ML16315A127

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Public Comment Resolutions for DG-1309
ML16315A127
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Issue date: 06/30/2018
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Office of Nuclear Regulatory Research
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Burton S
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ML16315A125 List:
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RG 1.207, Rev. 1 DG-1309
Download: ML16315A127 (27)
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NRC Staff Responses to Public Comments on DG-1309:

GUIDELINES FOR EVALUATING THE EFFECTS OF LIGHT-WATER REACTOR COOLANT ENVIRONMENTS IN FATIGUE ANALYSES OF METAL COMPONENTS I. INTRODUCTION This document presents the NRC staffs responses to written public comments received on Draft Guide (DG)-1309, Guidelines for Evaluating Fatigue Analyses Incorporating the Life Reduction of Metal Components Due to the Effects of the Light-Water Reactor Environment for New Reactors (ADAMS Accession No. ML1417A584), in response to a separate Federal Register entry (79 FR 69884) on November 24, 2014.

II. OVERVIEW OF COMMENTERS AND COMMENTS Comments on the subject draft report are available electronically at the NRC's electronic Reading Room at http://www.nrc.gov/reading-rm/adams.html. From this page, the public can gain entry into Agencywide Documents Access and Management System (ADAMS), which provides text and image files of NRC's public documents. Comments were received from the following individuals or groups:

Letter ADAMS Commenter Abbreviat Commenter Affiliation No. Accession No. Name ion 1 ML15023A569 Rolls Royce PLC, United Kingdom Keith Wright RR Camille 2 ML15023A570 Westinghouse Electric Company, USA Zozula WEST 3 ML15023A571 Nuclear Energy Institute, USA Jason Remer NEI David 4 ML15027A334 Union of Concerned Scientists, USA Lochbaum UCS Southern Nuclear Operating Company, Charles 5 ML15033A382 Inc., USA Pierce SNC 6 ML15033A383 AREVA, USA Morris Byram AREVA 7 ML15033A384 Electric Power Research Institute, USA Nathan Palm EPRI This table lists each public comment by Letter No. For each comment, the NRC staff has repeated the comment as written by the commenter. Each comment is referred to in the form

[XXX]-[YYY]-[ZZZ], where: [XXX] represents the Abbreviation from the above table, [YYY]

represents the Letter No. from the above table, and [ZZZ] represents the sequential comment number from that commenter.

RR COMMENTS AND RESPONSES RR-1-1 Comment The background information to the draft RG emphasizes that fatigue design curves using a "95/95" (%confidence/%probability) criterion are deemed acceptable to the NRC because the fatigue design curves are based on crack initiation, rather than component failure or through-wall leakage, and therefore, additional margin exists between crack initiation and actual component failure or leakage.

However, this additional margin is variable and will depend upon component thickness and the through wall stress distributions. Therefore, for thin wall components that are subjected to cyclic membrane stresses and strains the criterion of crack initiation (engineering significant) and through-wall leakage are, for all practical purposes, one and the same as they are coincident in terms of cyclic life. That is, the additional margin against through wall leakage is virtually zero in these circumstances.

The importance of this distinction is that when components are of a thicker wall section and have a significant through wall strain gradient then the additional margin between crack initiation and through-wall leakage will be a very significant proportion of a component total life.

Therefore, to penalize the design life of these components, in comparison to membrane loaded components, is considered excessive.

Unfortunately the ASME view of a fatigue criterion is not clearly articulated in the code books, the criteria papers or the Compendium Books. Nevertheless, the view that through-wall leakage is the intended criterion has been expressed on the basis that this is the stated 'end-of-life' for experimental stress analysis in ASME Section III, Appendix I1. Paragraph 11-1520 (b) states: "Failure is herein defined as a propagation through the entire thickness, such as would produce a measurable leak in a pressure retaining member".

It is noted that the implementation section of the draft RG states that methods or solutions that differ from those described in this RG may be deemed acceptable if sufficient basis and information is provided for the NRC staff to verify that the proposed alternative demonstrates compliance with the appropriate NRC regulations. It is assumed that this will enable the potential for this penalization of some component design lifetimes to be corrected, providing of course that adequate total reliability is demonstrated.

NRC Staff Response The commenter is correct that fatigue lives derived from small test specimens reflect membrane loading, and that there is generally a small difference between crack initiation and test specimen failure, and that additional margin exists between crack initiation and component failure under the presence of a significant though-wall strain gradient. The staff also agrees that Section III of the ASME Code is not clear on the meaning of fatigue failure. As articulated in Section 1.1 of NUREG/CR-6909, Revision 1, Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials, (ADAMS number ML16319A004) the NRC staff defines fatigue life as:

Based on the foregoing discussion and the results of the majority of the test data evaluated, fatigue life is defined in this report as the number of cycles of a specified strain amplitude that a specimen can sustain before the formation of a 3-mm-deep crack (i.e., an engineering crack). Thus, the fatigue life of a material can be described in terms of three parameters, viz., strain or stress, cycles, and crack depth. The best-fit curve to the existing fatigue -N data describes, for a given strain or stress amplitude, the number of cycles needed to develop a 3-mm deep crack. This number of cycles also equates to a 25% load drop in a solid 9.5 mm test specimen, and is assumed to equate to crack initiation in an actual component. Using this definition, a calculated fatigue CUF less than unity provides reasonable assurance that a fatigue crack has not formed in a component, and indicates that the probability of forming a crack in the component is low.

Item ii Definition of Fatigue Life in new Section 1.5 of NUREG/CR-6909, Revision 1, contains additional discussion which further expounds on the definition of fatigue life used within that report.

In addition, in response to this comment, the following statement was added in Part B (end of 4th paragraph on page 6) of the final RG:

The staff recognizes that additional margin may be present in components where the dominant loading consists of a significant though-wall strain gradient, as opposed to the membrane loading typically applied in small-scale laboratory fatigue test specimens.

Methods to account for this additional margin may be considered by the staff on a case-by-case basis, provided sufficient basis and information is provided to the staff to verify that the proposed alternative demonstrates compliance with all applicable NRC regulations.

RR-1-2 Comment The suggested use of the methodology presented in Rev 1 of NUREG/CR-6909 to describe the effects of fatigue loadings on nuclear plant materials, including correction using Fen for LWR coolant environments, is supported as the most appropriate currently-available set of fatigue design data relevant to small scale membrane loaded specimens under isothermal conditions. However, it must be considered that the NUREG/CR-6909 design data (including use of Fen) has been demonstrated in a number of papers published in open literature to be excessively conservative under a range of different operating conditions, leading to the calculation of usage factors for some components that do not reflect the satisfactory service experience of these components (even when considering the "95/95" basis of the design data).

Therefore it should be expected that use of these design data will lead to the calculation of unacceptable and unrealistic Cumulative Usage Factors for some components, particularly if operating profiles of civil nuclear plants evolve to meet the needs of a rapidly-changing energy market (e.g., increased transients due to load following) or if further rounds of plant life extension are pursued in future. In light of these concerns, it must be recognized that it is imperative to continue to support and fund research into the fatigue analysis and fatigue behavior of nuclear plant materials in order that improved models of this behavior may be developed to allow reductions in the over conservatisms.

NRC Staff Response The NRC staff agrees that continuing research into fatigue analysis is justified. The Fen approach for environmentally assisted fatigue (EAF) has been developed to bound 95% of the available small specimen data with 95% confidence. However, there are a variety of other factors that can lead to excessively conservative cumulative usage factor (CUF) calculations for nuclear components or structures. Some factors that contribute to significant differences in observed field behavior compared to CUF predictions of fatigue life include gradient loading (as discussed in RR-1-1), the fact that the ASME design curves for ferritic materials are applicable to a wide range of both carbon and low-alloy steels (instead of separate curves for specific materials), and conservatisms in the ASME Code,Section III CUF calculation methodology.

The NRC staff continues to encourage investigations into these and other topics as part of research efforts and remains open to accepting technically defensible alternative approaches that may result from such testing. With the publication of NUREG/CR-6909, Revision 1 (ADAMS number ML16319A004) and this RG, the NRC staff will have completed its currently planned research activities related to EAF. Because no significant technical problems have been encountered to-date by licensees pursuing 60 years of plant operation through license renewal applications, the NRC staff has not placed a high priority on continued EAF research activities, and is focusing resources on other, higher priority issues. However, the staff remains committed to following industry and other efforts related to EAF, and will reevaluate whether additional NRC-sponsored research is necessary as appropriate. The NRC staff will continue to maintain awareness of the results of new research activities, issues associated with operation beyond 60 years (i.e., subsequent license renewal), and issues arising from operating experience that are inconsistent with NRC guidance.

No changes were made to the draft RG in response to this comment.

WEST COMMENTS AND RESPONSES WEST-2-1 Comment Page 5, paragraph 1, states: The NRC staff deems this criterion acceptable because the NRC staff bases the fatigue design curves on crack initiation, rather than component failure or through-wall crack leakage, and, therefore, additional margin exists between crack initiation and actual component failure or leakage.

Comment: Specifying crack initiation vs. through-wall growth as a basis for margin is not consistent with the science of fatigue crack failures, where one case could have significant margin and another case essentially no margin. Other margins exist in the ASME Code design process which can be more consistently identified as applicable to all cases. We suggest that this sentence be deleted, or reworded similar to the following: The NRC staff deems this criterion acceptable because the staff recognizes the conservatism in the ASME Code design process.

NRC Staff Response The commenter is correct that there is conservatism in many aspects of the ASME Code design process. This comment is similar to comment RR-1-1. Revisions to this regulatory guide were made in response to this comment and RR-1-1, as discussed in response to comment RR-1-1.

WEST-2-2 Comment Page 5, paragraph 3, states: Fen calculations for carbon, low-alloy, austenitic stainless, and Ni-Cr-Fe alloy steels need only consider the types of stress cycles or load set pairs that exceed the strain threshold criteria.

Comment: As stated here in the draft Regulatory Guide, it is understood that the strain amplitude threshold may be used to exclude ASME fatigue pairs from Fen penalty (Fen = 1.0),

regardless of the strain rate method applied in the evaluation (modified rate or average strain rate). A cursory reading of Appendix A of Draft NUREG/CR-6909 Revision 1 could lead one to conclude that use of the strain amplitude threshold is not permissible when the modified rate approach will be employed. Section 4.1.14 of draft NUREG/CR-6909 Revision 1 seems to indicate that application of the modified rate approach to the remaining fatigue pairs should employ no further use of the strain amplitude threshold in the detailed integration method.

Please clarify and make consistent in both documents which applications of the strain amplitude threshold are permitted and not permitted with respect to the application of the modified rate approach.

A possible rewording of the first two sentences of the current paragraph of the draft Regulatory Guide would be: Fen calculations for carbon, low-alloy, austenitic stainless, and Ni-Cr-Fe alloy steels need only consider the types of stress cycles or load set pairs that exceed the strain threshold criteria, regardless of the strain rate method applied in the Fen evaluation.

For pairs that exceed the strain amplitude threshold criteria, the evaluation options to determine Fen depend on the complexity of the analyzed transient conditions and the details of the evaluation.

Another possibility is to remove this detail from the Regulatory Guide and include it with the suggested clarifications in Appendix A of NUREG/CR-6909 Revision 1.

With regard to this comment, please also see our comments that were provided in Westinghouse letter LTR-NRC-14-26 on draft NUREG/CR-6909 Revision 1, Section 4.1.14, related to application of the strain amplitude threshold.

NRC Staff Response With respect to NUREG/CR-6909, based on public comments received on draft NUREG/CR-6909, Revision 1, Section 4.1.14 was moved to a new Section 4.4. Both Appendix A and Section 4.4 of the final NUREG/CR-6909, Revision 1 (ADAMS number ML16319A004) state that the strain threshold should not be used when using the modified rate approach. For example, the second to last paragraph in Section A-2 of NUREG/CR-6909, Revision 1 states, the strain amplitude threshold should not be applied when using a modified rate approach, as it may yield non-conservative results.

Further, the NRC staff agrees that the first sentence of the third paragraph on page 5 of the draft RG could cause confusion. In response to this comment, this sentence was deleted, and the second sentence of the subject paragraph was modified in the final RG as follows: The Fen calculation options for carbon and low-alloy steels, austenitic stainless steels, and Ni-Cr-Fe materials depend on the complexity of the analyzed transient conditions and the details of the evaluation.

WEST-2-3 Comment Page 6, paragraph 2, states: These methods apply to those components exposed to reactor coolant that are required by regulation to have a fatigue CUF evaluation or have an existing CLB fatigue CUF evaluation.

Comment: The Fen factors have been developed for simulated primary reactor coolant environments. In some cases there are components that "have an existing CLB fatigue CUF evaluation" that are not exposed to reactor coolant, but are in secondary systems (e.g., PWR steam generator feedwater nozzles) exposed to secondary fluid. They are not required by regulation to have a fatigue CUF. If this statement is meant to apply to such components, then the applicability of the Fen factors to such components should be clearly stated. Otherwise, such components should be excluded from this requirement.

NRC Staff Response The NRC staff agrees that the scope of CUF evaluation should be clarified. The methods described in this regulatory guide are applicable to both primary and secondary systems because the associated technical basis in NUREG/CR-6909, Revision 1 is based on data from many diverse water environments that are representative of both primary and secondary systems in commercial nuclear power plants. Additionally, the staff agrees that the draft RG did not clearly define the term reactor coolant and that such a definition is needed. Therefore, the following sentence was added after the first sentence of the first paragraph of Part B: An LWR water environment is defined within this guidance as any transient or steady-state environment in a light water commercial nuclear power plant where the component of interest is exposed to water above 50oC. Additionally, the words coolant and reactor coolant were replaced with water throughout the document. The RG title was modified as well to reflect this comment.

Corresponding clarifications were made to NUREG/CR-6909, Revision 1, as discussed in the response to public comments for that document (ADAMS number ML16319A004).

With respect to the portion of the comment related to the scope of the guidance cited from Page 6, paragraph 2 of the draft RG, this regulatory guidance is applicable within the following three applications: 1) to address the effects of the LWR water environment in a CUF calculation in licensing applications associated with reactor designs submitted for NRC approval, 2) to address the effects of the LWR water environment in a CUF calculation in licensing applications associated with operating reactors pursuing license renewal, and 3) plants where addressing the effects of the LWR water environment is part of their licensing basis. In addition, the regulatory guidance can be used to address environmental effects in other applications other than those that are explicitly listed above, as appropriate.

In response to this comment and to clarify these requirements, the last sentence of the Background section of Part B, and the last sentence of the first paragraph of Part C was clarified in the final RG to state: The methods described in this regulatory guide may be used to address the effects of the LWR water environment in a CUF calculation in licensing applications associated with reactor designs submitted for NRC approval, operating reactors pursuing license renewal, and plants where addressing such effects is part of their licensing basis.

Please also see the responses to NEI-3-1, NEI-3-15, and NEI-3-18 for related information.

NEI COMMENTS AND RESPONSES NEI-3-1 Comment DG-1309 refers to NUREG/CR-6909, Revision 1, for the Fen formulas for evaluating environmental fatigue, which has not yet been formally approved and issued. However, the DG does not clarify if the use of NUREG/CR-6909, Revision 0 formulas remains acceptable.

Several LR applicants have used NUREG/CR-6909, Revision 0 methods and formulas for computing Fen values and would not wish to revise them just in order to meet NUREG/CR-6909, Revision 1 criteria.

NRC Staff Response The staff does not intend to change the EAF guidance applicable for plant operation up to 60 years that is documented in NUREG-1800, Revision 2, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, {ADAMS No. ML103490036} and NUREG-1801, Revision 2, Generic Aging Lessons Learned (GALL) Report {ADAMS No.

ML103490041}. That guidance continues to adequately address the effects of the LWR water environment on CUF calculations. This guidance allows the use of NUREG/CR-6909, Revision 0 for license renewal. Furthermore, NUREG/CR-6909, Revision 0 methods previously approved by the staff remain valid for the period of their intended use.

Subsequent license renewal (SLR)-specific Standard Review Plan (SRP) and GALL documents reference the use of NUREG/CR-6909, Revision 1 and RG 1.207, Revision 1 for addressing environmental effects on fatigue are forthcoming. These documents also clarify how NUREG/CR-6909, Revision 0 can be used to address environmental effects on fatigue. This guidance is contained in NUREG-2191, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report, Volumes 1 and 2, {The pre-final versions of these documents are found in ADAMS at ML17123A134 and ML17123A138, respectively} and NUREG-2192, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants {The pre-final version of this document is found in ADAMS at ML17123A140}. The following guidance related to this comment is provided within the pre-final versions of NUREG-2191 and NUREG-2192:

Environmental effects on fatigue for these critical components can be evaluated using the guidance in Regulatory Guide (RG) 1.207, Revision 1; NUREG/CR-6909, Revision 0 (with average temperature used consistent with the clarification that was added to NUREG/CR-6909, Revision 1); or other subsequent NRC-endorsed alternatives.

No changes were made to the draft RG in response to this comment.

NEI-3-2 Comment NUREG-1800 and NUREG-1801 will need to be revised to assure the guidance contained in the revised Reg Guide has been appropriately updated in the NUREGs.

This is necessary to avoid any differences which create inconsistencies between the documents. For example: the Reg Guide specifies using the 2013 Edition of the ASME Code.

The current NUREG's are silent.

NRC Staff Response The NRC staff intends for the existing guidance in NUREG-1800, Revision 2 and NUREG-1801, Revision 2 to remain applicable for 60 years of operation (i.e., through the initial license renewal period). For applicants that seek operation beyond 60 years (i.e., the subsequent license renewal period), SLR-specific guidance that references NUREG/CR-6909, Revision 1 and RG 1.207, Revision 1 is forthcoming. Guidance for the SLR period is contained in NUREG-2191, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report, Volumes 1 and 2, {The pre-final versions of these documents are found in ADAMS at ML17123A134 and ML17123A138, respectively} and NUREG-2192, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants {The pre-final version of this document is found in ADAMS at ML17123A140}.

No changes were made to the draft RG in response to this comment.

NEI-3-3 Comment Pg. 2; 5th Paragraph.

The paragraph states the guide contains information collections.

It is not clear that this is correct because the guide does not appear to be an information request.

NRC Staff Response This RG describes acceptable information for licensee use, including in license renewal and subsequent license renewal applications. The Federal Register Notice for the final RG provides the following:

PAPERWORK REDUCTION ACT STATEMENT This regulatory guide contains and references information collection requirements that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). These information collections were approved by the Office of Management and Budget (OMB),

approval numbers 3150-0011, 3150-0151, and 3150-0155.

PUBLIC PROTECTION NOTIFICATION The NRC may not conduct or sponsor, and a person is not required to respond to, a request for information or an information collection requirement unless the requesting document displays a currently valid OMB control number.

NEI-3-4 Comment Pg. 3; 1st Paragraph.

Change ...with the consideration of... to ...from...

"From" is an equivalent and much simpler Paragraph way to say "with the consideration of. Otherwise, could say with the consideration of risk,....

NRC Staff Response The staff agrees that the first sentence in the Background section of Part B of the draft version of this RG could cause confusion. Therefore, the first sentence was revised in the final RG as follows: This RG provides guidance for determining the acceptable fatigue lives of components using a CUF calculation that is performed in accordance with the fatigue design rules in Section III of the ASME Code, and incorporates the effects of LWR water environments. As discussed in the response to WEST-2-3, the word coolant in the draft guide was changed to water in the final RG to clarify the scope of the RG.

NEI-3-5 Comment Pg. 3; 1st Paragraph.

Change ...aging-management... to ...aging management...

Unnecessary hyphen should be removed.

NRC Staff Response The staff agrees with the comment. The hyphen was deleted from the first paragraph of the Background section of Part B on page 3 of the draft RG.

NEI-3-6 Comment Pg. 3; 1st Paragraph.

Change ...in ways consistent with existing requirements... to ...in ways consistent with requirements...

Eliminate unnecessary word.

NRC Staff Response The staff agrees with the comment that this sentence can be simplified by eliminating unnecessary wording. Therefore, the phrase, in ways consistent with existing requirements in, in the draft RG was changed to, consistent with the requirements in in the final RG.

NEI-3-7 Comment Pg. 3; 1st Paragraph.

Change ...findings were captured in the original... to ...findings were documented in the original...

Suggested wording enhancement. Documented is a more precise term than captured.

NRC Staff Response The staff agrees with the comment. The phrase, These findings were captured in the initial versions, in the draft version of this RG was changed to, These findings were documented in the initial versionsin the final RG.

NEI-3-8 Comment Pg. 3; 1st Paragraph.

Change ...cooling environments already forms a part... to ...cooling environments is part...

Simplified wording.

NRC Staff Response The staff agrees with the comment that the original wording in the draft version of this RG could be simplified. This sentence was changed in response to both this comment and WEST-2-3 as follows:

The methods described in this regulatory guide may be used to address the effects of the LWR water environment in a CUF calculation in licensing applications associated with reactor designs submitted for NRC approval, operating reactors pursuing license renewal, and plants where addressing such effects is part of their licensing basis.

Further, as discussed in the responses to WEST-2-3 and NEI-3-4, the word coolant in the original sentence was changed to water in the final RG to clarify the scope of the RG.

NEI-3-9 Comment Pg. 3; 2nd Paragraph.

Change ...ASME Code based on tests conducted in laboratory air environments... to

...ASME Code based on laboratory tests conducted in air environments...

Clarification. Laboratory should more appropriately modify test than air.

NRC Staff Response The staff agrees with the comment. The phrase ASME Code based on tests conducted in laboratory air environments, in the draft version of this RG was changed to ASME Code based on laboratory tests conducted in air environments in the final RG.

NEI-3-10 Comment Pg. 3; 2nd Paragraph.

Comment on ...NB-3121, Corrosion, in Section III of the ASME Codes...-Beginning with which edition?

It would be useful to identify the applicable edition or editions because wording and paragraph numbers have been known to change over time.

NRC Staff Response The staff agrees with the comment. The phrase NB-3121, Corrosion, in Section III of the ASME Code, in the draft version of this RG was changed to NB-3121, Corrosion, in Section III of editions of the ASME Code from 1971 onward, in the final RG.

NEI-3-11 Comment Pg. 3; 2nd Paragraph.

Change More recent fatigue-test data... to More recent fatigue test data...

Hyphen should be eliminated.

NRC Staff Response The staff agrees with the comment. The hyphen was deleted in the second paragraph of the Background section of Part B, which was on page 3 in draft version of this RG.

NEI-3-12 Comment Pg. 4; 1st Paragraph.

Change ...analyzed existing laboratory data... to ...analyzed laboratory data...

Existing adds no value. One could not have analyzed nonexistent laboratory data.

NRC Staff Response The staff agrees with the comment. The phrase, ...analyzed existing laboratory data..., in the draft version of this RG was changed to, ...analyzed laboratory data... in the final RG.

NEI-3-13 Comment Pg. 4; 1st Paragraph.

Change ...life in a room-temperature air to ...life in a room temperature air...

Hyphen should be eliminated.

NRC Staff Response The staff agrees with the comment. The hyphen was deleted in the fourth paragraph of the Background section of Part B, which was on page 4 of the draft version of this RG.

NEI-3-14 Comment Pg. 4; 2nd Paragraph.

Change ...generally considered to be conservative... to ...generally considered conservative...

Eliminate unnecessary words.

NRC Staff Response The staff agrees with the comment. The phrase, were generally considered to be conservative compared to, in the draft version of this RG was changed to, are generally conservative compared to in the final RG.

NEI-3-15 Comment Pg. 4; 2nd Paragraph.

The draft Reg Guide remains silent relative to the application and use of NUREG/CR-6583 and NUREG/CR-5704. Likewise NUREG/CR-6909, Revision 1, provides no discussion.

Therefore, for License Renewal, the Reg Guide should recognize the use of these NUREG's per NUREG's 1800 and 1801 as acceptable.

Currently NUREG-1800 and 1801 permit the use of NUREG/CR-6583, and NUREG/CR-5704 to determine the effects of LWR environment on the fatigue calculation. This will allow continued use of design basis CUF's to compute the CUFen using the appropriate Fen values.

Otherwise, based on the guidance of the Reg Guide new values would have to be computed using the ASME 2013 edition of the Code.

NRC Staff Response Future license renewal submittals (i.e., license applications to extend the original operating license up to 60 years) may continue to use the guidance set forth in NUREG-1801, Revision 2, as that guidance continues to adequately address the effects of the LWR water environmental on CUF calculations for that period. NUREG-1801, Revision 2, allows the use of NUREG/CR-6583 and NUREG/CR-5704 for license renewal.

SLR-specific Standard Review Plan (SRP) and GALL documents that reference the use of NUREG/CR-6909, Revision 1, and RG 1.207, Revision 1, and clarify the use of NUREG/CR-6583, NUREG/CR-5704 are forthcoming. This guidance is contained in NUREG-2191, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report, Volumes 1 and 2, {The pre-final versions of these documents are found in ADAMS at ML17123A134 and ML17123A138, respectively} and NUREG-2192, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants, {The pre-final version of this document is found in ADAMS at ML17123A140}. As discussed in the response to NEI-3-1, the following guidance related to this comment is provided within the pre-final versions of NUREG-2191 and NUREG-2192:

Environmental effects on fatigue for these critical components can be evaluated using the guidance in Regulatory Guide (RG) 1.207, Revision 1; NUREG/CR-6909, Revision 0 (with average temperature used consistent with the clarification that was added to NUREG/CR-6909, Revision 1); or other subsequent NRC-endorsed alternatives.

No changes were made to the draft version of this RG in response to this comment.

NEI-3-16 Comment Pg. 4; 3rd Paragraph.

Change ...additional available fatigue data... to ...additional fatigue data...

Available is unnecessary.

NRC Staff Response The staff agrees with the comment. The phrase, additional available fatigue data in the draft version of this RG was changed to, additional fatigue data in the final RG.

NEI-3-17 Comment Pg. 4; 4th Paragraph.

Change ...coolant environments in fatigue life... to ...coolant environments on fatigue life...

Suggested clarification. More common and correct English usage is to say effect on fatigue life rather than effect in fatigue life.

NRC Staff Response The staff does not agree with the comment. The complete sentence in the draft RG is as follows:

Those results identified the need to revise and consolidate the previously published guidance for incorporating the effects of LWR coolant environments in fatigue life evaluations. The effects are to be incorporated in fatigue life evaluations (evaluations is the noun), as stated in the original sentence.

No change to the draft version of this RG was made in response to this comment. However, as indicated in the responses to WEST-2-3 and NEI-3-4, coolant in the above sentence was changed to water in the final RG.

NEI-3-18 Comment Pg. 4; 4th Paragraph.

Explain the expected effect of the changes to the Fen equations provided in NUREG/CR-6909, Rev. 1, from those in NUREG/CR-6909, Rev. 0. Will the changes result in higher or lower Fen values?

This guidance is needed in order to determine if environmental fatigue calculations prepared in accordance with the Fen equations from NUREG/CR-6909, Revision 0, remain bounding of environmental fatigue calculations that would be prepared using the Fen equations from NUREG/CR-6909, Rev. 1.

NRC Staff Response Generally, if all inputs to the Fen equations are not changed, the revised Fen equations usually produce Fen factors that are similar or lower than the Revision 0 Fen factors. The continued use and applicability of environmental fatigue calculations based on a previous method (i.e.,

NUREG/CR-6583, NUREG/CR-5704, and NUREG/CR-6909, Revision 0) is addressed in the responses to comments NEI-3-1 and NEI-3-15. Please see those responses for additional information.

No changes were made to the draft RG in response to this comment.

NEI-3-19 Comment Pg. 4; 4th Paragraph.

Revision 1 of RG 1.207 maintains should say This Reg Guide revision...

Provides clarity.

NRC Staff Response The staff agrees that the phrase identified by the commenter can be clarified. Therefore, the phrase, Revision 1 of RG 1.207 maintains, in the draft RG was changed to, This revision of the regulatory guide maintains... in the final RG.

NEI-3-20 Comment Pg. 5; 1st Paragraph.

Comment on The NRC staff deems this criterion acceptable because the NRC staff bases the fatigue design curves on crack initiation, rather than component failure or through-wall crack leakage...

It's not clear how this is valid since the NRC is not the organization that developed the curves. It would be clearer if it said the NRC deems it acceptable because ANL based the fatigue design curves on crack initiation rather than component failure or throughwall cracking.

NRC Staff Response The staff agrees with the comment. The phrase, The NRC staff deems this criterion acceptable because the NRC staff bases the fatigue design curves on crack initiation , in the draft RG was changed to, The NRC staff finds this criterion acceptable because the fatigue design curves on crack initiation... in the final RG.

NEI-3-21 Comment Pg. 5; 2nd Paragraph.

Change ...the current ASME Code procedure... to ...the ASME Code procedure...

Eliminate unnecessary word. Current adds no value because it does not define a time at which the procedure was current. The procedure that was current at the time of the simulations will not necessarily be the same procedure that is in effect when the user reads the regulatory guide.

NRC Staff Response The staff agrees with the comment. The phrase, the current ASME Code procedure, in the draft RG was changed to, the ASME Code procedure... in the final RG.

NEI-3-22 Comment Pg. 5; Last Paragraph.

Change The NRC staff searched for available guidance... to The NRC staff searched for guidance...

It is understood that the NRC would not search for unavailable guidance.

NRC Staff Response The staff agrees with the comment. The phrase, The NRC staff searched for available guidance in the draft RG was changed to, The NRC staff searched for guidance... in the final RG.

NEI-3-23 Comment Pg. 6; Section C, 1st Paragraph.

Section C states: These methods apply to those components exposed to reactor coolant that are required by regulation to have a fatigue CUF evaluation or have an existing CLB fatigue CUF calculation. Imposing the methods to all locations which have a fatigue CUF evaluation goes beyond the Reactor Coolant Pressure Boundary. In some cases, this would include reactor vessel internals, heat exchangers in auxiliary systems, etc., which may include non safety-related components. Suggest the scope of the Reg Guide be applicable to Class 1 safety related components within the Reactor Coolant Pressure Boundary having existing fatigue CUF evaluations.

From a safety perspective, the application of more conservative requirements above those of the ASME code should be focused on the Reactor Coolant Pressure Boundary rather than imposing it on other systems and components which may not have a safety function.

NRC Staff Response The NRC staff agrees with this comment in part. The staff disagrees with the comment that the scope of the RG should be limited to class 1 safety related components within the reactor coolant pressure boundary. Environmental effects may be significant for any component subject to an LWR water environment, not just reactor coolant pressure boundary components.

However, the staff agrees, that this scope should be clarified in the final RG. As noted in the response to WEST-2-3, the scope stated in the first paragraph of Part B of the final RG has been modified accordingly. Please also see the responses to NEI-3-1, NEI-3-15, and NEI-3-18 for related information.

NEI-3-24 Comment Pg. 7; Sections 2.1 and 3.1.

Will (proposed design curve) be deleted after this Reg Guide is approved? Provide clarification of what (proposed design curve) means.

Clarification requested NRC Staff Response The staff agrees that the reason for the parenthetical phrase in the draft RG is unclear.

Therefore, the phrase, the stainless steel fatigue air curve (proposed design curve) provided in Figure A.3 and Table A.2, in Part C, Sections 2.1 and 3.1 of the draft RG was changed to, the fatigue design curve for austenitic stainless steels in air provided in Figure A.3 (i.e., curve identified as Based on ANL Model in figure legend) and Table A.2... in the final RG. This identifier is needed within the parentheses because there is more than one curve in Figure A.3.

Similarly, the phrase, the fatigue air curves (updated ANL model curves) provided in Figures A.1 and A.2 and Table A.1 in Part C, Section 1.1 of the draft RG, was changed to, the fatigue design curves in air provided in Figures A.1 and A.2 (i.e., curves identified as ANL Model and Eq. 27 in figure legend) and Table A.1 in the final RG. As above, this identifier is needed within the parentheses because there is more than one curve in Figures A.1 and A.2.

NEI-3-25 Comment Pg. 8; 2nd Paragraph.

Current licensees may continue to use guidance found acceptable for complying with the identified regulations as long as their current licensing basis remains unchanged. Need to clarify that current licensees approved for License Renewal and have entered in the Period of Extended Operation (PEO) are considered to be acceptable for complying with regulation.

Clarification requested.

NRC Staff Response Following license renewal, the original current licensing basis continues in effect except as modified during the license renewal process, per 10 CFR 54.29, 54.30, 54.31, and 54.33. As noted in the response to comment NEI-3-1, the NRC staff does not intend to make changes to applicable guidance for 60 years of operation (i.e., through the initial license renewal period).

Further, holders of renewed operating licenses are current licensees for purposes of the discussion in Part D of the regulatory guide.

No changes were made to the draft RG in response to this comment.

NEI-3-26 Comment Pg. 8; 2nd Paragraph.

Comment on ...with underlying NRC regulations.- It would be useful to define the underlying regulations.

The background mentions 54.21, but that would only apply to renewed license applicants. In addition, 54.21 addresses license renewal aging management reviews and TLAA evaluations in general, rather than addressing anything that is the basis for environmentally adjusted CUF analysis.

NRC Staff Response The commenter is correct that 10 CFR 54.21 only applies to license renewal applicants and does not specifically address the basis for performing an environmental adjusted CUF analysis.

The sentence in the draft version of this RG that address 10 CFR 54.21 has been deleted in response to this comment.

Applicable regulations are identified in the applicable rules and regulations portion of Part A, Introduction, and within Part C, Staff Regulatory Guidance. The purpose of Part D, Implementation, is to provide information on the 1) applicants and licensees use of the RG and

2) NRCs planned use of the RG. In addition, Part D describes how the NRC staff complies with 10 CFR 50.109, Backfitting, and any applicable issue finality provisions in 10 CFR Part 52. It is not the intent of Part D to repeat information provided earlier in the guidance.

NEI-3-27 Comment Pg. 8; 2nd Paragraph.

Change Methods or solutions that... to Methods that...

The regulatory guide appears to address acceptable methods, but not acceptable solutions.

NRC Staff Response In addition to describing methods acceptable to the NRC staff for meeting regulatory requirements, an RG may describe concepts, approaches, processes, and solutions that address a regulatory requirement. This is often the case with performance-based regulations where the NRC describes a conceptual approach or process to address a performance-based requirement, without prescribing a specific method. Thus, the term, solution, meaning an acceptable approach to address a regulatory requirement, is therefore appropriate.

No changes were made to the draft RG in response to this comment.

NEI-3-28 Comment Pg. 8; 2nd Paragraph.

Comment on ...the identified regulations...

It would be beneficial to identify the regulations because it does not appear that they have been identified.

NRC Staff Response This comment is similar to NEI-3-26. Regulations to which the RG is applicable are identified in Sections A and C of the RG, as noted in the response to NEI-3-26.

No changes were made to the draft RG in response to this comment.

NEI-3-29 Comment Pg. 8; 3rd Paragraph.

Change ...regulatory guide or applicable parts to resolve regulatory... to ...regulatory guide to resolve regulatory...

The phrase applicable parts is redundant to information in this regulatory guide unless it is meant to refer to applicable parts of federal regulations. It should be deleted or clarified depending on which is the case.

NRC Staff Response The staff disagrees that the term applicable parts is redundant because, for the fatigue evaluations pertaining to this guidance, portions of the guidance may be acceptable for use, depending upon the particular circumstance, in addition to the methodology specified in the entire guidance. Therefore, it is important to clarify that the guidance may not have to be used in its entirety.

However, the NRC staff agrees that the language in the implementation discussion could be clarified to convey the NRCs intent in this regard. Accordingly, the last sentence of the third paragraph of Part D in the draft RG was changed in the final RG to the following sentence:

Licensees may use the information in this regulatory guide (in whole or part, as applicable), to resolve regulatory or inspection issues.

NEI-3-30 Comment Pg. 9; 1st Paragraph.

Section D page 9 states: (2) the specific subject matter of this regulatory guide is an essential consideration in the staff's determination of the acceptability of the licensee's request, the staff may request that the licensee either follow the guidance in this regulatory guide or provide an equivalent alternative process that demonstrates compliance with the underlying NRC regulatory requirements. By making this statement, the Regulatory Guide seems to infer that since NUREG-1800 and NUREG-1801 offer an equivalent alternative process with NUREG/CR-6583 and NUREG/CR-5704, these alternate processes are acceptable to determine the effects of LWR environment on fatigue. This needs to be confirmed.

This clarification is required to confirm acceptability of using NUREG/CR-6583 and NUREG/CR-5704.

NRC Staff Response The staff disagrees that clarification on the acceptability of using NUREG/CR-5704 and NUREG/CR-6583 is needed in this regulatory guide. As discussed in the responses to NEI-3-2 and NEI-3-5, guidance on the use of these NUREGs and other alternative methods for the subsequent license renewal period is forthcoming and is contained in NUREG-2191, Volumes 1 and 2, {The pre-final versions of these documents are found in ADAMS at ML17123A134 and ML17123A138, respectively} and NUREG-2192 {The pre-final version of this document is found in ADAMS at ML17123A140}. Please also see the responses to NEI-3-1 and NEI-3-18 for related information.

No changes were made to the draft RG in response to this comment.

UCS COMMENTS AND RESPONSES UCS-4-1 Comment On behalf of UCS, I submitted FOIA/PA-2015-0082 seeking Final Safety Analysis Report records received after October 1, 2001, by the NRC from all nuclear plant owners, including these licensees. By letter dated January 7, 2015, the NRC administratively closed my request.

By email dated December 30, 2014, the NRC staff notified me that the estimated page count for the requested records was 448,000 just from Region IV. That email estimated approximately 4,000 pages per FSAR per reactor with multiple updates per reactor since October 1, 2001 --

accounting for the hundreds of thousands of pages per region.

By withholding this vast quantity of FSAR material from the public, the NRC is significantly impairing our ability to review and comment on this draft regulatory guide. The FSARs describe the safety-related structures, systems, and components at the plant and further describe their role in preventing or mitigating design basis transients and accidents. The FSARs describe the types of metal used in the reactor coolant pressure boundary and the number of fatigue cycles they are expected to experience. Without access to that vital information, we cannot assess whether the draft regulatory guide as sufficient scope and depth.

NRC regulation 10 CFR 50.59 requires licensees to screen proposed modifications and changes to plant operating procedures to see whether the proposed changes might reduce safety margins approved by the NRC in a significant way. If so, the changes cannot be made until after the NRC reviews and formally approves them. The 10 CFR 50.59 screenings and evaluations rely heavily on information in the FSARs. The FSARs are also extensively used by NRC's reviewers when evaluating licensee requests for licensing actions.

As stated above, by withholding hundreds of thousands of pages of FSARs from the public, the NRC is significantly impairing our ability to comment on this draft regulatory guide in a meaningful away. The licensees and the NRC staff rely heavily on the FSAR information while preparing and reviewing requests for licensing actions, but the NRC's withholding prevents the public from reviewing this information. Licensees have access to the FSARs, so they can comment on the draft regulatory guide. But we and the public lack this same opportunity.

The hundreds of thousands of pages of vital information being withheld by the NRC staff is being done so improperly. Entire FSARs, such as the recent updates for Beaver Valley Unit 2 and Watts Bar Unit 2, were made publicly available by the NRC in their entirety. This demonstrates that FSARs do not contain classified, safeguards, or sensitive information that must be withheld from public disclosure.

The NRC should not approve this draft regulatory guide until after the agency makes all the FSAR information for these reactors publicly available and gives the public sufficient time to review this information. Failing to do so makes a mockery of this public comment process.

NRC Staff Response The NRC staff does not agree with this comment, and notes that the commenters concerns about the resolution of a FOIA request are outside the scope of the draft regulatory guide, as the public availability of FSARs does not bear on the publics ability to review and comment on the draft Regulatory Guide. Neither DG-1309 nor NUREG/CR-6909, Revision 1, include FSAR references. The technical bases for DG-1309 were identified in the draft Regulatory Guide itself, and in the supporting draft technical basis document, NUREG/CR-6909, Revision 1, Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials, which was issued for a 45-day public comment period in the Federal Register (79 FR 21811) on April 17, 2014. No plant-specific FSAR information was used to develop either the technical basis supporting this RG in NUREG/CR-6909, Revision 1, or the provisions in the RG itself.

No changes were made to the draft RG in response to this comment.

SNC COMMENT AND RESPONSE SNC-5-1 Comment Representatives from SNC are working closely with the Nuclear Energy Institute (NEI) to address this very important change in regulatory guidance. SNC endorses the NEI comments on DG-1309 provided to the NRC in a letter dated January 22, 2015. In particular, DG-1309 references NUREG/CR-6909, Revision 1, as providing acceptable methodology for evaluating fatigue. Several NEI comments discuss the need to recognize other previously approved methods, not referenced in the draft revision, as providing acceptable calculational methods.

Additionally, the scope of the regulatory guide should be clarified as applicable to Class 1 safety related components within the reactor coolant pressure boundary (RCPB) having existing cumulative usage factor evaluations. The proposed wording in the draft could be interpreted as imposing these methods on components beyond the RCPB.

SNC encourages an open dialog to discuss resolution of the industry comments so that the final regulatory guide revision serves in the best interest of safety and provides clear guidance to licensees.

NRC Staff Response The NRC staff partially agrees and partially disagrees with this comment. The NRC disagrees with the portion of the comment related to the need to recognize other previously approved methods. For plant operation up to 60 years, the guidance documented in NUREG-1800, Revision 2, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, {ADAMS No. ML103490036} and NUREG-1801, Revision 2, Generic Aging Lessons Learned (GALL) Report {ADAMS No. ML103490041} remains applicable and acceptable. For plant operation beyond 60 years, guidance is forthcoming and is provided in NUREG-2191, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR)

Report, Volumes 1 and 2, {The pre-final versions of these documents are found in ADAMS at ML17123A134 and ML17123A138, respectively} and NUREG-2192, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants {The pre-final version of this document is found in ADAMS at ML17123A140}. Please see the responses to related comments NEI-3-1, NEI-3-2, NEI-3-15, NEI-3-18, and NEI-3-30 for additional information, The NRC staff agrees with the portion of the comment related to clarifying the scope of the regulatory guide. The RG is not limited to class 1 safety related components within the reactor coolant pressure boundary. Environmental effects may be significant for any component subject to an LWR water environment, not just reactor coolant pressure boundary components.

Therefore, the RG scope, as described in the first paragraph of Part B of the final RG has been modified accordingly. Please also see the responses to related comments WEST 2-3 and NEI-3-23 for additional information.

AREVA COMMENTS AND RESPONSES AREVA-6-1 Comment The Fen equations in NUREG/CR-6909 are highly questionable and not explicitly validated according to the extensive technical comments submitted by Devin Kelley on 6/2/2014 on behalf of AREVA (Docket ID NRC-2014-0023). Without being inclusive, the Fen expressions address only membrane-loaded specimens, the sub-factors used for the derivation of the in air fatigue curves, although correlated, are considered independent and the need for such over-conservative values poorly justified.

NRC Staff Response This comment is associated with the technical basis document, NUREG/CR-6909, Revision 1 (ADAMS number ML16319A004), used to develop the RG. The first part of the comment states that the Fen method has not been explicitly validated. Section 6 of NUREG/CR-6909, Revision 1 describes how independent specimen and component tests were used to validate the Fen method. The next part of this comment refers to extensive technical comments that have been provided by AREVA on the draft of NUREG/CR-6909, Revision 1. The complete set of AREVA comments and the staff response to these comments are contained in Appendix F of the final version of NUREG/CR-6909, Revision 1 (ADAMS number ML16319A004).

The final part of this comment identifies the following specific concerns related to NUREG/CR-6909, Revision 1: 1) the development of the Fen expressions using predominantly membrane-loaded specimen data, and 2) the sub-factors used to derive the air fatigue design curves. The comment states that both of these issues lead to over-conservatism in the Fen method. As illustrated by the agreement of experimental and predicted fatigue lives in Section 6 of NUREG/CR-6909, Revision 1, the staff does not agree that the Fen method is overly conservative. However, the staff agrees that, in certain applications, the Fen method can be conservative and, as stated in the response to RR-1-1, added the following statement in Part B of the final RG to address this point:

The staff recognizes that additional margin may be present in components where the dominant loading consists of a significant though-wall strain gradient, as opposed to the membrane loading typically applied in small-scale laboratory fatigue test specimens.

Methods to account for this additional margin may be considered by the staff on a case-by-case basis, provided sufficient basis and information is provided to the staff to verify that the proposed alternative demonstrates compliance with all applicable NRC regulations.

This statement does not explicitly mention the subfactors used to develop the air fatigue design curves. As stated in Section 5 of NUREG/CR-6909, Revision 1, the subfactors were analyzed to develop a single adjustment factor that bounds 95% of the individual analysis Monte Carlo runs. As mentioned above, the appropriateness of this approach is illustrated by the agreement between fatigue lives predicted by Fen method and those measured in the experiments depicted in Section 6 of NUREG/CR-6909, Revision 1.

AREVA-6-2 Comment Page 6 introduction of C Appendix A to that report includes details descriptions and additional guidance concerning the overall method and all the required calculations:

The staff s developed guidance in Appendix A is insufficient to develop practical engineering procedures. For example, not enough information is provided for the evaluation of the strain rate equations A.6, A.10, and A.15 to be used along with Equation A.19 per the modified strain rate approach. This is especially true for analyses according to ASME Sub-article NB-3600.

Insufficient guidance leads to the development of individual methods with a greater risk of higher operational and regulatory cost in the future.

NRC Staff Response This RG and Appendix A of NUREG/CR-6909, Revision 1, is not intended to provide prescriptive, step-by-step methods for evaluating Fen. Instead, this guidance is general so that it can be adapted into the wide range of methods that are used in existing fatigue CUF evaluations for nuclear components. This approach provides licensees with appropriate flexibility in applying the method to accommodate the large number of possible approaches, assumptions, and judgments that are employed in CUF evaluations.

Additionally, the example problem documented in Appendix C of NUREG/CR-6909, Revision 1, was specifically added to Revision 1 of NUREG/CR-6909 to provide more details and further guidance on an acceptable application of the Fen methods in a representative fatigue CUF evaluation. The example problem contains greater detail so that one can reproduce the solution to the sample problem using either the modified or average strain rate approaches. ASME Code,Section III, Subarticle NB-3200 methods were used by the NRC staff to solve the sample problem because those are the methods used in nearly all of the licensee fatigue CUF evaluations submitted to the NRC to-date. The NRC has also previously reviewed and approved fatigue CUF evaluations performed by licensees that used ASME Code,Section III, Subarticle NB-3600 methods1.

However, the NRC does agree that the phrase quoted in the comment could be clarified.

Therefore, in response to this comment, the phrase, includes detailed descriptions and additional guidance concerning the overall method and all the required calculations in the second paragraph on page 6 of the draft RG was changed to, includes the specifics of the Fen method for incorporating the effects of LWR water environments in CUF evaluations of metal components. This change is located in the first paragraph of Part C (page 8) in the final RG.

AREVA-6-3 Comment Page 5, for a varying factor of ten to twelve: ten is also a sufficient factor to account for the 95/95 criterion, but instead the factor of 12 is used. Why?

NRC Staff Response The factor of 12 on the allowable number of fatigue cycles, or life, was selected for use in developing the fatigue design curve to remain consistent with the current ASME Code Section III fatigue design curve for austenitic materials. However, NUREG/CR-6909, Revision 1 (ADAMS number ML16319A004) noted that the results indicated that the conservatism in the ASME Code Section III fatigue design air curves may be further reduced by using a factor of 10 on life instead of the factor of 12 on life. The basis for the factors of 10 to 12 are provided in Section 5.5 and Table 15 of NUREG/CR-6909, Revision 1.

To clarify this rationale, the following sentence has been added after the second sentence of the eleventh paragraph of Part B in the draft RG: Revision 1 of NUREG/CR-6909 chose to use a factor of 12 on life in the development of the air fatigue design curves to be consistent with the ASME Code Section III fatigue design curve for austenitic materials.

The final RG also retains the factor of 12 on life because it is associated with the factor of 2 assumption on stress, and that factor was not evaluated as part of NUREG/CR-6909, Revision

1. The factor of 2 on the allowable strain or stress has been an assumption adopted by ASME and has been consistently used since the first fatigue curves were developed for implementation into ASME Code,Section III. However, its basis is arbitrary and not well-established. The NRC staff believes that a more comprehensive approach to changing the factor of 12 would be to collectively review both the factors of 2 and 12, and to develop revised factors accordingly.

Based on the research results identified in NUREG/CR-6909, Revision 1, the staff would generally accept the use of factors of 2 and 10 in environmental fatigue calculations, provided that design fatigue curves were developed consistent with those factors and sufficient technical justification for the curves was provided.

AREVA-6-4 Comment Page 5: the sample example of Appendix C is not applicable to the NB-3600 analyses and not comprehensive for the NB-3200 analyses. ASME is revising this example to also address changes in the direction of the controlling principal stress.

1 See, for example, Safety Evaluation Report, Related to the License Renewal of Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, ADAMS Accession Number ML15182A051.

NRC Staff Response This comment applies only to the sample problem that is contained in Appendix C of NUREG/CR-6909, Revision 1 (ADAMS number ML16319A004), and not specifically to the information or guidance contained within this RG. First, the NRC staff notes that several changes were made to NUREG/CR-6909, Revision1to address these concerns. First, it is clarified throughout the several sections of the report that the example problem in Appendix C uses the NB-3200 methodology. Specifically, the end of Section 1.4 of NUREG/CR-6909, Revision 1, has been modified as follows:

Appendix C of this report presents a sample application of Fen using the methodology in ASME Code Section III, Subarticle NB-3200, that is intended to address some of the practical issues identified by interested stakeholders associated with the Fen calculations.

As stated, it is not possible for any single example to be comprehensive, but the example problem has been chosen to address some of the more significant issues for conducting an NB-3200 analysis. Further, items xxii, Use of Method and xxiii, Use of Fen in NB-3600 Analyses, have been added within new Section 1.5 in NUREG/CR-6909, Revision 1 to clarify the use of the method in an NB-3600 analysis and reference some examples.

To clarify the scope and intent of the NUREG/CR-6909, Revision 1, Appendix C sample problem within the RG, the last sentence of Page 5 in the draft RG (last paragraph of the Background section of Part B, page 7 in the final RG) has been modified as follows:

Appendix C to Revision 1 of NUREG/CR-6909 provides a sample problem showing one method of calculating and applying Fen to a CUF calculation using the NB-3200 methodology. This problem addresses some of the practical issues associated with Fen calculations.

The commenter is also referred to the responses to public comments submitted on draft NUREG/CR-6909, Revision 1. The responses to public comments are chronicled in Appendix F of the final version of NUREG/CR-6909, Revision 1. Specifically, the NRC staffs responses to comments FAIDY-2-4e and FAIDY-2-5g address related comments on the use of the Fen method in NB-3600 and NB-3200 analyses.

AREVA-6-5 Comment Page 3, end of second paragraph More recent fatigue test data from the fatigue lives of components: The word components should be replaced by the word specimens.

NRC Staff Response The NRC staff agrees with the comment. The phrase, can have a significant impact on the fatigue lives of components... in the last sentence of the second complete paragraph on page 3 of the draft RG was replaced with, have a significant impact on the fatigue lives of laboratory test specimens in the final RG.

AREVA-6-6 Comment ASME Code Cases need to be referenced as acceptable methods for the fatigue design in LWR and not research reports, as is the NUREG/CR-6909 Rev. 1, for the reason stated in point 2 above. Resolution of the concerns on the ASME Code Cases is necessary since the use of ASME Code Cases allows for the active participation of stakeholders, the frequent revision of the Code Cases to account for new research results and to address the designer's concerns.

NRC Staff Response In this RG, the NRC staff refers to NUREG/CR-6909, Revision 1 (i.e., a research report) as the basis for the guidance in this RG. There is sufficient information contained in NUREG/CR-6909, Revision 1, as well as this RG, to evaluate the effects of LWR water environments on the fatigue lives of metal components without being unreasonably prescriptive. This approach provides licensees with sufficient flexibility to address the wide ranging conditions, assumptions, margins and implications associated with CUF evaluations.

Further, the NRC staffs positions on the use of ASME Code Cases, including cases associated with environmentally assisted fatigue evaluation methods, is already documented in the following regulatory guidance documents:

The above documents are periodically updated to include Code Cases recently published by ASME.

No changes were made to the draft RG in response to this comment.

AREVA-6-7 Comment Many do not agree with the NUREG/CR-6909 direct multiplication of the in-air design usage factors (which already include the factor of 12.0 on the small-specimen mean fatigue curve) by F(en) factors that are based on membrane types of loadings of small specimens. This conservatism of having membrane types of loadings as a basis for the F(en) factors is in addition to a.) quick in-surge/out-surge types of loadings not being considered in the Rules from the NRC Staff, b.) favorable hold-time effects not being considered in those Rules, c.) load sequence penalty being incorporated into the factor of 12.0, although not justified for nuclear plant applications, etc. It is obvious that direct multiplication of these factors is very conservative. Why has an SRSS combination of these factors not been considered?

NRC Staff Response The first part of this comment questions the technical basis of the Fen method contained in NUREG/CR-6909, Revision 1 (ADAMS number ML16319A004). The basis of the method is presented in Sections 2 - 5 of that report, and the method is validated in Section 6 of that report by demonstrating that the method adequately predicts the results of independent tests conducted on both laboratory specimens and components.

The next part of the comment states that quick surge loading is not considered in the Fen method. The method considers the effect of the environment for all transients including surge loading. More information on this issue is provided in the response to AREVA-9-17k in Appendix F of NUREG/CR-6909, Revision 1. The comment next states that beneficial hold-time effects are not considered in the Fen method. The NRC staff agrees that hold time effects are not explicitly considered and such effects may be beneficial. However, more research is needed to quantify the effect, especially for transients that are representative of nuclear plant applications. More information on this issue is provided in the response to AREVA-9-15 in Appendix F of NUREG/CR-6909, Revision 1. Next, the commenter states that load sequence effects should not be incorporated into the adjustment factor of 12 on fatigue life. The NRC staff disagrees with this assertion as the staff believes that applicable data demonstrate the potential deleterious effects of high strain cycles followed by low strain cycles. More information on this issue is provided in the response to AREVA-9-7 in Appendix F of NUREG/CR-6909, Revision 1.

Finally, this comment asserts that direct multiplication of these factors is conservative and asks why an SRSS combination of factors has not been considered. The NRC staff disagrees with such an approach. As discussed in Section 5.5 of NUREG/CR-6909, Revision 1, a simple multiplication of these factors is not performed to get the adjustment factor. Rather, a Monte Carlo analysis performed to better consider the full range of possible combinations of these factors and bound the combined effect of these factors 95% of the time. Such an analysis is not possible using an SRSS approach. This point has been further clarified in Sections 1.5 and 5.5 of NUREG/CR-6909, Revision 1. More information on this issue is provided in the response to AREVA-9-7 in Appendix F of NUREG/CR-6909, Revision 1.

No changes were made to the draft RG in response to this comment.

AREVA-6-8 Comment When using the Design Fatigue Curves suggested by these Rules (from NUREG/CR-6909, Rev. 1), instead of the factor of 12.0 on the small-specimen mean fatigue curve, the following factors from Table 15 of the same document (NUREG/CR-6909, Rev. 1) should have been used, for example, 10.2 for carbon steel, 9.0 for low-allow steel and 9.6 for austenitic stainless steel. Therefore, to illustrate for austenitic stainless steel, all in-air usage factors should be multiplied by 9.6 / 12.0 = 0.80 before applying the F(en) factor noting that this F(en) factor is only required if the location is touching the water.

NRC Staff Response The NRC staff does not agree with the recommendation to use both more precise and material-specific factors (i.e., 10.2, 9.6, or 9.0) instead of the factor of 12 on fatigue life for the design fatigue curves. Such an approach is not justifiable given the variability in fatigue test results for the data that was analyzed in NUREG/CR-6909, Revision 1 (ADAMS number ML16319A004).

Section 5 of NUREG/CR-6909, Revision 1 demonstrates that either a factor of 10 to 12 on fatigue life is technically acceptable.

Further, as previously explained in the response to AREVA-6-3, the factor of 12 has been retained in this guidance for consistency with the ASME Code. In addition, further reduction of the factor of 12 on life should also simultaneously address the technical basis for the factor of 2 on stress or strain. Clarification of this point has been provided in Part B of the regulatory guide.

Therefore, there have been no changes to the draft RG in response to these comments.

AREVA-6-9 Comment The design fatigue curves and the F(en) factors are based on small specimens tested under membrane loading, although the plant components are not under membrane loading.

Therefore, the fatigue test results used for the methodologies referred to in this Reg. Guide are more severe than necessary. It is recommended that this observation be noted somewhere in this Reg. Guide. Since the fatigue tests are at membrane stress (in-air fatigue), they should as a minimum have been extended to failure and not just stop with the development of a flaw. This would have mitigated some of the conservatisms ingrained in the jump from membrane stress in the fatigue test to peak stress in the component.

NRC Staff Response The NRC staff agrees that in some applications, the methods described in the regulatory guide may be conservative. As stated in the response to RR-1, the following statement was added in Part B of the final regulatory guide to address this comment:

The staff recognizes that additional margin may be present in components where the dominant loading consists of a significant though-wall strain gradient, as opposed to the membrane loading typically applied in small-scale laboratory fatigue test specimens.

Methods to account for this additional margin may be considered by the staff on a case-by-case basis, provided sufficient basis and information is provided to the staff to verify that the proposed alternative demonstrates compliance with all applicable NRC regulations.

AREVA-6-10 Comment Page 6, Section C, first sentence: change This section describes the methods .... to This section describes, among others, the methods ..... The reason for adding among others is because Section C only refers to the newer NUREG/CR-6909 methodology, although Section B maintains the previously endorsed methods, such as NUREG/CR-6583 and NUREG/CR-5704. Although Section D mentions that Current licensees may continue to use guidance the NRC found acceptable for complying with the identified regulations as long as their current licensing basis remains unchanged, Section C should be much clearer regarding all the methodologies that are acceptable to the NRC.

NRC Staff Response The NRC staff disagrees with this comment. NUREG/CR-6583 and NUREG/CR-5704 are discussed in Part B to provide an historical background on the evolution of methods to evaluate environmental effects on the fatigue life of metal components and discuss the allowable use of these documents for initial license renewal. Section C of the RG provides the staffs regulatory position. The current regulatory position is clearly articulated in Section C and indicates that the method described in NUREG/CR-6909, Revision 1 should be used to comply with the regulatory guide. Alternatively, current licensees can propose to use NUREG/CR-5704, NUREG/CR-6583, and NUREG/CR-6909, Revision 0provided that a sufficient basis for their use is demonstrated as part of the application. However, these documents do not reflect the NRC staffs more recent studies and guidance, as further described in the responses to NEI-3-1, NEI-3-15, and NEI-3-30.

No changes were made to the draft RG in response to this comment.

AREVA-6-11 Comment Page 8, Section D, sub-section entitled Use by Applicants and Licensees: It is recommended to provide additional clarity between the current first and second paragraphs. A sentence such as this could be added: In general, NUREG/CR-5704 and NUREG/CR-6583 are the NUREG Documents that published previous approved methods (e.g. License Renewal) that were previously endorsed by the NRC staff. Per this Reg. Guide Revision, NUREG/CR-5704 and NUREG/CR-6583 remain applicable bases for current License Renewal evaluations.

It is only for the condition where the licensing bases are modified due to License Renewal extension that the EAF Methods also may be modified.

NRC Staff Response As discussed in the response to NEI-3-26, the purpose of Part D is to provide information on the

1) applicants and licensees use of the RG and 2) NRCs planned use of the RG. In addition, Part D describes how the NRC staff complies with 10 CFR 50.109, Backfitting, and any applicable issue finality provisions in 10 CFR Part 52.

Current licensees can propose to use NUREG/CR-5704, NUREG/CR-6583, and NUREG/CR-6909, Revision 0 provided that a sufficient basis forsuch use is demonstrated as part of the application. However, these documents do not reflect the NRC staffs more recent studies and guidance, as further described in the responses to NEI-3-1, NEI-3-15, and NEI-3-30.

No changes were made to the draft RG in response to this comment.

EPRI COMMENT AND RESPONSE EPRI-7-1 Comment Why does draft Regulatory Guide and NUREG 6909 not take into account radiation effects? It was published in 1970, under ASTM STP 484 and ASTM STP 529 that radiation reduces fatigue life by a factor of less than 2.5 for stainless steel. I would expect Carbon Steel such as A533B would have similar effects.

References:

ASTM STP 484 pp. 419-457, "Axial Fatigue of Irradiated Stainless Steels Tested at Elevated Temperatures", J. M. Beeston, and C. R. Brinkman.

ASTM STP 529, pp. 473-493, "Influence of Irradiation on Creep/Fatigue Behavior of Several Austenitic Stainless Steels and Incoloy 800 at 700 C" NRC Staff Response Radiation effects were considered in the development of this RG. Radiation effects are discussed in Section 1.3.2 of NUREG/CR-6909, Revision 1 (ADAMS number ML16319A004). In that section, available research is summarized, and used to demonstrate that the limited available data are inconclusive with regard to the significance of radiation effects on the fatigue lives of materials exposed to LWR environments.

ASTM STP 484 summarizes uniaxial fatigue properties from tests at 400, 500, 600, and 700°C on Type 304, 304L (titanium modified), and 316 stainless steel specimens irradiated at 450°C in sodium and 750°C in argon. ASTM STP 529 summarizes the results from strain controlled fatigue and constant-strain-rate tensile tests conducted at 700 to 750oC using specimens manufactured from Types 304, 304L (titanium modified), and 316 stainless steel, and Incoloy 800. The temperatures in these tests are significantly higher than LWR temperatures, which are typically less than 325oC. Further, the sodium or argon coolants used in these studies causes significantly different effects on fatigue life than LWR water environments. Therefore, neither of the cited studies are relevant to RG 1.207 because this RG, and the associated Fen methodology, is solely applicable to LWR water environments No changes were made to the draft RG in response to this comment.

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