ML18177A425

Relief from the Requirements of the Asme Code Regarding the Third 10-Year Inservice Inspection Interval

ML18177A425
Person / Time
Site:	South Texas
Issue date:	07/24/2018
From:	Robert Pascarelli Plant Licensing Branch IV
To:	Gerry Powell South Texas
Regner L
References
EPID L-2017-LLR-0010
Download: ML18177A425 (11)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 24, 2018 Mr. G. T. Powell Interim President and CEO/CNO STP Nuclear Operating Company South Texas Project P.O. Box 289

\Nadsworth,TX 77483

SUBJECT:

SOUTH TEXAS PROJECT, UNITS 1 AND 2 - RELIEF FROM THE REQUIREMENTS OF THE ASME CODE REGARDING THE THIRD 10-YEAR INSERVICE INSPECTION PROGRAM INTERVAL (EPID L-2018-LLR-0010)

Dear Mr. Powell:

By letter dated February 15, 2018, STP Nuclear Operating Company (STPNOC, the licensee) requested relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, to extend the inservice inspection (ISi) interval for Category B-A and B-D examinations for South Texas Project (STP), Units 1 and 2.

Specifically, pursuant to Title 10 of the Code of Federal Regulations ( 10 CFR) paragraph 50.55a(z)(1 ), the licensee requested to extend the interval for ASME Code Table I\NB-2500-1, Category B-A and B-D components examinations. The licensee requested to extend the inspection interval from September 25, 2020, to August 20, 2027, for Unit 1 and from October 19, 2020, to December 15, 2028, for Unit 2, on the basis that the proposed alternative provides an acceptable level of quality and safety.

The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation that STPNOC has demonstrated that the proposed alternative 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 staff authorizes the use of Relief Request RR-ENG-3-14 at STP Units 1 and 2, for the extended third ISi interval for ASME Categories B-A and B-D items until August 20, 2027, for Unit 1 and December 15, 2028, for Unit 2.

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

If you have any questions, please contact the Project Manager, Lisa Regner, at 301-415-1906 or via e-mail at Lisa.Regner@nrc.gov.

Sincerely, Robert J. Pascarelli, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-498 and 50-499

Enclosure:

Safety Evaluation cc: Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST RR-ENG-3-14 REGARDING THIRD 10-YEAR INSERVICE INSPECTION PROGRAM INTERVAL STP NUCLEAR OPERATING COMPANY SOUTH TEXAS PROJECT, UNITS 1 AND 2 DOCKET NOS. 50-498 AND 50-499

1.0 INTRODUCTION

By letter dated February 15, 2018 (Agencywide Document Access and Management System (ADAMS) Accession No. ML18047A039), STP Nuclear Operating Company (STPNOC or the licensee) requested relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, to extend the inservice inspection (ISi) interval for Category B-A and B-D examinations for South Texas Project (STP),

Units 1 and 2.

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) paragraph 50.55a(z)(1 ), the licensee requested to extend the interval for ASME Code Table IWB-2500-1, Category B-A and B-D components examinations from September 25, 2020, to August 20, 2027, for Unit 1, and from October 19, 2020, to December 15, 2028, for Unit 2, on the basis that the proposed alternative provides an acceptable level of quality and safety.

2.0 REGULATORY EVALUATION

Adherence to Section XI of the ASME Code is mandated by 10 CFR 50.55a(g)(4), "lnservice inspection standards requirements for operating plants," which states, in part, that ASME Code Class 1, 2, and 3 components will meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in Section XI of the ASME Code.

Section 50.55a(z) of 10 CFR, "Alternatives to codes and standards," 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 [this authorization has been delegated to the management of the Division of Operating Reactor Licensing].... A proposed alternative must be submitted and authorized prior to implementation. The applicant or licensee must demonstrate that:

Enclosure

( 1) Acceptable level of quality and safety. The proposed alternative would provide an acceptable level of quality and safety; or (2) Hardship without a compensating increase in quality and safety. 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 Based on the above, and subject to the following technical evaluation, the U.S. Nuclear Regulatory Commission (NRC) staff finds that regulatory authority exists for the licensee to request the use of an alternative and the NRC to authorize the proposed alternative.

3.0 TECHNICAL EVALUATION

3.1 Background The NRC staff's review of this proposed alternative assesses the consistency of the licensee's proposal with WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," dated October 2011 (ADAMS Accession No. ML11306A084).

Henceforth, WCAP-16168-NP-A, Revision 3, will be referred to as WCAP-A. WCAP-A provides a basis for the acceptability of the proposed inspection intervals for Category B-A and B-D components at U.S. pressurized-water reactors (PWRs) designed by Westinghouse, Combustion Engineering and Babcock and Wilcox (B&W) through the use of risk-informed analyses and probabilistic fracture mechanics for a pilot plant of each design. WCAP-A also contains the NRC staff's safety evaluation (SE) of the Westinghouse proposal. In its SE, the NRC staff found the proposal acceptable for use based on its consistency with the principles contained in Regulatory Guide (RG) 1.174, Revision 1, "An Approach For Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,"

dated November 2002 (ADAMS Accession No. ML023240437). However, the SE imposes a condition which states that licensees should provide plant-specific information in six areas to demonstrate the applicability of WCAP-A to the licensee's plant. The plant-specific information requested by the condition is as follows:

( 1) Licensees should provide the 95th percentile total through-wall cracking frequency (TWCFrnTAL) and the supporting material properties at the end of the proposed 20-year ISi interval. The 95th percentile TWCFrnTAL should be calculated using the methodology in NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS)" (ADAMS Accession No. ML070860156}, which is frequently referred as "the NRC PTS Risk Study."

The RTMAx-x (the material property which characterizes the reactor vessel's resistance to fracture initiating from flaws found along a specific RPV material) and the temperature shift in the Charpy transition temperature produced by irradiation defined at the 30 foot-pound energy level (tiT30), should be calculated using the latest revision of RG 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials" (ADAMS Accession No. ML003740284) or other NRC-approved methodology.

(2) Licensees should report whether the frequency of the limiting design basis transients during prior plant operation are less than the frequency of the design basis transients identified in the PWR Owners Group (PWROG) fatigue analysis as significant contributors to fatigue crack growth.

(3) Licensees should report the results of prior ISi of reactor pressure vessel (RPV) welds and the proposed schedule for the next 20-year ISi interval. Each licensee should identify the years in which future inspections will be performed, and the dates provided should be within plus or minus one refueling cycle of the dates identified in the implementation plan provided to the NRC in a letter by the PWROG dated July 12, 2010 (PWROG, Letter OG-10-238) (ADAMS Accession No. ML11153A033).

(4) Licensees with B&W plants should (a) verify that the fatigue crack growth of 12 heatup and cooldown transients per year used in the PWROG fatigue analysis bounds the fatigue crack growth for all of its design basis transients and (b) identify the design bases transients that contribute to significant fatigue crack growth.

(5) Licensees with RPV forgings that are susceptible to underclad cracking and with RTMAX Fo values exceeding 240 degrees Fahrenheit (°F) should submit a plant-specific evaluation because the analyses performed in WCAP-A are not applicable.

(6) Licensees seeking second or additional interval extensions should provide the information and analyses requested in Section 50.61a(e) of 10 CFR, "Examination and Flaw Assessment Requirements."

3.2 ASME Code Components Affected

The affected components are the RPV welds and full penetration nozzle welds. The following examination categories and item numbers from IWB-2500 and Table IWB-2500-1 of the ASME Code,Section XI, are listed in Relief Request RR-ENG-3-14:

Exam Category Item Number Description B-A B1 .11 Circumferential Shell Welds B-A B1.12 Longitudinal Shell Welds B-A B1.21 Circumferential Head Welds B-A B1.22 Meridional Head Welds B-A B1.30 Shell-to-Flange Weld B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inner Radius Section 3.3 Applicable Code Edition and Addenda For the third 10-year ISi intervals at STP Units 1 and 2, the Code of record for the inspection of ASME Code Class 1, 2, and 3 components is the 2004 Edition of the ASME Code,Section XI.

3.4 Applicable Code Requirements ASME Code,Section XI, paragraph IWB-2412, "Inspection Program B," requires volumetric examination of essentially 100 percent of the total number of RPV pressure-retaining welds identified in Table IWB-2500-1, once each 10-year interval.

3.5 Licensee's Proposed Alternative In Relief Request RR-ENG-3-14, the licensee proposed to extend the third ISi interval for the ASME Code Category B-A and B-D examination items from September 25, 2020, to August 20, 2027, for STP Unit 1 and October 19, 2020, to December 15, 2028, for STP Unit 2.

The licensee plans to perform the ASME Code required examination of the subject items no later than 2026 and 2027, which is not consistent with the schedule proposed in the PWROG Letter OG-10-238 dated July 12, 2010.

3.6 Licensee's Basis for Alternative The licensee stated that the alternative is based on a negligible change in risk, satisfying the risk criteria specified in RG 1.174, Revision 1. The licensee further stated that the methodology used to conduct this analysis is based on the study defined in WCAP-A. This study focuses on risk assessments of materials within the beltline region of the RPV wall. Appendix A of the WCAP-A identifies the parameters to be compared between an applicant's plant and the appropriate pilot plant. These items include:

• Dominant PTS Transients in the NRC PTS Risk Study

• Through Wall Cracking Frequency (TWCF)

• Frequency and Severity of Design Basis Transients

• Cladding Layers (Single/Multiple)

Table 1a "Critical Parameters for the Application of Bounding Analysis for STP Unit 1" of Relief Request RR-ENG-3-14 provides the above parameters for STP Unit 1 and the Westinghouse pilot plant. Table 1b, "Critical Parameters for the Application of Bounding Analysis for STP Unit 2," of Relief Request RR-ENG-3-14, provides similar information for STP Unit 2. Based on this information, the licensee concludes that the parameters for STP Units 1 and 2, are bound by the results of the Westinghouse pilot plant and, thus, qualify for ISi interval extensions.

For the most important parameter (TWCF), the licensee's calculated values are 8.27E-16 events per year for STP Unit 1 and 1.09E-16 for STP Unit 2, as compared to the WCAP-A TWCF of 1. 76E-08 events per year for the Westinghouse pilot plant.

Tables 2a and 2b of Relief Request RR-ENG-3-14 also contain inspection showing that RPV examinations have been performed with satisfactory results.

3. 7 Duration of Alternative The relief request, if granted, would extend the third 10-year ISi interval to August 20, 2027, for STP Unit 1 and December 15, 2028, for STP Unit 2, for ASME Categories B-A and B-D items listed in Section 3.1 of this SE.

3.8 NRC Staff Evaluation Since the WCAP-A methodology has already been accepted by the NRC staff, the current evaluation focused on the manner in which the licensee addresses the four critical parameters in Table A-1 of WCAP-A, Appendix A (the four critical parameters are provided in SE Section 3.6), and the six plant-specific information items specified in the NRC SE enclosed in WCAP-A (six plant-specific information items provided in SE Section 3.1 ). The NRC staff

reviewed the licensee's evaluation of the four critical parameters and the six plant-specific information items.

Regarding PTS transients, the licensee identified the NRC letter report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants" (ADAMS Accession No. ML042880482) as its plant-specific basis. This is acceptable because the SE in WCAP-A concludes that based on this letter report, the PTS transient characteristics are generally applicable for plants from the same reactor vendor. Regarding the cladding layers, the licensee states in its submittal that STP are "single layer" units. This is also acceptable because it is consistent with the Westinghouse pilot plant.

The remaining two critical parameters are among the six plant-specific information items discussed below.

3.8.1 Plant-Specific Information Item (1)

Plant-specific Information Item 1 addresses TWCFs. Tables 3a and 3b of the licensee's submittal pertain to this item. As specified in the guidance in Appendix A of WCAP-A, the licensee provided Tables 3a and 3b in its submittal, which contains a summary of the input parameters for all RPV materials and the resulting TWCFs for the controlling materials. The licensee proposed that the information specified in Tables 3a and 3b demonstrates that STP Units 1 and 2 are bounded by WCAP-A, and are therefore, acceptable. Specifically, Tables 3a and 3b of Relief Request RR-ENG-3-14 provide input chemistry data, unirradiated nil-ductility transition reference temperature (RTNor), and neutron fluence values for all RPV materials, and output shift and TWCF for controlling RPV materials of the two units.

The NRC staff compared Tables 3a and 3b information with that in the license renewal application (LRA) for STP Units 1 and 2 because these values were accepted in the SE dated June 2017 for the LRA, "Safety Evaluation Report Related to the License Renewal of South Texas Project, Units 1 and 2" (ADAMS Accession No. ML171466224) and are considered as the current licensing basis values. The NRC staff found three categories of discrepancies:

(1) fluence values in Tables 3a and 3b are lower than the LRA values, (2) the nickel values for all reported welds for STP Unit 2 are slightly higher than the LRA values, and (3) the chemistry factor for the intermediate plate R1606-2 for Unit 1 and the chemistry factor for the intermediate plate R2507-1 for Unit 2 are slightly higher than the LRA values. The discrepancies are evaluated below.

Lower Fluence Values The NRC staff noted that the LRA fluence values are for 54 effective full power years (EFPY) and the relief request fluence values are for 34 EFPY. The NRC staff finds this difference in reported EFPY to be acceptable, because the value provided in the LRA is required to predict the EFPY incurred by the end of the period of extended operation while the value used in this request needs only predict EFPY to the end of the request. Using linear interpolation, the NRC staff found that the highest LRA fluence value of 3.86E+19 neutrons per square centimeter (n/cm 2 ) for 54 EFPY for Unit 1 will become 2.43E+19 n/cm 2 for 34 EFPY, which is close to the value of 2.51 E+19 n/cm 2 reported in Relief Request RR-ENG-3-14 for 34 EFPY for all RPV materials of STP Unit 1. Similar findings were also observed in STP Unit 2. This small difference indicated that the fluence methodology, which was used to generate the fluence values for STP Units 1 and 2 in the relief request, is consistent with the LRA and, therefore, the lower fluence values reflecting lower EFPY reported in the relief request are acceptable.

Higher Nickel Values for all Reported Welds for STP Unit 2 Relief Request RR-ENG-3-14 reports higher nickel values for all reported welds for STP Unit 2 than the LRA. The NRC staff accepts these slightly higher nickel values for STP Unit 2 in this application because (1) they are more conservative than the LRA values, and (2) the welds will not become limiting material due to the much lower unirradiated RTNDT value and will not affect the resulting TWCF.

Slightly Higher Chemistry Factor for Plate R1606-2 (Unit 1) and Plate R2507-1 (Unit 2)

Relief Request RR-EN-3-14 reports a slightly higher chemistry factor for the intermediate plate R1606-2 for STP Unit 1 and the intermediate plate R2507-1 for STP Unit 2. The NRC staff found that the revised chemistry factors for these two plates are results of considering additional surveillance data from Capsule W for Unit 1 and another Capsule W for Unit 2. Complete surveillance data information for these two capsules are documented in WCAP-17482-NP, "Analysis of Capsule W from the South Texas Project Nuclear Operating Company South Texas Unit 1 Reactor Vessel Radiation Surveillance Program" dated May 2012 (ADAMS Accession No. ML12173A199) and WCAP-17636-NP, Revision 0, "Analysis of Capsule W from the South Texas Project Nuclear Operating Company South Texas Unit 2 Reactor Vessel Radiation Surveillance Program," dated October 2012 (ADAMS Accession No. ML13053A214). The NRC staff accepts the revised chemistry factors for these two plates because they are calculated in accordance with RG 1.99, Revision 2.

The parts of Tables 3a and 3b titled "Outputs" show that the calculated total TWCF is 8.27E-16 event per year for STP Unit 1 and 1.09E-16 event per year for STP Unit 2. These TWCF values were obtained by the licensee using the WCAP-A methodology with inputs from the parts of Tables 3a and 3b titled "Inputs." Tables 3a and 3b used RG 1.99, Revision 2, Position 1.1 (without surveillance data) or Position 2.1 (with surveillance data) to calculate RTMAX

(~T 30 + unirradiated RTNor + 460 °F) for 34 EFPY for all RPV beltline materials. Using Tables 3a and 3b input values, the NRC staff verified the licensee's calculated ~T30 values, RTMAX values, and the resulting TWCFs. The NRC staff determined that these TWCFs support Relief Request RR-ENG-3-14, because they are several orders of magnitude lower than the value of 1. 76E-08 for the Westinghouse pilot plant in the WCAP-A. Hence, the NRC staff determines that the embrittlement of the STP Units 1 and 2 RPVs is within the envelope used in the Westinghouse pilot plant analysis and thus, the NRC staff concludes that the licensee has acceptably addressed Plant-Specific Information Item (1 ).

3.8.2 Plant-Specific Information Item (2)

The NRC staff reviewed Plant-Specific Information Item (2) regarding the frequency of the limiting design basis transients. The licensee stated in Tables 1a and 1b of its submittal that the heatup and cooldown cycles per year are bounded by the heatup and cooldown cycles for the Westinghouse pilot plant. The NRC staff examined the heatup and cooldown design and projected cycles for 60 years of operation in Table 4.3-2 of the LRA. The NRC staff found that both frequencies (heatup and cooldown) are below the bounding value of 7 per year for the Westinghouse pilot plant. Therefore, the NRC staff concludes that the licensee has acceptably addressed Plant-Specific Information Item (2).

3.8.3 Plant-Specific Information Item (3)

The NRC staff reviewed Plant-Specific Information Item (3) regarding the results of prior ISi of RPV welds and the proposed schedule for the extended ISi interval. Tables 2a and 2b in the licensee's submittal provide additional information pertaining to previous RPV inspections and the schedule for the future inspection. Specifically, Tables 2a and 2b indicated that two 10-year ISls have been performed for each unit. One subsurface indication was identified in the intermediate to lower shell circumferential weld for STP Unit 1. Additionally, one indication was identified in the lower longitudinal weld for STP Unit 2 during the most recent ISi. Both indications were accepted per Table IWB-3510-1 of Section XI of the ASME Code. Both were also found acceptable with respect to the 10 CFR 50.61a requirement that prohibits flaws within the inner 1110th or 1 inch of the RPV wall thickness.

Because both ASME Code,Section XI and 10 CFR 50.61a requirements regarding detected flaws are met, the NRC staff determines that the licensee has adequately addressed the first part of Plant-Specific Information Item (3).

The licensee proposed to conduct the next RPV inspection in 2026 for STP Unit 1 and 2027 for STP Unit 2. Both dates represent a deviation from the RPV inspection proposed in the PWROG Letter OG-10-238 dated July 12, 2010. A change of the inspection date for Unit 1 would increase the number of industry inspections in 2026 from two to three and decrease the number of industry inspections in 2029 from five to four, making the number of inspections per year industrywide more uniform and favorable. The proposed change in the inspection date for Unit 2 would increase the number of inspections in 2027 from seven to eight and decrease the number of inspections in 2030 from five to four, making an unfavorable 14 percent increase in the industrywide inspections in 2027. However, the NRC staff determined that the deviation has an insignificant impact on the implementation plan in the PWROG Letter OG-10-238. Thus, the NRC staff concludes that the licensee has adequately addressed the second part of Plant-Specific Information Item (3). In summary, the licensee has adequately addressed Plant-Specific Information Item (3).

3.8.4 Plant Specific Information Items 4 through 6 The licensee did not address Plant-Specific Information Items (4), (5), and (6). The NRC staff examined the specifics in each of these Plant-Specific Information Items and confirmed that these information requirements are not applicable to STP, Units 1 and 2. Thus, the NRC staff concludes that the Plant-Specific Information Items (4), (5), and (6) do not apply to STP.

3.9 Summary The NRC staff has reviewed the licensee's submittal and performed independent calculations to verify the input data and output results in Tables 3a and 3b of the submittal. The difference between the licensee's and NRC staff's calculated TWCF95-TOTAL is insignificant. With this information, the NRC staff determined that the proposed alternative is based on the WCAP-A methodology and the TWCF95-TOTAL values in Tables 3a and 3b of the submittal are bounded by the corresponding pilot plant parameter in the WCAP-A. Consequently, the licensee has demonstrated that the proposed alternative will provide an acceptable level of quality and safety.

4.0 CONCLUSION

As set forth above, the NRC staff determines that the licensee has demonstrated that the proposed alternative 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 staff authorizes the use of RR-ENG-3-14 at STP Units 1 and 2 for the extended third ISi interval for ASME Categories B-A and B-D items until August 20, 2027, for Unit 1 and December 15, 2028, for Unit 2.

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

Principal Contributor: S. Sheng Date: July 24, 2018

G. Powell

SUBJECT:

SOUTH TEXAS PROJECT, UNITS 1 AND 2 - RELIEF FROM THE REQUIREMENTS OF THE ASME CODE REGARDING THE THIRD 10-YEAR INSERVICE INSPECTION PROGRAM INTERVAL (EPID L-2018-LLR-0010)

DATED JULY 24, 2018 DISTRIBUTION:

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