ML22061A253
| 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: |
| References | |
| RG 1.246 Rev 0 DG-1383 | |
| Download: ML22061A253 (31) | |
Text
October 2022 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 No.
ADAMS Accession No.
Commenter Affiliation Commenter Name 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 Letters No. 5 (ML21327A416) and No. 10 (ML21327A423) are duplicate copies of letters No. 4. and No. 9, respectively.
Letter/
Comment No.
Section of DG-1383 Specific Comments NRC Resolution 1-1 General A general comment is that traditional LWR considerations appears to permeate throughout and may have influenced the compilation of this draft Regulatory Guide. For, example, by assuming refueling outages and linking any reporting submission criteria to them is may be inappropriate for a RIM program developed for some advanced designs.
Some advanced reactor designs are anticipated to be fueled for life and the intended design life is 20 years. Further, while some of these designs do The NRC staff disagreed with this comment. The staff is aware of the differences in refueling outages between non-light water reactors and light water reactors (LWRs) and as such regulatory guidance position 4 addresses the alternative
2 anticipate having scheduled maintenance outages, they also anticipate conducting as much online maintenance and monitoring as possible.
Consequently, from the submittal of the OAR report, it is suggested that the Owner, based on the specifically developed RIM program, propose an appropriate submittal period for reports and documents such as the OAR to the USNRC.
It is unlikely there will be a one size fits all submission period that meaningfully would work for all advanced reactor designs.
for an applicant/licensee to propose an appropriate period for submitting the OAR report if there is not a refueling outage more frequently than every 5 years. In response to comment 1-6 below, the regulatory guidance position 4 was revised to delete reference to a refueling and used the term scheduled outage to be consistent with Appendix B of ASME Code,Section XI, Division 2.
1-2 Title (applicability to non-LWR)
It is understood that a Regulatory Guide cannot supersede existing regulations. It is also recognized that Title 10 CFR PART 50.55a establishes regulation for the use of ASME XI Division 1 as applied to light water reactors. However, a careful examination of ASME XI Division 1 reveals that the light water reactors that are appropriately addressed by ASME XI Division 1 consists of the existing fleet of PWR and BWR operating plant designs. Several advanced reactors design that are currently in development are light water cooled or moderated reactors, but for which the use of ASME XI Division 1 is neither adequate, nor appropriate. In some of these advanced LWR designs, even the traditional safety cases that are a foundational consideration for the application of ASME XI Division 1, may not be as relevant as the safety cases used in the existing fleet design (e.g., CDF, LERF, LOOP, etc.).
ASME XI Division 2 was developed to be technology neutral standard and can be equally applied to non-LWR as well as LWR technology. In fact, ASME XI Division 2 permits the use of Division 1 criteria if it is appropriate for a given reactor design and distinguishes which PRA standard should be used in the development of a RIM program when addressing a non-LWR versus a LWR advanced reactor design.
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 The NRC staff agreed that RIM was developed for any type of reactor design. However, the NRC staff reviewed and is endorsing ASME BPV Code,Section XI, Division 2 only for use by non-lightwater reactors because 10 CFR 50.55a(g) mandates the use of the ASME BPV Code,Section XI, Division 1 for boiling and pressurized water-cooled reactors.
If a boiling or pressurized water-cooled reactor licensee or applicant wishes to use RIM, they would need to request an exemption under 10 CFR 50.12 or 10 CFR 52.7 from 10 CFR 50.55a(g).
A footnote to this effect was added to the RG.
3 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 for Regulatory Guidance Position 1 (Page 6)
The term safety is used throughout this draft RG. Reference to 10CFR50 Appendix B is also a cited reference. It is not obvious that the term safety as it is used throughout this draft RG is meant to infer traditional Safety Related classification protocol (e.g., ASME Class 1, 2, 3 and Quality Groups A, B & C) or a broader use of the term safety (e.g., worker/public safety.) This is noted since ASME XI Division 2 applies no specific safety classifications to SSC within a RIM Program. Instead, RIM requires that any SSC that is risk significant for the safe operation of a plant and therefore worker and public safety is to be an SSC within the scope of the RIM developed program. This very well might include SSC that would otherwise not be traditionally classified as Safety Related.
Recommendation: A better description as to what is specifically intended by the use of the word SAFETY, so as to make it clear to users what SSC are specifically to be addressed should be provided for clarification within this Regulatory Guide.
The NRC staff agreed with this comment in part, relating to RIM application to risk significant SSCs and that not all risk significant SSCs would traditionally be classified as safety-related.
However, the NRC staff disagreed with the comment that the NRC staffs use of the term safety is confusing or that changes to the DG are needed. The term safety is used in several contexts in the DG. The majority of the time it is 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.
4 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 for Regulatory Guidance Position RIM establishes that an Inspection Interval shall not exceed 12 years (i.e.,
144 months) irrespective of the time between refueling (e.g., on-line refueling vs traditional off-line refueling.) This was created with the The NRC staff agreed with recommendation (1) not to reference refueling outage but did not agree that the end of the
5 1 (Page 7 - 3rd bullet) recognition that some advanced reactor designs may not be taken out of service (e.g., offline) to accommodate a refueling. From an ASME Code perspective it would be at the end of this 12-year interval that an OAR-1 Form would be prepared.
It is understandable why the USNRC may not wish to wait until the end of a 12 interval to receive information contained in a prepared OAR-1 Form for twelve years and hence has established that a five-year periodicity for the preparation of an OAR-1 be created and submitted.
There are however two clarifications that would be useful to include in DG-1383 regarding this matter.
The first is the descriptor of refueling outage. As noted, some advanced reactor designs may not have a traditional refueling outage, but instead a maintenance outage as the unit remains operational while refueling, or may be fueled for life.
The second is that with the 2019 Edition of ASME XI, both in Division 1 and Division 2, the FORM NIS-2 is merely an attestation for the completion of a Repair and Replacement Activity (RRA) signed by the Licensee and an Authorized Nuclear Inservice Inspector.
There is, however, no insightful or technical information contained on the FORM NIS-2 itself. This change 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.
Recommendations: (1) Provide a clarification to the presently used term 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.
inspection interval is the appropriate time to submit OAR forms. RIM-2.7.2 addresses the inspection interval but makes no mention of OAR form submission.
Non-Mandatory Appendix B recommends the OAR be submitted within 90 days of the completion of each scheduled outage. Regulatory guidance position 4 in this RG, addresses the frequency for OAR submittal.
Specifically, regulatory guidance position 4 addresses submittal of the OAR forms within 120 days after the completion of an outage, consistent with ASME Code,Section XI, Division 1. The term refueling was removed from the third bullet on page 7 and other locations in the regulatory guide.
The staff agreed with recommendation (2) that it is not necessary to submit the NIS-2 form, as the OAR form includes repair/replacement information.
Regulatory guide position 1 was revised to remove reference to the NIS-2.
6 1-5 Section B, Basis for Regulatory Guidance Position 1 (Page 7 - 4th bullet)
ASME III Division 1 is for traditional LWR reactor designs and the normal operating temperature ranges of LWRs. In contrast, ASME III Division 5 is for High temperature reactors and includes provisions for certain HAA materials to operate in the creep regime with a stated maximum number of permissible cycles. This same limitation is found in RIM Appendix VII Article VII-3.1. Consequently, it is not obvious why ASME Division 1 is cited for advanced reactor design flaw acceptance criteria.
Recommendation: Consider, revising this bullet so it is clearer that for advanced reactor designs that operate in the creep regime such as those designed in accordance and permitted by ASME Section III Division 5, that appropriate justification for flaw evaluation acceptance criteria shall be provided by the applicant.
The NRC staff agreed with this comment in part and disagreed in part. The NRC staff agreed that Section III, Division 1 may not be applicable to some designs that would implement RIM. However, the NRC staff has not reviewed the flaw evaluation acceptance criteria for use in temperature ranges above those described in Section III, Division 1, and cannot endorse their use at such temperatures.
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 for Regulatory Guidance Position 4 (Page 8, 5th paragraph)
Similar to the background and recommendation [for comment 1-4] noted above, the use of the term refueling outage may not specifically have a universally understood definition for some advanced designs, that may not have a discrete refueling outage or cycle. Additionally, this Regulatory Guidance Position states that: Licensees should submit the notification prior to the next refueling outage or within 3 years, whichever is less.. This 3-year periodicity appears to be contradictory to the Regulatory Guidance Position offered in Regulatory Guidance Position 1 (page 7 - third bullet) which reflects a five year or less periodicity.
Recommendation: Provide a clarification to the presently used term 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, and (2)
The staff agreed with this comment that the term refuling outage is not appropriate.
Reference to refueling was removed to make the regulatory guide consistent with Non-Mandatory Appendix B which specifies OAR forms to be submitted after scheduled outages.
However, the NRC staff disagreed that there is a contradiction between the timeframe for submitting an OAR and timeframe in position 1. The durations provided for submitting a
7 clarify what appears to be discrepancy in the desired periodicity reflected in 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 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 for Regulatory Guidance Position 5
This Regulatory Basis is understandable since the staff has not endorsed either Code Case N-788-1 nor ANDE-1 2015.
However, as written it is not clear whether the intent of the basis is to imply that the performance demonstration type of protocol of ASME XI Division 1 as is found in Division 1 Appendix VIII is meant to be employed.
Performance demonstration of any MANDE that may be selected for an SSC under a RIM program is an essential and imperative input and quantitatively factors into the establishment a Reliability Target assigned to an SSC for such considerations a Probability of Detection (POD) criteria.
It is believed that while the USNRC has not formally endorsed the use of either Code Case N-788-1 or ANDE-1 2015, the reservation by the staff is understood to be based on the fact that ANDE-1 describes a process for the qualification of NDE personnel. If that is in fact the existing reservation then the following recommendation is provided.
Recommendation: Provide guidance regarding the type of information that would be expected to be provided by a Licensee applicant that would allow for consideration of approval by the USNRC to use ANDE-1 for NDE personnel qualification under Division 2. This would assure that factors The NRC staff agreed with this comment in part and disagreed in part. The RG was revised to clarify that methods approved for qualification and certification of NDE personnel are not dependent on the reactor types and therefore non-LWRs should use any methods approved by the NRC for use by LWRs, e.g., in 10 CFR 50.55a. 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 staff 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
8 such as POD for any MANDE methods selected are established with consistency. The use of other already endorsed NDE personnal qualification criteria, such as CP-189 do not afford this essential criteria (i.e., POD) and potentially undermines a cornerstone to the development of a sound RIM Program.
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 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, Regulatory Guidance Position 1 (Page 14, 2nd and 8th bullets)
It is understood that the USNRC would seek to have summaries of:
(a) The bases for the scope of the program and (b) RIM strategies selected to achieve the reliability targets, as denoted in this portion of the Staff Regulatory Guidance at the time a Licensee/applicant makes an initial filing. However, both of these provisions may not be fully developed at the time of initial application and 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 program.
Recommendation: It is suggested that a clarification be provided to both of these items so as to better define what contents are expected to be provided by a Licensee/applicant at the time of license application.
The NRC staff agreed with this comment and revised the RG to identify a list of structures, systems, and components (SSCs) included in the scope of the RIM program rather than a summary of the bases for the scope of the RIM program.
In regard to the RIM strategies to be used, only a description of the types of factors as outlined in 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,
9 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, Regulatory Guidance Position 1 (Page 14, 12th bullet)
As previously outlined in comment [1-4] above, the FORM NIS-2 is merely an attestation for the completion of a Repair and Replacement Activity (RRA) signed by the Licensee and an Authorized Nuclear Inservice 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 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.
The NRC staff agreed with this comment and deleted reference to submitting Form NIS-2.
10 1-10 Section C, Regulatory Guidance Position 5
As cited in comment [1-7] above, performance demonstration of any MANDE that may be selected for an SSC under a RIM program is an essential and imperative input and quantitatively factors into the establishment of a Reliability Target assigned to an SSC by using considerations such as Probability of Detection (POD) criteria. The use of ANDE-1 requires such performance demonstrations that are not a mandate of ANSI/ASNT CP-189.
Recommendation: Consider providing guidance regarding the type of information that would be expected to be provided by a Licensee applicant that would allow for consideration of approval by the USNRC to use ANDE-1 for NDE personnel qualification under Division 2 as an alternative to advocating the use of ANSI/ASNT CP-189.
The NRC staff agreed with the comment in part that ANDE-1 and ANSI/ASNT CP-189 are personnel qualification and certification programs. Performance demonstration is a separate activity, governed by ASME Section XI, Division 1, Appendix VIII. While ANDE-1 and ANSI/ANS CP-189 include NDE demonstration activities as part of the qualification and certification 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
11 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, Applicability This comment requests that RG 1.246 include light water reactors (LWRs) within its scope to facilitate implementation of ASME Section XI Division 2 by LWR applicants and licensees via the code alternative process. The comment states:
The applicability of DG-1383 to non-LWRs is in contrast with the intended scope of ASME Section XI Division 2, which specifically provides applicability for LWRs. ASME Section XI Division 2, Section RIM-1.1(a) states that This Division provides the requirements for the creation of the Reliability and Integrity Management (RIM) Program for all types of nuclear power plants (emphasis added).
The NRC staff agrees with the comment that RIM was developed for any type of reactor design, but takes no position on the technical adequacy of RIM for LWRs. The NRC staff disagrees that RG 1.246 should be expanded to address LWRs. The purpose and scope of this RG is to provide guidance for non-LWRs that are not subject to the requirements of 50.55a to implement ASME BPV,Section XI, Division 1. The NRC staff
12 Notwithstanding, NuScale recognizes that current NRC regulations at 10 CFR 50.55a require the application of ASME Section XI, Division 1 for non-LWRs, and thus RG 1.246 cannot directly endorse ASME Section XI Division 2 for use by LWR applicants/licensees. NuScale seeks a future amendment of 10 CFR 50.55a to incorporate by reference ASME Section XI Division 2 for LWRs.
Until that rulemaking can occur, NRC should explicitly recognize in RG 1.246 that, from a technical perspective, ASME Section XI Division 2 is applicable and appropriate for implementation by LWR 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.
agrees that LWR applicants or licensees that wish to use RIM would need to request an appropriate exemption.
The NRC staff also takes no position on rulemaking to incorporate Division 2 into 50.55a.
Rulemaking is outside the scope of this RG.
13 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 Section A, Applicability ASME XI Division 2 is applicable as a technology neutral standard and has been developed to be applied to both non-LWR as well as LWR technology. ASME XI, Division permits the use of Division 1 criteria if it is appropriate for a given reactor design and distinguishes which PRA standard should be used in the development of a RIM program when addressing a non-LWR versus a LWR advanced reactor design.
It is understood that Title 10 CFR PART 50.55a establishes regulation for the use of ASME XI Division 1 as applied to light-water reactors. Further, ASME XI Division 1 addresses the existing fleet of PWR and BWR operating plant designs. Several advanced light-water reactors are currently in developments. The use of ASME XI Division 1 is neither adequate, nor appropriate for these newer designs. In some of these advanced LWR designs, the traditional safety cases that are a foundational consideration for the application of ASME XI Division 1, may not be as relevant as the safety cases used in the existing fleet design.
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 The NRC staff agrees with the comment that RIM was developed for any type of reactor design, but takes no position on the technical adequacy of RIM for LWRs. The NRC staff disagrees that RG 1.246 should be expanded to address LWRs. The purpose and scope of this RG is to provide guidance for non-LWRs that are not subject to 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.
14 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 for Regulatory Guidance Position 1
The term safety is used throughout this draft RG. Reference to 10CFR50 Appendix B is also a cited. It is not clear if the term safety is meant to address traditional Safety-Related classifications (e.g., ASME Class 1, 2, 3 and Quality Groups A, B & C) or a broader use (e.g., worker/public safety). ASME XI Division 2 applies no specific safety classifications to SSC within a RIM Program. Instead, RIM requires that any SSC that is risk significant for the safe operation of a plant and therefore worker and public safety is to be an SSC within the scope of the RIM developed program. This might include SSC that would otherwise not be traditionally classified as Safety-Related.
Recommendation: A description as to what is specifically intended by the use of the word safety would make it clearer as to what SSCs are specifically to be addressed.
The NRC staff agreed with this comment in part, relating to RIM application to risk significant SSCs and that not all risk significant SSCs would traditionally be classified as safety-related.
However, the NRC staff disagreed with the comment that the NRC staffs use of the term safety is confusing or that changes to the DG are needed. The term safety is used in several contexts in the DG. The majority of the time it is 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.
15 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 for Regulatory Guidance Position 1; Page 7, 3rd bullet RIM establishes that an Inspection Interval shall not exceed 12 years (i.e.,
144 months) irrespective of the time between refueling (e.g., on-line refueling vs traditional off-line refueling.) This was created with the recognition that some advanced reactor designs may not be taken out of service (e.g., offline) to accommodate a refueling outage. From an ASME The NRC staff agreed with recommendation (1) not to reference refueling outage but did not agree that the end of the inspection interval is the appropriate time to submit OAR
16 Code perspective it would be at the end of this 12-year interval that an OAR-1 Form would be prepared.
It is understandable why the USNRC may not wish to wait until the end of a 12 interval to receive information contained in a prepared OAR-1 Form for twelve years and hence has established that a five-year periodicity for the preparation of an OAR-1 be created and submitted.
There are however two clarifications that would be useful to include in DG-1383 regarding this matter:
One is the descriptor of refueling outage. As noted, some advanced reactor designs may not have a traditional refueling outage, but instead a maintenance outage where the unit remains operational while refueling.
Two is that with the 2019 Edition of ASME XI, both in Division 1 and Division 2, the FORM NIS-2 is an attestation for the completion of a Repair/Replacement Activity (RRA).
The FORM NIS-2 does not include any significant or technical information.
This change 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 and the FORM NIS-2 became abbreviated.
Recommendations:
- 1. Provide a clarification to the presently used term 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.
forms. RIM-2.7.2 addresses the inspection interval but makes no mention of OAR form submission.
Non-Mandatory Appendix B recommends the OAR be submitted within 90 days of the completion of each scheduled outage. Regulatory guidance position 4 in this RG, addresses the frequency for OAR submittal.
Specifically, regulatory guidance position 4 addresses submittal of the OAR forms within 120 days after the completion of an outage, consistent with ASME Code,Section XI, Division 1. The term refueling was removed from the third bullet on page 7 and other locations in the regulatory guide.
The staff agreed with recommendation (2) that it is not necessary to submit the NIS-2 form, as the OAR form includes repair/replacement information.
Regulatory guide position 1 was revised to remove reference to the NIS-2.
3-4 Section B, Basis for Regulatory Guidance Position 1; Page 7, 4th bullet ASME III Division 1 is for traditional LWR reactor designs and the normal operating temperature ranges of LWRs. In contrast, ASME III Division 5 is for High temperature reactors and includes provisions for certain HAA materials to operate in the creep regime with a stated maximum number of The NRC staff agreed with this comment in part and disagreed in part. The NRC staff agreed that Section III, Division 1 may not be
17 permissible cycles. This same limitation is found in RIM Appendix VII Article VII-3.1. Consequently, it is not obvious why ASME Division 1 is cited for advanced reactor design flaw acceptance criteria.
Recommendation: Consider, revising this bullet so it is clearer that for advanced reactor designs that operate in the creep regime such as those designed in accordance and permitted by ASME Section III Division 5, that appropriate justification for flaw evaluation acceptance criteria shall be provided by the applicant.
applicable to some designs that would implement RIM. However, the NRC staff has not yet endorsed Division 5, and as such it would be inappropriate to reference it in this RG. In any case, appropriate temperature limits should be determined by the applicants 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 for Regulatory Guidance Position 4; Page, 5th paragraph Similar to the background and recommendation [3-3] noted above, the use of the term refueling outage may not specifically have a universally understood definition for some advanced designs, that may not have a discrete refueling outage or cycle. Additionally, this Regulatory Guidance Position states that: Licensees should submit the notification prior to the next refueling outage or within 3 years, whichever is less. This 3-year periodicity appears to be contradictory to the Regulatory Guidance Position offered in Regulatory Guidance Position 1 (page 7 - third bullet) which reflects a five year or less periodicity.
Recommendation: Provide a clarification to the presently used term 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, and (2) clarify what appears to be discrepancy in the desired periodicity reflected in the two noted Regulatory Guidance Positions.
The staff agreed with this comment that the term refuling outage is not appropriate.
Reference to refueling was removed to make the regulatory guide consistent with Non-Mandatory Appendix B which specifies OAR forms to be submitted after scheduled outages.
However, the NRC staff disagreed that there is a contradiction between the timeframe for submitting an OAR and timeframe in position 1. The durations provided for submitting a 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
18 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 for Regulatory Guidance Position 5
The comment refers to conditions from 10 CFR 50.55a and Section XI, Division 1, IWA-2300, and suggests that such conditions should be addressed in the RG. The comment also provides background on activities that led to the development of the ANDE-1 Standard. Further, the comment describes the formation of a Certifying Body for the implementation of the ANDE-1 standard. The comment provided the following recommendations.
Recommendations:
ANSI/ASNT CP-189:
1a. Specify the conditions on Section XI, Division 1, addressed in 10 CFR 50.55a, (b)(2)(xv) and (b)(2).
1b. Reinstate the IE Bulletin 82-03 requirement that training for IGSCC detection, sizing and overlay weld repair examinations.
1c. Reinstate the requirement for 3-year recertification for PDI IGSCC detection, sizing and overly weld repair examinations. 1d. Require 3-year recertification for all Mandatory Appendix VIII PDI qualification examinations.
- 2. Condition the listed Section XI, Division 1, IWA-2300 amended requirements:
2a. IWA-2314 Certification and Recertification except that the ASNT Level III certificate is not required. It is recommended that the third-party ASNT Level III certificate, ACCP certificate or another recognized third-party qualification with a NQA-1 or Appendix B QA program, e.g., EPRI NDE Center be required. Additionally, contrary to the ASME Interpretation XI-1-10-34 an audit of the ASNT Level III or ACCP certificate program would be subject to audit by the licensee.
2b. IWA-2380 NDE INSTRUCTOR In lieu of the requirements of CP-189, It is recommended the CP-189, third-party requirements are maintained as compared to those amended by IWA-2380.
The NRC staff agreed with this comment in part and disagreed in part. The RG was revised to clarify that methods approved for qualification and certification of NDE personnel are not dependent on the reactor types and therefore non-LWRs should use any methods approved by the NRC for use by LWRs, e.g., in 10 CFR 50.55a. 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 staff 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 non-LWR licensees should apply the conditions in 10 CFR 50.55a(b)(2) when using
19 2c. Additionally, the DG does not address the conditions related to Division 1, Mandatory Appendices VI, VII or VIII.
ANDE-1:
Provide guidance regarding the type of information that would be expected to be provided by a Licensee applicant that would allow for consideration of approval by the USNRC to use ANDE-1 for NDE personnel qualification under Division 2. This would assure that factors such as POD for any MANDE methods selected are established with consistency. The use of other already endorsed NDE personnal qualification criteria, such as CP-189 do not afford this essential criteria (i.e., POD) and potentially undermines a cornerstone to the development of a sound RIM Program.
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.
3-7 Section C, Staff Regulatory Guidance 1, 2nd and 8th bullets It is understood that the USNRC would seek to have summaries of:
(a) The bases for the scope of the program and (b) RIM strategies selected to achieve the reliability targets, as denoted in this portion of the Staff Regulatory Guidance at the time a Licensee/applicant makes an initial filing. However, both of these provisions may not be fully developed at the time of initial application and 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 program.
Recommendation: It is suggested that a clarification be provided to both of these items so as to better define what contents are expected to be provided by a Licensee/applicant at the time of license application.
The NRC staff agreed with this comment and revised the RG to identify a list of SSCs included in the scope of the RIM program rather than a summary of the bases for the scope of the RIM program.
In regard to the RIM strategies to be used, only a description of the types of factors as outlined in RIM-2.5.1 that are used in the RIM strategies needs to be provided.
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
20 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 Regulatory Guidance 1, 12th bullet As previously outlined in comment [3-3] above, the FORM NIS-2 is merely an attestation for the completion of a Repair and Replacement Activity (RRA) signed by the Licensee and an Authorized Nuclear Inservice 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.
The NRC staff agreed with this comment and deleted reference to submitting Form NIS-2.
3-9 Section C, Staff Regulatory Guidance 5 As cited in comment [3-6] above, performance demonstration of any MANDE that may be selected for an SSC under a RIM program is an essential and imperative input and quantitatively factors into the establishment a Reliability Target assigned to an SSC by using considerations such as Probability of Detection (POD) criteria. The use of ANDE-1requires such performance demonstrations that are not a mandate of ANSI/ASNT CP-189.
Recommendation: Consider providing guidance regarding the type of information that would be expected to be provided by a Licensee applicant that would allow for consideration of approval by the USNRC to use ANDE-1 for NDE personnel qualification under Division 2 as an alternative to advocating the use of ANSI/ASNT CP-189.
The NRC staff agreed with the comment in part that ANDE-1 and ANSI/ASNT CP-189 are personnel qualification and certification programs. Performance demonstration is a separate activity, governed by ASME Section XI, Division 1, Appendix VIII. While ANDE-1 and ANSI/ANS CP-189 include NDE demonstration activities as part of the qualification and certification process, performance
21 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
22 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 for Regulatory Guidance Position 5
As background, the comment provided a summary of industry experience, field failures and conclusions of round robin studies, based on which the comment stated that it was extradentary to find NRC position 5 supporting the continuance of CP-189. The comment also stated the following points as a summary:
For over 50 years the employer based self-certification process has been unreliable to consistently qualify NDE personnel that meet industry expectations (detect flaws before failure).
All round robin studies have shown detection rates at best are 50%
No known study has ever shown the process to be effective The unstructured training and experience do not include the key factors essential for learning INPO and Academia agree the process does not have a path for consistent reliability and success Further, the comment requested the information and provided the recommendations below.
Request for Additional Information:
In response to position 5 [above], the NRC has been involved with the ANDE project since the beginning and has invested both manpower and taxpayer dollars. ANDE-1 has successfully completed both revision 1 and 2 through the ANSI Standards review process with no unresolved NRC comments. It is essential for the NRC to specifically identify what is needed to address the following in position 5: Code Case N-788-1 and the ANDE The NRC staff disagreed with this comment because approval of ANDE-1 is outside the scope of the RG. However, in response to other ANDE-1 and CP-189 related comments, 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.
Thus, 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, if the NRC approves the use of ANDE-1 for LWRs, that approval would extend to non-LWRs. This means that non-LWRs applicants and licensees can use ANDE-1 without need for further NRC approvals, once that work has been completed and approved by the NRC in other processes.
23 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) incorporates the term risk-informedand performance-based (RIPB) in several places. NEIMA does not offer a definition for RIPB. The staff should use the Commission developed definitions in SRM-SECY-98-0144 and use the work that has been done over the past 20 years employing these This comment appears to be outside the scope of this RG.. The intent of this review was not to develop a new program but to review an existing consensus
24 definitions. Although risk-informed (RI) and performance-based (PB) have progressed on separate tracks for most of this time the rulemaking for Part 53 should be taken as an opportunity to establish RIPB practices within an integrated framework. Using NUREG/BR-0303 and other supporting work to ensure that there is consistent application for RI, PB and RIPB will goa long way toward making progress on NEIMA in a consistent and coherent manner.
standard for use. This standard, ASME Section XI, Division 2 is a performance-based standard which establishes reliability targets for components that could adversely affect plant safety and reliability.
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 comments provided, the approach that the staff has chosen for inspection and testing requirements falls short of delivering on the expectations of NEIMA. The staff should adopt a systems-based approach to establishing requirements in 10 CFR Part 53. From the perspective of such an approach the stated functional objective of guidance related to In-Service Inspection (ISI) and In-Service Testing (IST) would be seen as means to validate and verify on a continuing basis the fitness for service and operational readiness of some of the key design features and programmatic controls (Part 53 language) that provide the technical justification for the safety evaluation of a design. This logic should extend to relevant phases where pre-service inspection and testing as well as post-construction inspection and testing are considered.
The NRC staff disagreed with this comment. The approach using ASME Section XI, Division 2 has not been mandated. It is one means that applicants and licensees can choose to ensure that passive components operate at an acceptable level of performance, utilizing reliability targets to measure performance. The RIM program does include preservice inspection and post construction 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 recommended in September 2000 when reviewing work related to PB guidance. This consideration relates to the concept that performance levels and reliability parameters should be set at the highest practical level. It is important for the staff to bear in mind that the purpose of ISI/IST is to validate and verify design provisions using a PB approach starting at the functional level and flowing down to systems and components. It is only at the component level that the prescriptive aspects of the ASME code become significant.
The NRC staff disagreed with this comment. The staff finds that using Probabilistic Risk Assessment (PRA) to develop the reliability targets of SSCs included in the RIM Program is consistent with Commission direction on the use of PRA (60 FR 42622) and appropriate for such purpose. As a
25 result, no changes were made to the regulatory guide.
6-4 The first attachment to the comment submission, provided a detailed discussion supporting the general comments above. The above general comments summarized the main points provided in the detail discussion.
Additionally, the comment states:
[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 applicants and licensees to employ a needlessly prescriptive approach that will quite clearly limit flexibility that should be available to the developers of a novel design concept.
In a systems-based framework design requirements for ISI and IST would be PB to provide the maximum flexibility for the designer consistent with the safety objectives conforming to regulatory objectives. At this level, the focus is on validation and verification of design requirements as opposed to the details prescribed in Section C of DG-1383. Such details occur elsewhere as well, for example in the Interim Staff Guidance (ISG) in ADAMS Accession Number ML21216A051 related to Risk-Informed ISI/IST Programs in the Advanced Reactor Content of Application Project.
The staff should recognize that the ASME code is prescriptive because it needs to provide detailed rules, requirements, and criteria for the manufacture of specific components.
[T]he staff has prepared DG-1380 to support high temperature application of components covered by ASME Section III, Division 5. I have provided comments on DG-1380 consistent with the approach taken in this submittal as part of Public Comment for NRC-2021-0117, which includes technical review of NUREG-2245 (ADAMS Accession No. ML21286A738). I request that my Public Comment for NRC-2021-0117 be incorporated by reference into this submittal.
The NRC staff disagreed with the comment. The NRC staff found the details in the first attachment to be consistent with the comments 6-1-6-3 above, which are responded to above.
In addition, the NRC staff disagrees that this RG forces applicants and licensees to do anything. Use of RIM is one means the staff finds acceptable for the development of an inspection program for nuclear power plant components. It is not mandated for use.
The RIM program is not prescriptive. The RIM program provides flexibility for licensees to determine now to meet reliability targets, which are developed from the PRA. RIM is a program that is used in conjunction with the design process to develop an inspection program that will ensure component reliability. The details provided in Section C of DG-1383 are to provide a regulatory framework if a licensee or an applicant desire to implement a RIM program, identify what information should be provided to
26 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
27 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, Applicability Thank you for preparing this draft regulatory guide. This guide will help move the industry forward. Please consider the enhancements below.
The purpose of the regulatory guide is specific to non-LWRs (as defined on page 1, paragraph 1). While the regulatory directive may be specific to non-LWRs,Section XI, Division 2 is also applicable to LWRs. A user could misinterpret the current regulatory guide to preclude the use of Division 2 for reactors using light water. The final regulatory guide should be enhanced to include a brief reference to the process for an applicant to meet the requirements of Division 2 for LWRs in lieu of the requirements of Division 1(such as through 10CFR50.55a).
The NRC staff agrees with the comment that RIM was developed for any type of reactor design, but takes no position on the technical adequacy of RIM for LWRs. The NRC staff disagrees that RG 1.246 should be expanded to address LWRs. The purpose and scope of this RG is to provide guidance for non-LWRs that are not subject to 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, Regulatory Guidance Position 8
Section C.8 states that: The provisions of ASME Code,Section XI, Division 2, should not be used to depart from matters governed by the construction code of the plant. This item should be clarified by amending without prior notification and NRC review.
The NRC staff agreed with this comment and revised the regulatory guide to clarify regulatory guidance position 8 by amending it with the phrase without prior NRC review and approval.
28 7-3 Section C, Regulatory Guidance Position 11 Section C.11 states that: Mandatory Appendix V tables should be considered in the development of MANDE requirements and RIM strategies for components designed to very low or atmospheric pressures.
This item should be clarified by amending and document how the tables were considered.
The NRC staff agreed with this comment and revised the regulatory guide to document how the tables were considered in the RIM program.
8-1 Section B, Page 4
- 1. Draft DG-138[3] states (as does RG 1.232):
- a. ARDC-14 states that the reactor coolant boundary shall be tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.
- b. ARDC-30 indicates that the components that are part of the reactor coolant boundary shall be tested to the highest quality standards practical.
- c. ARDC-32 provides that the components that are part of the reactor coolant boundary shall be designed to permit periodic inspection and functional testing of important areas and features to assess their structural and leak tight integrity.
These guidance statements may not be applicable to an 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.
The NRC staff disagreed with this comment. While the Advanced Reactor Design Criteria (ARDC) mentioned in the comment, may not be applicable for all advanced reactor designs, they may be applicable to some designs and therefore the staff finds it appropriate to include reference to these ARDC in the background section of the regulatory guide.
No changes were made to the regulatory guide as a result of the comment.
9-1 General The Staff has done an excellent and thorough review of the 2019 Edition of Section XI, Division 2 and have identified areas requiring regulatory guidance and my comments will be primarily addressing those items in 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 The NRC staff appreciates the comment. No changes were made as a result of the comment.
29 (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, Regulatory Guidance Position 1
The 10th bullet appears to be redundant with Regulatory Guidance 3.
Recommendation: Unless redundancy is the goal, one of these should probably be deleted and my suggestion is to delete the 10th bullet.
The NRC staff agrees that the bullet is redundant. This bullet was meant to denote that this 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, Regulatory Guidance Position 5
There is confusion regarding the personnel requirements because ASME Section III requires meeting SNT-TC-1A,Section XI Division 1 has upgraded from SNT-TC-1A and requires meeting CP-189 and Section XI Division 2 identified issues with both of these standards and requires meeting ANDE-1-2015. Having three different personnel standards is going to lead to significant inconsistencies from owner to owner. The NRC needs to reduce the inconsistency by adopting one of these standards.
Recommendation: The NRC needs to adopt ANDE-1-2015 and make this uniform for all Section XI, Division 2 nuclear applications. The question then becomes what evidence does the NRC need in order to endorse ANDE-1-2015?
The NRC staff partially agreed with this comment. The RG was revised to clarify that methods approved for qualification and certification of NDE personnel are not dependent on the reactor types and therefore would be acceptable for non-LWRs. In regard to ANDE-1, the RG was revised to clarify that ANDE-1 may be used once a revision of Code Case N-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
30 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, Regulatory Guidance Position 8
This position is a contradiction in the use of Section XI Division 2 since RIM is to be actively involved in nuclear power plants entire life span beginning with design and construction. Current construction practices are based on Section III workmanship standards that are outdated and result in needless repairs which result in putting lower quality components into service because the repairs change many things including the residual weld stresses, weld chemistries, ISI inspection volumes, etc.Section XI has 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.
The NRC staff agreed with this comment and revised the regulatory guide to clarify regulatory guidance position 8 by amending it with the phrase without prior NRC review and approval.
9-5 Section C, Regulatory Guidance Position 11 It is somewhat unclear as to what is needed regarding the use or nonuse of Mandatory Appendix V.
Recommendation: Regulatory Guidance 11 needs to be revised to require that the rationale for the use or nonuse of Mandatory Appendix V be documented as part of the RIM Program.
The NRC staff agreed with this comment and revised the regulatory guide to document how the tables were considered in the RIM program.
9-6 General Since this new Section XI Division 2 is for non-LWRs where there is no operating experience and there will be new materials, new designs, new fabrication methods and new operating conditions thus creating extensive uncertainties. The RIM Program is going to use the latest tools such as The NRC staff agrees with this comment. The NRC plans to provide oversight of advanced reactor operation and utilize
31 PRAs as well as any laboratory studies, non-nuclear applications and expert judgement for identifying all safety related structures, systems, and components (SSCs) that are to be included in the RIM Program and hopefully to address these uncertainties. Based on experience for the operating fleet of reactors, there have been surprises which occurred during operation (see for example NUREG-0531 (primarily BWRs) and NUREG-0691 (PWRs)). It is reasonable to expect that with the new advanced non-LWRs without any operating experience, these uncertainties may lead to surprises. It is not clear in DG-1383 how the NRC will address these uncertainties regarding safety significant SSCs in the RIM Program and the majority of SSCs that are not included in the RIM Program because they were not considered safety significant?
information from operating experience and the OAR form to assess program performance.
When noted, the staff will work with ASME to make changes to the ASME Code,Section XI. As appropriate information notices or other generic communications will be used to provide information to licensees. If necessary, the NRC retains the authority to order licensees to make any changes necessary for safety. No changes were made to the regulatory guide as a result of this comment.