ML15226A412
| ML15226A412 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 10/30/2015 |
| From: | Travis Tate, Bhalchandra Vaidya Plant Licensing Branch III |
| To: | Bryan Hanson Exelon Generation Co, Exelon Nuclear |
| Vaidya B, NRR/DORL/LPL3-2, 415-3308 | |
| References | |
| TAC MF5654, TAC MF5655 | |
| Download: ML15226A412 (23) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 30, 2015 Mr. Bryan C. Hanson Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer (CNO)
Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
LASALLE COUNTY STATION, UNITS 1 AND 2, RELIEF FROM THE REQUIREMENTS OF THE ASME CODE RE: RR 13R14, PROPOSED ALTERNATIVE TO THE EXAMINATION REQUIREMENTS FOR NOZZLE-TO-VESSEL WELDS AND INNER RADII SECTIONS IN ACCORDANCE WITH 10 CFR 50.55a(z)(1) (TAC NOS. MF5654 AND MF5655)
Dear Mr. Hanson:
By letter dated January 29, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15030A175), as supplemented by letters dated June 8, and July 30, 2015 (ADAMS Accession Nos. ML15160A620 and ML15211A530, respectively), Exelon Generation Company, LLC (EGC, the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) for the use of alternatives to certain American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (B&PV Code),Section XI, requirements at LaSalle County Station, Units 1 and 2 (LSCS).
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a(a)(3)(i), the licensee requested to use a proposed alternative on the basis that the alternative provides an acceptable level of quality and safety. The proposed changes in Relief Request (RR) 13R-14 would revise the inspection requirements for certain reactor pressure vessel (RPV) nozzle-to-vessel welds and nozzle inner radii from those based on ASME B&PV Code,Section XI, to an alternative based on ASME Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds."
The NRC staff has reviewed the subject request, as supplemented, and concludes, as set forth in the enclosed safety evaluation (SE), that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1) and is in compliance with the ASME Code requirements.
Therefore, the NRC authorizes the licensee's proposed alternative for inspection of nozzle-to-vessel shell welds and nozzle inner radii sections of reactor pressure vessel nozzles listed in Section 3.1 of this SE for the third 10-year inservice inspection interval for LSCS, Units 1 and 2.
Use of the ASME Code Case is authorized until such time as the ASME Code Case is published in a future version of Regulatory Guide (RG) 1.147 and incorporated by reference in
B. Hanson 10 CFR 50.55a(b). At that time, if the licensee intends to continue implementing this ASME Code Case, it must follow all provisions of ASME Code Case N-702 with conditions as specified in RG 1.147 and limitations as specified in 10 CFR 50.55a(b)(4), (b)(5), and (b)(6), if any.
All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector.
Sincerely, Travis L. Tate, Chief Plant Licensing 111-2 and Planning and Analysis Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-373 and 50-37 4
Enclosures:
As+ stated cc w/encl: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST 13R-14 FOR OPERATING LICENSE NOS. NPF-21 AND NPF-18 LASALLE COUNTY STATION. UNITS 1 AND 2 EXELON GENERATION COMPANY. LLC DOCKET NOS. 50-373 AND 50-37 4
1.0 INTRODUCTION
By letter dated January 29, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15030A175), as supplemented by letters dated June 8, and July 30, 2015 (ADAMS Accession Nos. ML15160A620 and ML15211A530, respectively), Exelon Generation Company, LLC (EGC, the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC or Commission) for the use of alternatives to certain American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (B&PV Code),Section XI, requirements at LaSalle County Station, Units 1 and 2 (L$CS).
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR),
Section 50.55a(a)(3)(i), the licensee requested to use a proposed alternative on the basis that it provides an acceptable level of quality and safety. The relief requested in Relief Request (RR) 13R-14 revision of the inspection requirements for certain reactor pressure vessel (RPV) nozzle-to-vessel welds and nozzle inner radii from those requirement based on ASME B&PV Code,Section XI, to an alternative based on ASME Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds."
2.0 REGULATORY EVALUATION
lnservice inspection (ISi) of the ASME Code Class 1, 2, and 3 components is performed in accordance with Section XI of the ASME Code and applicable addenda as a way to detect anomaly and degradation indications so that structural integrity of these components can be maintained. This is required by 10 CFR, Section 50.55a(g), except where specific relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). Section 50.55a(z) of 10 CFR states that alternatives to the requirements of paragraphs (b) through (h) of 10 CFR 50.55a or portions thereof may be used when authorized by the Director, Office of Nuclear Reactor Regulation. A proposed alternative must be submitted and authorized prior to implementation.
The applicant or licensee must demonstrate that: (1) the proposed alternative would provide an Enclosure
acceptable level of quality and safety; or (2) compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
For all RPV nozzle-to-vessel shell welds and nozzle inner radii, ASME Code,Section XI, requires 100 percent inspection during each 10-year ISi interval. However, ASME Code Case N-702 provides an alternative which reduces the inspection of RPV nozzle-to-vessel shell welds and nozzle inner radius areas from 100 percent to 25 percent of the nozzles for each nozzle type during each 10-year interval. The NRC approved the Electric Power Research Institute (EPRI) Technical Report (TR) 1003557 report, "BWRVIP [Boiling Water Reacator Vessel and Internals Project]-108: BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radii" and the BWRVIP-241 report, "BWRVIP-241: BWR Vessel and Internals Project, Probabilistic Fracture Mechanics [PFM] Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," which contain the technical basis supporting ASME Code Case N-702. The BWRVIP-241 report contains additional PFM results supporting revision of the evaluation criteria in the BWRVIP-108 report. Hence, the conditions and limitations specified in the April 19, 2013, safety evaluation (SE) (ADAMS Accession No. ML13071A240) for the BWRVIP-241 report supersede those in the SE for the BWRVIP-108 report. The NRC requires that each licensee demonstrate the plant-specific applicability of the BWRVI P-241 report to its units in the RR by demonstrating that the general and nozzle-specific criteria are satisfied.
The ASME Code of record for LSCS, Units 1 and 2, for the third 120-month interval ISi program is the ASME Code,Section XI, 2001 Edition of the through the 2003 Addenda.
3.0 TECHNICAL EVALUATION
3.1 The Licensee's Request for Alternative Component(s) for which Alternative is Requested (ASME Code Class 1)
Reactor Vessel Nozzles: N1, N2, N3, NS, N6, N7, N8, N9, N16, and N18 Examination Category B-D, "Full Penetration Welded Nozzles in Vessels" Examination Item Number B3.90, "Nozzle-to-Vessel Welds" and B3.100, "Nozzle Inside Radius Section"
Applicable Code Edition and Addenda
The third 10-year lnservice Inspection Program at LSCS, Units 1 and 2, is based on the ASME B&PV Code,Section XI, 2001 Edition through the 2003 Addenda. Additionally, for ultrasonic examinations, ASME Section XI, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," of the 2001 Edition is implemented, as required and modified by 10 CFR 50.55a(b)(2)(xv).
ASME Code Requirement for which Alternative is Requested (as stated)
The applicable requirement is contained in Table IWB-2500-1, "Examination Category B-O, Full Penetration Welded Nozzle in Vessels - Inspection Program B." Class 1 nozzle-to-vessel weld and nozzle inner radii examination requirements are delineated in Item Number B3.90, "Nozzle-to-Vessel Welds," and B3.100, "Nozzle Inside Radius Section."
The required method of examination is volumetric. All nozzles with full penetration welds to the vessel shell (or head) and integrally cast nozzles are examined each interval.
Licensee's Proposed Alternative to the.ASME Code (as stated)
In accordance with 10 CFR 50.55a(z)(1 ), relief is requested from performing the required examinations on 100 percent of the nozzle assemblies identified in Tables 5-1 [1l and 5-2[2l below (see Enclosure 1 for a list of RPV Examination Category B-D Nozzles for which this relief request is applicable). As an alternative for all welds and inner radii identified in Tables 5-1 and 5-2, EGC proposes to examine a minimum of 25 percent of the LSCS, Units 1 and 2, nozzle-to-vessel welds and inner radii sections, including at least one nozzle from each system and nominal pipe size, in accordance with Code Case N-702.
Code Case N-702 stipulates that a visual test (VT)-1 examination may be used in lieu of the volumetric examination for the inner radii (i.e., Item No. B3.100, "Nozzle Inside Radius Section"). EGC may utilize Code Case N-648-1 with associated RG 1.147 conditions for the nozzles selected for examination. Volumetric examinations ofthe inside radius section of those reactor vessel nozzles selected for examination will be completed if Code Case N-648-1 is not applied.
Licensee's Basis for Proposed Alternative (as stated)
Electric Power Research Institute (EPRI), Technical Report (TR) 1003557, "BWRVIP-108: Boiling Water Reactor Vessel and Internals Project Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," provides the technical basis for Code Case N-702.
BWRVIP-108 determined that failure probabilities at the nozzle blend radius region and nozzle-to-vessel shell weld due to a Low Temperature Overpressure (LTOP) event are very low (i.e., <1 x 106 for 40 years) with or without an ISi. The report concluded that inspection of 25 percent of each nozzle type is technically justified. The BWRVIP-108 report was approved by the NRC in an SE dated December 19, 2007 (ADAMS Accession No. ML073600374), and requires additional criteria to be met in order to apply the technical basis of BWRVIP-108 for the reduction of inspection coverage of the RPV nozzles and nozzle-to-vessel shell welds.
[1] This refers to Table 5-1 of the licensee's January 29, 2015, submittal.
[2] This refers to Table 5-2 of the licensee's January 29, 2015, submittal.
BWRVIP-108 was supplemented by EPRI TR 1021005, "BWRVIP-241: Boiling Water Reactor Ve.ssel and Internals Project Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," and was approved by the NRC in an SE dated April 19, 2013 (ADAMS Accession No. ML13071A240). This report revised the acceptance criteria associated with the NRC additional criteria ..
As stated in the BWRVIP-241 NRC SE, Section 5.0, "Conditions and Limitations," each licensee who plans to request relief from ASME Code,Section XI, requirements for RPV nozzle-to-vessel shell welds and nozzle inner radii sections may reference the BWRVIP-241 report as the technical basis for the use of ASME Code Case N-702 as an alternative. However, each licensee should demonstrate the plant-specific applicability of the BWRVIP-241 report to its units in the RR by demonstrating that the following general and nozzle-specific criteria are satisfied:
(1) The maximum RPV heatup/cooldown rate is limited to less than 115 °F/hour.
LSCS, Units 1 and 2, Technical Specifications (TS) 3.4.11, "Reactor Coolant System (RCS) Pressure and Temperature (PIT) Limits," provides a Surveillance Requirement limiting heatup and cooldown rates to s 100 °F in any 1-hour period. This heatup/cooldown rate is also described in the LSCS Updated Final Safety Analysis Report (UFSAR), Section 5.2.3.3.1. 7, "Operating Limits During Heatup, Cooldown and Core Operations."
For the recirculation inlet nozzles (N2), the following criteria must be met:
(2) (pr/t)/CRPV ~ 1.15 p = RPV normal operating pressure (psi),
r = RPV inner radius (inch),
t = RPV wall thickness (inch), and CRPV *= 19332.
The calculation for the LSCS, Units 1 and 2, N2 nozzle results in a maximum value of 1.064, which satisfies this criteria.
(3) [p(re2+ n2)/( re 2 - fj 2
)]/CNOZZLE ~ 1.47 p = RPV normal operating pressure (psi},
re =nozzle outer radius (inch),
n = nozzle inner radius (inch), and CNozzLE = 1637.
The calculation for the LSCS, Units 1 and 2, N2 nozzle results in a maximum value of 1.134, which satisfies this criteria;-
For the Recirculation Outlet Nozzles (N1 ), the following criteria must be met:
- (4) (pr/t)/ CRPv::; 1.15 p = RPV normal operating pressure (psi),
r = RPV inner radius (inch),
t = RPV wall thickness (inch), and CRPV = 16171 .
The calculation for the LSCS, Units 1 and 2, N1 nozzle results in a value of 1.025 for Unit 1 and a value of 1.272 for Unit 2. The Unit 1 results satisfy the criteria; however, the Unit 2 results are greater than 1.15.
(5) (p(ro2+ fi2)/(ro 2 - fj 2
)]/CNOZZLE ::; 1.59 p = RPV normal operating pressure (psi),
re =nozzle outer radius (inch),
n = nozzle inner radius (inch), and CNOZZLE = 1977.
The calculation for the LSCS, Units 1 and 2, N1 nozzle results in a maximum value of 1.114, which satisfies the criteria.
Based upon the above information, all LSCS RPV nozzle-to-vessel shell or head full penetration welds and nozzle inner radii sections, with the exception of the recirculation outlet nozzles on Unit 2, meet the general and nozzle-specific criteria in BWRVIP-241.
Because the Unit 2 N1 nozzles did not meet the BWRVIP-241 criteria, a bounding analysis was performed to qualify all the Units 1 and 2, nozzles. This analysis is contained in Design Analysis L-003976, "Probability of Failure Analysis for Reactor Pressure Ves~el Nozzles." The methods approved in BWRVIP-108 and BWRVIP-241 were applied, using the U2 Nt nozzle as the limiting nozzle geometry. A finite element model was developed, and the stresses caused by thermal transients and internal pressure were determined. The thermal transients evaluated were those associated with the LSCS reactor vessels, and the bounding transients were chosen based on their temperature ranges and rate of change. The resultant through-wall stresses at the locations of interest were used in a PFM calculation. The probability of failure was calculated based on operation for 60 years and assumes no inspections were performed in the initial 40 years of operation. Although the current licenses for LSCS, Units 1 and 2, expire after 40 years, the use of the 60-year duration provides additional margin in the analysis.
Since the bounding nozzle, the N1 on Unit 2, has been shown to meet the NRC safety goal of 5E-6 per year, and all other nozzles meet the plant-specific applicability criteria from the BWRVIP-241 report and are bounded by the Unit 2 N1 nozzle analysis, the application of Code Case N-702 to all the Unit 1 and Unit 2 nozzles listed in Enclosure 1 [of January 29, 2015, submission] is acceptable.
Period of application (as stated)
The third interval for LSCS, Units 1 and 2, began on October 1, 2007, and will conclude September 30, 2017. The proposed alternative will be used for the remainder of the third 10-year interval of the LSCS lnservice Inspection Program.
3.2 NRC Staff Evaluation 3.2.1 BWRVIP-108, BWRVIP-241, and NRC Requirements The NRC staff's SE for the BWRVI P-241 report specified plant-specific requirements which must be met for applicants proposing to use this alternative. Relief Request 13R-14 intended to demonstrate that the relevant LSCS, Units 1 and 2, RPV nozzle-to-vessel welds and the inner radii meet these plant-specific requirements.
As stated before, in the BWRVIP-241 NRC SE, Section 5.0, "Conditions and Limitations," each licensee who plans to request relief from ASME Code,Section X, requirements for RPV nozzle-to-vessel shell welds and nozzle inner radii sections may reference the BWRVIP-241 report as the technical basis for the use of ASME Code Case N-702 as an alternative. However, each licensee should demonstrate the plant-specific applicability of the BWRVI P-241 report to its units in the RR by demonstrating that the general and nozzle-specific criteria are satisfied as described in Section 3.1 of this SE.
The BWRVIP-241 report documents additional PFM results supporting revision of the five evaluation criteria in the BWRVIP-108 report. Since the objective of the BWRVIP-241 report is limited, i.e., revision of the limitations and conditions specified in the December 19, 2007, SE for the BWRVIP-108 report, it is considered as a supplement to the BWRVIP-108 report, not a replacement. Nonetheless, the conditions and limitations specified in the SE for the BWRVIP-241 report supersede those in the SE for the BWRVI P-108 report. Applicants requesting relief from the ASME Code,Section XI, inspection requirements on the subject RPV nozzles for their plants must demonstrate that the five plant-specific criteria are satisfied, so that BWRVIP-241 report results apply to their plants.
The December 19, 2007, NRC staff SE, established that: (1) the fracture toughness-related reference temperature (RT Nor) used in the PFM analyses were based on data from the entire fleet of BWR RPV, making the PFM analyses bounding with respect to fracture resistance and leaving the driving force of the underlying PFM analyses the only item to be evaluated, and (2) except for the RPV heatup/cooldown rate, the plant-specific criteria are for the recirculation inlet and outlet nozzles only because the probabilities of failure, P(FIE)s, for other nozzles* are an order of magnitude lower. Based on the above, the BWRVIP-241 report documents additional PFM analyses on the recirculation inlet and outlet nozzles having the highest driving force among the BWR fleet to demonstrate that the associated vessel failure probability for the normal operation is still consistent with the NRC safety goal, thus supporting the proposed revision of the five evaluation criteria. The NRC staff SE for the BWRVIP-241 report accepted the proposed revision of the five evaluation criteria in the BWRVIP-108 report.
3.2.2 NRC Staff Evaluation of the Licensee Submissions The licensee provided in its submittal dated January 29, 2015, the EGC's evaluation of the five driving force factors, or ratios, using plant-specific RPV data, and compared them against the criteria established in the BWRVIP-241 SE dated April 19, 2013. The NRC staff verified the licensee's evaluation and confirmed that Criterion 4 for the recirculation outlet nozzle (Unit 2 N1 nozzle) was not met. As a result, the licensee performed a plant-specific PFM analysis using the methodology and input variables (with generic design data replaced by plant-specific design data) approved in the BWRVIP-108 and BWRVIP-241 SEs and provided a summary of the plant-specific results in the submittal. Since a plant-specific PFM analysis was used, the NRC staff issued a request for additional information (RAI) requesting the details of the PFM analysis.
The proprietary and nonproprietary reports regarding the stress analysis and the PFM analysis were provided to the NRC in a letter dated June 8, 2015.
The NRC staff reviewed the stress analysis and the PFM analysis reports based on the plant-specific design data and found that the PFM approach and the input variables that are discussed in the reports are consistent with BWRVIP-108 and BWRVIP-241 SEs. However, the PFM report did not discuss the mean stress corrosion cracking (SCC) initiation curve, which is an important parameter to the PFM analysis. The December 19, 2007, NRC staff SE on BWRVIP-108 states, "The original BWRVIP-108 report assumed that the nozzle weld cladding is non-susceptible to SCC and applied a factor of 5 to the curve based on cast austenitic stainless steel weld data as the mean SCC initiation curve in the PFM analyses. Since this assumption was not justified, the NRC staff requested that the BWRVIP use a mean curve without the factor of 5 in this evaluation. This change increased the P(FIE) for nozzles significantly." Since the December 19, 2007, NRC staff SE on BWRVIP-108 is based on the PFM results using a mean SCC initiation curve without the factor of 5, the NRC staff issued a follow-up RAI requesting the licensee to confirm tha~ the plant-specific PFM analyses are consistent with the December 19, 2007, NRC staff SE approach regarding the SCC initiation time.
The licensee's response dated July 30, 2015, confirmed that the licensee's plant-specific PFM analysis is in accordance with the December 19, 2007, the NRC staff SE on BWRVIP-108.
Therefore, although Criterion 4 for the recirculation outlet nozzle, which is based on the generic analysis of BWRVIP-108 and BWRVIP-241, was not met, the plant-specific PFM results indicated that the LSCS, Units 1 and 2, recirculation outlet nozzles have P(FIE) below the NRC criterion of 5 x 10-5 per year. Therefore, based on the plant-specific PFM results, the NRC staff determined that the reduced inspection requirements in ASME Code Case N-702 apply to all proposed LSCS RPV nozzles (see Section 3.1 of this SE). The proposed alternative also provides an acceptable level of quality and safety because the plant-specific PFM results meet the NRC safety goal on P(FIE).
It should be noted that RPV feedwater nozzles and control rod drive return line nozzles are outside the scope of ASME Code Case N-702 and are, accordingly, outside the scope of this application.
ASME Code Case N-702 stipulates that a VT-1 examination may be used in lieu of the volumetric examination for the inner radii (i.e., Item No. 83.100, "Nozzle Inside Radius Section"). This is not consistent with the NRC position for VT-1 examination that was established in RG 1.147, Revision 15, regarding ASME Code Case N-648-1, "Alternative
Requirements for Inner Radius Examinations of Class 1 Reactor Vessel Nozzles." As indicated in RR 13R-14, "EGC may utilize Code Case N-648-1 with associated RG 1.147 conditions for the nozzles selected for examination. Volumetric examinations of the inside radius section of those reactor vessel nozzles selected for examination will be completed if Code Case N-648-1 is not applied." Therefore, the NRC staff finds that inconsistency between ASME Code Case N-702 and RG 1.147 regarding VT-1 examination is not an issue in this application and, therefore, is acceptable.
4.0 CONCLUSION
The NRC staff has reviewed the submittal regarding the licensee's evaluation of the five plant-specific criteria specified in the April 19, 2013, NRC staff SE for the BWRVIP-241 report, which provides*technical bases for use of ASME Code Case N-702, to examine RPV nozzle-to-vessel welds and nozzle inner radii at LSCS. Based on the evaluation in Section 3.2 of this SE, the NRC staff determined that the licensee's proposed alternative provides an acceptable level of quality and safety and applies to all requested LSCS RPV nozzles. However, this RR does not include feedwater nozzles and control rod drive return nozzles.
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) and is in compliance with the ASME Code's requirements.
Therefore, the NRC authorizes the licensee's proposed alternative for inspection of nozzle-to-vessel shell welds and nozzle inner radii sections of RPV nozzles listed in Section 3.1 of this SE for the third 10-year ISi interval for LSCS, Units 1 and 2.
Use of the ASME Code Case is authorized until such time as the ASME Code Case is published in a future version of RG 1.147 and incorporated by reference in 10 CFR 50.55a(b). At that time, if the licensee intends to continue implementing this ASME Code Case, it must follow all provisions of ASME Code Case N-702 with conditions as specified in RG 1.147 and limitations as specified in 10 CFR50.55a(b)(4), (b)(S), and (b)(6), if §lny."
All other ASME Code,Section XI, requirements for which relief was not specifically requested
- and approved remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector.
Principal Contributor: Simon C.F. Sheng, NRR/EVIB Date of issuance: October 30, 201 5
B. Hanson 10 CFR 50.55a(b). At that time, if the licensee intends to continue implementing this ASME Code Case, it must follow all provisions of ASME Code Case N-702 with conditions as specified in RG 1.147 and limitations as specified in 10 CFR 50.55a(b)(4}, (b)(5), and (b)(6}, if any.
All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved remain applicable, including third-party review by the.Authorized Nuclear lnservice Inspector.
Sincerely,
/RAJ Travis L. Tate, Chief Plant Licensing 111-2 and Planning and Analysis Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-373 and 50-37 4
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