ML18192C183

Fifth 10-Year Inservice Relief Request RR-05

ML18192C183
Person / Time
Site:	Duane Arnold
Issue date:	07/24/2018
From:	David Wrona Plant Licensing Branch III
To:	Nazar M Nextera Energy
Chawla M
References
EPID L-2017-LLR-0140
Download: ML18192C183 (13)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 24, 2018 Mr. Mano Nazar President and Chief Nuclear Officer Nuclear Division NextEra Energy Duane Arnold, LLC Mail Stop: NT3/JW 15430 Endeavor Drive Jupiter, FL 33478

SUBJECT:

DUANE ARNOLD ENERGY CENTER- FIFTH 10-YEAR INSERVICE RELIEF REQUEST RR-05 (EPID L-2017-LLR-0140)

Dear Mr. Nazar:

By letter dated November 16, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML173258215), as supplemented by letter dated April 27, 2018 (ADAMS Accession No. ML18117A204), NextEra Energy Duane Arnold, LLC (or the licensee) submitted a request for relief from certain requirements of American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," at Duane Arnold Energy Center (DAEC).

In this relief request (RR-05), the licensee proposed to use its risk-informed inservice inspection (RI-ISi) program for the fifth 10-year inservice inspection (ISi) interval as an alternative to the inspection requirements of ASME Code,Section XI. The U.S. Nuclear Regulatory Commission (NRC) staff's approval for the previous fourth interval RI-ISi program is documented in the letter dated January 31, 2007 (ADAMS Accession No. ML070090357).

Specifically, pursuant to Title 10 of the Code of Federal Regulations ( 10 CFR) 50.55a(z)( 1), the licensee requested to use the proposed RI-ISi program on the basis that the alternative would provide an acceptable level of quality and safety.

As described in the enclosed safety evaluation, the NRC staff has determined that the proposed alternative to the requirements of the ASME Code,Section XI, provides an acceptable level of quality and safety.

Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1). Therefore, the NRC authorizes the proposed alternative for the fifth ISi interval at DAEC that is scheduled to end on October 31, 2026.

All other requirements of ASME Code,Section XI, for which relief has not been specifically requested remain applicable, including a third-party review by the Authorized Nuclear lnservice Inspector.

M. Nazar If you have any questions, please contact the Project Manager, Mahesh Chawla at 301-415-8371 or via e-mail at Mahesh.chawla@nrc.gov.

Sincerely, Ju 9- ¥._

David J. Wrona, Chief Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-331

Enclosure:

Safety Evaluation cc: ListServ

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REGULATORY REGULATION RELIEF REQUEST NO. RR-05 FIFTH 10-YEAR INSERVICE INSPECTION INTERVAL NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER DOCKET NO. 50-331

1.0 INTRODUCTION

By letter dated November 16, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML173258215), as supplemented by letter dated April 27, 2018 (ADAMS Accession No. ML18117A204), NextEra Energy Duane Arnold, LLC (the licensee) submitted a request for relief from certain requirements of American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components" at Duane Arnold Energy Center (DAEC). In this relief request (RR-05), the licensee proposed to use its risk-informed inservice inspection (RI-ISi) program for the fifth 10-year inservice inspection (ISi) interval as an alternative to the inspection requirements of ASME Code,Section XI. The U.S. Nuclear Regulatory Commission (NRC) staff's approval for the previous fourth interval RI-ISi program is documented in the letter dated January 31, 2007 (ADAMS Accession No. ML070090357).

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z){1), the licensee requested to use the proposed RI-ISi program on the basis that the alternative would provide an acceptable level of quality and safety.

2.0 REGULATORY EVALUATION

Pursuant to 10 CFR 50.55a(g), ASME Code Class 1, 2, and 3 components (including supports) must meet the requirements "except design and access provisions and pre-service examination requirements," set forth in the ASME Code to the extent practical within the limitations of the design, geometry, and materials of construction of the components. The regulation in 10 CFR 50.55a(g) also states that ISi of the ASME Code Class 1, 2, and 3 components is to be performed in accordance with Section XI of the ASME Code and applicable addenda except where specific written relief has been granted by the NRC.

Enclosure

The regulations in 10 CFR 50.55a(g)(4), also require during the first 10-year ISi interval and during subsequent intervals, the licensee's ISi program comply with the requirements in the latest edition and addenda of the ASME Code incorporated by reference into 10 CFR 50.55a(b) 12 months before the start of the 120-month inspection interval, subject to the conditions listed in 10 CFR 50.55a(b). DAEC is currently in its fifth 10-year ISi interval.

In accordance with the ASME Code (as incorporated by reference in 10 CFR 50.55a), certain percentages of ASME Code Category B-F, B-J, C-F-1, and C-F-2 pressure retaining piping welds must receive ISi during each 10-year ISi interval.

As stated in 10 CFR 50.55a(z), alternatives to the requirements of paragraphs (b) through (h) of 10 CFR 50.55a or portions thereof may be used, when authorized by the NRC, and the proposed alternate must be submitted and authorized prior to implementation. The regulation in 10 CFR 50.55a(z)(1) requires that the submittal must demonstrate that the proposed alternative would provide an acceptable level of quality and safety.

The NRC staff evaluated the proposed RI-ISi program using the following guidance documents:

• Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment

[PRA] In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" (ADAMS Accession No. ML100910006). RG 1.174 provides guidance on the use of PRA findings and risk insights in support of licensee requests for changes to a plant's current licensing basis (CLB). NRC RG 1.174 also defines an acceptable approach to analyzing and evaluating proposed LB changes. The approach includes traditional engineering evaluations supported by insights derived from the use of PRA methods about the risk significance of the proposed changes. In implementing risk-informed decision making, the NRC expects CLB changes to meet the acceptance guidelines and key principles of risk-informed guidance specified in NRC RG 1.174.

• RG 1.178, "An Approach for Plant-Specific Risk-Informed Decision making - In-service Inspection of Piping" (ADAMS Accession No. ML032510128). RG 1.178 describes methods acceptable to the NRC for integrating insights from PRA techniques with traditional engineering analyses into ISi programs for piping. Incorporating risk insights into the programs can focus inspections on the more important locations and reduce personnel exposure, while at the same time maintaining or improving public health and safety.

• RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities" (ADAMS Accession No. ML090410014). RG 1.200 describes one acceptable approach for determining whether the quality of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results such that the PRA can be used in regulatory decision-making.

• The Electric Power Research Institute (EPRI) Topical Report (TR)-1021467-A, "Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs," (ADAMS Accession No. ML12171A450).

• NUREG-0800, Chapter 3.9.8, "Standard Review Plan [SRP] for the Review of Risk-Informed In-service Inspection of Piping" (ADAMS Accession No. ML032510135). SRP 3.9.8 describes review procedures and acceptance guidelines for NRC staff reviews of proposed plant-specific, risk-informed, changes to a licensee's ISi program for piping.

SRP provides guidance for evaluating the licensee's requests for changes to the CLB due to use of risk insights.

The ASME Code requires 100 percent of B-F welds and 25 percent of 8-J welds greater than 1 inch nominal pipe size be selected for volumetric or surface examination, or both, on the basis of existing stress analyses. For Categories C-F-1 and C-F-2 piping welds, 7.5 percent of non-exempt welds are selected for volumetric or surface examination, or both. According to 10 CFR 50.55a(z), the NRC may authorize alternatives to the requirements of 10 CFR 50.55a(g),

if an applicant or licensee demonstrates that (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified two requirements of 10 CFR 50.55a would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Based on the above and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request and the NRC to authorize the licensee-proposed alternative.

3.0 TECHNICAL EVALUTION 3.1 Components Affected All ASME Code Class 1 and 2 piping welds subject to the requirements of ASME Section XI, Table IWB-2500-1, Examination Categories B-F and B-J, and Table IWC-2500-1, Examination Categories C-F-1 and C-F-2.

3.2 Applicable ASME Code Edition and Addenda The current Code of Record for DAEC is the 2007 Edition of ASME Code,Section XI, with the 2008 Addenda.

3.3 Applicable ASME Code Requirements The selection process for ASME Code Class 1 and Code Class 2 piping welds to be examined in the fifth inspection interval is required to be prescriptively determined in accordance with ASME Code,Section XI, Table IWB-2500-1, Examination Categories B-F and B-J, and Table IWC-2500-1, Examination Categories C-F-1 and C-F-2.

3.4 Duration of Propose Alternative The licensee stated that the proposed alternative is applicable for the fifth 10-year ISi interval that commenced on November 1, 2016, and is scheduled to end on October 31, 2026.

3.5 Licensee's Proposed Alternative The licensee proposed to update the RI-ISi program approved for the previous 10-year ISi interval and apply the updated program to the fifth 10-year ISi interval. Other nonrelated portions of the ASME Code,Section XI, are unaffected.

The proposed updated RI-ISi program is based on EPRI TR-112657, Revision B-A, "Revised Risk-Informed lnservice Inspection Evaluation Procedure (ADAMS Accession No. ML013470102)," and RG 1.174 "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decision on Plant Specific Changes to the Licensing Basis" risk acceptance criteria.

The licensee proposed to use its RI-ISi program at DAEC during the fifth 10-year ISi interval as an alternative to the inspection requirements of ASME Code,Section XI, for Class 1 Examination Category B-F and B-J piping welds and Class 2 Examination Category C-F-1 and C-F-2 piping welds. In accordance with 10 CFR 50.55a(z)(1 ), the licensee requested to continue to use a risk-informed process based on the guidance in EPRI TR-112657, Revision B-A. The risk-informed process considers the potential for pipe rupture associated with the degradation mechanisms and the consequences of pipe rupture to determine the risk categories of piping locations. These risk categories (high; medium, and low) are used to select inspection locations.

The RI-ISi program is a living program requiring feedback of new relevant information (e.g.,

plant design changes and operating experience) to ensure the appropriate identification of safety significant piping locations.

3.6 Basis of the Proposed Alternative The licensee's basis of the alternative is summarized as follows. The DAEC RI-ISi program is based on the methodology in EPRI TR 112657 that was reviewed and approved by the NRC staff. For the previous fourth ISi interval, the licensee submitted RR NDE-R005 to use its RI-ISi program by letter dated June 30, 2006 (ADAMS Accession No. ML061870230), and the NRC staff reviewed and approved the RR in the letter dated January 31, 2007 (ADAMS Accession No. ML070090357).

The DAEC RI-ISi program is a living program in accordance with Nuclear Energy Institute (NEI) 04-05, "Living Program Guidance To Maintain Risk-Informed lnservice Inspection Programs For Nuclear Plant Piping Systems." An updated risk impact analysis was performed for the fifth interval RI-ISi program in consideration of the contributions to risk from all the systems. The updated analysis indicates that the change in risk associated with the implementation of the RI-ISi Program meets the requirements of RGs 1.174 and 1.178. Table 1 summarizes the total changes in core damage frequency (CDF) and large early release frequency (LERF) without crediting the enhanced probability of detection (POD) of the RI-ISi examinations. Table 1 also indicates that the changes in CDF and LERF meet the acceptance criteria of EPRI TR-112675, Revision B-A.

Table 1. Total Changes in Risk from ASME Code,Section XI Inspections:

Fifth Interval RI-ISi (without crediting the enhanced POD)

Delta-CDF 2.15E-08 Delta-LE RF 2.14E-08 Criteria for Delta-CDF 1.00E-06 Criteria for Delta-LERF 1.00E-07

3. 7 NRC Staff Evaluation The NRC staff has evaluated this RR pursuant to 10 CFR 50.55a(z)(1). The NRC staff focused on whether the proposed alternative provides an acceptable level of quality and safety.

The proposed RI-ISi program is based on the methodology in EPRI TR-112657, Revision B-A.

The EPRI report provides technical guidance on categorizing the risk significance of piping components and selecting inspection locations for the purpose of developing an RI-ISi program.

The licensee will inspect at least 25 percent of the welds in high-risk categories (Categories 1, 2, and 3) and at least 10 percent of the welds in medium-risk categories (Categories 4 and 5),

as indicated in attachment 1 to the RR. The NRC staff finds that the inspection sample sizes are consistent with the guidance in EPRl TR-112657, Revision B-A, and, therefore, are acceptable. The NRC staff also noted that the changes in CDF and LERF with the implementation of the RI-ISi program meet the acceptance criteria of EPRI TR-112675, Revision B-A, as summarized in Table 1.

In addition, the NRC staff confirmed that the proposed RI-ISi program does not affect the augmented inspections for intergranular stress corrosion cracking (IGSCC) and flow-accelerated corrosion (FAC) and the existing augmented inspections will continue to be implemented as described below.

DAEC applies the NRC-approved Boiling Water Reactor Vessel and Internals Project-75-A, "Technical Basis for Revision to Generic Letter [GL] 88-01 Inspection Schedules," which provides alternative frequency requirements to GL 88-01 for the examination of welds subject to IGSCC. The RI-ISi program subsumes Category A welds (resistant to IGSCC) in accordance with EPRI TR-112657, Revision B-A. These IGSCC-resistant welds are assigned to the low failure potential if no other damage mechanisms are present. The NRC staff noted that the augmented inspection program for the other piping welds susceptible to IGSCC (Categories B through G) is not affected by the RI-ISi program in accordance with EPRI TR-112657, Revision B-A. The NRC staff also noted that the augmented inspection program for FAC per GL 89-08 is relied upon to manage the effect of the damage mechanism and is not affected by the RI-ISi program, and is consistent with EPRI TR-112657, Revision B-A.

The NRC staff reviewed and evaluated the licensee's proposed RI-ISi program, including those portions related to the applicable methodology and processes based on guidance and acceptance guidelines provided in RGs 1.174 and 1.178, SRP 3.9.8, and EPRI TR-112657, Revision B-A. An acceptable RI-ISi program is expected to meet the five key principles of risk-informed decision-making discussed in RGs 1.174 and 1.178, SRP 3.9.8, and EPRI TR-112657, as stated below:

1. The proposed change meets the current regulations unless it is explicitly related to a requested exemption or rule change.

2. The proposed change is consistent with the defense-in-depth philosophy.

3. The proposed change maintains sufficient safety margins.

4. When proposed changes result in an increase in CDF or risk, the increases should be small and consistent with the intent of the Commission's Safety Goal Policy Statement.

5. The impact of the proposed change should be monitored by using performance measurement strategies.

The first key principle is met in this RR because an alternative ISi program may be authorized pursuant to 10 CFR 50.55a(z) and, therefore, an exemption request is not required.

The second and third principles are met because (a) the inservice inspection is an integral part of defense-in-depth; (b) as part of the RI-ISi process, the risk significance categorization and specification of the number and location for inspections will maintain the intent of ISi (i.e.,

identifying and repairing flaws before pipe integrity is challenged); and (c) the RI-ISi process does not affect the adequacy of design basis accident evaluations such that sufficient safety margins will be maintained.

The fourth key principle requires an estimate of the change in risk. The change in risk estimate is dependent on the location of inspections in the proposed RI-ISi program compared to the location of inspections that would be performed using the deterministic requirements of the ASME Code,Section XI. The NRC staff has previously determined that it is not necessary to develop a new deterministic ASME Code program for each new 10-year ISi interval. Instead, it is acceptable to compare the new proposed RI-ISi Program with the last deterministic ASME Code program. In Section 5 of the RR, the licensee stated that the change in risk of implementing the RI-ISi program was determined to represent a negligible change in risk when compared to the last deterministic Section XI inspection program; therefore, the NRC staff finds that implementation of the RI-ISi program will have a small and acceptable impact on risk consistent with the acceptance guidelines in RG 1.174.

The fourth key principle also requires demonstration of the technical adequacy of the PRA. As discussed in RGs 1.178 and 1.200, an acceptable change in risk evaluation (and risk-ranking evaluation used to identify the most risk significant locations) requires the use of a PRA of appropriate technical adequacy that models the as-built and as-operated plant.

EPRI TR-1021467-A provides guidance on the minimum acceptable quality requirement for a PRA used to support a RI-ISi program. EPRI TR-1021467-A was developed to provide guidance and supporting requirements (SRs) for the traditional methodology of EPRI TR-112657. The licensee stated that in March 2011 a focused PRA peer review was held with a review focused on the disposition of findings and suggestions as a result of the full scope PRA peer review conducted in 2007. The full scope PRA peer review conducted in 2007 used version RA-Sb-2005 of the ASME Code Standard as endorsed and clarified in RG 1.200, Revision 1.

The focused scope PRA peer review conducted in March 2011 utilized version RA-Sa-2009 of the ASME Code Standard as endorsed and clarified by the NRC in RG 1.200, Revision 2. The licensee stated that the focused scope peer review team found that most of the previously identified gaps had been incorporated into Revision 6 of the PRA. The 12 findings that were identified to not fully meet Capability Category I or II requirements have been addressed and summarized in Table C-1 of the RR. The licensee stated that all findings were addressed, therefore, there is no impact on the quantitative risk results and the PRA conclusions of the RI-ISi consequence evaluation.

In Table C-1 of the RR, the licensee provided the focused scope PRA peer review finding with the description of any gaps and the disposition of the findings. The NRC staff identified during its review the need for the licensee to further address several previously identified gaps that were incorporated into the DAEC PRA.

In a request for additional information (RAl)-01, the NRC staff requested additional information stating that the revised program continues to be less than the EPRI criterion with a total change in CDF and LERF of 4.34E-09 and4.3E-09, respectively. The NRC staff requested the licensee to explain why the total changes in CDF and LERF are essentially equal and why the 03RWCU system parameter has a change in LERF exceeding that of CDF. In response to RAl-01 (Reference 6), the licensee states the change in LERF due to application of the RI-ISi process was estimated by substituting the conditional large early release probability (CLERP) for conditional core damage probability (CCDP). In response to RAl-01, the licensee also states when equal values for the maximum CCDP and CLERP are inserted into the equation, they result in a change in risk that is nearly the same for LERF as it is for CDF.

Therefore, the reason the changes in CDF and LERF are so close is because the maximum CCDP and CLERP at DAEC are equal. The staff concludes that this resolution is acceptable for this application and implementation of the EPRI traditional RI-ISi methodology.

In RAl-02, the NRC staff requested the licensee to confirm that all technical elements in Part 2 of the ASME Code Standard, including internal flooding, were included in the March 2011 review or identify any gaps between the peer review and the requirements in RG 1.200, Revision 2, particularly focusing on those technical elements for which the most current peer review remains from 2007 against Revision 1 of RG 1.200. In response to RAl-02, the licensee stated that the 2011 DAEC PRA Focused Peer Review is a review of the technical elements of the internal events, including internal flood at-power, PRA with a focus on DAEC's disposition offindings and suggestions and upgrades as a result of the 2007 peer review.

Previously, the NRC staff concluded that DAEC's internal events PRA is adequate in its "Issuance of Amendment Regarding Transition to a Risk-Informed, Performance Based Fire Protection Program in Accordance with 10 CFR 50.48(c)" (ADAMS Accession No. ML13210A449). The NRC staff stated that the DAEC internal events PRA full scope peer review was performed in December 2007 using the NEI 05-04 process, the combined PRA standard, ASME/ANS [American Nuclear Society]-RA-Sa-2005, and RG 1.200, Revision 1.

Additionally, the NRC staff described the focused scope peer review of the internal events PRA was conducted in March 2011 using the combined standard, ASME/ANS-RA-Sa-2009, and RG 1.200, Revision 2. As a result of this review, the staff concluded that the DAEC internal events PRA is sufficiently technically adequate. As a result of review of the information provided in the RI-ISi submittal, the NRC concludes that the information has not changed since the National Fire Prevention Association-805 amendment and is, therefore, technically adequate for use in the RI-ISi program.

In RAl-03(a), regarding the reasonableness of the prior and posterior distributions in Fact and Observation (F&O) DA-D4-01a, the NRC staff requested the results of the final Bayesian analysis and the conclusion by the peer review team that SR DA-D4-01 a is now met. In response to RAl-03(a) (Reference 6), the licensee clarified that this SR was graded at Capability Category I by the peer review team. The licensee also performed a sensitivity PRA case and judged the original posterior mean values to be acceptable.

Furthermore, an independent assessment team reviewed this F&O for closure and concluded that the licensee's resolution was adequate and complete. The NRC staff concludes that this resolution is acceptable for this application and implementation of the EPRI traditional RI-ISi methodology.

In RAl-03(b), regarding the F&O IE-B3-01a dealing with subsuming and screening not being met, the NRC staff requested that the licensee provide information that includes the applicable risk metrics and determines the effects. The licensee responded (Reference 6) that F&O IE-B3-01A was assessed during the 2017 independent closure review of open F&Os for DAEC's Internal Events and Fire PRAs. The review team concluded that the licensee's resolution is adequate, but the F&O remains open pending update of associated documentation. The NRC staff concludes that this resolution is acceptable for this application and implementation of the EPRI traditional RI-ISi methodology.

In RAl-03(c), regarding F&O HR-A1-01A and F&O HR-A2-01A, the NRC staff requested why the use of a different approach to accurately identify pre-initiators. The licensee responded (Reference 6) that review of procedures alone would result in overlooking pre-initiator human-error probabilities that would otherwise be identified by understanding and reviewing system configuration and operational aspects. The licensee also explained further that F&O HR-A1-01A and F&O HR-A2-01A was assessed during the 2017 independent closure review of open F&Os for DAEC's Internal Events and Fire PRAs. The review team concluded that NextEra Energy Duane Arnold's resolution is adequate but the F&O remains open pending update of associated documentation. The NRC staff concludes that this resolution is acceptable for this application and implementation of the EPRI traditional RI-ISi methodology.

Regarding RAl-03(d), the F&O HR-G1-01A disposition cites addition of train-level pre-initiating human factors engineering (HFE) to the model. However, it is was not clear to the NRC staff if this deficiency regarding "corresponding system-level dependent HFEs" has been corrected as well. The licensee responded (Reference 6) that DAEC's documented process for selecting pre-initiator human error probability events for inclusion in the PRA is judged to be adequately addressed in the PRA Human Reliability Analysis Notebook .. The licensee also explained that F&O HR-G1-01A was assessed during the 2017 independent closure review of open F&Os for DAEC's Internal Events and Fire PRAs. The review team concluded that the licensee's resolution is adequate, but the F&O remains open pending update of associated documentation.

The NRC staff concludes that this resolution is acceptable for this application and implementation of the EPRI traditional RI-ISi methodology.

In RAl-04(a) regarding Item No. 49, the statement that Level 2 results had little influence in the RI-ISi evaluation, the NRC staff requested the licensee to provide additional information to explain why such influence would not affect the risk metrics for this application. The licensee stated (Reference 6) that since uncertainty associated with recovery of reactor pressure vessel injection systems following onset of core damage is treated conservatively, through use of high failure probabilities in the Level 2 PRA, Item 49 of Table C-2, is judged to have no adverse impact on the RI-ISi application. The NRC staff concludes that this resolution is acceptable for this application and implementation of the EPRI traditional RI-ISi methodology.

In RAl-04(b), the NRC staff requested the licensee provide information for Item No. 58. The licensee in response (Reference 6) stated that Item 58 was inadvertently omitted and provided the required information. The staff concludes that this resolution is acceptable for this application and implementation of the EPRI traditional RI-ISi methodology.

In RAl-04(c), the NRC staff requested that the licensee provide the NRC staff with information regarding any effect from the retention of gate reactor core isolation coolant (RCIC)-07-01 that could lead to an underestimate of any of the applicable risk metrics for this application.

In RAl-04(d), the NRC staff asked the licensee to provide information regarding the basis for the statement for Item No. 63. For this item, the licensee stated in part that modeling of water hammer is a very low contributor to HPCI and RCIC system failure. The licensee's response to both RAl-04(c) and RAl-04(d) (Reference 6) were similar as the resolution affected both. The licensee demonstrated that the risk significance of the water hammer failure mode is very low compared to other failure modes for RCIC and high-pressure cooling injection (HPCI). As such, including this failure mode for RCIC and HPCI in the PRA is unlikely to lead to the underestimation of any applicable risk metric for the RI-ISi application. The staff concludes that this resolution is acceptable for this application and implementation of the EPRI traditional RI-ISi methodology.

In RAl-04(e), the NRC staff asked the licensee to provide information regarding any potential dependencies among the human action events of Item No. 64. The licensee responded (Reference 6) that increasing the dependency between the two feedwater related actions is not expected to cause the value of these combination events to increase above their floor values; therefore, calculation of CCDP and CLERP for the RI-ISi application would not be impacted. The staff concludes that this resolution is acceptable for this application and implementation of the EPRI traditional RI-ISi methodology.

In RAl-05, the NRC staff asked the licensee to provide the NRC with information regarding Table C-3, EC No. 156110. The PRA disposition of this item is based on it being of low safety significance to well water system isolation valves. The licensee responded (Reference 6) that the risk importance of individual valves in the system is low due to the relatively high level of redundancy provided by its four pumps. The NRC staff concludes that this resolution is acceptable for this application and implementation of the EPRI traditional RI-ISi methodology.

Therefore, the NRC staff finds that the licensee has assessed the technical adequacy of its PRA using an appropriate version of RG 1.200 and the PRA is consistent with the methodology in EPRI TR-112657.

The fifth key principle requires that the impact of the proposed change be monitored by using performance measurement strategies. The RI-ISi program is a living program and, as such, is subject to periodic reviews.

Based on the above, the NRC staff determined that the proposed RI-ISi Program for the fifth ISi interval met the five key principles of risk-informed regulation, and therefore, provides an acceptable level of quality and safety.

4.0 CONCLUSION

Based on the above discussion of this safety evaluation, the NRC staff determines that the five key principles of risk-informed decision-making for the licensee's proposed fifth 10-year RI-ISi program are met. Therefore, the licensee's proposed fifth 10-year RI-ISi program is acceptable.

The NRC staff finds that the proposed alternative to the requirements of the ASME Code,Section XI, provides an acceptable level of quality and safety.

Accordingly, the NRC staff concludes that the licensee has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(z)(1 ), and is in compliance with the ASME Code's requirements.

Therefore, the NRC staff authorizes continued use of the proposed alternative for the fifth ISi interval at DEAC RI-ISi program as described in RR-05 at DAEC, for the fifth 10-year ISi interval scheduled to end on October 31, 2026.

All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved remains applicable, including third-party review by Authorized Nuclear In-service Inspector.

5.0 REFERENCES

1. EPRI TR-112657, Revision B-A, Revised Risk-Informed In-service Inspection Evaluation Procedure, December 1999 (ADAMS Accession No. ML013470102).

2. ASME Code Case N-578-1, Risk-Informed Requirements for Class 1, 2, or 3 Piping, Method B,Section XI, Division 1, (c) ASME, New York, New York, March 28, 2000.

3. ASME Code Case N-770-1, Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities,Section XI, Division 1, (c) ASME, New York, New York, December 25, 2009.

4. Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" (ADAMS Accession No. ML100910006)

5. NextEra Energy Duane Arnold "NG-18-0046 (signed) Response to RAI regarding Relief Request RR-05." April 27, 2018. (ADAMS Accession No. ML18117A204)

6. USNRC to R. M. Anderso, "Duane Arnold Energy Center - Issuance Of Amendment Regarding Transition To A Risk-Informed, Performance-Based Fire Protection Program In Accordance with 10 CFR 50.48(c) (TAC NO. ME6818)," September 10, 2013 (ADAMS Accession No. ML13210A449).

Principal Contributors: C. Spore, DSS/APLB S. Min, DMLR/MPHB

ML18192C183 *via memorandum OFFICE NRR/LPL3/PM NRR/LPL3/LA NRR/MHPB/BC(A)*

NAME MChawla SRohrer $Cumblidge DATE 07/16/18 07/17/18 06/15/18 OFFICE NRR/DRA/BC* NRR/LPL3/BC NAME KHsueh DWrona DATE 07/18/18 07/24/18