ML110590921

Relief Request from ASME Code Case N-729-1 Requirements for Examination of Reactor Vessel Head Penetration Welds

ML110590921
Person / Time
Site:	Byron, Braidwood
Issue date:	03/03/2011
From:	Robert Carlson Plant Licensing Branch III
To:	Pacilio M Exelon Nuclear, Exelon Generation Co
DiFrancesco N, NRR/DORL/LPL3-2, 415-1115
References
TAC ME3510, TAC ME3511, TAC ME3512, TAC ME3513
Download: ML110590921 (12)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 20555-0001 March 3, 2011 Mr. Michael J. Pacilio President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

BRAIDWOOD STATION, UNITS 1 AND 2 AND BYRON STATION, UNIT NOS. 1 AND 2 - RELIEF REQUEST FROM ASME CODE CASE N-729-1 REQUIREMENTS FOR EXAMINATION OF REACTOR VESSEL HEAD PENETRATION WELDS (TAC NOS. ME3510, ME3511, ME3512, and ME3513)

Dear Mr. Pacilio:

By letter to the U.S. Nuclear Regulatory Commission (NRC), dated March 12, 2010 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML100710764), as supplement by letter dated December 6,2010 (ADAMS Accession No. ML103410394). Exelon Generation Company, LLC (the licensee), submitted a relief request from certain requirements of Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g)(6)(ii)(D) for the third 10-year inservice inspection (lSI) interval of Braidwood Station (Braidwood), Units 1 and 2, and Byron Station (Byron), Units Nos. 1 and 2.

The submittal requests authorization of a proposed alternative from the inspection requirement of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, ASME Code Case N-729-1, for the examination of reactor pressure vessel upper head penetrations for the control rod drive mechanism (CRDM).

Specifically, pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee requested to use the alternative on the basis that complying with the specified requirement would result in hardship or unusual difficulty.

The NRC staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety, and that the alternative proposed will provide reasonable assurance of structural integrity of the CRDM nozzles. Therefore, the alternative proposed is authorized pursuant to 10 CFR 50.55a(a)(3)(ii), for the third 10-year lSI interval for Braidwood, Units 1 and 2, and Byron, Unit Nos. 1 and 2, or until the respective head for each unit is replaced, whichever comes first.

All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved remain applicable, including a third-party review by the Authorized Nuclear Inservice Inspector. The NRC staff's safety evaluation is enclosed.

M. Pacilio -2 Please contact Mr. Nicholas DiFrancesco at (301) 415-1115 if you have any questions on this action.

Sincerely, Robert D. Carlson, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-456, STN 50-457, STN 50-454 and STN 50-455

Enclosure:

Safety Evaluation cc w/encl: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR RELIEF FROM REQUIREMENTS FOR EXAMINATION OF REACTOR VESSEL HEAD PENETRATION NOZZLES EXELON GENERATION COMPANY, LLC BRAIDWOOD STATION. UNITS 1 AND 2. AND BYRON STATION. UNIT NOS. 1 AND 2 DOCKET NOS. STN 50-456, STN 50-457. STN 50-454. AND STN 50-455.

1.0 INTRODUCTION

By letter to the U.S. Nuclear Regulatory Commission (NRC, the Commission) dated March 12, 2010 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML100710764), as supplement by letter dated December 6,2010 (ADAMS Accession No. ML103410394). Exelon Generation Company. LLC (the licensee), submitted a relief request from certain requirements of Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g)(6)(ii)(D) for the third 1O-year inservice inspection (lSI) interval of Braidwood Station (Braidwood), Units 1 and 2, and Byron Station (Byron). Unit Nos. 1 and 2.

The submittal requests authorization of a proposed alternative from the inspection requirement of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI. ASME Code Case N-729-1. for the examination of reactor pressure vessel upper head penetrations for the control rod drive mechanism (CRDM). SpeCifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(a)(3)(ii). the licensee requested to use an alternative on the basis that complying with the specified requirement would result in hardship or unusual difficulty.

2.0 REGULATORY EVALUATION

The ASME Boiler and Pressure Vessel Code,Section XI, states that lSI of Class 1. 2 and 3 components shall be performed in accordance with the requirements of Section XI, "Rules for In service Inspection of Nuclear Power Plant Components." of the ASME Code and applicable editions and addenda as required by 10 CFR 50.55a(g), except where specific written relief has been granted by the Commission. Pursuant to 10 CFR 50.55a(g)(4), throughout the service-life of a pressurized water reactor (PWR). components which are classified as ASME Code Class 1.

2 and 3 must meet the requirements, except design and access provisions and pre-service examination requirements. set forth in the ASME Code,Section XI, to the extent practical within the limitations of design, geometry and materials of construction of the components.

Further regulations under 10 CFR 50.55a(g)(4)(i) require that lSI of components and system pressure tests conducted during the first 10-year lSI interval and subsequent intervals shall comply with the requirements in the latest edition and addenda of the ASME Code,Section XI,

-2 incorporated by reference in paragraph (b) of 10 CFR 50.55a on the date 12 months prior to the start of the 120-month 151 interval subject to the limitations and modifications listed herein. The Section XI, ASME Code of Record for the third 10-year 151 interval at Braidwood and Byron is the 2001 Edition through the 2003 Addenda. For Braidwood, the third 10-year 151 interval end dates are July 28,2018 for Unit 1 and October 16,2018 for Unit 2. For Byron, the current third 10-year 151 interval end date is July 15, 2016 for both Unit Nos. 1 and 2.

In 10 CFR 50.55a(g)(6)(ii), it states that the Commission may require the licensee to follow an augmented 151 program for systems and components for which the Commission deems that added assurance of structural reliability is necessary. 10 CFR 50.55a(g)(6)(ii)(D) requires augmented 151 of reactor pressure vessel (RPV) head penetration nozzles of PWRs in accordance with ASME Code Case N-729-1, "Alternative Examination Requirements for PWR Reactor Pressure Vessel Upper Heads with Nozzles Having Pressure-Retaining Partial Penetration WeldsSection XI, Division 1", subject to the conditions specified in paragraphs (2) through (6) of 10 CFR 50.55a(g)(6)(ii)(D).

Pursuant to 10 CFR 50.55a(a)(3), proposed alternatives to the requirements of 10 CFR 50.55a(g) may be used when authorized by the Commission if: (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

3.0 TECHNICAL EVALUATION

3.1 Affected Systems and Components The CRDM penetrations nozzles designated as Item No. 84.20, "UNS N06600 nozzles and LlNS N06082 or UNS W86182 partial-penetration welds in head," in Table 1 of Code Case N-729-1, for Braidwood and Byron are as follows; Braidwood, Unit 1 Penetration nozzle Nos. 1,2,4,9,14-17,22,25,26,28-37,42,49, 54, 63, 65, 66, 71, 72, 77 and 78; Braidwood, Unit 2 Penetration nozzle Nos. 1, 4, 5, 8, 32, 34, 35, 40, 46, 66-68, 70, 71, 73,74 and 76-78; Byron, Unit No.1 Penetration nozzle Nos. 1-4,6-8, 15, 17,30-35,37,62,66,68,69, 74 and 75; Byron, Unit No.2 Penetration nozzle Nos. 1-9, 14-17,30-37,39,42,44,45,51-53,55,56, 58, 63, 68, 69 and 71.

-3 3.2 Proposed Alternative The licensee proposes to perform ultrasonic examination of each RPV head CRDM penetration nozzle (Le., nozzle base material) for a distance equal to "a" above the J-groove weld on the uphill slope, as defined by Figure 2 of ASME Code Case N-729-1, and to the minimum required inspection distances below the J-groove weld on the downhill slope as identified in Table 1 through 4, in this safety evaluation for Braidwood and Byron, respectfully. For all other penetrations, the required examination coverage dimension "a" reflected in Figure 2 of ASME Code Case N-729-1, will be met or exceeded.

Table 1: Braidwood 1 CRDM Nozzle Minimum Required Inspection Coverage I. Penetration No.

~~~:Ut~el~~:~; ~~~~~~: i pen~tration Minimum Inspection Coverage Below the J-Groove Weld Toe on the Downhill Side (in) o. on the Downhill Side (in) 1 1.28 33 1.48 I

2 1.36 34 1.28

! 4 1.48 35 1.28 9 1.44 36 1.20 14 1.40 37 i 1.08 15 1.28 42 0.92 16 1.28 49 0.76 17 1.36 54 0.92 22 1.40 63 0.92 25 1.48 65 0.92 26 1.48 66 0.92 28 1.40 71 0.88 29 1.16 72 0.92 30 1.24 77 0.92 31 1.48 78 ~. 0.84 32 I 1.36 I

-4 Table 2: Braidwood 2 CRDM Nozzle Minimum Required Inspection Coverage Minimum Inspection Minimum Inspection Penetration Coverage Below the J- Penetration Coverage Below the J-No. Groove Weld Toe on the No. Groove Weld Toe on the Downhill Side (in) Downhill Side (in) 1 1.40 67 0.68 4 1.44 68 0.92 L 5 1.48 70 0.64 8 1.44 71 0.56 32 1.28 73 0.84 I 34 1.32 74 0.80 I 35 1.28 76 0.92 40 0.96 77 0.96 46 0.88 78 0.68 66 0.92 Table 3: Byron 1 CRDM Nozzle Minimum Required Inspection Coverage Minimum Inspection Minimum Inspection Penetration Coverage Below the J- Penetration Coverage Below the J-No. Groove Weld Toe on the No. Groove Weld Toe on the Downhill Side (in) Downhill Side (in) 1 1.36 32 1.44 2 1.48 33 1.44 I 3 1.32 34 1.36 4 1.40 I 35 1.40 6 1.40 37 1.48 7 1.40 62 0.84 8 1.48 66 0.96 15 1.40 68 0.50 17 1.48 69 0.68 L 30 1.36 74 0.72 31 1.36 75 0.56

- 5 Table 4: Byron 2 CRDM Nozzle Minimum Required Inspection Coverage Minimum Inspection Minimum Inspection Penetration Coverage Below the J- Penetration Coverage Below the J-No. Groove Weld Toe on the No. Groove Weld Toe on the Downhill Side (in) Downhill Side (in) 1 1.16 35 1.28 2 1.40 36 1.16 3 1.28 37 1.48 4 1.28 39 0.80 5 1.20 42 0.88 6 1.16 44 0.84 7 1.12 45 0.80 8 1.16 51 0.96 9 1.20 52 0.88 14 1.16 53 0.88 15 1.16 55 0.92 16 1.44 56 0.64 17 1.24 58 0.84 30 1.12 63 0.76 31 1.20 68 0.84 32 1.08 69 0.80 33 0.92 71 0.96 34 0.88 3.3 Licensee's Basis for Request The bottom end of all of the Braidwood and Byron RPV head CRDM penetrations are externally (outside diameter or "00") threaded, internally (inside diameter or "lO") tapered, and have an ultrasonic corner shadow zone produced by the thread relief, precluding ultrasonic or eddy current data acquisition in a zone extending up approximately 1.18 inches from the bottom of each nozzle. For the majority of the penetrations, these geometric limitations reduce the inspectable distance from the bottom of the J-groove weld fillet to the top of the thread.

Therefore, the licensee requests relief for some CRDM penetrations that cannot meet the required inspection coverage dimension "a" shown in Figure 2 of ASME Code Case N-729-1.

During the previous refueling outages for each unit at Braidwood and Byron, the licensee obtained examination coverage data on all 78 CRDM penetrations in the RPV head. This information was used to support the licensee's previous NRC Order EA-03-09 relaxation requests regarding examination coverage below the J-groove weld which was approved by the NRC in letters dated February 7,2008 (ADAMS Accession No. ML080230600) for Byron, Unit No.2, and September 11, 2006 (ADAMS Accession No. ML062430598) for Byron, Unit No.1 and Braidwood, Unit 2, and September 26, 2007 (ADAMS Accession No. ML072430457) for Braidwood, Unit 1. However, these relaxation authorizations were rescinded with the issuance

-6 of 10 CFR 50.55a(g)(6)(ii)(0), on September 10, 2008 (73 FR 52724), which required implementation of ASME Code Case N-729-1, as conditioned, by December 31, 2008.

The licensee supported the proposed alternative with a stress analysis and deterministic fracture mechanics analysis. The plant-specific stress analysis demonstrated that the hoop and axial stresses remain below 20 kips per square inch (ksi) over the entire region outside the alternative examination zone defined by the licensee's proposed alternative. The stress analysis was provided for NRC staff review. The licensee also provided a plant-specific fracture mechanics analysis that demonstrated that a potential axial crack in the unexamined zone will not grow to the toe of the J-groove weld within six effective full power years (EFPYs). The licensee noted that the volumetric re-inspection frequency for each unit would be less than six EFPYs. Hence, the licensee's analysis noted, for all cases, each unit's head would be re-inspected prior to a postulated crack, in the uninspected region of the nozzle, growing to the toe of a penetration nozzle's associated J-groove weld and the reactor coolant system (RCS) pressure boundary.

By letter dated December 6, 2010, the licensee also noted that performance of dye penetrant testing may be possible in lieu of eddy current or ultrasonic testing to cover the area of missed inspection coverage. However, compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

The presence of thermal sleeves in the vast majority of the CROM penetrations prohibits dye penetrant testing of the tapered 10 surface of the tube. Dye penetrant testing of threaded surfaces, like the tube 00, is difficult due to physical restraints and the need to properly clean the surface to provide accurate test results. As a result, performing dye penetrant testing on the bottom nozzle area would require thermal sleeve removal, extensive manpower, and would result in significant radiation exposure to licensee personnel. The radiation exposure is estimated to be in excess of 100 person-rem, without a compensating increase in the level of quality or safety.

3.4 NRC Staff's Evaluation The NRC staff's review of this request was based on 10 CFR 50.55a(a)(3)(ii) which states that:

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.

The specifiC regulatory requirements for which relief is requested are defined in 10 CFR 50.55a(g)(6)(ii)(0)(3), which states in part:

Instead of the specified 'examination method' requirements for volumetric and surface examinations in Note 6 of Table 1 of Code Case N-729-1, the licensee shall perform volumetric and/or surface examination of essentially 100 percent of the required volume or equivalent surfaces of the nozzle tube, as identified by Figure 2 of ASME Code Case N-729-1.

-7 The extent of the examination of the nozzle tube in question is determined by the incidence angle, e, and the distance "a" below the J-groove weld, as defined in Figure 2 of ASME Code Case N-729-1, "a = 1.5 in. (38mm) for incidence angle, e, S 30 deg and for all nozzles ~ 4.5 in.

(115 mm) 00 or 1 in. (25 mm) for incidence angle, e, ~ 30 deg; or to the end of the tube, whichever is less." The licensee has identified 106 CRDM penetration nozzles, in Tables 1 through 4 above, for which this inspection coverage is not physically obtainable with ultrasonic inspection.

The licensee has shown a physical and radiological hardship which would be incurred in order to be within compliance with the specified requirements. The NRC staff finds that a physical hardship exists due to the inability of ultrasonic or eddy current inspection to effectively scan the bottom end of each CRDM penetration as each nozzle is threaded on the 00 and internally tapered. While dye penetrant inspection would be a viable option for the licensee, the inspection would require manual application in a high radiation area. Furthermore, additional setup work would require additional accumulation of dose for each nozzle. Therefore, the NRC staff finds that the radiological dose required to perform the additional inspection would be a significant radiological hardship for the limited additional inspection coverage.

The NRC staff then compared the regulatory requirements to the proposed alternative to ensure that given this hardship, compliance with the regulations did not provide a compensating increase in the level of quality and safety. The NRC staff reviewed the licensee's basis for the proposed alternative through a review of the licensee's stress and fracture mechanics analysis.

The NRC staffs review of the stress analysis was based on the degradation phenomenon of concern being primary water stress corrosion cracking (PWSCC). PWSCC typically initiates in the areas of the highest tensile stress in susceptible materials, such as alloy 600 materials, and propagates in response to time, environment (Le., temperature) and stress intensity. The NRC staff reviewed the licensee's stress analysis and conclusions by comparison of the licensee's supporting data for various nozzle angles (Figures 2 through 6 of the March 12, 2010, letter), the conservative analysis performed to support Materials Reliability Program Report, MRP-95R 1, "Generic Evaluation of Examination Coverage Requirements for Reactor Pressure Vessel Head Penetration Nozzles, Revision 1," dated September 2004 (ADAMS Accession No. ML043200602), and Pacific Northwest National Laboratory Report 17763, "Final Report Inspection Limit Confirmation for Upper Head Penetration Nozzle Cracking," dated August of 2008. The results of the NRC staffs review support the licensee's stress analysis and the NRC staff finds the areas of missed inspection coverage are in a reduced stress area, less than 20 ksi.

The licensee's fracture mechanics analysis showed that a conservative through-wall axial flaw located in the uninspected region of the nozzle would not grow to the toe of the J-groove weld, in this case the edge of the RCS pressure boundary, in less than six EFPYs. The NRC staff's assessment of the licensee's fracture mechanics analysis conclusions is based on data analysiS of the supporting Figures 7 through 12 of the crack growth predictions for various nozzle angles, as provided in the licensee's March 12, 2010, letter. In addition, the NRC staff performed an independent crack growth calculation, and the results of which support the licensee's analysis.

- 8 However, the NRC staff notes that the licensee's calculated allowed operational time period of six EFPYs as defined in Figure 12 and Section 5.1 of the March 12, 2010, letter, may not bound the current regulation's allowed time interval between inspections, which is 8 calendar years or 2.25 re-inspection years, whichever is less, in accordance with Item No. B4.20 of Table 1 of ASME Code Case N-729-1 and NRC regulation 10 CFR 50.55a(g)(6)(ii)(D)(2). The licensee, by letter dated December 6, 2010, in response to a request for additional information by the NRC, stated that none of the Braidwood or Byron units would exceed four 18-month cycles of operation between inspections. Due to the operational conditions, four 18-month cycles is less than six EFPYs. Therefore, given the licensee's technical basis, the NRC staff finds the maximum time between volumetric re-inspection in accordance with the requirements of Item No. B4.20 of Table 1 of ASME Code Case N-729-1 for Braidwood, Units 1 and 2 and Byron, Unit No.1 is 8 calendar years, 2.25 re-inspection years or 6 EFPYs, whichever is less. Due to the identification of PWSCC in one penetration nozzle at Byron, Unit No.2, the re-inspection frequency of Byron, Unit No.2 will remain in accordance with a letter dated January 28,2010, from the NRC to the licensee (ADAMS Accession No. ML100210231). Given the above, the NRC staff finds sufficient technical basis, such that the re-inspection frequencies above will provide reasonable assurance of structural integrity of each nozzle, due to the area of missed inspection coverage, as defined in Tables 1 through 4 above.

The safety issues that are addressed by 10 CFR 55a(g)(6)(ii)(D) are degradation (corrosion) of the low-alloy steel RPV upper head, RCS pressure boundary integrity, and ejection of the RPV upper head penetration nozzle due to circumferential cracking of the nozzle above the J-groove weld. The licensee's proposed alternative inspection provides reasonable assurance that these safety issues are addressed at Braidwood and Byron. The licensee has noted that while surface examination could be performed to increase the inspection coverage for the nozzle, these additional inspections would be of limited value and require extensive work in very high radiation fields. The NRC staff finds that performing these additional surface examinations would result in hardship through significant radiation exposure without a compensating increase in the level of quality or safety.

4.0 CONCLUSION

As set forth above, the NRC staff concludes that the proposed alternative provides reasonable assurance of structural integrity of the subject components. Based on the information provided in the licensee's submittals and given the maximum time between volumetric reinspection in accordance with the requirements of Item No. B4.20 of Table 1 of ASME Code Case N-729-1 for Braidwood Units 1 and 2 and Byron Unit No.1 is 8 calendar years, 2.25 reinspection years or 6 EFPYs, whichever is less, the NRC staff has determined that the licensee's proposed alternative demonstrates that compliance with the specified requirements of 10 CFR 50.55a(g)(6)(ii)(D) would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(ii), the use of the proposed alternative to define an alternate examination zone below the J-groove weld is authorized for Braidwood Units 1 and 2, and Byron Unit Nos. 1 and 2, for the third 10-year lSI interval, or until the respective head for each unit is replaced, whichever occurs first.

- 9 All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.

Principal Contributor: Jay Collins, NRR Date of issuance: March 3, 2011

M. Pacilio - 2 Please contact Mr. Nicholas DiFrancesco at (301) 415-1115 if you have any questions on this action.

Sincerely, IRA!

Robert D. Carlson, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-456, STN 50-457, STN 50-454 and STN 50-455

Enclosure:

Safety Evaluation cc w/encl: Distribution via Listserv DISTRIBUTION:

PUBLIC LPL3-2 RlF RidsNrrDirsltsb Resource RidsOgcRp Resource RidsRgn3MailCenter Resource RidsNrrDorlLpl3-2 Resource RidsNrrLASRohrer Resource RidsNrrPMByron Resource RidsNrrDorlDpr Resource JCollins, NRR RidsNrrPMBraidwood Resource RidsAcrsAcnw_MaitCTR Resource ADAMS Accession No ML110590921 NRR-028 *SE memo date OFFICE LPL3-2/PM LPL3-21LA DCIICPNB/BC* LPL3-2/BC NAME NDiFrancesco SRohrer TLupold RCarlson DATE 02/28/11 03/01/11 2115111 03/3/11 OFFICIAL RECORD COpy