees with the comment. C.1.j(1)
Horin of C.1.j (1) / The staffs research results from NUREG/CR-0275, NUREG/CR-4301, and has been rewritten to be presented more NUGEQ p12 NUREG/CR-4091 did not identify a significant effect on performance based clearly. See the staffs response to comment 8 on simultaneous vs. sequential test sequences. Consistency with 50.49(e)(7). for additional information.
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.
94
Commenter Section of Specific Comments NRC Resolution DG-1361 William Section Comment 130 The staff partially agrees with the comment.
Horin of C.1.j.3 / The proposed wording represents a backfit for current plant licensing bases, The staff understands that certain data may not NUGEQ p12 would constitute a forward fit for current plants taking actions that would be immediately available to licensees since appropriately be managed premised on the existing licensing basis or under some information may be maintained by a Section the specific regulatory terms of 10 CFR 50.49, related to the selection and vendor/manufacturer. For this reason, the staff C.1.j.3 / justification for selecting an activation energy value for a material. has modified the RG to include reference to p13 This information is generally unavailable to the licensee and where it does Attachment 2, Select Topics Regarding the exist, it is often considered proprietary by many manufacturers. Environmental Qualification Process, of This new staff position: Inspection Procedure 71111 Attachment 21N, a) Could be interpreted as meaning that the basis for the activation energy Design Bases Assurance Inspection must include data within the temperature range that the equipment or (Programs), (ADAMS Accession No.
component is exposed to during normal operation. This would directly ML19036A556) dated February 5, 2019, which conflict with IEEE Std. 101 provides additional clarification on select b) Doesnt address or provide options for justifying an activation energy environmental qualification topics, including (either by the manufacturer, vendor, or licensee) when activation energy data activation energy.
for the specific formulation is not available or is only available for a generic material family. The staff disagrees with the comment that the c) Doesnt reflect the position in IP 71111.21N regarding validation of information in the RG, including the above information in EQ Reports (e.g., Activation Energy) from approved 10CFR50 reference conflicts with the intent of IEEE Std.
Appendix B suppliers. 101. The RG, and Attachment 2 of IP71111.21N note that For organic materials, A significant portion of the activation energies selected by manufacturers is a regression line (IEEE Std. 101, IEEE Guide derived from materials databases, academic research, and testing performed for the Statistical Analysis of Thermal Life for or by other organizations, which the vendors may consider proprietary or Test Data (Ref. 32)), may be used as a basis otherwise retained in their record system. 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 /> This is a new staff position that would impose very specific expectations for should not be used.
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 The staff disagrees that this position represents changes to current guidance in Regulatory Guide 1.89, Rev. 1, Section C.5.c a backfit, forward fit, or NRC staff as has been applied to currently licensed plants [and potentially to expectations. This guidance does not meet the replacement equipment in those currently licensed plants.]. 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 95
Commenter Section of Specific Comments NRC Resolution DG-1361 The staff position that activation energies should be based on the data from the applicable NRC regulations. The NRC staff the specific compound is new. While this may be the optimum approach, the is not imposing or expecting specific licensees guidance needs to account for situations where the specific formulation is to implement the guidance in RG 1.89, Rev. 2.
proprietary or not currently available. Further, forward fitting occurs only in the context of a licensing action, and the issuance The staff position that activation energy should be selected based on the of a RG is not a licensing action.
temperature range of the equipment in service is also new. Activation energies that are based on isothermal testing and analysis per IEEE 101 are See the staffs response to comments 9 and 50 not based on data points within the range of normal operating temperatures for additional information.
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.
William Section Comment 131 The sentence that used the term must ensure Horin of C.1.f / p11 Consistency with Management Directive MD 6.6. The use of the term must has been removed from the guidance. See the NUGEQ ensure should be clarified since it could be interpreted as going beyond staffs response to comment 6 for additional providing reasonable assurance that certain important to safety electrical information.
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.
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Commenter Section of Specific Comments NRC Resolution DG-1361 William Section Comment 132 The staff partially agrees with the comment.
Horin of C.1.l / p13- The environmental parameters discussed in RG 1.183 is limited to dose. As The reference to RG 1.183 in Section C.1.l was NUGEQ 14 discussed in Comment 2.16, the source term used for EQ has margins that are made in error. The second sentence in C.1.l inherently conservative. has been removed.
The proposed wording is not consistent with the staffs resolution of the Nuclear Power Engineering Committee (NPEC) comment number 73 to The staff does not intend to require the use of NUREG-0588 that clarifies the intent of margin requirements in Section RG 1.183 or any other source term 6.3.1.5 of IEEE Std. 323-1974. methodology as part of the update to RG 1.89.
Nor does the staff intend to require any The reference to RG 1.183 is confusing as well as a change from the reference changes to the source term methodology used to Position C.4 in RG 1.89 R1. The current staff position in C.4 of RG 1.89 in currently licensed plants. The source term R1 covers margin and is not specific or limited to the application of methodology used must be applicable to the quantified margin for dose. This staff position is effectively invoking AST or specific plant and adequate to meet applicable RG 1.183 on Part 50 plants. requirements.
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 William Comment 133 The staff disagrees with the comment. While Horin of NUREG-0588 (1.4 (1)) states that additional radiation margin identified in the source term methodology used for many NUGEQ Section 6.3.1.5 of IEEE Std. 323-1974 are not required if the required components may contain significant accident radiation dose is developed in accordance with the methodology in conservatisms, for others it may not (e.g.,
Appendix D of 0588. This position recognizes the inherent conservatism in components that receive a maximum accident the methodology used to define the integrated accident doses used for EQ. dose from an accident other than the maximum Since these conservatisms are quantifiable, this approach would satisfy hypothetical accident LOCA). In addition, EQ 50.49(e)(8). The source term for EQ (based on earlier EQ guidance, RG 1.31, analysis vary based on the specific plant and RG 1.42, RG 1.1833 or RG 1.1954) are significantly more severe than the licensee. Also, RG 1.89 has recommended allowable level of fuel failure under design basis accident conditions. The certain potential non-conservatisms, with the level of conservatism in the assumptions used to define the accident dose inclusion of the 10% margin added to the should be more than sufficient to eliminate the need to arbitrarily add an accident dose calculations. For example, beta additional 10%. This is further supported by NUREG/CR-53135 which dose may not need to be considered if it is less than 10% of the gamma dose (see the response 97
Commenter Section of Specific Comments NRC Resolution DG-1361 indicates in Section VII.6.8 that core melt in-containment radiation conditions to comment 7). Therefore, the staff does not have yet to be calculated to this accuracy (e.g., within a factor of 2). find that the 10% margin should be removed.
As a result, no changes to DG-1361 were made Recommendation as a result of this comment.
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 See the staffs response to comment 101 for developed consistent with RG 1.183 or RG 1.195 (or earlier analysis additional details.
performed per Appendix D of NUREG-0588, Appendix D of RG 1.89 R1, RG 1.3, or 1.4).
William Section Comment 134 See the staffs responses to comment 4.
Horin of C.2.a(3) / Compared to RG 1.89 R1, this new guidance appears to narrow the focus on NUGEQ p15 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.
William Section Comment 135 The staff disagrees with the comment. The Horin of C.2.c / p16 Consistent with the staffs response to Comment 37 to NUREG-0588 and guidance provided, distinguishing mild and NUGEQ Table C-1 of the DOR Guidelines, there should be some recognition or harsh radiation environments, is consistent with allowance for equipment to be classified as being in a mild radiation only previous staff positions documented in environment when it can be demonstrated that the total integrated dose for NUREG-0800, NUREG-1793, and NUREG-which the equipment is being qualified is below the lowest radiation damage 1503.
threshold for any of the items that are relied upon for the equipment to perform the credited important to safety function(s). In addition, demonstrating that the total integrated dose that the equipment is exposed The addition of radiation thresholds changes the definition of mild as to is less than the damage threshold for the provided in 50.49. The proposed revision to RG 1.89 should recognize that equipment is part of demonstrating that the the distinction between mild and harsh radiation environments is directly equipment is appropriately qualified. The staff related to elevated stressors under DBA conditions that could result in does not intend to identify different thresholds environmentally induced common cause failures. Equipment items made of for a harsh radiation environment for different 98
Commenter Section of Specific Comments NRC Resolution DG-1361 materials with radiation damage thresholds above the total integrated dose types of equipment other than those specified levels are not subject to radiation induced common cause failures. in Section C.2. (This is now C.2.b)
Recommendation See the staffs response to comment 17 for Proposed Change: Add the following to the end of the statement in C.2.c: additional information on total integrated dose Total integrated dose requirements that are above these thresholds may also as it pertains to the environmental qualification be considered a mild radiation environment when it can be demonstrated that of electrical equipment.
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 No changes were made to DG-1361 as a result Comment 3.3. of this comment.
William Section Comment 136 See the staffs response to comments 48, 67, Horin of C.2.d (2) / As noted in Comment 2.4, the imposition of an aging sequence that produces 96, and 121.
NUGEQ p16 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.
William General Comment 137 The staff partially agrees with the comment.
Horin of It would be helpful to end users of the revised RG if the guidance was While the staff agrees that it would be helpful NUGEQ structured in a way that clearly differentiated between the guidance for Part to be able to delineate between each type of 52 plants vs. Part 50 plants. Having clear and concise delineation of applicant/licensee, some of the criteria apply to requirements that apply would be consistent with NRC Management both Part 50 and 52 applicants and licensees.
Directive MD 6.6. The Applicable Regulations section lists the applicable regulations without specifying the The following comments address the need for DG-1361 to clearly criteria each type of applicant/licensee differentiate between Part 50 and Part 52 guidance. There are several (construction permit, operating license, 99
Commenter Section of Specific Comments NRC Resolution DG-1361 examples in DG-1361 where the distinction between specific guidance combined license, manufacturing license, relative to Part 50 or Part 52 plants is not clearly delineated. See following standard design approval, design certification) comments. 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 General Comment 138 The staff agrees that the current license/design Horin of Consistency with Management Directive MD 6.6. basis remains applicable. This RG, as with the NUGEQ previous version, provides one acceptable The relevant guidance that would be applicable to Part 50 plants should be means of complying with the regulations.
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 No changes were made to DG-1361 as a result standard. For existing plants, this should be able to be done using CLB/CDB of this comment.
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.
William Section Comment 139 The staff disagrees with the comment. The RG Horin of C.2.c / p16 Consistency with Management Directive MD 6.6. update is applicable to any nuclear power plant NUGEQ These NUREGs are used as the reference for accepting a mild radiation that references it. Existing facilities are not environment for electronic equipment as 103 rad and 104 rad for other required to conform to the updated RG unless equipment. As worded, it is not clear if this position is specific to just Part 52 they chose to. Note that RG 1.89, Revision 1 plants. Additional discussion should be provided to clarify why regulatory provided the value of 104 rads for non-positions in SERs for Part 52 plants is applicable to Part 50 plants. electronic components and indicated that Note that this citation may be impacted depending on the resolution of electronic components may have a lower Comment 2.18. threshold. Therefore, the inclusion of the 103 rad threshold for electronic equipment adds a threshold when there was none specified in RG 100
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Applicable Comment 140 The staff disagrees with the comment.
Horin of Guidance / The text should be revised or clarified. The proposed revision specifically NUGEQ p4 and focuses on AST, which would be appropriate for a Part 52 plant, but not This revision to RG 1.89 focuses on AST (and Appendix necessarily applicable to EQ for a Part 50 plant. The Draft Guide is not clear therefore, the guidance provided in RG 1.183)
D / D-2 whether Technical Information Document (TID) 14844, Calculation of for radiological accident guidance because that Distance Factors for Power and Test Reactor Sites remains a valid is the most recent radiological source term methodology for calculating source terms. The proposed revision to RG 1.89 guidance. Other approved source term does not recognize or reference RG 1.195, Methods and Assumptions for methodologies and EQ methodologies, Evaluating Radiological Consequences of Design Basis Accidents at Light- including those provided in TID-14844 and RG Water Nuclear Power Reactors or the acceptability for Part 50 plants to use 1.89, Revision 1, continue to be acceptable to source terms developed in accordance with NUREG-0588, Appendix D of licensees currently using them, provided that RG 1.89 R1, or RGs 1.3 or 1.4 which are based on TID-14844. design changes affecting the source term are not being made. Additional justification is The proposed change to the guidance and Appendix D, Qualification in the needed for use of TID-14844 or other Radiation Environment regarding the use of alternate source term (AST) is previously approved methodologies for fuel an example where the DG is primarily focused on Part 52 plants. enrichment greater than 5% and peak burnup There is no need to repeat the information in Appendix D, Section D-2, that is greater than 62,000 MWD/MTU.
already contained in Appendix I to RG 1.183.
In addition, Section D-2.2 of Appendix D of Recommendation DG-1361 indicates that approved alternative Proposed Change: Appendix D should simply refer to Appendix I of RG assumptions may be used. This is intended to 1.183 and expand upon the continued acceptability of source terms based on specify that the source terms and source term TID-14844 for EQ purposes, even if the plant has adopted AST for other methodologies in RG 1.183 are not the only radiological analysis. Appendix D-2 should also cover the resolution of GSI- acceptable source terms or methodologies that 187, which concluded that licensees may continue to use TID-14844 for EQ can be used.
even if they adopt AST (See ML011210348). Also See Comment 5.2.
101
Commenter Section of Specific Comments NRC Resolution DG-1361 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 References Comment 141 The staff agrees with the comment. As noted Horin of / p19-20 As noted in Section 3.11 of currently issued DSRS, these documents contain in the Design Specific Review Standard NUGEQ guidance acceptable to the staff for environmental design and qualification of (DSRS) for the NuScale Small Modular computer-specific requirements that should be used in conjunction with Reactor design, there are applicable NUREG-0588 and RG 1.89, as appropriate, for evaluating computer specific environmental criteria associated with the use requirements. of computer-based equipment in safety related applications. The staff augmented the Related DG 1361 doesnt reference or point to RG 1.152 or IEEE Std 7-4.3.2. At a Guidance section in RG 1.89, Rev. 2 to include minimum, inclusion of these references this would seem appropriate for Part appropriate documents outlining criteria for 52 plants. 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 Section Comment 142 See the staffs response to comments 101 and Horin of C.2.f / p17 While Part 52 plants will utilize AST, Part 50 plants continue to use a 140.
NUGEQ 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.
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Commenter Section of Specific Comments NRC Resolution DG-1361 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 General Comment 143 The staff partially agrees with the comment.
Horin of The proposed revision to RG 1.89 does not address or cover some of the The staff reviewed the February 20, 2004, NUGEQ potential burden reduction areas that the staff previously indicated would be letter from the NRC to William A. Horin considered in the next revision to RG 1.89. The staffs prior indication to (ADAMS Accession No. ML040510309), in consider these in a future revision of RG 1.89 was not captured or identified response to the comment. DG-1361 does not in the staffs periodic review that concluded in 2018 that a revision to this prohibit alternative qualification methods from regulatory guide was warranted. being used nor does it prevent qualifying for select harsh environmental conditions. For For example: example, if equipment is only exposed to a
- 1) Graded Qualification Methods Based on Severity of Accident Environment radiation harsh environment and temperature, NRC response: We are considering whether to clarify the option to qualify a pressure, etc. do not cause significant aging component specific to the environment through a revision to Regulatory degradation of equipment as result of the Guide (RG) 1.89. environmental conditions, then only radiation
- 2) Alternative Qualification Methods for Equipment Exposed to Radiation- would have to be considered when qualifying Only Harsh Conditions the equipment. The staff finds that the NRC response: We are considering whether to clarify the option to address proposed RG and IEC/IEEE 60780-323-2016, EQ for radiation only environments through a revision to RG 1.89. contain adequate guidance on qualifying Reference Accession No. ML040510309 equipment subject to a radiation environment.
The NUGEQ believes that these examples could be achievable under existing regulatory direction with license amendments (e.g., application of 10 CFR With regard to alternative qualification 50.59, or application of existing guidance). methods, applicants or licensees may use The Group would like to see such positions referenced in DG-1361, perhaps alternative approaches, if appropriately in an Appendix. justified, and consistent with current regulatory practice and applicable NRC requirements.
103
Commenter Section of Specific Comments NRC Resolution DG-1361 Based on this, the staff determined that no additional information needed to be incorporated into the revision to RG 1.89.
William Section Comment 144 The staff partially agrees with the comment.
Horin of C.1.g (1) / The guidance in Appendix C doesnt cover Part 52 plants.
NUGEQ p11 The staff revised the guidance to include a App C / C-1 The cited methods in Appendix C are unchanged from RG 1.89 R1 and only footnote to Appendix C in RG 1.89, Rev. 2.
cover B&W, CE, Westinghouse and GE designs that were licensed under Part that states:
- 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 Mass and energy releases are developed using methodology used to define the containment response to design basis a methodology that is consistent with the accidents for the safety analysis or consistent with the methodology used for licensing basis of the plant. The listed methods HELB analysis. This approach would result in the appendix being applicable are examples for existing designs.
to both Part 50 and Part 52 plants.
See the staffs responses to comments 90 and Recommendation 99 for additional details.
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.
William Applicable Comment 145 See the staffs response to comments 101 and Horin of Guidance / For Part 50 plants, GSI-187 (which deals with the potential impact of 140.
NUGEQ p4 and postulated cesium concentration on Equipment Qualification) was closed out Appendix based on the conclusion by the staff that there was no clear basis for D / D-2 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 104
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Appendix Comment 146 The staff agrees with the comment. The staff Horin of A Consistency between regulatory guidance documents. inserted the suggested information as a NUGEQ footnote in Appendix A of RG 1.89, Rev. 2.
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.
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Commenter Section of Specific Comments NRC Resolution DG-1361 William Appendix B Comment 147 See the staffs response to comment 99.
Horin of / B-1 Since the DG will still apply to Part 50 plants, the guidance should be NUGEQ 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.
William App D, Comment 148 The staff partially agrees with the comment.
Horin of Section D-1 The period of exposure may be limited to a qualified life that is less than the The staff deleted the sentence of concern as it NUGEQ / D-1 plant license. 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 Recommendation equipment is installed in the plant.
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.
William Missing Comment 149 See the staffs response to comment 103.
Horin of App E The guidance in Appendix E specifically defines EQ categories that are not NUGEQ 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.
106
Commenter Section of Specific Comments NRC Resolution DG-1361 Recommendation Suggested Change: Reinstate and retain Appendix E.
William Related Comment 150 See the staffs response to comment 72.
Horin of Guidance / 10CFR50.69 is relevant since it can eliminate safety-related electrical NUGEQ p3 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.
William Related Comment 151 The staff agrees with the comment. The staff Horin of Guidance / Section 6.2 of IEC/IEEE 60780-323 is specific to reassessing qualified life, added a staff position in Section C.1.f of RG NUGEQ p3 but there is no specific reference to the existing guidance in X.E1 related to 1.89, Rev. 2, to supplement Section 6.2 of Section C.1 the reanalysis of TLAA. Consistency with existing regulatory guidance. IEC/IEEE Std. 60780-323, Edition 1, 2016-02, with the following:
Staff has provided acceptable methods related to reassessing qualified life in NUREG-1801 & NUREG-2191.Section X.E1 contains specific guidance X.E1, Environmental Qualification of related to the reanalysis of an EQ aging evaluations or TLAA. Electric Equipment, of NUREG-2191, Generic Aging Lessons Learned for Recommendation Subsequent License Renewal (GALL-SLR)
Proposed change - Add a staff position in Section C.1 that covers Section 6.2 Report, and X.E1, Environmental of the dual logo standard by providing a reference to the existing guidance in Qualification (EQ) of Electric Components, of Section X.E1 of NUREG-1801 & NUREG-2191. 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 107
Commenter Section of Specific Comments NRC Resolution DG-1361 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 Reason for Comment 152 Horin of Revision / Consistency between RG 1.89 and RG 1.209. The staff agrees with the comment. A NUGEQ p7 clarification has been included to identify that Note that the NRC has endorsed the use of IEEE 323-2003 for qualifying guidance for ensuring that digital and computer-based I&C equipment in mild environment applications in RG computer-based devices in mild environments 1.209. may be found in RG 1.209. Digital devices proposed for installation in harsh environments Recommendation and used in SSCs for achieving design basis Proposed Change: Reword to clarify that the NRC did not officially endorse functions should meet the requirements of 10 these standards for qualification of electrical equipment to harsh design basis CFR 50.49.
accident conditions.
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Commenter Section of Specific Comments NRC Resolution DG-1361 William Section Comment 153 The staff disagrees with the comment and with Horin of C.1.b / p10 Consistency with 10CFR50.69. To clarify that not all safety-related electrical incorporating the specific suggested language.
NUGEQ equipment that must function during or following exposure to harsh accident However, the staff revised the guidance to conditions is subject to 50.49 if they are classified as RISC-3 under include information on 10 CFR 50.69. See the 10CFR50.69. staffs response to comment 72 for additional details.
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.
William Section Comment 154 The staff agrees with the comment that the Horin of C.2.f / p17 This statement is confusing, particularly the interjection of before testing statement could lead to confusion.
NUGEQ when discussing various elements of a test sequence.
See the response to comments to 16 and 17.
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.
William Applicable Comment 155 See the staffs response to comment 83.
Horin of Regulations The reference to 10 CFR 52.157, Contents of applications; technical NUGEQ /p2 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 109
Commenter Section of Specific Comments NRC Resolution DG-1361 10 CFR 52.167. This may be intended to reference 10 CFR 52.167 for the definition of a manufacturing license.
William References Comment 156 See the staffs response to comment 97.
Horin of / p 20 The citation to EPRI Report 1021067 should not include NEI.
NUGEQ Recommendation Suggested citation: Electric Power Research Institute (EPRI), Nuclear Power Plant Equipment Qualification Reference Manual, Revision 1. EPRI, Palo Alto, CA: 2010. 1021067 William Appendix C Comment 157 The staff agrees with the comment and has Horin of Footnote 10 Correct the embedded hyperlink address to remove the word and at the end. fixed the link.
NUGEQ / C-1 The embedded hyperlink to Doc Collections includes the word and as part of the address.
Comment Document 13: ML21131A005 Jeremy C.1.a Comment 158 See the staffs responses to comment 33.
Owen of In the context of IEC/IEEE 60780-323 end condition means the condition of Kinetrics 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.
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Commenter Section of Specific Comments NRC Resolution DG-1361 Jeremy C.1.b Comment 159 See the staffs response to comment 2.
Owen of This section is intended to provide clarity about the phrase important to Kinetrics 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.
Jeremy C.1.d Comment 160 See the staffs response to comment 4.
Owen of The term service life in IEC/IEEE 60780- 323 does not imply anything about Kinetrics 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.
Jeremy C.1.j(3) Comment 161 The staff partially agrees with the comment.
Owen of While it is preferred to use the activation for the actual compound being The staff is not implying that the lowest Kinetrics tested, it is not always practical, and the accepted industry approach has been activation energy for a material must be used.
to use available conservative values. More often than not, activation energies To clarify this, RG 1.89, Rev. 2 includes the for a specific compound are not available. following: The selected activation energy should be representative of the most limiting Recommendation material in a component/sub-component when Given the importance of activation energies for qualified life of equipment, as determining qualified life.
much guidance as possible should be given on selection of activation energies. The staff modified the guidance to include references to IEEE Std. 98-2016 and IEEE Std.
It is recommended to refer to IEEE 98, 99 and/or UL Std 746B for the 99-2019 as discussed in the response to determination of activation energies. The section should also indicate that comment 65.
while the activation energy for the specific material being considered is 111
Commenter Section of Specific Comments NRC Resolution DG-1361 sometimes required, such as in IEEE 383, the conservative approach is also The staff did not have adequate time to review acceptable if properly justified. the referenced UL standard to consider it for incorporation into the RG.
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.
Jeremy C.1.l Comment 162 The staff partially agrees with the comment.
Owen of This section endorses the margins presented in Section 7.4.1.7 of IEC/IEEE While there is no guidance on how to Kinetrics Std. 60780-323 are acceptable. The margins presented are only applicable to incorporate margin under normal conditions, the accident conditions. Is there any guidance for margins applied to normal conservatisms are generally applied, which conditions? inherently add margin. The staff finds that the margins added to accident conditions is Recommendation adequate to satisfy the requirements for the Provide guidance for margins to use for normal conditions. 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 C.1.n Comment 163 See the staffs response to comment 12. The Owen of It is not clear how a double peak should be used for equipment vulnerable to portion of Section C.1.n that comment 163 is Kinetrics thermal binding or when there are limitations of the steam supply during associated with has been deleted.
testing.
Recommendation Recommend to providing additional guidance or specific examples.
Jeremy C.2.c Comment 164 See the staffs response to comment 14.
Owen of This section refers to additional stressors such as smoke exposure. This type Kinetrics of aging mechanism is not part of the scope of IEC/IEEE 60780-323 112
Commenter Section of Specific Comments NRC Resolution DG-1361 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 C.2.d(1) Comment 165 The staff agrees that Section 7.4.1.9.3 of Owen of This section refers to preconditioning of test samples employing the IEC/IEEE 60780-323-2016 discusses the use of Kinetrics Arrhenius methodology. It is not clear as to what aspect of preconditioning Arrhenius aging and the sequence of age this statement refers to and what the reader should consider. IEC/IEEE conditioning considers sequential, 60780-323 clearly describes the use of the Arrhenius methodology. simultaneous, and synergistic effects.
Recommendation The intent of Section C.2.d in DG-1361 was to This statement should be clarified to indicate what it alludes to, otherwise it address uncertainties with regard to the should either refer to the discussion in IEC/IEEE 60780- 323 on that topic or processes and environmental factors that could be removed. Item #2 is also discussed in IEC/IEEE 60780-323, so this result in such degradation. This Section statement does not provide any additional clarity. 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 C.2.e(4) Comment 166 The staff disagrees with the comment. While Owen of The 7th line bullet # 4 states: analyses taking into account arrangements of the LOCA or HELB is the most severe accident Kinetrics equipment and radiation sources may be necessary to determine whether for most equipment, the design basis accident equipment needed for mitigation of design basis accidents other than LOCA that is most severe to a particular piece of or high-energy line breaks (HELB) could be exposed to a more severe equipment is dependent on the plant and the environment than the plant specific LOCA or HELB environments. location of the equipment. For example, if the equipment is located near the main steam Recommendation piping, the limiting design basis accident for This could be clarified. In order to be clear, suggest defining the other DBAs that equipment may be a main steam line break that are more severe than the plant specific LOCA or HELB. or steam generator tube rupture.
Since the limiting accident is specific to the plant and the location of the equipment, 113
Commenter Section of Specific Comments NRC Resolution DG-1361 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.
114