0CAN052004, Request for Exemption from 10 CFR 50, Appendix B Requirements to Support Application of ASME Code Case N-752, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 Systems
| ML20148M344 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 05/27/2020 |
| From: | Gaston R Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation |
| References | |
| 0CAN052004 | |
| Download: ML20148M344 (12) | |
Text
10 CFR 50.12 0CAN052004 May 27, 2020 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Request for Exemption from 10 CFR 50, Appendix B Requirements to Support Application of ASME Code Case N-752, "Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 Systems,Section XI, Division 1" Arkansas Nuclear One, Units 1 and 2 NRC Docket Nos. 50-313, 50-368, and 72-13 Renewed Facility Operating License Nos. DPR-51 and NPF-6
Reference:
Entergy Operations, Inc. (Entergy) letter to the U. S. Nuclear Regulatory Commission (NRC), Relief Request Number EN-20-RR-001 - Proposed Alternative to Use ASME Code Case N-752, Risk-Informed Categorization and Treatment for Repair/ Replacement Activities in Class 2 and 3 Systems,Section XI, Division 1, (0CAN052003), dated May 27, 2020.
In the reference submittal, Entergy Operations, Inc. (Entergy) submitted a request for the use of an alternative to the American Society of Mechanical Engineers (ASME),Section XI requirements at Arkansas Nuclear One, Units 1 and 2 (ANO-1 and ANO-2, respectively).
Specifically, the request is to use the alternate requirements of ASME Code Case N-752. In conjunction with that request, Entergy is submitting this request for an exemption to the requirements of 10 CFR 50, Appendix B, for the ASME Section XI, Repair/Replacement Programs that are in the scope of the referenced request in accordance with 10 CFR 50.12, "Specific exemptions." Special circumstance 10 CFR 50.12(a)(2)(vi) is applicable to this request.
Appendix B to 10 CFR 50 provides Quality Assurance (QA) requirements for the design, construction, and operation of structures, systems and components (SSCs) that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public. These requirements apply to all activities that affect SSC safety-related functions. The general requirements contained in 10 CFR 50, Appendix B, are supplemented Entergy Operations, Inc.
1340 Echelon Parkway Jackson, MS 39213 Tel 601-368-5138 Ron Gaston Director, Nuclear Licensing
0CAN052004 Page 2 of 3 by industry standards and NRC regulatory guides that describe specific practices that have been found acceptable by the industry and the staff.
While both 10 CFR 50, Appendix B, and the industry standards include provisions for the flexible application of the QA practices commensurate with the importance to safety of the SSCs for which these practices are applied, other regulations require that the explicit criteria of 10 CFR 50, Appendix B, still be applied. Entergy considers these requirements to be unduly burdensome for components of Low Safety Significance (LSS) and that alternative practices, supplemented as appropriate, are adequate to provide reasonable confidence that the LSS components will perform their safety related functions under design basis condition.
10 CFR 50.55a, "Codes and Standards," lists the standards that have been approved for incorporation by reference. One of the standards listed is ASME,Section XI (i.e., as specified in 10 CFR 50.55a(a)(1)(ii)). ASME Code Case N-752 includes provisions for exempting Class 2 or Class 3 items categorized as LSS from the QA requirements of the 2007 Edition/ 2008 Addenda of ASME Section XI, paragraph IWA-1400(n), which invokes the requirement for a QA Program in accordance with either 10 CFR 50, Appendix B, or ASME NQA-1. However, approval to implement this Code Case cannot waive the 10 CFR 50, Appendix B, licensing basis requirement. The purpose for this request is to obtain approval for exempting ANO-1 and ANO-2, from the requirements of 10 CFR 50, Appendix B, for those safety-related Class 2 or Class 3 items categorized as LSS using the risk-informed methodology found in Code Case N-752. The details of the exemption request are provided in the enclosure.
Implementation of Code Case N-752, including this requested exemption, will allow ANO-1 and ANO-2 to improve focus on equipment that has greater safety significance, resulting in improved plant safety. This also enhances the effectiveness and efficiency of the NRCs oversight of the activities at ANO by focusing its resources on those SSCs that are most significant to maintaining public health and safety. Further, the reactor oversight process relies on the application of risk insights to determining significance of issues or events identified at licensee facilities.
Entergy requests authorization by June 27, 2021, to support planning activities associated with the ANO-2 refueling outage in the Fall 2021 (2R28).
No new regulatory commitments are included in this submittal.
If there are any questions or if additional information is needed, please contact Riley Keele, Manager, Regulatory Assurance, Arkansas Nuclear One, at 479-858-7826.
Respectfully, ORIGINAL SIGNED BY RON GASTON Ron Gaston RWG/rwc
0CAN052004 Page 3 of 3
Enclosure:
Request for Exemption from 10 CFR 50, Appendix B, to Support the Implementation of Code Case N-752 cc:
NRC Region IV Regional Administrator NRC Senior Resident Inspector - Arkansas Nuclear One NRC Project Manager - Arkansas Nuclear One
ENCLOSURE 0CAN052004 REQUEST FOR EXEMPTION FROM 10 CFR 50, APPENDIX B TO SUPPORT THE IMPLEMENTATION OF CODE CASE N-752
0CAN052004 Enclosure Page 1 of 8 REQUEST FOR EXEMPTION FROM 10 CFR 50, APPENDIX B TO SUPPORT THE IMPLEMENTATION OF CODE CASE N-752 1
REQUEST FOR EXEMPTION Pursuant to 10 CFR 50.12, "Specific exemptions," Entergy Operations, Inc. (Entergy), requests an exemption from the requirements of 10 CFR 50, Appendix B for the Arkansas Nuclear One, Units 1 and 2 (ANO-1 and ANO-2, respectively). The proposed exemption would allow the quality assurance (QA) requirements listed in 10 CFR 50, Appendix B, for items categorized as Low Safety Significant (LSS) during implementation of the American Society of Mechanical Engineers (ASME) Code Case N-752, "Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 Systems,Section XI, Division 1" (Reference 1),
to be exempted in accordance with 10 CFR 50.12, "Specific exemptions." Special circumstance 10 CFR 50.12(a)(2)(vi) is applicable to this request. This specific circumstance is consistent with References 2 and 3. With respect to Code Case N-752, LSS is analogous to risk-informed safety class (RISC)-3 in 10 CFR 50.69.
Appendix B to 10 CFR 50 provides QA requirements for the design, construction, and operation of structures, systems, and components (SSCs) that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public. These requirements apply to all activities that affect SSC safety-related functions. The general requirements contained in 10 CFR 50, Appendix B, are supplemented by industry standards and NRC regulatory guides that describe specific practices that have been found acceptable by the industry and the staff.
While both 10 CFR 50, Appendix B, and the industry standards include provisions for the application of the QA practices commensurate with the importance to safety of the SSCs for which these practices are applied, other regulations require that the explicit criteria of 10 CFR 50, Appendix B, still be applied. Entergy considers these requirements to be unduly burdensome for LSS components and that alternative practices may be applied to provide reasonable confidence that the LSS components will perform their safety related functions under design basis conditions.
10 CFR 50.55a, "Codes and Standards," lists the standards that have been approved for incorporation by reference. One of the standards listed is ASME,Section XI (i.e., as specified in 10 CFR 50.55a(a)(1)(ii)). ASME Code Case N-752 includes provisions for exempting Class 2 or Class 3 items categorized as LSS from the QA requirements of ASME Section XI, IWA-1400(n),
which invokes the requirement for a QA Program in accordance with either 10 CFR 50, Appendix B, or ASME NQA-1.
The provisions of Code Case N-752 are similar to those of 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Plants" (Reference 4), in that the provisions allow adjustment of the scope of items subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For items determined to be of LSS, alternative treatment requirements may be implemented in accordance with this Code Case. For equipment determined to be of High Safety Significance (HSS), requirements will not be changed.
0CAN052004 Enclosure Page 2 of 8 This application allows improved focus on equipment that has safety significance resulting in improved plant safety. This also enhances the effectiveness and efficiency of the NRCs oversight of the activities at ANO by focusing its resources on those SSCs that are most significant to maintaining public health and safety. Further, the reactor oversight process relies on the application of risk insights to determining significance of issues or events identified at licensee facilities.
The purpose for this request is to obtain approval for exempting ANO-1 and ANO-2, from the requirements of 10 CFR 50, Appendix B, for those safety-related Class 2 or Class 3 items categorized as LSS using the risk-informed methodology provided in Code Case N-752.
Implementation of Code Case N-752 including this requested exemption will allow ANO-1 and ANO-2 to improve focus on equipment that has greater safety significance, resulting in improved plant safety.
- 2.
BACKGROUND The NRC has established a set of regulatory requirements for commercial nuclear reactors to ensure that a reactor facility does not impose an undue risk to the health and safety of the public, thereby providing reasonable assurance of adequate protection to public health and safety. The current body of NRC regulations and their implementation are largely based on a "deterministic" approach.
This deterministic approach establishes requirements for engineering margin and quality assurance in design, manufacture, and construction. In addition, it assumes that adverse conditions can exist (e.g., equipment failures and human errors) and establishes a specific set of design basis events (DBEs). The deterministic approach then requires that the facility include safety systems capable of preventing or mitigating the consequences of those DBEs to protect public health and safety. Those items necessary to defend against the DBEs are defined as "safety-related," and are the subject of many regulatory requirements, herein referred to as "special treatments," designed to ensure that the SSCs are of high quality, high reliability, and have the capability to perform during postulated design basis conditions. Special treatment includes, but is not limited to, quality assurance, testing, inspection, condition monitoring, assessment, evaluation, and resolution of deviations. Typically, the regulations establish the scope of SSCs that receive special treatment using one of three different terms: "safety-related,"
"important to safety," or "basic component." Note that the terms "safety-related" and "basic component" are defined in the regulations, while "important to safety," used principally in the General Design Criteria (GDC) of 10 CFR 50, Appendix A, is not explicitly defined.
A probabilistic approach to regulation enhances and extends the traditional deterministic approach by allowing consideration of a broader set of potential challenges to safety, providing a logical means for prioritizing these challenges based on safety significance, and allowing consideration of a broader set of resources to defend against these challenges. In contrast to the deterministic approach, Probabilistic Risk Assessment (PRA) addresses credible initiating events by assessing the event frequency. Mitigating system reliability is then assessed, including the potential for common cause failures. The probabilistic approach to regulation is an extension and enhancement of traditional regulation by considering risk in a comprehensive manner.
0CAN052004 Enclosure Page 3 of 8 To take advantage of the safety enhancements available using PRA, in 2004 the NRC published a new regulation, 10 CFR 50.69, "Risk-informed Categorization and Treatment of SSCs for Nuclear Power Plants." The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of LSS, alternative treatment requirements may be implemented in accordance with the regulation. For equipment determined to be of HSS, requirements will not be changed or enhanced. This allows improved focus on equipment that has safety significance resulting in improved plant safety.
3 TECHNICAL EVALUATION In 2019, consistent with intent of 10 CFR 50.69, the ASME approved Code Case N-752, which is applicable to repair/replacement activities performed on items, including portions of a Class 2 or Class 3 system and/or the associated supports, that provide a passive (pressure boundary) safety function without a need for performing a risk-informed evaluation of the entire system.
The Code Case contains requirements on how a licensee categorizes these items using a risk-informed process, permits alternative treatments in lieu of certain ASME Section XI Repair/Replacement requirements consistent with the relative significance of items categorized as LSS, and manages the process over the lifetime of the plant. A risk-informed categorization process is employed to determine the safety significance of items of a system, including portions of a system, and place the SSCs into two risk-informed safety class categories, either HSS or LSS. The determination of safety significance is performed by an integrated decision-making process which uses both risk insights and traditional engineering insights. The safety functions include the passive design basis functions, as well as passive functions credited for severe accidents (including external events). The provisions may not be applied to the active safety function of any item. Special or alternative treatment for the items is applied as necessary to maintain functionality and reliability and is a function of how an item or portion of system is categorized. Finally, assessment activities are conducted to adjust the categorization and treatment processes as needed so that all items continue to meet all applicable requirements.
The Code Case does not allow for the elimination of functional requirements or allow items required by the deterministic design basis to be removed from the facility. Instead, the Code Case enables licensees to focus their resources on items that make a significant contribution to plant safety. For items that do not significantly contribute to plant safety on an individual basis, the Code Case allows a reasonable, though reduced, level of confidence that these items will continue to satisfy pressure boundary functional requirements.
Code Case N-752 includes provisions for exempting Class 2 or Class 3 items categorized as LSS from the QA requirements of ASME Section XI, IWA-1400(n), which invokes the requirement for a QA Program in accordance with either 10 CFR 50, Appendix B, or ASME NQA-1.
The categorization process of Code Case N-752 is delineated in Appendix I of the Code Case.
As mentioned above, this risk-informed categorization process is employed to determine the safety significance of items of a system, including portions of a system, and place the SSCs into two risk-informed safety class categories, either HSS or LSS. Details of the proposed categorization process is provided in Entergy Relief Request Number EN-20-RR-001,
0CAN052004 Enclosure Page 4 of 8 "Proposed Alternative to Use ASME Code Case N-752, Risk-Informed Categorization and Treatment for Repair/ Replacement Activities in Class 2 and 3 Systems,Section XI, Division 1,"
(Reference 5).
Appendix I, Section I-3.2 of Code Case N-752 requires that the plant-specific PRA shall be assessed to confirm it is applicable to the safety significant categorization of Code Case N-752 including verification of assumptions on equipment reliability for equipment not within the scope of the code case. Details of the PRA assessment process are provided in Reference 5.
Both ANO-1 and ANO-2 continue to demonstrate that the unit specific PRA is applicable to the safety significant categorization of Code Case N-752 including verification of assumptions on equipment reliability for equipment not within the scope of the code case. While this is the case, Entergy intends to review and assess the existing ANO-1 and ANO-2 PRA used to support the evaluations required by Code Case N-752 to verify their technical adequacy.
Code Case N-752 exempts Class 2 or Class 3 items, which have been categorized as LSS, from having to comply with the repair/replacement requirements of ASME Section XI. In lieu of these requirements, Code Case N-752, Paragraph -1420, requires the Owner to define alternative treatment requirements which confirm with reasonable confidence that each LSS item remains capable of performing its safety-related functions under design basis conditions.
These Owner defined treatment requirements must address or include all the provisions stipulated in Paragraphs -1420(a) through (j) of the code case. This approach to treatment is consistent with RISC-3 treatment requirements specified in 10 CFR 50.69(d)(2). Reference 5 provides additional details.
Entergy will develop new and/or revise existing procedures and documents to define treatment requirements for performing repair/replacement activities on LSS items in accordance with Code Case N-752. These procedures and the commitments to revise or develop new procedures are detailed in Reference 5.
4 REGULATORY EVALUATION 10 CFR 50.12(a), "Specific exemptions," states:
(a)
The Commission may, upon application by any interested person or upon its own initiative, grant exemptions from the requirements of the regulations of this part, which are (1)
Authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security.
(2)
The Commission will not consider granting an exemption unless special circumstances are present.
Proposed exemption authorized by law 10 CFR 50.12 was issued by the NRC under the authority granted to it pursuant to the Atomic Energy Act of 1954, as amended (68 Stat. 919), and Title II of the Energy Reorganization Act of 1974 (88 Stat. 1242), to provide for the licensing of production and utilization facilities.
Section 50.12 allows the NRC to grant specific exemptions from the requirements of 10 CFR 50.
0CAN052004 Enclosure Page 5 of 8 Granting the proposed exemption provides adequate protection to public health and safety. As described in this request, the proposed exemption will not present an undue risk to the public health and safety is consistent with the common defense and security. Therefore, the exemption is authorized by law.
Proposed exemption will not present an undue risk to the public health and safety, is consistent with the common defense and security In Reference 2, the staff noted:
The staff found that the licensees application of a risk-informed categorization process has identified a class of SSCs that have little or no safety significance with respect to protecting the health and safety of the public. The staff also found that the proposed treatment processes to be applied to activities associated with LSS and NRS SSCs, as described by the licensee, if effectively implement, will provide reasonable confidence that these SSCs remain capable of performing their safety functions under design-basis conditions. Based on these findings, the staff concluded that granting of the requested exemptions from 10 CFR Part 50, Appendix B, for LSS and NRS SSCs would pose no undue risk to public health and safety.
The staff found that the exemptions to be granted that relax the treatment requirements for safety-related LSS and NRS SSCs do not pose undue risk to public health and safety. This is supported by the staffs finding on categorization and treatment. Therefore, the staff concluded that it could reduce unnecessary regulatory burden without comprising safety.
This also enhances the effectiveness and efficiency of the NRCs oversight of the licensees activities at STP by focusing its resources on those SSCs that are most significant to maintaining public health and safety. Likewise, the licenses resources and attention can be focused on those SSCs that have the highest contribution to plant risk. Further, the reactor oversight process relies on the application of risk insights to determining significance of issues or events identified at licensee facilities. The licensees categorization process provides a method for establishing a licensing basis for STP that is consistent with the framework under which STPNOC operates its facility and by which the NRC oversees the licensees activities.
Using the risk-informed categorization process outlined in ASME Code Case N-752 presented in References 1 and 5, Entergy expects to achieve results consistent with 10 CFR 50.69.
Therefore, based on the proposed categorization process and the above information, the proposed exemption will not present an undue risk to the public health and safety. It is consistent with the common defense and security.
Special circumstances Under 10 CFR 50.12(a)(2) there is a list of special circumstances for which the NRC will consider for granting an exemption. In this request, 10 CFR 50.12(a)(2)(vi) is applicable special circumstances. 10 CFR 50.12(a)(2)(vi) states:
There is present any other material circumstance not considered when the regulation was adopted for which it would be in the public interest to grant an exemption. If such condition is relied on exclusively for satisfying paragraph (a)(2) of this section, the exemption may not be granted until the Executive Director for Operations has consulted with the Commission.
0CAN052004 Enclosure Page 6 of 8 As noted previously, consistent with the intent of 10 CFR 50.69, Code Case N-752 is applicable to repair/replacement activities performed on items, including portions of a Class 2 or Class 3 system and/or the associated supports, that provide a passive (pressure boundary) safety function. The provisions of Code Case N-752 may be applied on a system basis or on individual items categorized as LSS within selected systems. This included the exemption from the requirements of 10 CFR 50, Appendix B.
In Section IV.2.0 of SECY-04-0190, "Risk-informed Categorization and Treatment of SSCs for Nuclear Power Plants," which published the Final Rule for 10 CFR 50.69 (Reference 3), the NRC described the basis for approval of a similar exemption request (i.e., a risk-informed categorization process), by stating:
A major element of the rulemaking plan described in SECY-99-256 was the review of the STPNOC exemption request. The review of the STPNOC exemption request was viewed as a proof-of-concept prototype for this rulemaking rather than a pilot because it preceded development of draft rule language or related implementation guidance.
By letter dated July 13, 1999, STPNOC required approval of exemption requests to enable implementation of processes for categorizing the safety significance of SSCs and treatment of those SSCs consistent with its categorization process. The STPNOC process included many similar elements to that described in this rulemaking, but with some differences. Their process identified SSCs as being either high, medium, low or non-risk significant. The scope of the exemptions requested included only those safety-related SSCs that have been categorized as low safety significant or as non-risk significant using STPNOCs categorization process. The licensee indicated that the categorization and treatment processes would be implemented over the remaining licensed period of the facility. Thus, the basis for the exemptions granted was the NRC staffs approval of the licensees categorization process and alternative treatment elements, rather than a comprehensive review of the final categorization and treatment of each SSC (review of the process rather than the results is also the approach planned under the rulemaking). As a result of discussions with the staff on a number of topics, STPNOC submitted a revised exemption request on August 31, 2000.
On November 15, 2000, the NRC staff issued a draft safety evaluation (SE)(ADAMS accession number ML003761558), based on the revised exemption requests. Following the licensees response to the draft SE, the staff prepared SECY-01-0103 dated June 1, 2001 (ADAMS accession number ML011560317), to inform the Commission of the staffs finding regarding the STPNOC exemption review. The staff approved the STPNOC exemption requests by letter dated August 3, 2001 (ADAMS accession number ML011990368).
As noted in Section 20.2 of the STPNOC exemption request (Reference 2), the staff provided their basis for applying this special circumstance.
As discussed in Section 3.0 of this SE, the staff determined that the licensees categorization process provides a method for identifying safety-related SSCs that do not have a significant contribution to risk (LSS and NRS SSCs). The categorization process was found to use both a probabilistic and a deterministic (based on expert judgement) methodology that addressed the issues of defense-in-depth, safety margins, and aggregate risk impacts. In categorizing SSCs, the staff found that the licensee appropriately used the risk insights derived from the
0CAN052004 Enclosure Page 7 of 8 site specific PRA with the insights derived through the expert judgement of the GQA Working Group and Expert Panel. The probabilistic analysis included the plants PRA-based sensitivity studies which showed that the increases in risk were small and were consistent with the guidelines of RG 1.174. Therefore, the staff concluded that the licensees categorization process can be used as a method for deducing the scope of safety-related SSCs subject to special treatment.
As discussed in Section 4.0 of this SE, the staff found that the treatment processes included the programmatic elements for design control; procurement; installation; maintenance, inspection, test, and surveillance; corrective action; management and oversight; and configuration control. The staff concluded that these programmatic elements, if properly implemented, will provide reasonable confidence that safety-related LSS and NRS SSCs remain capable of performing their safety function under design-basis conditions.
The finding related to categorization, which combined with the finding on functionality, provides confidence that relaxing the special treatment requirements for safety-related LSS and NRS SSCs does not have a significant impact on public health and safety. Further, while risk insights may have been considered for some of the regulations from which the licensee is seeking exemption, the licensees categorization process was not considered when these regulations were adopted. STPNOCs categorization process relies on the risk insights derived from the application of the licensees PRA in determining the safety significance of SSCs along with insights derived from a deterministic methodology that incorporated the expert judgment of cognizant STPNOC technical management and staff. As such, the NRC determined that STPNOCs categorization process is a material circumstance not considered when the regulations were adopted.
The proposed categorization method for ASME Section XI repair/replacement activities can be found in Code Case N-752. Reference 5 provides the request to use the alternatives presented in Code Case N-752. These alternatives were not considered when the rules related to ASME XI repair/replacement activities were developed.
5 ENVIRONMENTAL CONSIDERATION A review of this requested has determined that the proposed exemption would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20. However, the proposed exemption does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed exemption meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(22). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed exemption.
0CAN052004 Enclosure Page 8 of 8
- 6.
SUMMARY
Code Case N-752 specifies requirements for performing risk-informed categorization and treatment for performing repair/replacement activities on Class 2 and 3 pressure retaining items or associated supports. The Code Case N-752 categorization process provides a comprehensive methodology for determining the safety significance of items - HSS or LSS.
Repair/replacement activities performed on items determined to be HSS must continue to comply with the ASME Section XI Code. Repair/replacement activities performed on LSS items may comply with alternative treatment requirements that are defined by the Owner but must comply with all provisions of paragraph -1420 of Code Case N-752. Finally, categorization and treatment requirements of Code Case N-752 applicable to repair/replacement activities are consistent with NRC requirements specified in 10 CFR 50.69.
Specifically, ASME Code Case N-752 includes provisions for exempting Class 2 or Class 3 items categorized as LSS from the QA requirements of ASME Section XI, IWA-1400(n), which invokes the requirement for a Quality Assurance Program in accordance with either 10 CFR 50, Appendix B, or ASME NQA-1.
In conclusion, Entergy requests approval for exempting ANO-1 and ANO-2, from the requirements of 10 CFR 50, Appendix B, for those safety-related Class 2 or Class 3 items categorized as LSS using the risk-informed methodology found in Code Case N-752. As discussed previously, this exemption has low nuclear safety significance and will not pose an undue risk to public health and safety and is consistent with the common defense and security.
Entergy requests authorization by June 1, 2021, to support planning activities associated with ANO-2, refueling outage in the Fall 2021 (2R28).
7 REFERENCES
- 1.
Code Case N-752, "Risk-Informed Categorization and Treatment for Repair/
Replacement Activities in Class 2 and 3 Systems,Section XI, Division 1," American Society of Mechanical Engineers, 2019.
- 2.
South Texas Project, Units 1 and 2, Cover Letter, STPNOC Exemption Request from Special Treatment Requirements of 10 CFR Part 21, 50, and 100 (TAC NOS MA6057 and MA6058) (ML011990368), dated August 3, 2001.
- 3.
SECY-04-0190 - Attachment 2 - FRN - Final Rule - 10 CFR 50.69, "Risk-informed Categorization and Treatment of SSCs for Nuclear Power Plants," (ML041000458)
- 4.
10 CFR 50.69, "Risk-informed Categorization and Treatment of SSCs for Nuclear Power Plants," (69 FR 68008).
- 5.
Entergy Operations, Inc. (Entergy) letter to the U. S. Nuclear Regulatory Commission (NRC), "Relief Request Number EN-20-RR-001 - Proposed Alternative to Use ASME Code Case N-752, Risk-Informed Categorization and Treatment for Repair/
Replacement Activities in Class 2 and 3 Systems,Section XI, Division 1,"
(0CAN052003) dated May 27, 2020.