ML15035A148

Relief Request 13R-17, Proposed Alternative to ASME Code,Section XI Requirements, Which Extends Reactor Vessel Inservice Inspection Frequency from 10 to 20 Years, Third 10-Year Interval

ML15035A148
Person / Time
Site:	Callaway
Issue date:	02/10/2015
From:	Eric Oesterle Plant Licensing Branch IV
To:	Diya F Union Electric Co
Lyon C
References
TAC MF3876
Download: ML15035A148 (11)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 10, 2015 Mr. Fadi Diya Senior Vice President and Chief Nuclear Officer Union Electric Company P.O. Box 620 Fulton, MO 65251

SUBJECT:

CALLAWAY PLANT, UNIT 1- REQUEST FOR RELIEF 13R-17, ALTERNATIVE TO ASME CODE REQUIREMENTS WHICH EXTENDS THE REACTOR VESSEL INSPECTION INTERVAL FROM 10 TO 20 YEARS (TAC NO. MF3876)

Dear Mr. Diya:

By letter dated April 8, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14098A428), as supplemented by letter dated August 20, 2014 (ADAMS Accession No. ML14232A776), Union Electric Company (dba Ameren Missouri, the licensee) submitted request for relief 13R-17 for Callaway Plant, Unit 1 (Callaway Plant), to the U.S. Nuclear Regulatory Commission (NRC) proposing an alternative to certain requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI, for the third and fourth 10-year inservice inspection (lSI) program intervals.

Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) Part 50, paragraph 50.55a(a)(3)(i), the licensee requested to implement an alternative requirement to the requirement of IWB-2412, Inspection Program B, that volumetric examination of essentially 100 percent of reactor vessel pressure-retaining Examination Category B-A and B-D welds be performed once each 10-year interval. In lieu of this requirement, the licensee proposed not to perform the ASME Code required volumetric examination of the Callaway Plant reactor vessel full penetration pressure-retaining Examination Category B-A and B-D welds for the third lSI, currently scheduled for 2014, and instead proposed to perform the third ASME Code required volumetric examination of the Callaway Plant reactor vessel full penetration pressure-retaining Examination Category B-A and B-D welds in the fourth lSI interval in 2023 plus or minus one refueling outage.

The paragraph headings in 10 CFR 50.55a were changed by Federal Register notice dated November 5, 2014 (79 FR 65776), which became effective on December 5, 2014 (e.g., 10 CFR 50.55a(a)(3)(i) is now 50.55a(z)(1), and 50.55a(a)(3)(ii) is now 50.55a(z)(2)). See the cross reference tables, which are cited in the notice, at ADAMS Accession No. ML14015A191 and ADAMS package Accession No. ML14211A050.

Based on the enclosed safety evaluation, the NRC staff determines 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 13R-17 at Callaway

F. Diya Plant for the extended third 10-year lSI interval, which ends on December 18, 2024, for the subject components.

All other ASME Code requirements for which relief was not specifically requested and approved remain applicable, including the third-party review by the Authorized Nuclear lnservice Inspector. If you have any questions, please contact Fred Lyon at 301-415-2296 or via e-mail at fred.lyon@nrc.gov.

Sincerely, Eric R. Oesterle, Acting Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-483

Enclosure:

Safety Evaluation 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 REQUEST FOR RELIEF 13R-17 RELATED TO THE INSERVICE INSPECTION PROGRAM FOR THE REACTOR VESSEL UNION ELECTRIC COMPANY CALLAWAY PLANT. UNIT 1 DOCKET NO. 50-483

1.0 INTRODUCTION

By letter dated April 8, 2014, as supplemented by letter dated August 20, 2014, in response to the NRC staff's request for additional information (RAI) (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML14098A428 and ML14232A776, respectively), Union Electric Company (dba Ameren Missouri, the licensee) proposed an extension of the inservice inspection (lSI) interval requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, lSI Program for Callaway Plant, Unit 1 (Callaway Plant). Enclosure 1 of the submittal contains Relief Request 13R-17, requesting to use an alternative to the requirements of the ASME Code,Section XI, Table IWB-2500-1.

Specifically, Relief for Request 13R-17 proposes an alternative pursuant to paragraph 50.55a(a)(3)(i) of Title 10 of the Code of Federal Regulations (1 0 CFR) to extend the lSI interval for examinations of the reactor pressure vessel (RPV) welds (Category B-A) as well as the nozzle-to-vessel welds and inner radius sections (Category B-D) from 10 to 20 years.

The paragraph headings in 10 CFR 50.55a were changed by Federal Register notice dated November 5, 2014 (79 FR 65776), which became effective on December 5, 2014 (e.g., 10 CFR 50.55a(a)(3)(i) is now 50.55a(z)(1), and 50.55a(a)(3)(ii) is now 50.55a(z)(2)). See the cross-reference tables, which are cited in the notice, at ADAMS Accession No. ML14015A191 and ADAMS package Accession No. ML14211A050.

The current third 10-year interval is scheduled to end on December 18, 2014, for Callaway Plant, but, for the subject examinations of this request, it has been extended to December 18, 2015, using the provisions of IWA-2430(d) in order to allow adequate evaluation of this request.

Enclosure

2.0 REGULATORY EVALUATION

2.1 Regulations and Guidance In accordance with 10 CFR 50.55a(g)(4), the licensee is required to perform lSI of ASME Code Class 1, 2, and 3 components and system pressure tests during the first 10-year interval and subsequent 10-year intervals that comply with the requirements in the latest edition and addenda of Section XI of the ASME Code, incorporated by reference in 10 CFR 50.55a(a)(1 ),

subject to the limitations and modifications listed therein.

For the third 10-year lSI interval at Callaway Plant, the Code of record for the inspection of ASME Code Class 1, 2, and 3 components is the 1998 Edition through the 2000 Addenda of the ASME Code,Section XI. The regulation in 10 CFR 50.55a(z) states, in part, that the Director of the Office of Nuclear Reactor Regulation may authorize an alternative to the requirements of 10 CFR 50.55a(b)-(h). There are two justifications for an alternative to be authorized. First, per 10 CFR 50.55a(z)(1 ), the licensee must demonstrate that the proposed alternative would provide an acceptable level of quality and safety. For the second possible justification for an alternative to be authorized, described in 10 CFR 50.55a(z)(2),

the licensee must show that following the ASME Code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Regulatory Guide (RG) 1.99, Revision (Rev.) 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988 (ADAMS Accession No. ML003740284), describes general procedures acceptable to the staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled RPVs.

RG 1.174, Rev. 1, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," November 2002 (ADAMS Accession No. ML023240437), describes a risk-informed approach, acceptable to the NRC, for assessing the nature and impact of proposed licensing basis changes by considering engineering issues and applying risk insights.

RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001 (ADAMS Accession No. ML010890301 ), describes methods and assumptions acceptable to the staff for determining the RPV neutron fluence.

2.2 Background The lSI of Categories B-A and B-D components consists of visual and ultrasonic examinations intended to discover whether flaws have initiated, whether pre-existing flaws have extended, and whether pre-existing flaws may have been missed in prior examinations. These examinations are required to be performed at regular intervals, as defined in Section XI of the ASME Code.

2.3 Summary ofWCAP-16168-NP-A. Rev. 2 In June 2008, the Pressurized Water Reactor Owners Group (PWROG) issued the NRC-approved topical report WCAP-16168-NP-A, Rev. 2, "Risk-Informed Extension of the

Reactor Vessel In-Service Inspection Interval," June 2008 (ADAMS Accession No. ML082820046), which is in support of a risk-informed assessment of extensions to the lSI intervals for Categories B-A and B-D components. Specifically, WCAP-16168-NP-A, Rev. 2 took data associated with three different pressurized-water reactor (PWR) plants (referred to as the pilot plants), one designed by each of the three main vendors (Westinghouse, Combustion Engineering, and Babcock and Wilcox (B&W)) for PWR nuclear power plants in the United States, and performed studies on these pilot plants to justify the proposed extension of the lSI interval for Categories B-A and B-D components from 10 to 20 years.

The analyses in WCAP-16168-NP-A, Rev. 2 used probabilistic fracture mechanics (PFM) tools and inputs from the work described in NUREG-1806, "Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61): Summary Report" (ADAMS Accession No. ML061580318) and NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS)" (ADAMS Accession No. ML070860156). The PWROG analyses incorporated the effects of fatigue crack growth and lSI. Design basis transient data was used as input to the fatigue crack growth evaluation. The effects of lSI were modeled consistent with a previously-approved PFM Code in WCAP-14572-NP-A, "Westinghouse Owners Group Application of Risk-Informed Methods to Piping lnservice Inspection" (ADAMS Accession Nos. ML012630327, ML012630349, and ML012630313).

These effects were considered in the PFM evaluations, using the Fracture Analysis of Vessels- Oak Ridge (FAVOR) computer code (ADAMS Accession No. ML042960391). All other inputs were identical to those used in the PTS risk re-evaluation underlying 10 CFR 50.61a.

From the results of the studies, the PWROG concluded that the ASME Code,Section XI 10-year inspection interval for Categories B-A and B-D components in PWR RPVs can be extended to 20 years. Their conclusion from the results for the pilot plants was considered to apply to any plant designed by the three vendors as long as the critical, plant-specific parameters (defined in Appendix A of WCAP-16168-NP-A, Rev. 2) are bounded by the pilot plants.

2.4 Summary of the July 26. 2011, NRC Safety Evaluation (SE) for WCAP-16168-NP-A. Rev. 2 The NRC original SE in WCAP-16168-NP-A, Rev. 2, published in 2008, was superseded by the SE dated July 26, 2011 (ADAMS Accession No. ML111600303), to address the PWROG's request for clarification of the information needed in applications utilizing WCAP-16168-NP-A, Rev. 2. The staff's conclusion in the July 26, 2011, SE indicates that the methodology presented in WCAP-16168-NP-A, Rev. 2 is consistent with RG 1.174, Rev. 1 and is acceptable for referencing in requests to implement alternatives to ASME Code inspection requirements for PWR plants in accordance with the limitations and conditions in the SE. As stated in the SE dated July 26, 2011, in addition to showing that the subject plant parameters and inspection history are bounded by the critical parameters identified in Appendix A in WCAP-16168-NP-A, Rev. 2, the licensee's application must provide the following plant-specific information:

1) Licensees must demonstrate that the embrittlement of their [RPV] is within the envelope used in the supporting analyses. Licensees must provide the 951h percentile [total through-wall cracking frequency

(TWCFToTAL)] and its supporting material properties at the end of the period in which the relief is requested to extend the [lSI] from 10 to 20 years. The 95th percentile TWCFTOTAL must be calculated using the methodology in NUREG-1874 .. The RTMAx-x and the shift in the Charpy transition temperature produced by irradiation defined at the 30 ft-lb energy level, flT 30 , must be calculated using the methodology documented in the latest revision of RG 1.99 or other NRC-approved methodology.

2) Licensees must 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 PWROG fatigue analysis that are considered to significantly contribute to fatigue crack growth.

3) Licensees must report the results of prior lSI of RPV welds and the proposed schedule for the next 20-year lSI interval. The 20-year inspection interval is a maximum interval. In its request for an alternative, each licensee shall identify the years in which future inspections will be performed. The dates provided must be within plus or minus one refueling cycle of the dates identified in the implementation plan provided to the NRC in PWROG letter OG-10-238 [(ADAMS Accession No. ML11153A033)].

4) Licensees with 8&W plants must (a) verify that the fatigue crack growth of 12 heat-up/cool-down transients per year that was used in the PWROG fatigue analysis bound 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 RPVs having forgings that are susceptible to underclad cracking and with RT MAx-Fo values exceeding 240 oF must submit a plant-specific evaluation to extend the inspection interval for ASME Code,Section XI, Category 8-A and 8-D RPV welds from 10 to a maximum of 20 years because the analyses performed in the WCAP-A are not applicable.

6) Licensees seeking second or additional interval extensions shall provide the information and analyses requested in Section (e) of 10 CFR 50.61 a.

WCAP-16168-NP-A, Rev. 3, which contains this latter SE for WCAP-16168-NP-A, Rev. 2, was issued in October 2011 (ADAMS Accession No. ML11306A084, referred to as the WCAP-A in the rest of this SE).

3.0 PROPOSED ALTERNATES FOR CALLAWAY PLANT

3. 1 Description of Proposed Alternatives In 13R-17, the licensee proposed to defer the ASME Code required Categories 8-A and 8-D weld lSI for Callaway Plant until 2023. This schedule is consistent with the schedule proposed in the revision to PWROG Letter OG-10-238.

3.2 Components for Which Relief is Requested The affected components are the subject plant RPV and their interior attachments and core support structures. The following examination categories and item numbers from IWB-2500 and Table IWB-2500-1 of the ASME Code,Section XI, are addressed in this request:

For Relief Request 13R-17:

Exam Category Item Number Description 8-A 81.11 Circumferential Shell Welds 8-A 81.12 Longitudinal Shell Welds 8-A 81.21 Circumferential Head Welds 8-A 81.22 Meridional Shell Welds 8-A 81.30 Shell-to-Flange Weld 8-A 81.40 Head-to-Flange Weld 8-A 81.50 Repair Welds 8-A 81.51 Beltline Region 8-D 83.90 Nozzle-to-Vessel Welds 8-D 83.100 Nozzle Inside Radius Section 3.3 Basis for Proposed Alternatives The licensee stated that the methodology used to demonstrate the acceptability of extending the inspection intervals for Examination Category 8-A and 8-D components is contained in the WCAP-A. This methodology used the estimated TWCF as a measure of the risk of RPV failure, and it was demonstrated that the inspection interval for the affected components can be extended from 10 to 20 years, meeting the change in risk guidelines in RG 1.174. The licensee addressed the plant-specific information discussed in Section 2.4 of this SE as follows:

(1) A plant-specific analysis, with identified critical parameters and detailed TWCF calculation demonstrated that the Callaway Plant RPV's parameter is bounded by corresponding pilot plant parameters. The total TWCFs were calculated as 3.98E-14 for Callaway Plant, which is much less than the value of 1.76E-08 for the Westinghouse pilot plant in the WCAP-A.

(2) The frequencies of the Callaway Plant RPV's limiting design basis transients are bounded by the frequencies identified in the PWROG fatigue analysis.

(3) The results of the previous RPV inspection for the Callaway Plant are provided, which confirm that satisfactory examinations have been performed on the

Callaway Plant RPV. The RPV examination currently scheduled for 2015 will be deferred until 2023, which matches the date identified in PWROG letter OG-10-238, dated July 12, 2010.

Plant-specific information items (4), (5), and (6) have not been addressed by the licensee because they do not apply to Callaway Plant. Since Callaway Plant is bounded by the pilot plant application, the licensee concluded that use of this proposed alternative will provide an acceptable level of quality and safety and requested, pursuant to 10 CFR 50.55a(z)(1 ), that would justify the NRC staff authorizing the alternative.

3.4 Duration of Proposed Alternatives The licensee's letter dated August 20, 2014, revised the duration of proposed alternatives to the third lSI interval with the next ASME Categories B-A and B-D RPV weld inspections scheduled for 2023.

4.0 STAFF TECHNICAL EVALUATION In the relief request 13R-17, the licensee stated, in part, that Ameren Missouri proposes not to perform the ASME Code required volumetric examination of the Callaway Plant reactor vessel full penetration pressure-retaining Examination Category B-A and B-D welds for the third inservice inspection, currently scheduled for 2014, and instead proposes to perform the third ASME Code required volumetric examination of the Callaway Plant reactor vessel full penetration pressure-retaining Examination Category B-A and B-D welds in the fourth inservice inspection interval in 2023 plus or minus one refueling outage.

It was unclear to the NRC staff that the examination would be performed by the end of an additional10 years from scheduled end of third lSI interval (December 18, 2014). In addition, the lSI interval for which the exemption is requested should be "the third lSI interval," which is extended from 10 years to 20 years (December 18, 2024). By letter dated July 25, 2014 (ADAMS Accession No. ML14203A063), the staff requested that the licensee confirm that the next ASME Code-required volumetric examination of category B-A and B-D welds must be performed by the end of the extended third lSI interval (December 18, 2024). The licensee's RAI response dated August 20, 2014, stated that the examinations will be performed prior to December 18, 2024, and is satisfactory.

For Plant-Specific Information 1 (see SE Section 2.4), the NRC staff confirmed that the "Frequency and Severity of Design Basis Transients" of Callaway Plant is below the number in the conditions and limitations described in the WCAP-A in Table 1 of the relief request. Also, the Callaway Plant RPV has a single-layer cladding on the inside like the assumption used in the WCAP-A analysis. Considering these critical plant-specific parameters, the licensee performed TWCF calculation using the WCAP-A methodology in its address of Plant-Specific Information 1.

This TWCF calculation used inputs from Table 3 of the relief request. The relief request used RG 1.99, Rev. 2, Position 1.1 and Position 2.1 when credible surveillance data was available to calculate ~T 3o for all RPV beltline materials for the Callaway Plant. The NRC staff noted that there is a different value of material property between this relief request and, in the recent license renewal application from Callaway Plant (ADAMS Accession No. ML113530372). The licensee's August 20, 2014, response to RAI-3 explained that the subject weld was fabricated with three different weld metal heats. The relief request lists details for each heat, while the license renewal application shows a composite of the details, keeping only the most limiting value for each parameter from among the heats as explained in Table 3.5 ofWCAP-17168-NP, Revision 0. The licensee's response clarified the information satisfactorily.

Table 2 in the relief request contains additional information pertaining to previous RPV inspection and the schedule for future ones. Table 2 was provided to address Plant-Specific Information 3. Specifically, Table 2 reveals that thirteen indications were identified in the most recent lSI of the Callaway Plant RPV beltline region. The licensee concluded that the indications for both units are acceptable per Table IWB-351 0-1 of Section XI of the ASME Code and the number of flaws is acceptable per the requirements of the Alternate PTS Rule, 10 CFR 50.61 a. The allowable number of flaws per the requirements depends on the area of plate actually inspected. The licensee's August 20, 2014, response to RAI-2 indicated that the total length of welds inspected was 4950 inches and the area of plate actually inspected was 46,600 square inches. With these inputs, the NRC staff verified that the number of flaws would be acceptable per the requirements of the Alternate PTS Rule, 10 CFR 50.61 a. The licensee's response to RAI-2 clarified the information satisfactorily.

Regarding the requirements of 10 CFR 50.61 a for allowable flaws within the inner 1/1 Oth of the RPV wall thickness, one indication associated with Callaway Plant is within the thickness range and is required to be evaluated. This indication is characterized as an embedded axial flaw, 0.55 inches from the cladding-metal interface, 1.6 inches long, and 0.32 inch in through-wall extent. The licensee reproduced in Table 2 in CALLAWAY-ISI-ALT-05 information from the 10 CFR 50.61 table for allowable number of flaws in various through-wall extents for plates and compare this information with the number of relevant detected flaws associated with Callaway Plant. This comparison indicated that the number of detected flaws from the Callaway Plant RPV is bounded by those specified in 10 CFR 50.61a. All flaws associated with Callaway Plant are outside the thickness range and, therefore, are not required to be evaluated.

The next inspection for Callaway Plant is planned for 2023. The NRC staff has reviewed the revised PWROG plan and finds that the proposed alternatives match the inspection plan for the PWR fleet.

Based on the above evaluation, the NRC staff concludes that the licensee has addressed Plant-Specific Information 3 satisfactorily, because the licensee demonstrated that the plant-specific flaw information for Callaway Plant in 13R-17 is bounded by the WCAP-A, supporting the plant-specific applicability of the WCAP-A to the Callaway Plant.

In summary, the NRC staff reviewed the licensee's submittal and performed independent calculations to verify the input data and output results in Table 3 of the relief request. The difference between the licensee's and staff's calculated TWCF9s-rorAL is insignificant. With this information, the NRC staff concluded that the TWCFss-rorAL value in Table 3 of relief request

is bounded by the WCAP-A results. Consequently, the licensee has demonstrated that the proposed alternatives will provide an acceptable level of quality and safety and meets the guidance provided by RG 1.174, Rev. 1, for risk-informed decisions.

5.0 CONCLUSION

The NRC staff has completed its review of Relief Request 13R-17 for Callaway Plant. The staff concludes that increasing the lSI interval for Categories 8-A and 8-D components from 10 to 20 years will result in no appreciable increase in risk. This conclusion is based on the fact that the plant-specific information provided by the licensee is bounded by the data in the WCAP-A, and the request meets all the conditions and limitations described in the WCAP-A. Therefore, Relief Request 13R-17 provides an acceptable level of quality and safety, and the alternative can be authorized for Categories 8-A and 8-D components pursuant to 10 CFR 50.55a(z)(1) until the end of the third interval, which has been extended to December 18, 2024, for Callaway Plant for the subject components. Further, the NRC staff accepts the proposed alternative examination date of 2023 for Categories 8-A and 8-D components for Callaway Plant.

All other ASME Code,Section XI, requirements for which relief was not specifically requested and authorized herein by the staff remain applicable, including the third-party review by the Authorized Nuclear lnservice Inspector.

Principal Contributor: P. Purtscher, NRR/DE/EVI8 Date: February 10, 2015

• .. ML15035A148 *email dated OFFICE NRR/DORULPL4-1/PM N RR/DORULPL4-1/LA NRR/DE/EVIB/BC* NRR/DORULPL4-1/BC(A)

NAME Flyon JBurkhardt SRosenberg EOesterle DATE 2/6/15 2/6/15 1/30/15 2/10/15