ML22061A253

DG-1383 (RG 1.246 Rev 0) Public Comment Response

ML22061A253
Person / Time
Issue date:	10/26/2022
From:	Margaret Audrain, Bill Lin, Timothy Lupold, Robert Roche-Rivera NRC/NRR/DANU, NRC/RES/DE
To:
Roche-Rivera, R
Shared Package
ML22061A243	List: ML22061A244 ML22061A246 ML22061A253 ML22136A233 ML22136A238 ML22139A137
References
RG 1.246 Rev 0 DG-1383
Download: ML22061A253 (31)

Text

Response to Public Comments on Draft Regulatory Guide (DG)-1383 Acceptability of ASME Code Section XI, Division 2, Requirements for Reliability and Integrity Management (RIM) Programs for Nuclear Power Plants, for Non-Light Water Reactors On September 30, 2021, the NRC published a notice in the Federal Register (86 FR 54253) that Draft Regulatory Guide, DG-1383 (proposed new RG 1.246; Agencywide Documents Access and Management System [ADAMS] Accession No. ML21120A185) was available for public comment. The public comment period ended on November 15, 2021. The NRC received comments from the organizations listed below. The NRC has combined the comments and NRC staff responses in the following table. This document lists each public comment by letter and comment number. For example, Comment 1-1 would be the first comment provided in Letter No. 1 listed in the table below.

Letter ADAMS Accession No. Commenter Affiliation Commenter Name No.

1 ML21327A409 POMO18 Consult LLC A. Thomas Roberts III 2 ML21327A410 NuScale Power LLC Ross Snuggerud 3 ML21335A063 -- Henry Stephens 4* ML21327A412 -- Michael Turnbow 6 ML21327A417 -- N. Prasad Kadambi 7 ML21327A418 MPR Associates Robert Vayda 8 ML21327A420 NEI Mark Richter 9* ML21327A422 Anonymous Anonymous Letter/ Section of DG- Specific Comments NRC Resolution Comment 1383 No.

1-1 General A general comment is that traditional LWR considerations appears to The NRC staff disagreed with this permeate throughout and may have influenced the compilation of this draft comment. The staff is aware of the Regulatory Guide. For, example, by assuming refueling outages and linking differences in refueling outages any reporting submission criteria to them is may be inappropriate for a RIM between non-light water reactors program developed for some advanced designs. and light water reactors (LWRs)

Some advanced reactor designs are anticipated to be fueled for life and the and as such regulatory guidance intended design life is 20 years. Further, while some of these designs do position 4 addresses the alternative

Letters No. 5 (ML21327A416) and No. 10 (ML21327A423) are duplicate copies of letters No. 4. and No. 9, respectively.

October 2022

anticipate having scheduled maintenance outages, they also anticipate for an applicant/licensee to conducting as much online maintenance and monitoring as possible. propose an appropriate period for submitting the OAR report if there Consequently, from the submittal of the OAR report, it is suggested that the is not a refueling outage more Owner, based on the specifically developed RIM program, propose an frequently than every 5 years. In appropriate submittal period for reports and documents such as the OAR to response to comment 1-6 below, the USNRC. the regulatory guidance position 4 was revised to delete reference to a It is unlikely there will be a one size fits all submission period that refueling and used the term meaningfully would work for all advanced reactor designs. scheduled outage to be consistent with Appendix B of ASME Code,Section XI, Division 2.

1-2 Title (applicability It is understood that a Regulatory Guide cannot supersede existing The NRC staff agreed that RIM to non-LWR) regulations. It is also recognized that Title 10 CFR PART 50.55a was developed for any type of establishes regulation for the use of ASME XI Division 1 as applied to light reactor design. However, the NRC water reactors. However, a careful examination of ASME XI Division 1 staff reviewed and is endorsing reveals that the light water reactors that are appropriately addressed by ASME BPV Code,Section XI, ASME XI Division 1 consists of the existing fleet of PWR and BWR Division 2 only for use by non-operating plant designs. Several advanced reactors design that are currently lightwater reactors because 10 in development are light water cooled or moderated reactors, but for which CFR 50.55a(g) mandates the use of the use of ASME XI Division 1 is neither adequate, nor appropriate. In the ASME BPV Code,Section XI, some of these advanced LWR designs, even the traditional safety cases that Division 1 for boiling and are a foundational consideration for the application of ASME XI Division 1, pressurized water-cooled reactors.

may not be as relevant as the safety cases used in the existing fleet design (e.g., CDF, LERF, LOOP, etc.). If a boiling or pressurized water-cooled reactor licensee or applicant ASME XI Division 2 was developed to be technology neutral standard and wishes to use RIM, they would can be equally applied to non-LWR as well as LWR technology. In fact, need to request an exemption ASME XI Division 2 permits the use of Division 1 criteria if it is under 10 CFR 50.12 or 10 CFR appropriate for a given reactor design and distinguishes which PRA 52.7 from 10 CFR 50.55a(g).

standard should be used in the development of a RIM program when addressing a non-LWR versus a LWR advanced reactor design. A footnote to this effect was added to the RG.

It is recognized that a license applicant for any advanced LWR design could request an alternative to 10CFR 50.55a but this may not be immediately 2

obvious to some new reactor designers who may not have familiarity with the provisions of 10CFR50.55a (z).

Recommendation: It is recommended that DG-1383 be amended to provide a clarification to prospective future users that the use of ASME XI Division 2 is a potential option for advanced LWR designs that are not well suited to the use of ASME XI Division 1, and an explanation regarding the regulatory provisions that would need to be addressed to make use this option.

1-3 Section B, Basis The term safety is used throughout this draft RG. Reference to 10CFR50 The NRC staff agreed with this for Regulatory Appendix B is also a cited reference. It is not obvious that the term safety as comment in part, relating to RIM Guidance Position it is used throughout this draft RG is meant to infer traditional Safety application to risk significant SSCs 1 (Page 6) Related classification protocol (e.g., ASME Class 1, 2, 3 and Quality and that not all risk significant Groups A, B & C) or a broader use of the term safety (e.g., worker/public SSCs would traditionally be safety.) This is noted since ASME XI Division 2 applies no specific safety classified as safety-related.

classifications to SSC within a RIM Program. Instead, RIM requires that However, the NRC staff disagreed any SSC that is risk significant for the safe operation of a plant and with the comment that the NRC therefore worker and public safety is to be an SSC within the scope of the staffs use of the term safety is RIM developed program. This very well might include SSC that would confusing or that changes to the otherwise not be traditionally classified as Safety Related. DG are needed. The term safety is used in several contexts in the Recommendation: A better description as to what is specifically intended by DG. The majority of the time it is the use of the word SAFETY, so as to make it clear to users what SSC are related to the general concept of specifically to be addressed should be provided for clarification within this safety of from exposure to Regulatory Guide. radiation. The NRC staff have reviewed the DG for the use of the term safety and determined that it is not used in a manner that has any bearing on the guidance for applicants desiring to use RIM. In particular, the DG does not use the term safety in relationship to the protocol of classifying equipment such as designating components as ASME Class 1, 2, or 3, or Quality Groups A, B, or C.

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It is important to note that the terms important to safety and safety-significant are NRC regulatory classifications independent of RIM. For example, important to safety is a key term in the Genderal Design Criteria (GDC) of Appendix A to 10 CFR part 50. In general, when the NRC uses the term safety, it does not necessarily reference a regulatory classification and should be interpreted based on context. In DG-1383, where the term important to safety is used, it is meant in the context of the referenced documents, specifically in the advanced reactor design criteria (ARDC), which were developed in reference to the GDC. The term safety-related is used in the definition of Inservice Inspection (ISI) Program, which is identical to the definition in IWA-9000 for inservice inspection.

No changes were made to the regulatory guide as a result of this comment.

1-4 Section B, Basis RIM establishes that an Inspection Interval shall not exceed 12 years (i.e., The NRC staff agreed with for Regulatory 144 months) irrespective of the time between refueling (e.g., on-line recommendation (1) not to Guidance Position refueling vs traditional off-line refueling.) This was created with the reference refueling outage but did not agree that the end of the 4

1 (Page 7 - 3rd recognition that some advanced reactor designs may not be taken out of inspection interval is the bullet) service (e.g., offline) to accommodate a refueling. From an ASME Code appropriate time to submit OAR perspective it would be at the end of this 12-year interval that an OAR-1 forms. RIM-2.7.2 addresses the Form would be prepared. inspection interval but makes no mention of OAR form submission.

It is understandable why the USNRC may not wish to wait until the end of a Non-Mandatory Appendix B 12 interval to receive information contained in a prepared OAR-1 Form for recommends the OAR be twelve years and hence has established that a five-year periodicity for the submitted within 90 days of the preparation of an OAR-1 be created and submitted. completion of each scheduled outage. Regulatory guidance There are however two clarifications that would be useful to include in DG- position 4 in this RG, addresses the 1383 regarding this matter. frequency for OAR submittal.

Specifically, regulatory guidance The first is the descriptor of refueling outage. As noted, some advanced position 4 addresses submittal of reactor designs may not have a traditional refueling outage, but instead a the OAR forms within 120 days maintenance outage as the unit remains operational while refueling, or may after the completion of an outage, be fueled for life. consistent with ASME Code,Section XI, Division 1. The term The second is that with the 2019 Edition of ASME XI, both in Division 1 refueling was removed from the and Division 2, the FORM NIS-2 is merely an attestation for the completion third bullet on page 7 and other of a Repair and Replacement Activity (RRA) signed by the Licensee and an locations in the regulatory guide.

Authorized Nuclear Inservice Inspector.

The staff agreed with There is, however, no insightful or technical information contained on the recommendation (2) that it is not FORM NIS-2 itself. This change represents the incorporation of ASME necessary to submit the NIS-2 Code N-532, where RRA that were required because an SSC had failed to form, as the OAR form includes an established ASME XI acceptance criteria would be documented on the repair/replacement information.

OAR-1 Form but the FORM NIS-2 became abbreviated. Regulatory guide position 1 was revised to remove reference to the Recommendations: (1) Provide a clarification to the presently used term NIS-2.

refueling outage to be more inclusive to evolutions such as maintenance outages or a description of a time frame for the desired submittal of an OAR-1 Form, exclusive of using the terminology refueling outage.

(2) Consider deleting the requirement to submit NIS-2 Forms as suggested, since the information that is likely of interest to the USNRC will already be contained in the OAR-1 Form.

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1-5 Section B, Basis ASME III Division 1 is for traditional LWR reactor designs and the normal The NRC staff agreed with this for Regulatory operating temperature ranges of LWRs. In contrast, ASME III Division 5 is comment in part and disagreed in Guidance Position for High temperature reactors and includes provisions for certain HAA part. The NRC staff agreed that 1 (Page 7 - 4th materials to operate in the creep regime with a stated maximum number of Section III, Division 1 may not be bullet) permissible cycles. This same limitation is found in RIM Appendix VII applicable to some designs that Article VII-3 .1. Consequently, it is not obvious why ASME Division 1 is would implement RIM. However, cited for advanced reactor design flaw acceptance criteria. the NRC staff has not reviewed the flaw evaluation acceptance criteria Recommendation: Consider, revising this bullet so it is clearer that for for use in temperature ranges advanced reactor designs that operate in the creep regime such as those above those described in Section designed in accordance and permitted by ASME Section III Division 5, that III, Division 1, and cannot endorse appropriate justification for flaw evaluation acceptance criteria shall be their use at such temperatures.

provided by the applicant. Should the applicable temperatures for a component exceed these temperature ranges, the applicant or licensee will need to establish the acceptance criteria and submit to the NRC for review and approval. No changes were made to the regulatory guide as a result of this comment.

1-6 Section B, Basis Similar to the background and recommendation [for comment 1-4] noted The staff agreed with this for Regulatory above, the use of the term refueling outage may not specifically have a comment that the term refuling Guidance Position universally understood definition for some advanced designs, that may not outage is not appropriate.

4 (Page 8, 5th have a discrete refueling outage or cycle. Additionally, this Regulatory Reference to refueling was paragraph) Guidance Position states that: Licensees should submit the notification prior removed to make the regulatory to the next refueling outage or within 3 years, whichever is less.. This 3- guide consistent with Non-year periodicity appears to be contradictory to the Regulatory Guidance Mandatory Appendix B which Position offered in Regulatory Guidance Position 1 (page 7 - third bullet) specifies OAR forms to be which reflects a five year or less periodicity. submitted after scheduled outages.

However, the NRC staff disagreed Recommendation: Provide a clarification to the presently used term that there is a contradiction refueling outage to be more inclusive to evolutions such as maintenance between the timeframe for outages or a description of a time frame for the desired submittal of an submitting an OAR and timeframe OAR-1 Form, exclusive of using the terminology refueling outage, and (2) in position 1. The durations provided for submitting a 6

clarify what appears to be discrepancy in the desired periodicity reflected in notification of a change to the RIM the two noted Regulatory Guidance Positions. program is different than the frequency for submitting an OAR.

The NRC staff suggested that an OAR may be used to notify the NRC of a RIM program change for convenience to the licensee, but the time frames are for different aspects, one for the notification of program changes and one for the results of activities such as inspections. No change was made related to this portion of the comment.

1-7 Section B, Basis This Regulatory Basis is understandable since the staff has not endorsed The NRC staff agreed with this for Regulatory either Code Case N-788-1 nor ANDE-1 2015. comment in part and disagreed in Guidance Position part. The RG was revised to clarify 5 However, as written it is not clear whether the intent of the basis is to imply that methods approved for that the performance demonstration type of protocol of ASME XI Division qualification and certification of 1 as is found in Division 1 Appendix VIII is meant to be employed. NDE personnel are not dependent on the reactor types and therefore Performance demonstration of any MANDE that may be selected for an non-LWRs should use any SSC under a RIM program is an essential and imperative input and methods approved by the NRC for quantitatively factors into the establishment a Reliability Target assigned to use by LWRs, e.g., in 10 CFR an SSC for such considerations a Probability of Detection (POD) criteria. 50.55a. However, the NRC staff disagrees that guidance can be It is believed that while the USNRC has not formally endorsed the use of provided on how licensees or either Code Case N-788-1 or ANDE-1 2015, the reservation by the staff is applicants can request approval to understood to be based on the fact that ANDE-1 describes a process for the use ANDE-1, which the NRC staff qualification of NDE personnel. If that is in fact the existing reservation has previously found too then the following recommendation is provided. incomplete to approve for use.

Nevertheless, because the NRC Recommendation: Provide guidance regarding the type of information that staff finds that it is appropriate for would be expected to be provided by a Licensee applicant that would allow non-LWR licensees to use any for consideration of approval by the USNRC to use ANDE-1 for NDE method of qualification and personnel qualification under Division 2. This would assure that factors certification of NDE approved for 7

such as POD for any MANDE methods selected are established with use by LWRs, then if the NRC consistency. The use of other already endorsed NDE personnal qualification approves the use of ANDE-1 for criteria, such as CP-189 do not afford this essential criteria (i.e., POD) and LWRs, that approval would extend potentially undermines a cornerstone to the development of a sound RIM to non-LWRs.

Program.

Further, the RG was revised to specify that non-LWR licensees should apply the conditions in 10 CFR 50.55a(b)(2) when using ASNT/CP-189. Additionally, the RG was revised to clarify the requirements for performance demonstrations. The staff considers using Section XI, Division 1 Appendix VIII for performance demonstration, when applying the conditions established within 10 CFR 50.55a(b)(2),

acceptable.

1-8 Section C, It is understood that the USNRC would seek to have summaries of: The NRC staff agreed with this Regulatory (a) The bases for the scope of the program and comment and revised the RG to Guidance Position (b) RIM strategies selected to achieve the reliability targets, as denoted in identify a list of structures, 1 (Page 14, 2nd and this portion of the Staff Regulatory Guidance at the time a systems, and components (SSCs) 8th bullets) Licensee/applicant makes an initial filing. However, both of these included in the scope of the RIM provisions may not be fully developed at the time of initial application and program rather than a summary of would not be fully vetted to be able to provide a detailed listing of all bases the bases for the scope of the RIM or strategies that may apply to each SSC selected to be within a final RIM program.

program.

In regard to the RIM strategies to Recommendation: It is suggested that a clarification be provided to both of be used, only a description of the these items so as to better define what contents are expected to be provided types of factors as outlined in by a Licensee/applicant at the time of license application. RIM-2.5.1 that are used in the RIM strategies needs to be provided.

For example, identifying the monitoring or NDE methods that will be used, any testing strategies, 8

periodic maintenance, repair, or replacements, etc. A listing for each SSC is not necessary. If the staff needs additional information, the staff may audit the applicant/licensees RIM program.

If information is not complete at the time of application, then the application itself should contain a schedule for when the information will be completed and submitted to the NRC. For example, if the RIM program is being submitted for a Construction Permit then such application should describe the information that would be included in an Operating License application.

1-9 Section C, As previously outlined in comment [1-4] above, the FORM NIS-2 is merely The NRC staff agreed with this Regulatory an attestation for the completion of a Repair and Replacement Activity comment and deleted reference to Guidance Position (RRA) signed by the Licensee and an Authorized Nuclear Inservice submitting Form NIS-2.

1 (Page 14, 12th Inspector.

bullet)

There is, however, no insightful or technical information contained on the FORM NIS-2 itself because the 2019 Edition of ASME XI Division 1 and Division 2 represents the incorporation of ASME Code N-532, where RRA that were required because an SSC had failed to an established ASME XI acceptance criteria would be documented on the OAR-1 Form, but the FORM NIS-2 became abbreviated.

Recommendation: Consider deleting the requirement to submit NIS-2 Forms as suggested, since the information that is likely of interest to the USNRC will already be contained in the OAR-1 Form.

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1-10 Section C, As cited in comment [1-7] above, performance demonstration of any The NRC staff agreed with the Regulatory MANDE that may be selected for an SSC under a RIM program is an comment in part that ANDE-1 and Guidance Position essential and imperative input and quantitatively factors into the ANSI/ASNT CP-189 are personnel 5 establishment of a Reliability Target assigned to an SSC by using qualification and certification considerations such as Probability of Detection (POD) criteria. The use of programs. Performance ANDE-1 requires such performance demonstrations that are not a mandate demonstration is a separate of ANSI/ASNT CP-189. activity, governed by ASME Section XI, Division 1, Appendix Recommendation: Consider providing guidance regarding the type of VIII. While ANDE-1 and information that would be expected to be provided by a Licensee applicant ANSI/ANS CP-189 include NDE that would allow for consideration of approval by the USNRC to use demonstration activities as part of ANDE-1 for NDE personnel qualification under Division 2 as an alternative the qualification and certification to advocating the use of ANSI/ASNT CP-189. process, performance demonstration is a separate task, to ensure that equipment, procedures, and personnel will be capable of properly examining and finding flaws in components. As outlined in the RIM program, it is the MANDEEP function to establish performance demonstration requirements.

The RG was clarified to indicate that non-LWRs should use any methods approved by the NRC for use by LWRs, e.g., in 10 CFR 50.55a, for qualification and certification of NDE personnel.

However, the NRC staff disagrees that guidance can be provided on how licensees or applicants can request approval to use ANDE-1, which the NRC has previously found too incomplete to approve for use. Nevertheless, because the 10

NRC staff finds that it is appropriate for non-LWR licensees to use any method of qualification and certification of NDE approved for use by LWRs, then if the NRC approves the use of ANDE-1 for LWRs, that approval would extend to non-LWRs.

Further, the RG was revised to specify that the conditions in 10 CFR 50.55a (b)(2) are applicable to personnel qualification when using ASNT/CP-189. Additionally, the RG was revised to clarify the requirements for performance demonstrations. The staff considers using Section XI, Division 1 Appendix VIII for performance demonstration when applying the conditions established within 10 CFR 50.55a(b)(2),

acceptable.

2-1 Section A, This comment requests that RG 1.246 include light water reactors (LWRs) The NRC staff agrees with the Applicability within its scope to facilitate implementation of ASME Section XI Division comment that RIM was developed 2 by LWR applicants and licensees via the code alternative process. The for any type of reactor design, but comment states: takes no position on the technical adequacy of RIM for LWRs. The The applicability of DG-1383 to non-LWRs is in contrast with the NRC staff disagrees that RG 1.246 intended scope of ASME Section XI Division 2, which specifically should be expanded to address provides applicability for LWRs. ASME Section XI Division 2, Section LWRs. The purpose and scope of RIM-1.1(a) states that This Division provides the requirements for the this RG is to provide guidance for creation of the Reliability and Integrity Management (RIM) Program for all non-LWRs that are not subject to types of nuclear power plants (emphasis added). the requirements of 50.55a to implement ASME BPV,Section XI, Division 1. The NRC staff 11

Notwithstanding, NuScale recognizes that current NRC regulations at 10 agrees that LWR applicants or CFR 50.55a require the application of ASME Section XI, Division 1 for licensees that wish to use RIM non-LWRs, and thus RG 1.246 cannot directly endorse ASME Section XI would need to request an Division 2 for use by LWR applicants/licensees. NuScale seeks a future appropriate exemption.

amendment of 10 CFR 50.55a to incorporate by reference ASME Section XI Division 2 for LWRs. The NRC staff also takes no position on rulemaking to Until that rulemaking can occur, NRC should explicitly recognize in RG incorporate Division 2 into 50.55a.

1.246 that, from a technical perspective, ASME Section XI Division 2 is Rulemaking is outside the scope of applicable and appropriate for implementation by LWR this RG.

applicants/licensees. In order for a LWR applicant/licensee to implement it, they would still need to seek approval of a code alternative under 10 CFR 50.55a(z) and/or an exemption under the controlling regulations for the facility (10 CFR 50.12, 10 CFR 52.7, and/or applicable design certification rule provisions), but the technical basis and conditions for doing so would be established in advance.

Recommendation:

• Expand the Applicability of RG 1.246 to LWR designs;

• Include in the Applicable Regulations 10 CFR 50.55a(z);

• Address LWRs within the Background discussion to explain that similar considerations for non-LWRs apply to LWRs and the satisfaction of the General Design Criteria (GDCs) therefor;

• Include an additional regulatory guidance position and basis for that position to address applicability of the guidance to LWRs. This proposed Regulatory Guidance Position would recognize that ASME Section XI Division 2, as conditioned by RG 1.246, provides for LWRs an acceptable approach to satisfy 10 CFR 50.34(b)(6)(iv), 10 CFR Part 50 Appendix B, 10 CFR 52.79(a)(5), 10 CFR 52.79(a)(29)(i), and the relevant GDCs. This proposed Regulatory Guidance Position would impose the preceding 15 positions on LWRs, and further condition implementation of ASME

• Section XI Division 2 for LWRs on the applicant/licensee seeking and receiving authorization of a code alternative under 10 CFR 50.55a(z) and/or an exemption under the controlling regulations for the facility.

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Extending the technical basis and regulatory positions of RG 1.246 to LWRs would substantially reduce the burden associated with a LWR applicant/licensees request to implement ASME Section XI Division 2 as a code alternative; the RG would support a conclusion under 10 CFR 50.55a(z) that RIM achieves an acceptable level of quality and safety for LWRs. NRC has followed a similar approach in other contexts. For example, NUREG-1791 guides the Staff in reviewingand by extension supports an applicant in seekingan exemption from the licensed operator staffing requirements of 10 CFR 50.54(m).

In conclusion, NuScale requests that, until 10 CFR 50.55a is amended to incorporate by reference ASME Section XI Division 2 for LWRs, NRC recognize the technical acceptability of RIM for LWRs by including LWR applicants and licensees within the scope of RG 1.246, subject to NRCs specific approval of an alternative to the required ASME Section XI Division 1.

3-1 RG Title and ASME XI Division 2 is applicable as a technology neutral standard and has The NRC staff agrees with the Section A, been developed to be applied to both non-LWR as well as LWR comment that RIM was developed Applicability technology. ASME XI, Division permits the use of Division 1 criteria if it is for any type of reactor design, but appropriate for a given reactor design and distinguishes which PRA takes no position on the technical standard should be used in the development of a RIM program when adequacy of RIM for LWRs. The addressing a non-LWR versus a LWR advanced reactor design. NRC staff disagrees that RG 1.246 should be expanded to address It is understood that Title 10 CFR PART 50.55a establishes regulation for LWRs. The purpose and scope of the use of ASME XI Division 1 as applied to light-water reactors. Further, this RG is to provide guidance for ASME XI Division 1 addresses the existing fleet of PWR and BWR non-LWRs that are not subject to operating plant designs. Several advanced light-water reactors are currently the requirements of 50.55a to in developments. The use of ASME XI Division 1 is neither adequate, nor implement ASME BPV, Section appropriate for these newer designs. In some of these advanced LWR XI, Division 1. LWR applicants or designs, the traditional safety cases that are a foundational consideration for licensees that wish to use RIM the application of ASME XI Division 1, may not be as relevant as the safety would need to request an cases used in the existing fleet design. appropriate exemption.

It is recognized that a license applicant for any advanced LWR design could request an alternative to 10CFR 50.55a but this may not be immediately 13

obvious to some new reactor designers who may not have familiarity with the provisions of 10CFR50.55a (z).

Recommendation: It is recommended that DG-1383 be amended to provide guidance to users that the use of ASME XI Division 2 is a potential option for advanced LWR designs when ASME XI Division 1 is not applicable and directs them to the provisions of 10CFR50.55a (z).

3-2 Section B, Basis The term safety is used throughout this draft RG. Reference to 10CFR50 The NRC staff agreed with this for Regulatory Appendix B is also a cited. It is not clear if the term safety is meant to comment in part, relating to RIM Guidance Position address traditional Safety-Related classifications (e.g., ASME Class 1, 2, application to risk significant SSCs 1 3 and Quality Groups A, B & C) or a broader use (e.g., worker/public and that not all risk significant safety). ASME XI Division 2 applies no specific safety classifications to SSCs would traditionally be SSC within a RIM Program. Instead, RIM requires that any SSC that is classified as safety-related.

risk significant for the safe operation of a plant and therefore worker and However, the NRC staff disagreed public safety is to be an SSC within the scope of the RIM developed with the comment that the NRC program. This might include SSC that would otherwise not be traditionally staffs use of the term safety is classified as Safety-Related. confusing or that changes to the DG are needed. The term safety Recommendation: A description as to what is specifically intended by the is used in several contexts in the use of the word safety would make it clearer as to what SSCs are DG. The majority of the time it is specifically to be addressed. related to the general concept of safety of from exposure to radiation. The NRC staff have reviewed the DG for the use of the term safety and determined that it is not used in a manner that has any bearing on the guidance for applicants desiring to use RIM. In particular, the DG does not use the term safety in relationship to the protocol of classifying equipment such as designating components as ASME Class 1, 2, or 3, or Quality Groups A, B, or C.

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It is important to note that the terms important to safety and safety-significant are NRC regulatory classifications independent of RIM. For example, important to safety is a key term in the Genderal Design Criteria (GDC) of Appendix A to 10 CFR part 50. In general, when the NRC uses the term safety, it does not necessarily reference a regulatory classification and should be interpreted based on context. In DG-1383, where the term important to safety is used, it is meant in the context of the referenced documents, specifically in the advanced reactor design criteria (ARDC), which were developed in reference to the GDC. The term safety-related is used in the definition of Inservice Inspection (ISI) Program, which is identical to the definition in IWA-9000 for inservice inspection.

No changes were made to the regulatory guide as a result of this comment.

3-3 Section B, Basis RIM establishes that an Inspection Interval shall not exceed 12 years (i.e., The NRC staff agreed with for Regulatory 144 months) irrespective of the time between refueling (e.g., on-line recommendation (1) not to Guidance Position refueling vs traditional off-line refueling.) This was created with the reference refueling outage but 1; Page 7, 3rd recognition that some advanced reactor designs may not be taken out of did not agree that the end of the bullet service (e.g., offline) to accommodate a refueling outage. From an ASME inspection interval is the appropriate time to submit OAR 15

Code perspective it would be at the end of this 12-year interval that an forms. RIM-2.7.2 addresses the OAR-1 Form would be prepared. inspection interval but makes no mention of OAR form submission.

It is understandable why the USNRC may not wish to wait until the end of a Non-Mandatory Appendix B 12 interval to receive information contained in a prepared OAR-1 Form for recommends the OAR be twelve years and hence has established that a five-year periodicity for the submitted within 90 days of the preparation of an OAR-1 be created and submitted. completion of each scheduled outage. Regulatory guidance There are however two clarifications that would be useful to include in DG- position 4 in this RG, addresses the 1383 regarding this matter: frequency for OAR submittal.

Specifically, regulatory guidance One is the descriptor of refueling outage. As noted, some advanced reactor position 4 addresses submittal of designs may not have a traditional refueling outage, but instead a the OAR forms within 120 days maintenance outage where the unit remains operational while refueling. after the completion of an outage, consistent with ASME Code, Two is that with the 2019 Edition of ASME XI, both in Division 1 and Section XI, Division 1. The term Division 2, the FORM NIS-2 is an attestation for the completion of a refueling was removed from the Repair/Replacement Activity (RRA). third bullet on page 7 and other locations in the regulatory guide.

The FORM NIS-2 does not include any significant or technical information.

This change represents the incorporation of ASME Code Case N-532, The staff agreed with where RRA that were required because an SSC had failed to an established recommendation (2) that it is not ASME XI acceptance criteria would be documented on the OAR-1 Form necessary to submit the NIS-2 and the FORM NIS-2 became abbreviated. form, as the OAR form includes repair/replacement information.

Recommendations: Regulatory guide position 1 was

1. Provide a clarification to the presently used term refueling outage to be revised to remove reference to the more inclusive to evolutions such as maintenance outages or a NIS-2.

description of a time frame for the desired submittal of an OAR-1 Form, exclusive of using the terminology refueling outage.

2. Consider deleting the requirement to submit NIS-2 Forms.

3-4 Section B, Basis ASME III Division 1 is for traditional LWR reactor designs and the normal The NRC staff agreed with this for Regulatory operating temperature ranges of LWRs. In contrast, ASME III Division 5 is comment in part and disagreed in Guidance Position for High temperature reactors and includes provisions for certain HAA part. The NRC staff agreed that 1; Page 7, 4th bullet materials to operate in the creep regime with a stated maximum number of Section III, Division 1 may not be 16

permissible cycles. This same limitation is found in RIM Appendix VII applicable to some designs that Article VII-3 .1. Consequently, it is not obvious why ASME Division 1 is would implement RIM. However, cited for advanced reactor design flaw acceptance criteria. the NRC staff has not yet endorsed Division 5, and as such it would be Recommendation: Consider, revising this bullet so it is clearer that for inappropriate to reference it in this advanced reactor designs that operate in the creep regime such as those RG. In any case, appropriate designed in accordance and permitted by ASME Section III Division 5, that temperature limits should be appropriate justification for flaw evaluation acceptance criteria shall be determined by the applicants provided by the applicant. construction code. Therefore, the NRC staff revised the draft guide to refer to the temperature limits of the applicants construction code 3-5 Section B, Basis Similar to the background and recommendation [3-3] noted above, the use The staff agreed with this for Regulatory of the term refueling outage may not specifically have a universally comment that the term refuling Guidance Position understood definition for some advanced designs, that may not have a outage is not appropriate.

4; Page, 5th discrete refueling outage or cycle. Additionally, this Regulatory Guidance Reference to refueling was paragraph Position states that: Licensees should submit the notification prior to the removed to make the regulatory next refueling outage or within 3 years, whichever is less. This 3-year guide consistent with Non-periodicity appears to be contradictory to the Regulatory Guidance Position Mandatory Appendix B which offered in Regulatory Guidance Position 1 (page 7 - third bullet) which specifies OAR forms to be reflects a five year or less periodicity. submitted after scheduled outages.

However, the NRC staff disagreed Recommendation: Provide a clarification to the presently used term that there is a contradiction refueling outage to be more inclusive to evolutions such as maintenance between the timeframe for outages or a description of a time frame for the desired submittal of an submitting an OAR and timeframe OAR-1 Form, exclusive of using the terminology refueling outage, and (2) in position 1. The durations clarify what appears to be discrepancy in the desired periodicity reflected in provided for submitting a the two noted Regulatory Guidance Positions. notification of a change to the RIM program is different than the frequency for submitting an OAR.

The NRC staff suggested that an OAR may be used to notify the NRC of a RIM program change for convenience to the licensee, but the time frames are for different aspects, one for the notification of 17

program changes and one for the results of activities such as inspections. No change was made related to this portion of the comment.

3-6 Section B, Basis The comment refers to conditions from 10 CFR 50.55a and Section XI, The NRC staff agreed with this for Regulatory Division 1, IWA-2300, and suggests that such conditions should be comment in part and disagreed in Guidance Position addressed in the RG. The comment also provides background on activities part. The RG was revised to clarify 5 that led to the development of the ANDE-1 Standard. Further, the comment that methods approved for describes the formation of a Certifying Body for the implementation of the qualification and certification of ANDE-1 standard. The comment provided the following recommendations. NDE personnel are not dependent on the reactor types and therefore Recommendations: non-LWRs should use any ANSI/ASNT CP-189: methods approved by the NRC for 1a. Specify the conditions on Section XI, Division 1, addressed in 10 CFR use by LWRs, e.g., in 10 CFR 50.55a, (b)(2)(xv) and (b)(2). 50.55a. However, the NRC staff 1b. Reinstate the IE Bulletin 82-03 requirement that training for IGSCC disagrees that guidance can be detection, sizing and overlay weld repair examinations. provided on how licensees or 1c. Reinstate the requirement for 3-year recertification for PDI IGSCC applicants can request approval to detection, sizing and overly weld repair examinations. 1d. Require 3-year use ANDE-1, which the NRC staff recertification for all Mandatory Appendix VIII PDI qualification has previously found too examinations. incomplete to approve for use.

Nevertheless, because the NRC

2. Condition the listed Section XI, Division 1, IWA-2300 amended staff finds that it is appropriate for requirements: non-LWR licensees to use any 2a. IWA-2314 Certification and Recertification except that the ASNT method of qualification and Level III certificate is not required. It is recommended that the third-party certification of NDE approved for ASNT Level III certificate, ACCP certificate or another recognized third- use by LWRs, then if the NRC party qualification with a NQA-1 or Appendix B QA program, e.g., EPRI approves the use of ANDE-1 for NDE Center be required. Additionally, contrary to the ASME Interpretation LWRs, that approval would extend XI-1-10-34 an audit of the ASNT Level III or ACCP certificate program to non-LWRs.

would be subject to audit by the licensee.

2b. IWA-2380 NDE INSTRUCTOR In lieu of the requirements of CP- Further, the RG was revised to 189, It is recommended the CP-189, third-party requirements are specify that non-LWR licensees maintained as compared to those amended by IWA-2380. should apply the conditions in 10 CFR 50.55a(b)(2) when using 18

2c. Additionally, the DG does not address the conditions related to Division ASNT/CP-189. Additionally, the 1, Mandatory Appendices VI, VII or VIII. RG was revised to clarify the requirements for performance ANDE-1: demonstrations. The staff Provide guidance regarding the type of information that would be expected considers using Section XI, to be provided by a Licensee applicant that would allow for consideration of Division 1 Appendix VIII for approval by the USNRC to use ANDE-1 for NDE personnel qualification performance demonstration, when under Division 2. This would assure that factors such as POD for any applying the conditions established MANDE methods selected are established with consistency. The use of within 10 CFR 50.55a(b)(2),

other already endorsed NDE personnal qualification criteria, such as CP- acceptable.

189 do not afford this essential criteria (i.e., POD) and potentially undermines a cornerstone to the development of a sound RIM Program.

3-7 Section C, Staff It is understood that the USNRC would seek to have summaries of: The NRC staff agreed with this Regulatory (a) The bases for the scope of the program and comment and revised the RG to Guidance 1, 2nd (b) RIM strategies selected to achieve the reliability targets, as denoted in identify a list of SSCs included in and 8th bullets this portion of the Staff Regulatory Guidance at the time a the scope of the RIM program Licensee/applicant makes an initial filing. However, both of these rather than a summary of the bases provisions may not be fully developed at the time of initial application and for the scope of the RIM program.

would not be fully vetted to be able to provide a detailed listing of all bases or strategies that may apply to each SSC selected to be within a final RIM In regard to the RIM strategies to program. be used, only a description of the types of factors as outlined in Recommendation: It is suggested that a clarification be provided to both of RIM-2.5.1 that are used in the RIM these items so as to better define what contents are expected to be provided strategies needs to be provided.

by a Licensee/applicant at the time of license application. For example, identify the monitoring or NDE methods that will be used, any testing strategies, periodic maintenance, repair, or replacements, etc. A listing for each SSC is not necessary. If the staff needs additional information, the staff may audit the applicant/licensees RIM program.

If information is not complete at the time of application, then the 19

application itself should contain a schedule for when the information will be completed and submitted to the NRC. For example, if the RIM program is being submitted for a CP then such application should describe the information that would be included in an OL application 3-8 Section C, Staff As previously outlined in comment [3-3] above, the FORM NIS-2 is merely The NRC staff agreed with this Regulatory an attestation for the completion of a Repair and Replacement Activity comment and deleted reference to Guidance 1, 12th (RRA) signed by the Licensee and an Authorized Nuclear Inservice submitting Form NIS-2.

bullet Inspector. There is, however, no insightful or technical information contained on the FORM NIS-2 itself because the 2019 Edition of ASME XI Division 1 and Division 2 represents the incorporation of ASME Code Case N-532, where RRA that were required because an SSC had failed to an established ASME XI acceptance criteria would be documented on the OAR-1 Form, but the FORM NIS-2 became abbreviated.

Recommendation: Consider deleting the requirement to submit NIS-2 Forms as suggested, since the information that is likely of interest to the USNRC will already be contained in the OAR-1 Form.

3-9 Section C, Staff As cited in comment [3-6] above, performance demonstration of any The NRC staff agreed with the Regulatory MANDE that may be selected for an SSC under a RIM program is an comment in part that ANDE-1 and Guidance 5 essential and imperative input and quantitatively factors into the ANSI/ASNT CP-189 are personnel establishment a Reliability Target assigned to an SSC by using qualification and certification considerations such as Probability of Detection (POD) criteria. The use of programs. Performance ANDE-1requires such performance demonstrations that are not a mandate demonstration is a separate of ANSI/ASNT CP-189. activity, governed by ASME Section XI, Division 1, Appendix Recommendation: Consider providing guidance regarding the type of VIII. While ANDE-1 and information that would be expected to be provided by a Licensee applicant ANSI/ANS CP-189 include NDE that would allow for consideration of approval by the USNRC to use demonstration activities as part of ANDE-1 for NDE personnel qualification under Division 2 as an alternative the qualification and certification to advocating the use of ANSI/ASNT CP-189. process, performance 20

demonstration is a separate task, to ensure that equipment, procedures, and personnel will be capable of properly examining and finding flaws in components. As outlined in the RIM program, it is the MANDEEP function to establish performance demonstration requirements.

The RG was clarified to indicate that non-LWRs should use any methods approved by the NRC for use by LWRs, e.g., in 10 CFR 50.55a, for qualification and certification of NDE personnel.

However, the NRC staff disagrees that guidance can be provided on how licensees or applicants can request approval to use ANDE-1, which the NRC has previously found too incomplete to approve for use. Nevertheless, because the NRC staff finds that it is appropriate for non-LWR licensees to use any method of qualification and certification of NDE approved for use by LWRs, then if the NRC approves the use of ANDE-1 for LWRs, that approval would extend to non-LWRs.

Further, the RG was revised to specify that the conditions in 10 CFR 50.55a (b)(2) are applicable to personnel qualification when 21

using ASNT/CP-189. Additionally, the RG was revised to clarify the requirements for performance demonstrations. The staff considers using Section XI, Division 1 Appendix VIII for performance demonstration when applying the conditions established within 10 CFR 50.55a(b)(2),

acceptable.

4-1 Section B, Basis As background, the comment provided a summary of industry experience, The NRC staff disagreed with this for Regulatory field failures and conclusions of round robin studies, based on which the comment because approval of Guidance Position comment stated that it was extradentary to find NRC position 5 supporting ANDE-1 is outside the scope of 5 the continuance of CP-189. The comment also stated the following points as the RG. However, in response to a summary: other ANDE-1 and CP-189 related

• For over 50 years the employer based self-certification process has been comments, the RG was clarified to unreliable to consistently qualify NDE personnel that meet industry indicate that non-LWRs should use expectations (detect flaws before failure). any methods approved by the NRC

• All round robin studies have shown detection rates at best are 50% for use by LWRs, e.g., in 10 CFR

• No known study has ever shown the process to be effective 50.55a, for qualification and

• The unstructured training and experience do not include the key factors certification of NDE personnel.

essential for learning Thus, because the NRC staff finds

• INPO and Academia agree the process does not have a path for that it is appropriate for non-LWR consistent reliability and success licensees to use any method of qualification and certification of Further, the comment requested the information and provided the NDE approved for use by LWRs, recommendations below. if the NRC approves the use of ANDE-1 for LWRs, that approval Request for Additional Information: would extend to non-LWRs. This In response to position 5 [above], the NRC has been involved with the means that non-LWRs applicants ANDE project since the beginning and has invested both manpower and and licensees can use ANDE-1 taxpayer dollars. ANDE-1 has successfully completed both revision 1 and 2 without need for further NRC through the ANSI Standards review process with no unresolved NRC approvals, once that work has been comments. It is essential for the NRC to specifically identify what is needed completed and approved by the to address the following in position 5: Code Case N-788-1 and the ANDE NRC in other processes.

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standard do not contain sufficient specificity for use as a qualification or certification program. Several important sections of ASME ANDE-1-2015 are not defined and are to be determined in the future by specific industry sector committees. It is not possible for the NRC to evaluate a certification and qualification program that has not been defined.

• What specificity?

• What is not defined?

• Several important sections of ASME ANDE-1-2015 are not defined.

• Considering CP-189 defines very little except Time. Where ANDE-1 defines knowledge and skills through Job Task Analysis with a qual card to document learning/experience and demonstration of proficiency.

Specifically, what is not defined?

Recommendations:

Recommendation #1: Include ANDE-1 in Regulator Guide DG-1383 as the only option.

Recommendation #2: In accordance with NRC concerns documented in January 12, 2010 letter to ASME (ML10040091) including industry experience, field failures and round robin study results, communicate to the industry that SNT-TC-1A and CP-189 to be phased out of the nuclear codes and that a process and schedule to be developed for an orderly transition to ANDE-1.

Recommendation #3: To assure NDE performance and continued nuclear power plant safety, the NRC should require all PDI supplement qualifications to requalify every 3 years. Based on over 30 years of IGSCC requalification pass rates at approximately 50%, it can only be expected that PDI initially qualified examiners have also lost continuity and proficiency with expected requalification pass rates to be at 50% or less.

6-1 General The Nuclear Energy Innovation and Modernization Act (NEIMA) This comment appears to be incorporates the term risk-informedand performance-based (RIPB) in outside the scope of this RG.. The several places. NEIMA does not offer a definition for RIPB. The staff intent of this review was not to should use the Commission developed definitions in SRM-SECY-98-0144 develop a new program but to and use the work that has been done over the past 20 years employing these review an existing consensus 23

definitions. Although risk-informed (RI) and performance-based (PB) have standard for use. This standard, progressed on separate tracks for most of this time the rulemaking for Part ASME Section XI, Division 2 is a 53 should be taken as an opportunity to establish RIPB practices within an performance-based standard which integrated framework. Using NUREG/BR-0303 and other supporting work establishes reliability targets for to ensure that there is consistent application for RI, PB and RIPB will goa components that could adversely long way toward making progress on NEIMA in a consistent and coherent affect plant safety and reliability.

manner. This is consistent with NUREG/BR-0303. No changes were made to the regulatory guide as a result of this comment.

6-2 General As currently expressed by DG-1383 and other guidance mentioned in the The NRC staff disagreed with this comments provided, the approach that the staff has chosen for inspection comment. The approach using and testing requirements falls short of delivering on the expectations of ASME Section XI, Division 2 has NEIMA. The staff should adopt a systems-based approach to establishing not been mandated. It is one means requirements in 10 CFR Part 53. From the perspective of such an approach that applicants and licensees can the stated functional objective of guidance related to In-Service Inspection choose to ensure that passive (ISI) and In-Service Testing (IST) would be seen as means to validate and components operate at an verify on a continuing basis the fitness for service and operational readiness acceptable level of performance, of some of the key design features and programmatic controls (Part 53 utilizing reliability targets to language) that provide the technical justification for the safety evaluation of measure performance. The RIM a design. This logic should extend to relevant phases where pre-service program does include preservice inspection and testing as well as post-construction inspection and testing are inspection and post construction considered. monitoring and non-destructive examination. As a result, no changes have been made to the regulatory guide.

6-3 General NUREG/BR-0303 incorporates a consideration that the ACRS The NRC staff disagreed with this recommended in September 2000 when reviewing work related to PB comment. The staff finds that guidance. This consideration relates to the concept that performance levels using Probabilistic Risk and reliability parameters should be set at the highest practical level. It is Assessment (PRA) to develop the important for the staff to bear in mind that the purpose of ISI/IST is to reliability targets of SSCs included validate and verify design provisions using a PB approach starting at the in the RIM Program is consistent functional level and flowing down to systems and components. It is only at with Commission direction on the the component level that the prescriptive aspects of the ASME code become use of PRA (60 FR 42622) and significant. appropriate for such purpose . As a 24

result, no changes were made to the regulatory guide.

6-4 The first attachment to the comment submission, provided a detailed The NRC staff disagreed with the discussion supporting the general comments above. The above general comment. The NRC staff found comments summarized the main points provided in the detail discussion. the details in the first attachment to be consistent with the comments 6-Additionally, the comment states: 1-6-3 above, which are responded to above.

[T]he guidance that is being developed in the area of ISI and IST (of which DG-1383 is a part) takes a compartmentalized approach that forces In addition, the NRC staff applicants and licensees to employ a needlessly prescriptive approach that disagrees that this RG forces will quite clearly limit flexibility that should be available to the developers applicants and licensees to do of a novel design concept. anything. Use of RIM is one means the staff finds acceptable In a systems-based framework design requirements for ISI and IST would for the development of an be PB to provide the maximum flexibility for the designer consistent with inspection program for nuclear the safety objectives conforming to regulatory objectives. At this level, the power plant components. It is not focus is on validation and verification of design requirements as opposed to mandated for use.

the details prescribed in Section C of DG-1383. Such details occur elsewhere as well, for example in the Interim Staff Guidance (ISG) in The RIM program is not ADAMS Accession Number ML21216A051 related to Risk-Informed prescriptive. The RIM program ISI/IST Programs in the Advanced Reactor Content of Application provides flexibility for licensees to Project. determine now to meet reliability targets, which are developed from The staff should recognize that the ASME code is prescriptive because the PRA. RIM is a program that is it needs to provide detailed rules, requirements, and criteria for the used in conjunction with the manufacture of specific components. design process to develop an inspection program that will ensure

[T]he staff has prepared DG-1380 to support high temperature component reliability. The details application of components covered by ASME Section III, Division 5. I provided in Section C of DG-1383 have provided comments on DG-1380 consistent with the approach are to provide a regulatory taken in this submittal as part of Public Comment for NRC-2021-0117, framework if a licensee or an which includes technical review of NUREG-2245 (ADAMS Accession applicant desire to implement a No. ML21286A738). I request that my Public Comment for NRC-2021- RIM program, identify what information should be provided to 0117 be incorporated by reference into this submittal.

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the NRC, and identify when approval is needed from the NRC.

The RIM process is also not prescriptive in establishing what inspections are to be conducted. It starts by using the PRA to establish reliability targets for the SSCs. The RIM process then develops strategies to monitor, examine, or assess SSC performance in meeting the reliability targets. There is significant flexibility to develop the strategies. If targets are not being met, the process includes feedback and adjustments.

Further, the staff notes that the ASME Code,Section XI, Division 2 is a paradigm shift from previously developed standards in that Division 2 is not prescriptive and provides much flexibility in the development and execution of the program.

The comments on DG-1380 refer to DG-1380 as being prescriptive rather than risk-informed and performance-based (RIPB) and suggests that the guidance should be developed using established RIPB approaches. The staff finds the comments on DG-1380 with respect to the suggested use of 26

RIPB approaches to be consistent with the comments provided for DG-1383 in comments 6-1 thru 6-4 and adequately responded to by the staff responses to the comments provided for DG-1383.

No changes were made to the regulatory guide as a result of this comment.

7-1 Section A, Thank you for preparing this draft regulatory guide. This guide will help The NRC staff agrees with the Applicability move the industry forward. Please consider the enhancements below. comment that RIM was developed for any type of reactor design, but The purpose of the regulatory guide is specific to non-LWRs (as defined on takes no position on the technical page 1, paragraph 1). While the regulatory directive may be specific to non- adequacy of RIM for LWRs. The LWRs,Section XI, Division 2 is also applicable to LWRs. A user could NRC staff disagrees that RG 1.246 misinterpret the current regulatory guide to preclude the use of Division 2 should be expanded to address for reactors using light water. The final regulatory guide should be LWRs. The purpose and scope of enhanced to include a brief reference to the process for an applicant to meet this RG is to provide guidance for the requirements of Division 2 for LWRs in lieu of the requirements of non-LWRs that are not subject to Division 1(such as through 10CFR50.55a). the requirements of 50.55a to implement ASME BPV,Section XI, Division 1. LWR applicants or licensees that wish to use RIM would need to request an appropriate exemption.

7-2 Section C, Section C.8 states that: The provisions of ASME Code,Section XI, The NRC staff agreed with this Regulatory Division 2, should not be used to depart from matters governed by the comment and revised the Guidance Position construction code of the plant. This item should be clarified by amending regulatory guide to clarify 8 without prior notification and NRC review. regulatory guidance position 8 by amending it with the phrase without prior NRC review and approval.

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7-3 Section C, Section C.11 states that: Mandatory Appendix V tables should be The NRC staff agreed with this Regulatory considered in the development of MANDE requirements and RIM comment and revised the Guidance Position strategies for components designed to very low or atmospheric pressures. regulatory guide to document how 11 This item should be clarified by amending and document how the tables the tables were considered in the were considered. RIM program.

8-1 Section B, Page 4 1. Draft DG-138[3] states (as does RG 1.232): The NRC staff disagreed with this

a. ARDC-14 states that the reactor coolant boundary comment. While the Advanced shall be tested so as to have an extremely low Reactor Design Criteria (ARDC) probability of abnormal leakage, of rapidly mentioned in the comment, may propagating failure, and of gross rupture. not be applicable for all advanced

b. ARDC-30 indicates that the components that are reactor designs, they may be part of the reactor coolant boundary shall be tested applicable to some designs and to the highest quality standards practical. therefore the staff finds it

c. ARDC-32 provides that the components that are appropriate to include reference to part of the reactor coolant boundary shall be these ARDC in the background designed to permit periodic inspection and section of the regulatory guide.

functional testing of important areas and features to assess their structural and leak tight integrity. No changes were made to the regulatory guide as a result of the These guidance statements may not be applicable to an comment.

AR design that does not rely on the reactor cooling system (and particularly the pressure boundary) for safe shutdown and prevention of large releases. While footnote 2 acknowledges the requirements are based on water-cooled plants, no basis is provided why these are applicable to what may be non-safety components. Consider removing the statements, particularly references to pressure boundary, and add clarification and basis for why these ARDC statements were chosen to be applicable.

9-1 General The Staff has done an excellent and thorough review of the 2019 Edition of The NRC staff appreciates the Section XI, Division 2 and have identified areas requiring regulatory comment. No changes were made guidance and my comments will be primarily addressing those items in as a result of the comment.

Section C of the DG-1383. As background, I have been involved in the development of RIM for more than 15 years and followed as well as provided input to its evolution. Several other RIM committee members 28

(Tom Roberts and Henry Stephens) have provided comments and recommendations. I strongly support their input and will not repeat them in my response.

9-2 Section C, The 10th bullet appears to be redundant with Regulatory Guidance 3. The NRC staff agrees that the Regulatory bullet is redundant. This bullet was Guidance Position Recommendation: Unless redundancy is the goal, one of these should meant to denote that this 1 probably be deleted and my suggestion is to delete the 10th bullet. information is to be included in the submittal to the NRC. However, in evaluating this recommendation the staff has determined that this level of information is not appropriate for submittal and can be obtained by the staff via audits of the RIM program plan.

Therefore, the staff removed this bullet from the regulatory guide.

9-3 Section C, There is confusion regarding the personnel requirements because ASME The NRC staff partially agreed Regulatory Section III requires meeting SNT-TC-1A,Section XI Division 1 has with this comment. The RG was Guidance Position upgraded from SNT-TC-1A and requires meeting CP-189 and Section XI revised to clarify that methods 5 Division 2 identified issues with both of these standards and requires approved for qualification and meeting ANDE-1-2015. Having three different personnel standards is going certification of NDE personnel are to lead to significant inconsistencies from owner to owner. The NRC needs not dependent on the reactor types to reduce the inconsistency by adopting one of these standards. and therefore would be acceptable for non-LWRs. In regard to Recommendation: The NRC needs to adopt ANDE-1-2015 and make this ANDE-1, the RG was revised to uniform for all Section XI, Division 2 nuclear applications. The question clarify that ANDE-1 may be used then becomes what evidence does the NRC need in order to endorse once a revision of Code Case N-ANDE-1-2015? 788 or other code cases, if any, are approved by NRC staff for Section XI Division 1.

Further, the RG was revised to specify that the conditions in 10 CFR 50.55a (b)(2) are applicable to personnel qualification when 29

using ASNT/CP-189. Additionally, the RG was revised to clarify the requirements for performance demonstrations. The staff considers using Section XI, Division 1 Appendix VIII for performance demonstration, when applying the conditions established within 10 CFR 50.55a(b)(2),

acceptable.

9-4 Section C, This position is a contradiction in the use of Section XI Division 2 since The NRC staff agreed with this Regulatory RIM is to be actively involved in nuclear power plants entire life span comment and revised the Guidance Position beginning with design and construction. Current construction practices are regulatory guide to clarify 8 based on Section III workmanship standards that are outdated and result in regulatory guidance position 8 by needless repairs which result in putting lower quality components into amending it with the phrase service because the repairs change many things including the residual weld without prior NRC review and stresses, weld chemistries, ISI inspection volumes, etc.Section XI has approval.

evolved and uses a fitness for service approach for addressing the significance and recommended management of flaws.

Recommendation: Section XI Division 2 must be an integral part of the construction code and this regulatory guidance item needs to be changed to reflect this. The rationale for changes to the construction code in the RIM Program must be developed and documented for review and approval by the regulators.

9-5 Section C, It is somewhat unclear as to what is needed regarding the use or nonuse of The NRC staff agreed with this Regulatory Mandatory Appendix V. comment and revised the Guidance Position regulatory guide to document how 11 Recommendation: Regulatory Guidance 11 needs to be revised to require the tables were considered in the that the rationale for the use or nonuse of Mandatory Appendix V be RIM program.

documented as part of the RIM Program.

9-6 General Since this new Section XI Division 2 is for non-LWRs where there is no The NRC staff agrees with this operating experience and there will be new materials, new designs, new comment. The NRC plans to fabrication methods and new operating conditions thus creating extensive provide oversight of advanced uncertainties. The RIM Program is going to use the latest tools such as reactor operation and utilize 30

PRAs as well as any laboratory studies, non-nuclear applications and expert information from operating judgement for identifying all safety related structures, systems, and experience and the OAR form to components (SSCs) that are to be included in the RIM Program and assess program performance.

hopefully to address these uncertainties. Based on experience for the When noted, the staff will work operating fleet of reactors, there have been surprises which occurred during with ASME to make changes to operation (see for example NUREG-0531 (primarily BWRs) and NUREG- the ASME Code,Section XI. As 0691 (PWRs)). It is reasonable to expect that with the new advanced non- appropriate information notices or LWRs without any operating experience, these uncertainties may lead to other generic communications will surprises. It is not clear in DG-1383 how the NRC will address these be used to provide information to uncertainties regarding safety significant SSCs in the RIM Program and the licensees. If necessary, the NRC majority of SSCs that are not included in the RIM Program because they retains the authority to order were not considered safety significant? licensees to make any changes necessary for safety. No changes were made to the regulatory guide as a result of this comment.

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