ML13016A454

Safety Evaluation in Support of Request for Relief Associated with the Reactor Pressure Vessel Nozzle Repairs

ML13016A454
Person / Time
Site:	Quad Cities
Issue date:	01/30/2013
From:	Joel Wiebe Plant Licensing Branch III
To:	Pacilio M Exelon Generation Co
Michael Mahoney, NRR/DORL 415-3867
References
TAC ME8347
Download: ML13016A454 (15)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 30, 2013 Mr. Michael J. Pacilio Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer (CNO)

Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

QUAD CITIES NUCLEAR POWER STATION, UNIT 2 - SAFETY EVALUATION IN SUPPORT OF REQUEST FOR RELIEF ASSOCIATED WITH THE REACTOR PRESSURE VESSEL NOZZLE REPAIRS (TAC NO. ME8347)

Dear Mr. Pacilio:

By letter to the U.S. Nuclear Regulatory Commission (NRC) dated April 6, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12100A012). as supplemented by letters dated April 11 ,2012 {ADAMS Accession No. ML12103A009}, April 12, 2012 (ADAMS Accession Nos. ML12104A021, ML12104A067, and ML12104A068 ,

respectively), April 13, 2012 (ADAMS Accession No. ML12107A361), and April 14, 2012 (ADAMS Accession Nos. ML12107A362 , ML12108A028, and ML12108A026, respectively),

pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a(a)(3)(i),

Exelon Generation Company, LLC (EGC, the licensee), submitted request for relief (RR) 14R-19 for the fourth 10-year inservice inspection (lSI) interval at the Quad Cities Nuclear Power Station (QCNPS), Unit 2. The licensee requested relief from several provisions of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI.

On April 15, 2012, the NRC staff verbally authorized the use of RR 14R-19 (ADAMS Accession No. ML12107A472).

Accordingly, the NRC staff concludes that EGC has adequately addressed all of the regulatory requirements set forth in 10 CFR, Section SO.SSa(a)(3)(i), and is in compliance with the requirements of 10 CFR 50.55a with the authorizing of this alternative described in RR 14R-19 for the duration of QCNPS, Unit 2, operating cycle 22; which will end at the next refueling outage. Therefore, the NRC staff authorizes the licensee's proposed alternative contained in RR 14R-19.

However, the NRC staff does not authorize relief from the requirements for subsequent examination of flaws in accordance with IWB-2420 (b) and (c), since the duration of the current relief is only for one operating cycle, and any subsequent examination would occur beyond this time frame. The NRC staffs safety evaluation is enclosed.

M. Pacilio - 2 Please contact the QCNPS, Unit 2, Project Manager Brenda Mozafari at (301) 415-2020 if you have any questions on this action.

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Joel Wiebe, Acting Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-265

Enclosure:

Safety Evaluation cc w/encl: Distribution via Listserv

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST 14R-19 REGARDING FOURTH 10-YEAR INTERVAL INSPECTION PROGRAM VESSEL PENETRATION REPAIR EXELON GENERATION COMPANY. LLC QUAD CITIES NUCLEAR POWER STATION, UNIT 2 DOCKET NO. 50-265 (TAC NO. ME8347)

1.0 INTRODUCTION

By letter to the U.S. Nuclear Regulatory Commission (NRC) dated April 6, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12100A012), as supplemented by letters dated April 11, 2012 (ADAMS Accession No. ML12103A009), April 12, 2012 (ADAMS Accession Nos. ML12104A021, ML12104A067, and ML12104A068, respectively), April 13, 2012 (ADAMS Accession No. ML12107A361), and April 14, 2012 (ADAMS Accession Nos. ML12107A362, ML12108A028, and ML121080063, respectively),

pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a(a)(3)(i),

Exelon Generation Company, LLC (EGC, the licensee), submitted the relief request (RR) 14R-19 for the fourth 1O-year inservice inspection (lSI) interval at the Quad Cities Nuclear Power Station (QCNPS), Unit 2. The licensee requested relief from several provisions of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI. The ASME Code of record for QCNPS, Unit 2, is the 1995 Edition through 1996 Addenda.

During the QCNPS, Unit 2 spring 2012 refueling outage (Q2R21), the licensee discovered signs of leakage through visual examinations consisting of water exiting the reactor pressure vessel (RPV) at a rate of 60 drops per minute from the N-11 B RPV penetration. This penetration is part of the reference leg of the RPV water level instrument system. This observation necessitated the repair of this penetration during refueling outage Q2R21.

Approval of this request would allow the licensee to repair the RPV penetration without performing non-destructive examinations (NDE) to determine the size, orientation, or exact location of the flaw in the penetration. Approval of this request also exempts the licensee from the requirement to remove or reduce the size of the flaw. Instead, the licensee proposes to relocate the pressure-retaining weld to the outside of the RPV as opposed to the original boundary on the inside of the RPV.

On April 15, 2012, the NRC staff granted verbal authorization for the use of RR 14R-19 for the repair of the N-118 RPV penetration during the Q2R21 refueling outage (ADAMS Accession No. ML12107A472).

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2.0 REGULATORY EVALUATION

Pursuant to 10 CFR SO.SSa(g)(4), ASME Code Class 1, 2, and 3 components (including supports) must meet the requirements, except the design and access provisions and the preservice 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. The regulations require that lSI of components and system pressure tests conducted during the 10-year intervals be in compliance with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR SO.SSa(b), 12 months prior to the start of the 120-month interval. The ASME Code of record for QCNPS, Unit 2, is the 1995 Edition through 1996 Addenda.

The regulations in 10 CFR SO.SSa(g)(6)(i) state that the Commission may grant such relief and may impose such alternative requirements as it determines is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest, given due consideration of the burden upon the licensee.

The regulations in 10 CFR SO.SSa(a){3) state that alternatives to the requirements of paragraph (g) of 10 CFR SO.SSa may be used, when authorized by the NRC, if: (i) the proposed alternatives would provide an acceptable level of quality and safety, or Oi) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

The regulations in 10 CFR SO.SSa(g)(S)(iii) state that if the licensee has determined that conformance with certain code requirements is impractical for its facility, the licensee shall notify the Commission and submit, as specified in 10 CFR SO.4, information to support the determinations.

3.0 TECHNICAL EVALUATION

3.1 Licensee's Relief Request ASME Code Components Affected Code Class: 1 Component Number: RPV Penetration N-11 B Examination Category: B-P Item Number: B1S.10

==

Description:==

RPV Water Level Instrument Penetration- 2-inch nominal pipe size

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Applicable Code Edition and Addenda

The ASME Code of record for QCNPS, Unit 2, is the 1995 Edition through 1996 Addenda.

The code of construction for the RPV is the 1965 Edition through summer 1965 Addenda of ASME Code,Section III. The code of construction for the penetration is likewise the 1965 Edition through summer 1965 Addenda of ASME Code,Section III.

Applicable Code Requirements:

Flaw Removal

• IWA-5250(a)(3) states, "Components requiring correction shall have repair/replacement activities performed in accordance with IWA-4000."

• IWA-4410(a) states, "Repair replacement activities shall be performed in accordance with the Owner's Requirements and the original Construction Code of the component or system, except as provided in IWA-4410(b), (c), (d)."

• The applicable requirements of the construction code required by IWA-4410(a) for the removal of defects.

o N-528 requires repair of weld defects including removal of defects detected by leakage tests.

• IWA-4611.1 states in part, "Defects shall be removed or reduced in size in accordance with this Paragraph."

Flaw Evaluation

• IWB-3522.1 states, " ... relevant conditions that may be detected during the conduct of system pressure tests shall require correction to meet the requirements of IWB-3142 and IWA-5250 prior to continued service ... "

• IWB-3142.1 (b) states, "A component whose visual examination detects the relevant conditions described in the standards of Table IWB-3410-1 shall be unacceptable for continued service, unless such components meet the requirements of IW8-3142.2, 3142.3, or 3142.4."

• IWB-3142.4 states, "A component containing relevant conditions is acceptable for continued service if an analytical evaluation demonstrates the component's acceptability_

The evaluation analysis and evaluation acceptance criteria shall be specified by the Owner. A component accepted for continued service based on analytical evaluation shall be subsequently examined in accordance with IWB-2420(b) and (c)."

• IWB-2420(b) and (c) requires reexamination of the flaw during the next three inspection periods and IWV-2420(c) allows the examination schedule to revert to the original schedule provided the flaws remain essentially unchanged.

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• IWA-3300(a) states in part, "Flaws detected by the pre-service and inservice examinations shall be sized ... "

• IWA-3300(b) states in part, "Flaws shall be characterized in accordance with IWA-3310 through IWA-3390, as applicable ... II

• IWB-3610(b) states, "For purposes of evaluation by analysis, the depth of flaws in clad components shall be defined in accordance with Fig. IWB-361 0-1."

• IWB-3420 states, "Each detected flaw or group of flaws shall be characterized by the rules of IWA-3300 to establish the dimensions of the flaws. These dimensions shall be used in conjunction with the acceptance standards of IWB-3S00."

ASME Code Requirement for Which Relief is Requested

• The licensee requested relief from:

(1) Requirements for removal and/or reduction in size of the flaws of IWA-4410 and IWA 4611; (2) Requirements to characterize the flaw of IWB-3420 and IWB-3610(b); and (3) Requirement for subsequent reexamination of flaws in accordance with IWB-2420 (b) and (c) for components that have been accepted by an analytical evaluation as allowed by IW8-3132.3 (for flaws detected by volumetric examinations) or IWB-3142.4 (for flaws detected by a visual examination).

Licensee's Proposed Alternative

• The licensee states that:

Pursuant to 10 CFR SO.SSa(a)(3){i), EGC proposes the following alternatives to the ASME Code,Section XI, requirements specified [above].

(1) As an alternative to flaw removal or reduction in size to meet the applicable acceptance standards, EGC proposes to implement the 00 [outer diameter] repair of the reactor vessel instrument nozzle N-11 B utilizing an 00 weld pad as described below in the discussion of the repair of nozzle penetrations.

(2) As an alternative to performing the NOE required to characterize the flaw under IWB 3420 and IW8-3610(b) in penetration N-11B, EGC proposes analyzing a maximum postulated flaw that bounds the range of flaw sizes that could exist in the J-groove weld and nozzle.

(3) As an alternative to performing the subsequent NOE required by IWB-3142.4 in accordance with IWB-2420(b) and (c) to assess potential growth of the flaw in

- 5 penetration N-11 B, EGC proposes analyzing a maximum postulated flaw that bounds the potential growth of the existing flaw.

The licensee provided further details regarding the repair design and flaw analysis. The repair design consists of the application of a weld pad on the 00 of the RPV made from Alloy 52M weld material, a new section of nozzle fabricated from Alloy 690, and a partial penetration J groove weld between the new nozzle section and the weld pad. The licensee provided a flaw analysis bounding the growth of a postulated flaw until the refueling outage estimated to begin in April 2014.

Licensee's Basis for Requesting Relief The licensee stated that the flaw location was such that no qualified or demonstrated NDE testing methods are available to meet all of the requirements of the ASME Code relative to the repair and analysis procedures. The licensee indicated that they were confident that repairing the penetration as described above would provide adequate assurance of vessel integrity when coupled with their conservative flaw evaluation. The flaw evaluation consists of two parts: (1) a one-cycle evaluation as part of RR 14R-19, and (2) a long-term evaluation using "detailed residual stress analysis and fatigue crack growth analysis" as a separate RR. The licensee provided a proprietary flaw evaluation (ADAMS Accession No. ML12104A068, not publically available) as Attachment 3 to the letter dated April 12, 2012 (ADAMS Accession No. ML 1204A067), as supplemented by a request for additional information (RAI) response dated April 14, 2012 (ADAMS Accession No. ML12108A028). The flaw evaluation employed linear elastic fracture mechanics (LEFM) techniques to demonstrate that a radially-oriented crack encompassing the entire cross section of the J-groove weld was acceptable for one operating cycle (approximately two years). The postulated flaw met the acceptance criteria of the ASME Code, IWB-3600, through the end of the operating cycle.

3.2 NRC Staff Evaluation The licensee identified leakage of 60 drops per minute from the N-11 B RPV penetration during the Q2R21 outage. This observation necessitated the repair of the leaking nozzle N-11B during the refueling outage for which the licensee requested NRC authorization for the use of RR 14R-19 in order to support the Q2R21 outage. On April 15, 2012, the NRC staff granted verbal authorization for use of RR 14R-19 during the QCNPS, Unit 2, Q2R21 outage.

3.2.1 Repair. Replacement. and NDE of Nozzle N-11 B The NRC staff notes that the QCNPS, Unit 2, RPV is made of carbon steel with the inside surface clad with stainless steel. The instrumentation nozzles and the attachment J-groove dissimilar metal (OM) welds were fabricated from Alloy 600/821182 and are susceptible to intergranular stress corrosion cracking (IGSCC). The NRC staff questioned whether this might only be a leading indicator of instrument nozzle degradation. In their April 11, 2012, response, the licensee stated that this particular nozzle had been repaired but not heat-treated during fabrication and therefore had significantly more cold-work than the other similar nozzles. The licensee stated that as a consequence, this nozzle was more susceptible to IGSCC than the

- 6 other nozzles. The NRC staff finds this a credible explanation, and concludes that the other nozzles would likely be more IGSCC-resistant than N-11 B nozzle.

To repair the leaking nozzle N-11 B, the licensee proposed to utilize the "half-nozzle" repair technique where a portion of the existing nozzle assembly at or near the 00 surface of the RPV will be cut and removed by machining, and replaced with an Alloy 690 (SB-166, UNS N06690) material resistant to primary water stress corrosion cracking. The Alloy 690 replacement process consists of applying a weld pad onto the 00 of the RPV using Alloy 52M (ERNiCrFe 7A, UNS N06054) filler metal, and using the same type of Alloy 52M to perform the final partial penetration welding between the nozzle and the weld pad. The remnant of the original nozzle N-11 B, and the original partial penetration attachment weld that contained the flaw, will be left in place.

The weld pad process utilizes ASME Code Case N-638-4, "Similar and Dissimilar Metal Welding Using Ambient Temperature Machine (GTAW) Temper Bead Technique,Section XI, Division 1."

ASME Code Case N-638-4 establishes requirements for performing ambient temperature temper bead welding as an alternative to preheat and post-weld heat treatment requirements of the construction code. ASME Code Case N-638-4 requires that the final weld surface shall be examined using a surface and volumetric methods when the completed weld has been at ambient temperature for at least 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. ASME Code Case N-638-4 has been conditionally approved for use by the NRC in Regulatory Guide (RG) 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," Revision 16, dated October 2010, with the conditions that: (1) the ultrasonic (UT) examinations shall be demonstrated for the repaired volume using representative samples which contain construction type flaws, and that (2) the provisions of 3(e)(2) of ASME Code Case N-638-4 only be used when 3(e)(1) is found to be impractical.

The Alloy 52M welded pad is welded on the carbon steel vessel using the machine gas tungsten arc weld (GTAW) ambient temperature temper bead technique in accordance with ASME Code Case N-638-4. The NRC staff noted that ASME Code Case N-638-4 prohibits its use when the base metal specification is SA-302, Grade B, such as the QCNPS, Unit 2, RPV, except under specific conditions. The licensee responded in their April 11, 2012, letter that the vessel conformed to ASME Code Case 638-4 conditions for use with SA-302, Grade B, materials. In the same letter, the licensee confirmed that the required NDE was performed and that industry experience indicated that the weld pad itself would introduce additional stresses "not expected to be significantly higher than the stresses were at the original weld site."

The Alloy 690 replacement nozzle is attached to the new weld pad with a partial penetration weld using a non-temper bead manual welding technique. The licensee detailed this technique in their April 11 ,2012 letter, specifying that the weld procedure was qualified in accordance with the ASME Code 2007 Edition through 2008 Addenda,Section III, NB-4000, and ASME Code 2010 Edition through 2011 Addendum,Section XI requirements. The licensee further noted, in response to NRC staff RAls, that the process consisted of a manual GTAW process, and that liquid penetrant (PT) examination was performed on the J-groove weld in preparation for laying the connecting weld in it. This examination ensures that no pre-existing cracks in the weld pad would threaten the integrity of the partial penetration weld.

-7 NRC staff requested that the licensee detail the examinations to be performed as part of the repair process, and the licensee responded in detail in the April 11, 2012 letter. The licensee performed significant volumetric and surface examinations throughout the process, including visual, UT, PT, and magnetic particle examinations as appropriate to each step of the repair process. Examinations are performed before, during, and after the repair.

The licensee also referenced ASME Code Case N-416-3, "Alternative Pressure Test Requirement for Welded or Brazed Repairs, Fabrication Welds, or Brazed Joints for Replacement Parts and Piping Subassemblies, or Installation of Replacement Items by Welding or Brazing, Classes 1, 2, and 3,Section XI, Division 1," and ASME Code Case N-405-1, "Socket Welds,"Section III, Division 1.

In response to NRC staff RAls, the licensee discussed their use and implementation of these ASME Code Cases in their April 12, 2012, letter (ADAMS Accession No. ML12104A021). Relief is not required to use these ASME Code Cases as ASME Code Case N-405-1 is unconditionally approved in RG 1.84 "Design, Fabrication, and Materials Code Case acceptability, ASME Section III," Revision 35, dated October 2010, and ASME Code Case N-416-3, was approved unconditionally in RG 1.147. Both ASME Code Cases are referenced in the QCNPS, Unit 2, lSI program plan.

Although the latest version of RG 1.147 (Revision 16 dated October 2010) lists ASME Code Case N-416-4, it is acceptable to use the earlier revision of the ASME Code Case as 10 CFR 50.55a (b}(5}(ii) states that if a licensee has previously applied a ASME Code Case and a later version of the ASME Code Case is incorporated by reference, the licensee may continue to apply, to the end of the current 120-month interval, the previous version of the ASME Code Case as authorized or may apply the later version of the ASME Code Case, including any NRC specified conditions placed on its use.

ASME Code Case N-416-3 allows the performance of a system leakage test following weld repairs in lieu of a hydrostatic test and requires NDE in accordance with the 1992 Edition of the ASME Code,Section XI, including a VT-2 visual examination in conjunction with the system leakage test.

The licensee further confirmed that the examinations would be performed according to NRC acceptable ASME Code requirements, specifically, the examinations would be completed according to the ASME Code of Record, and acceptance criteria would be drawn from the construction code (1965 Edition with summer 1965 Addenda), or the ASME Code 1992 Edition,Section III. In their April 12, 2012 RAI response, the licensee indicated that surface examination of the base metal weld area prior to welding activities referenced the original code of construction, ASME Section III, 1965 Edition, summer 1965 Addenda, since no welding has taken place in the area. Once welding activities occur on the weld area, which is part of the pressure-retaining boundary, the conditions within ASME Code Case N-416-3 require NDE methods in accordance with ASME Section III, 1992 Edition. This limitation is written into this revision of ASME Code Case N-416-3.

- 8 3.2.2 Corrosion Evaluation In response to an NRC staff RAI regarding the potential of crevice corrosion due to the gap left between the original and the repaired nozzle tubes, the licensee provided a corrosion evaluation in its April 12, 2012, letter (ADAMS Accession No. ML12104A067). The corrosion evaluation addressed general corrosion, crevice corrosion, and galvanic corrosion of the exposed LAS in the gap. The licensee stated that the repair area would be exposed to steam, which has passed through the dryers and separators, and possibly some condensate, and that QCNPS, Unit 2, operation injects hydrogen, zinc, and utilizes noble metal chemical addition (NMCA).

However, the licensee noted that the hydrogen water chemistry (HWC) is not effective above the top guide (thus is not effective at the repair location). The licensee provided data on corrosion by the four mechanisms listed above. Based on the data, the licensee concluded the repairs are acceptable with respect to corrosion.

The NRC staff requested that the licensee confirm that the water chemistry conditions represented by the corrosion data bounds the QCNPS, Unit 2 water chemistry conditions with respect to the potential for causing the various corrosion mechanisms. In response to an RAI in letter dated April 13, 2012 (ADAMS Accession No. ML12107A361), the licensee stated that their corrosion evaluation references report NEDC-21120, "Monticello Feedwater Nozzle Cladding Crack Repair Report," as the basis for low general corrosion rates and the absence of galvanic corrosion. The corrosion rate used in the licensee's corrosion evaluation is 70 percent faster than the maximum corrosion rate from all of the testing in NEDC-21120. NEDC-21120 does not provide specific water chemistry parameters. However, since NEDC-21120 dates from 1975, the licensee inferred that the boiling-water reactor (BWR) chemistry conditions simulated in the testing were less stringent than the chemistry conditions under which QCNPS, Unit 2, currently operates. The licensee stated that the reactor water chemistry of QCNPS, Unit 2, meets the requirements of the latest BWRVIP Water Chemistry Guidelines (BWRVIP-190), and that the requirements of BWRVIP-190 are much more restrictive than the reactor water chemistry requirements of the 1970's.

Therefore, the licensee concluded that the QCNPS, Unit 2, water chemistry is less aggressive than the chemistry simulated in the testing documented in NEDC-21120, thus the corrosion rates used in the licensee's corrosion evaluation are conservative.

HWC with NMCA reduces dissolved oxygen concentration in the reactor water to very low levels. While dissolved oxygen concentration would tend to increase due to radiolysis in upper parts of the RPV where HWC loses its effectiveness, the dissolved oxygen should still be less than in a BWR that does not employ HWC, but instead employs normal water chemistry, which would have included all BWRs in 1975.

In a letter dated April 14, 2012 (ADAMS Accession No. ML12107A362), the licensee provided the cycle average reactor water and final feedwater oxygen concentrations of 0.6 parts per billion (ppb) and 53 ppb respectively. While the licensee does not directly sample or estimate the oxygen content in main steam or at the N-11 B nozzle location, the oxygen content will be bounded by the oxygen content of the final feedwater and the impurity concentrations will be bounded by those in the reactor water, which meet the stringent requirements of BWRVIP-190.

-9 Therefore, the maximum dissolved oxygen to which the nozzle would be exposed should be around 53 ppb which is far below the saturated oxygen content, and the impurity concentrations will be very low. Based on this information, the NRC staff concludes the corrosion rates for the QCPNS, Unit 2, N-11 B nozzle location should be much lower than the corrosion rate for oxygen saturated conditions and should be bounded by the corrosion rates reported in the licensee's corrosion evaluation.

The NRC staff finds the licensee's evaluation of the potential for general, crevice and galvanic corrosion of the exposed low-alloy steel (LAS) is based on laboratory testing and field experience in operating BWRs under conditions that conservatively bound those at QCNPS, Unit 2, with respect to the susceptibility to these types of corrosion, and is therefore acceptable.

The corrosion evaluation also included an evaluation of the potential for stress corrosion cracking (SCC) of the LAS, which concluded the potential for SCC was minimal.

Recommended crack growth rate for assumed SCC growth into LAS are addressed in the discussion of the NRC staff evaluation of the licensee's flaw evaluation.

3.2.3 Flaw Evaluation Final Flaw Size To justify not removing the flaw in the J-groove weld of the remnant of the original N-11 B nozzle, the licensee performed a flaw evaluation (ADAMS Accession No. ML12104A068) to demonstrate the structural integrity of the RPV for one operating cycle. A typical flaw evaluation requires determination of the initial flaw size, the applied stress intensity factor

(!<applied) values, crack growth, and stability of the final crack-size. As the flaw was not directly examined, the licensee postulated an assumed-flaw.

The initial flaw was modeled as a double-sided crack that has propagated through the J-groove weld and touches the RPV LAS material that comprises the pressure boundary. This assumed initial flaw size which represents the radial cross section of the J-groove weld (the worst possible radial crack that could exist in the weld). This is consistent with those in approved applications of a similar nature and has become standard industry practice now, as referenced in a safety evaluation (SE) dated January 12, 2005, for Topical Report WCAP-15973-P, Revision 01, "Low-Alloy Steel Component Corrosion Analysis Supporting Small-Diameter Alloy 600/690 Nozzle Repair/Replacement Program" (ADAMS Accession No. ML050180528).

The licensee's postulated final flaw-size considers flaw-growth. In a letter dated April 13, 2012, the licensee provides a quantitative assessment of the crack growth using a bounding crack growth-rate in the LAS from BWRVIP-60-A, "BWR Vessel and Internals Project, Evaluation of Stress Corrosion Crack Growth in Low Alloy Steel Vessel Materials in the BWR Environment,>>

dated June 2003. Since the response supports the assumed crack increment over the 2-year operating cycle, the NRC staff determined that the licensee's final flaw size is appropriate for the subsequent stress and fracture mechanics analysis.

- 10 Operating and Residual Stresses The thermal and pressure stresses used in the current flaw evaluation are from the three transients: heatup and cooldown, safety relief valve blowdown, and SCRAM. The licensee states in their April 13, 2012 response, that these transients are determined to be controlling because, based on the ASME Code,Section III, design analysis of record, they have significant temperature excursions. This is acceptable, considering that the significant temperature excursions of these transients would lead to the highest stress ranges. Also, depending on where the crack plane is assumed along the circumference of the nozzle hole, the thermal and pressure stresses will be different. The licensee confirmed in their April 13, 2012 response, that the crack plane is assumed to exist at a circumferential location with respect to the nozzle hole such that the maximum remote membrane and bending stresses are applied.

The NRC staff also reviewed Reference 5 of AREVA Document No. 32-9181076-001, in letter dated April 14, 2012 (ADAMS Accession No. ML12108A026) , of the licensee's flaw evaluation regarding the calculation of operating stresses used in the LEFM analysis. The licensee explained in its RAI response dated April 14, 2012 (ADAMS Accession No. ML12108A028), that use of a radius of 2.5 times the RPV radius for the aXi-symmetric finite element model (FEM) model for the stress analysis is sufficient to account for the varying stresses due to pressure in the actual RPV. This is acceptable based on the provided justification that the circumferential stress due to pressure in a thin-shell cylinder is twice the stress in a spherical geometry.

It should be noted that the licensee's simplified approach is in line with that discussed in Oak Ridge National Laboratory's (ORNL) report (ORNLffM-2010/246), "Stress and Fracture Mechanics Analyses of Boiling Water Reactor and Pressurized Water Reactor Pressure Vessel Nozzles," dated December 2010, and is more conservative because the ORNL approach used a radius of only two times the actual RPV radius. Further, the NRC staff found the FEM model has sufficient element mesh to simulate the essential details of the RPV nozzle configuration, and considered the operating stresses credible.

In addition to thermal and pressure stresses, the licensee's evaluation also considered the residual stresses and the stresses due to crack face pressure. For residual stresses, the licensee further clarified in the April 13, 2012 response to NRC staff's RAI that assuming an average compressive stress in the LAS material will underestimate the compressive stresses near the crack front, resulting in a conservative analysis. The NRC staff concurs with this additional information and accepts the through-wall residual stress profile used in the licensee's flaw evaluation. This acceptance is further supported by the NRC staff's finding that the licensee's theoretical residual stress model through the weld and RPV is consistent with the residual stress solutions for control rod drive mechanism nozzles, reported in an ASME pressure vessel and piping conference paper by D. Rudland, et at, "Comparison of Welding Residual Stress Solutions for Control Rod Drive Mechanism Nozzles," ASME 2007 Pressure Vessels and Piping Conference Proceedings Volume 6: Materials and Fabrication, dated July 2007. The equivalent membrane and bending stresses for the crack face pressure case are generated similarly using an appropriate through-wall stress profile and are, therefore, acceptable.

- 11 LEFM Evaluation The licensee utilized the Raju-Newman formulation documented in National Aeronautics and Space Administration Technical Memorandum 85793, UStress-lntensity Factor Equations for Cracks in Three-Dimensional Finite Bodies Subjected to Tension and Bending Loads." The NRC staff's evaluation and acceptance of this FEM-based LEFM model for a similar application was documented in the January 12, 2005 SE mentioned above. Applying the stresses (operating, residual, and crack face) discussed above to the Raju-Newman model with the final flaw size and the RPV nozzle geometry documented in Section 4.2 of the licensees flaw evaluation, the licensee calculated the Kapplied values during the three transients and summarized the results in Table 5-1 of the licensee's flaw evaluation. This table also contains the corresponding fracture resistance, calculated margins, and required margins.

For fracture resistance, the licensee used the lower bound crack arrest stress intensity factor (Kia) and crack initiation stress intensity factor (Kid from the ASME Code,Section XI. For the required margins for the three transients, the licensee used the ASME Code,Section XI, required values. The NRC staff found both to be acceptable.

For the crack stability analysis, Table 5-1 of the licensee's flaw evaluation provides a comparison of calculated margins versus required margins during each of the three transients, with the most limiting values for the three transients presented in Table 6.1 of the flaw evaluation. Since all the calculated margins exceed the ASME Code,Section XI, required margins during each of the three transients, the flaw evaluation results meet the ASME Code,Section XI, requirements for demonstrating structural integrity of the RPV with the flaw in the J groove weld of the remaining original nozzle for continued operation for one operating cycle.

The NRC staff concludes that the licensee's flaw evaluation, as explained above, is acceptable because a conservative initial crack size is assumed and the flaw evaluation has demonstrated that the assumed worst-case crack size will not grow to an unacceptable depth into the LAS RPV during the period for which relief will be granted (one cycle of operation).

4.0 CONCLUSION

Based on the review and evaluation of the licensee's submittals, the NRC staff concludes that the proposed alternatives related to the repair of QCPNS, Unit 2, RPV penetration N-11 B would provide an acceptable level of quality and safety until the next refueling outage estimated to begin in April 2014.

Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the NRC staff authorizes the proposed alternative described in RR 14R-19 for the duration of QCNPS, Unit 2, operating cycle 22, which will end at the next refueling outage (April 2014). However, the NRC staff does not authorize relief from the requirements for subsequent examination of flaws in accordance with IWB-2420 (b) and (c), since the duration of the current relief is only for one operating cycle, and any subsequent examination would occur beyond operating cycle 22. The licensee will need to submit a revised RR for application of this alternative beyond the end of current operating cycle 22.

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

Principal Contributors: J. Poehler, NRR S. Sheng, NRR D. Widrevitz, NRR Date of issuance: January 30, 2013

M. Pacilio -2 Please contact the QCNPS, Unit 2, Project Manager Brenda Mozafari at (301) 415-2020 if you have any questions on this action.

Sincerely, I RA I Joel Wiebe, Acting Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-265

Enclosure:

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

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