HNP-15-034, Relief Request I3R-15 Revision and Supplement, Reactor Vessel Closure Head Nozzle Repair Technique, Inservice Inspection Program - Third Ten-Year Interval

Relief Request I3R-15 Revision and Supplement, Reactor Vessel Closure Head Nozzle Repair Technique, Inservice Inspection Program - Third Ten-Year Interval

ML15105A521
Person / Time
Site:	Harris
Issue date:	04/15/2015
From:	Waldrep B Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
HNP-15-034 RR I3R-15
Download: ML15105A521 (66)

Text

Benjamin C. Waldrep Vice President Harris Nuclear Plant 5413 Shearon Harris Rd New Hill NC 27562-9300 919-362-2502 Contains Proprietary Information Withhold from Public Disclosure Under 10 CFR 2.390 Trade Secrets and Commercial or Financial Information Uncontrolled When Separated from Enclosure 3 10 CFR 50.55a April 15, 2015 Serial: HNP-15-034 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Shearon Harris Nuclear Power Plant, Unit No. 1 Docket No. 50-400/Renewed License No. NPF-63

Subject:

Relief Request I3R-15 Revision and Supplement, Reactor Vessel Closure Head Nozzle Repair Technique, Inservice Inspection Program - Third Ten-Year Interval Ladies and Gentlemen:

Duke Energy Progress, Inc. (Duke Energy), requested NRC approval of relief request I3R-15 for the Shearon Harris Nuclear Power Plant, Unit 1 (HNP) inservice inspection program in a letter dated April 2, 2015 (ADAMS Accession No. ML15092A236). At the time of the request, there were no known flaws that required repair in the reactor vessel closure head nozzles. HNP examines the reactor head each refueling outage, and the examination performed in the current outage identified three flaws that require repair.

Duke Energy revises and supplements relief request I3R-15 as follows:

Identify three flaws that require repair in the current refueling outage to which I3R-15 will be applicable.

Revise the basic steps for the Inner Diameter Temper Bead repair technique to make step 8, abrasive water jet machining remediation, optional.

Remove credit for abrasive water jet machining in the design life expectancy of the repair. Calculations supporting the change are enclosed.

Expedite the requested approval schedule for the request.

The above changes in the revised relief request are indicated with revision bars in Enclosure 1.

The original request for relief was to be used for repair of flaws that may be discovered in the future. This revision and supplement maintains the provision that the relief would be applicable to repair of flaws not yet identified that may occur in the future, to preclude the need for another emergent request on an expedited schedule.

Calculations supporting this relief request were previously docketed supporting relief request I3R-13 in Duke Energy letter dated November 25, 2013 (ADAMS Accession No. ML13330A996). The calculations that were revised to support this request are provided in, and they contain information considered proprietary to AREVA Inc. On behalf of AREVA Inc., Duke Energy requests that the NRC withhold this information in accordance with 10 CFR 2.390. Enclosure 2 contains an affidavit supporting withholding of the proprietary

U.S. NuclearRegulatoryCommission HNP-15-034 Page 2 information. Upon removal of the proprietary information in Enclosure 3, the balance of this letter is decontrolled.

HNP requests approval of this request by April28, 2015,to support startup from the current refuelingoutage.

This document contains no new regulatorycommitments.

Please refer any questions regarding this submittal to Dave Corlett,HNP Regulatory Affairs Manager, at (919) 362-3137.

Sincerely,

Enclosures:

1. Revised Relief Request

2. Affidavit Supporting Withholding of Proprietary Information

3. Proprietary Calculations

4. Redacted/Non-Proprietary Calculations cc:

Mr. J. D. Austin, NRC Sr. Resident Inspector, HNP Ms. M. Barillas, NRC Project Manager, HNP Mr. V. M. McCree, NRC Regional Administrator, Region II

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 1 of 27 HNP-15-034 Shearon Harris Nuclear Power Plant, Unit No. 1 Docket No. 50-400/Renewed License No. NPF-63 Relief Request I3R-15 Revision and Supplement Reactor Vessel Closure Head Nozzle Repair Technique Inservice Inspection Program - Third Ten-Year Interval Revised Relief Request

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 2 of 27 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400/Renewed License No. NPF-63 Relief Request I3R-15 Revision and Supplement Reactor Vessel Closure Head Nozzle Repair Technique Inservice Inspection Program - Third Ten-Year Interval Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

Alternative Provides Acceptable Level of Quality and Safety

1.

ASME Code Components Affected Components:

Reactor Vessel Closure Head (RVCH) Penetration Nozzles Code Class:

Class 1 Examination Category:

B-P Code Item Number:

B4.20 (Code Case N-729-1, Alternative Examination Requirements for PWR Reactor Vessel Upper Heads with Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1)

==

Description:==

RVCH Penetration Nozzles Size:

4 Inch Nominal Outside Diameter Material:

Inconel SB-167

2.

Applicable Code Edition and Addenda

Shearon Harris Nuclear Power Plant, Unit No. 1 (HNP), Inservice Inspection Program (ISI) - Third Interval American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Section Xl, 2001 Edition through 2003 Addenda Shearon Harris Nuclear Power Plant, Unit No. 1, RVCH Code of Construction American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section III, 1971 Edition through Winter 1971 Addenda

3.

Applicable Code Requirements ASME Code, Section Xl, 2001 Edition through 2003 Addenda IWA-4221(b) states:

An item to be used for repair/replacement activities shall meet the Construction Code specified in accordance with (1), (2) or (3) below.

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 3 of 27 ASME Code, Section Xl, 2001 Edition through 2003 Addenda IWA-4221(c) states in part:

As an alternative to (b) above, the item may meet all or portions of the requirements of different Editions and Addenda of the Construction Code, or Section IIIprovided the requirements of IWA-4222 through IWA-4226, as applicable, are met..

ASME Code, Section Xl, 2001 Edition through 2003 Addenda, IWA-4400 provides welding, brazing, metal removal, and installation requirements related to repair/replacement activities.

ASME Code, Section Xl, 2001 Edition through 2003 Addenda IWA-4411 states:

Welding, brazing, and installation shall be performed in accordance with the Owners Requirements and, except as modified below, in accordance with the Construction Code of the item.

ASME Code, Section Xl, 2001 Edition through 2003 Addenda IWA-4411(a) states in part:

Later editions and addenda of the Construction Code, or a later different Construction Code, either in its entirety or portions thereof, and Code Cases may be used, provided the substitution is as listed in IWA-4221(c).

ASME Code, Section Xl, 2001 Edition through 2003 Addenda IWA-4610(a) states in part:

Thermocouples and recording instruments shall be used to monitor the process temperatures.

Code Case N-638-1, Similar and Dissimilar Metal Welding Using Ambient Temperature Machine GTAW Temper Bead Technique, provides requirements for automatic or machine gas tungsten arc welding (GTAW) of Class 1 components without the use of preheat or postweld heat treatment.

Code Case N-638-1 paragraph 3.0(d) states:

The maximum interpass temperature for field applications shall be 350° F regardless of the interpass temperature during qualification.

Code Case N-638-1 paragraph 4.0(b) states:

The final weld surface and the band around the area defined in paragraph 1.0(d) shall be examined using a surface and ultrasonic 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 />. The ultrasonic examination shall be in accordance with Appendix I.

Code Case N-729-1, Alternative Examination Requirements for PWR Reactor Vessel Upper Heads with Nozzles Having Pressure-Retaining Partial-Penetration WeldsSection XI, Division 1, Fig. 2, Examination Volume for Nozzle Base Metal and Examination Area for Weld and Nozzle Base Metal, is applicable to the RVCH nozzle penetrations.

ASME Code, Section Xl, 2001 Edition through 2003 Addenda IWA-4611.1(a) states:

Defects shall be removed in accordance with IWA-4422.1. A defect is considered removed when it has been reduced to an acceptable size.

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 4 of 27 ASME Code, Section Xl, 2001 Edition through 2003 Addenda, IWA-3300 specifies requirements for characterization of flaws detected by inservice examination.

ASME Code, Section Xl, 2001 Edition through 2003 Addenda, 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-3500.

ASME Code, Section Xl, 2001 Edition through 2003 Addenda IWB-3132.3 states:

A component whose volumetric or surface examination detects flaws that exceed the acceptance standards of Table IWB-3410-1 is acceptable for continued service without a repair/replacement activity if an analytical evaluation, as described in IWB-3600, meets the acceptance criteria of IWB-3600. The area containing the flaw shall be subsequently reexamined in accordance with IWB-2420 (b) and (c).

4.

Reason for Request

Flaw indications requiring repair were detected during ISI program ultrasonic (UT) examination of the HNP RVCH nozzle penetrations. The flaws are in the tube outside diameter (OD) surface extending inward toward the tube inside diameter (ID) and are axially oriented at the lower toe side of the weld. Three nozzles (14, 18, and 23) will be repaired under this request. Figure 10 shows the location of the axial indications and Figure 11 shows the relative locations of the nozzles on the RVCH. Table 1 provides sizing and characterization information on the flaws leading to the repair activities.

HNP inspects the RVCH each refueling outage. In the event that flaws are detected during future inspections, HNP would be required to resolve the inspection issues prior to startup from the outage. Relief is also requested at this time for repairs to nozzles discovered during future inspections.

The repair technique is intended to be the same as was used previously for nozzles 5, 17, 37, 38, 49, and 63, which is sometimes referred to as a half-nozzle repair. The half-nozzle repair involves machining away the lower section of the nozzle containing the flaw, then welding the remaining portion of the nozzle to the RVCH to form the new pressure boundary. This technique requires relief from certain aspects of the ASME Boiler and Pressure Vessel code as described below. Approval of this request will permit repair of nozzles 14, 18, and 23 and would preclude the need for an emergent request on an expedited schedule in the event that a flaw is discovered and repair required prior to startup at some time in the future.

Because of the risk of damage to the RVCH material properties or dimensions, it is not feasible to apply the post welding heat treatment requirements of the original Construction Code. As an alternative to the requirements of the RVCH Code of Construction, ASME Section III, 1971 including Addenda through Winter 1971, HNP proposes to perform the repair of the RVCH nozzle penetration utilizing the Inside Diameter Temper Bead (IDTB) welding method to restore the pressure boundary of the degraded nozzle penetration. The IDTB welding method is performed with a remotely operated weld tool, utilizing the machine GTAW process and the ambient temperature temper bead method with 50° F minimum preheat temperature and no post weld heat treatment. The repair will be performed in accordance with the 2001 Edition

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 5 of 27 through the 2003 Addenda of ASME Section XI, Code Case N-638-1, Code Case N-729-1, and the alternatives discussed below.

Basic steps for the IDTB repair are:

1. Removal of lower portion of existing Thermal Sleeve Assembly, if present, to provide access for IDTB weld repair.

2. Roll expansion above the area of repair. This stabilizes the nozzle to prevent any movement when the nozzle is separated from the nozzle to RVCH J-groove weld.

3. Machining to remove the nozzle to above the J-groove weld eliminating the portions of the nozzle containing the unacceptable indication. This machining operation also establishes the weld prep area (Refer to Figure 1).

4. Liquid penetrant (PT) examination of the machined area (Refer to Figure 3).

5. Welding the remaining portion of the nozzle to the RVCH using primary water stress corrosion cracking (PWSCC) resistant Alloy 52M weld material (Refer to Figure 2). Alloy 82 weld material may be used at the interface between the Alloy 182 existing weld and the Alloy 52M new weld if necessary.

6. Machining the weld and nozzle to provide a surface suitable for nondestructive examination (NDE).

7. PT and UT examination of the weld and adjacent area (Refer to Figure 3).

8. Optional - abrasive water jet machining (AWJM) remediation on the portion of the remaining nozzle most susceptible to PWSCC. The AWJM process removes a small amount of material thickness while imposing compressive residual stress on the nozzle surface.

9. PT examination of the weld and adjacent area, if AWJM performed.

10. Welding of new Lower Thermal Sleeve Assembly, if the location has a thermal sleeve.

Note that the figures included in this request are provided to assist in clarifying the description above. The location of the weld relative to the inner and outer radii of the head and the existing J-groove weld will vary depending upon the location of the nozzle and as-found dimensions.

HNP has determined that repair of the RVCH nozzle penetrations utilizing the alternatives specified in this request will provide an acceptable level of quality and safety. Relief is requested in accordance with 10 CFR 50.55a(z)(1).

5.

Proposed Alternative and Basis for Use

a. Monitoring of Interpass Temperature Code Case N-638-1 paragraph 3.0(d) states:

The maximum interpass temperature for field applications shall be 350° F regardless of the interpass temperature during qualification.

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 6 of 27 Code Case N-638-1 states that all other requirements of IWA-4000 must be met when using this Case. IWA-4610(a) requires that thermocouples and recording instruments be used to monitor process temperatures. Direct interpass temperature measurement is impractical to perform during welding operations from inside the RVCH nozzle penetration bore. The maximum interpass temperature will be determined by one of the following methods:

(1)

Heat-flow calculations, using at least the variables listed below.

(a)

Welding heat input (b)

Initial base material temperature (c)

Configuration, thickness, and mass of the item being welded (d)

Thermal conductivity and diffusivity of the materials being welded (e)

Arc time per weld pass and delay time between each pass (f)

Arc time to complete the weld (2)

Measurement of the maximum interpass temperature on a test coupon that is no thicker than the item to be welded. The maximum heat input of the welding procedure shall be used in welding the test coupon.

This methodology is consistent with the associated requirements specified in Code Case N-638-2 and subsequent versions. Alternatives to Code Case N-638-1 interpass temperature monitoring requirements have been previously approved by the NRC for dissimilar metal weld overlays in HNP Inservice Inspection Relief Request I3R-1, ADAMS Accession Number ML072760737.

HNP requests relief from using thermocouples and recording instruments to verify process temperatures.

Method 1, the use of heat flow calculations, will be used to determine a conservative maximum anticipated interpass temperature to ensure interpass temperature limits are not exceeded. In the IDTB repair scenario, the maximum heat input of 32,200 Joules per inch with an average time of 1 minute between subsequent weld passes results in a calculated base material temperature increase of approximately 6° F. Based on AREVAs experience with over 128 IDTB reactor vessel head nozzle repairs, the typical time between weld passes will be significantly greater than a minute as a result of weld sequencing, viewing previously deposited weld passes, completing paperwork, independent verifications, and routine equipment maintenance including tungsten electrode replacement.

b. Acceptance Examination Area Code Case N-638-1 paragraph 4.0(b) states in part:

The final weld surface and the band around the area defined in paragraph 1.0(d) shall be examined using a surface and ultrasonic methods Code Case N-638-1 paragraph 1.0(d) defines the area requiring preheat, and therefore examination, as the area to be welded and the band around the area of at least 1.5 times the component thickness or five inches, whichever is less.

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 7 of 27 The band includes an annular area extending five inches around the penetration bore on the inside surface of the RVCH. The purpose for the examination of the band is to ensure all flaws associated with the weld repair area have been removed, or addressed, since these flaws may be associated with the original flaw and may have been overlooked. For this modification, the repair welding is performed remote from the known flaw.

The band around the area defined in paragraph 1.0(d) cannot be examined due to the physical configuration of the partial penetration weld. The alternative final examination of the new weld and immediate surrounding area within the bore will be sufficient to verify that defects have not been induced in the low alloy steel RVCH material due to the welding process and will assure integrity of the nozzle and the new weld. Figure 3 identifies the areas for PT and UT examination of the modified nozzle penetration. UT examination will be performed by scanning from the inner diameter surface of the weld. The UT examination is qualified to detect construction type flaws in the new weld and base metal interface beneath the new weld. UT examination acceptance criteria will be in accordance with ASME Section III, 2001 Edition, including Addenda through 2003, NB-5330. The extent of the examination is consistent with Construction Code requirements.

Scanning is performed from the inside surface of the new weld and the adjacent portion of the nozzle, excluding the weld taper. The volume of interest for UT examination extends from at least one inch above the new weld and into the RVCH low alloy steel base material beneath the weld, to at least one-quarter inch depth. The PT examination area includes the weld surface and extends upward on the nozzle inside surface to include the area required by Code Case N-729-1, Figure 2, and at least one-half inch below the new weld. Figure 3 of this request identifies the area for PT examination of the modified nozzle penetration after machining and before welding.

ASME Section III, 2001 Edition including Addenda through 2003, NB-5245, specifies progressive surface examination of partial penetration welds. The original Construction Code requirement for progressive PT examination, in lieu of volumetric examination, was because volumetric examination is not practical for the conventional partial penetration weld configurations. For this modification, the weld, except for the taper transition, is suitable for UT examination and a final surface PT examination can be performed as shown in Figure 3. Liquid penetrant examination will be performed on the entire weld, including the taper transition. In addition, 70L and 45L axial UT examination scans looking down (see Figures 5 and 7) will interrogate the taper transition volume. The performance of the surface and UT examinations provides assurance of structural integrity.

Code Case N-638-1, paragraph 4.0(b) requires that the specified volumetric examination be in accordance with Section XI, Appendix I. Paragraph 4.0(e) specifies acceptance criteria to be in accordance with IWB-3000.

ASME Code, Section Xl, 2001 Edition through 2003 Addenda, IWB-3000 does not have any acceptance criteria that directly apply to the partial penetration weld configuration. Regulatory Guide 1.147, Rev. 15, has conditionally approved Code Case N-638-1 with the condition that UT volumetric examinations be performed with personnel and procedures qualified for the repaired volume and qualified by demonstration using representative samples which contain construction type flaws. The acceptance criteria of NB-5330, in ASME Section III, 2001 Edition through 2003 Addenda, will apply to all flaws identified within the repaired volume.

ASME Section III, 2001 Edition including Addenda through 2003, NB-5245 requires incremental and final surface examination of partial penetration welds. Due to the welding layer deposition

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 8 of 27 sequence (i.e., each layer is deposited parallel to the penetration centerline), the specific requirements of NB-5245 cannot be met. The Construction Code requirement for progressive surface examination is because volumetric examination is not practical for conventional partial penetration weld configurations. For this modification, the repair weld is suitable, except for the taper transition, for UT examination and a final surface examination.

The final examination of the repair weld and immediate surrounding area will be sufficient to verify that defects have not been induced in the ferritic low alloy steel RVCH base material due to the welding process. PT examination coverage is shown in Figure 3. UT examination will be performed scanning from the inside surface of the weld, excluding the transition taper portion at the bottom of the weld, and adjacent portion of the nozzle bore. The UT examination is qualified to detect flaws in the new weld and base metal interface in the repair region, to the maximum practical extent.

The UT transducers and delivery tooling are capable of scanning from cylindrical surfaces with inside diameters near 2.75 inches. The UT equipment is not capable of scanning from the face of the weld taper. The scanning is performed using 0° L-wave, 45° L-wave, and 70° L-wave transducers. Approximately 70% of the weld surface will be scanned by UT. Approximately 83%

of the RVCH ferritic steel heat affected zone will be covered by UT. The UT examination coverage volumes are shown in Figures 4 through 8 for the various scans.

The repair weld produces a region that limits the examination volume. The downward aimed angle beam transducers (45L and 70L) are used to interrogate this area for defects (planar defects normal to the beam, cracking, lack-of-fusion, etc.). The UT is being performed in addition to the surface examinations. There is no portion of the repair volume that does not receive at least single direction UT coverage. The actual volume examined will be calculated after the as-built dimensions of the weld are known and the examination is performed. It is anticipated that greater than 80% of the examination volume will obtain two-directional coverage.

PT examination will be performed on the entire surface area. In addition, the volume in question will be examined to the extent practical using the 70L and 45L (see Figures 5 and 7) axial UT examination scans (looking down). There is no portion of the repair that does not receive surface liquid penetrant examination and at least single-direction UT coverage of the volume.

Examination of the area depicted in Figure 3 will assure that all unacceptable flaws associated with the weld repair area have been removed.

HNP requests relief from examination of the area defined in Code Case N-638-1, paragraph 1.0(d).

c. 48 Hour Hold Code Case N-638-1 paragraph 4.0(b) states in part:

The final weld surface and the band around the area defined in paragraph 1.0(d) shall be examined using a surface and ultrasonic 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 /> Hydrogen cracking is a form of cold cracking. It is produced by the action of internal tensile stresses acting on low toughness heat affected zones. The internal stresses are produced from localized build-ups of monatomic hydrogen. Monatomic hydrogen forms when moisture or hydrocarbons interact with the welding arc and molten weld pool. The monatomic hydrogen can

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 9 of 27 be entrapped during weld solidification and tends to migrate to transformation boundaries or other microstructure defect locations. As concentrations build, the monatomic hydrogen recombines to form molecular hydrogen - thus generating localized internal stresses at these internal defect locations. If these stresses exceed the fracture toughness of the material, hydrogen induced cracking occurs. This form of cracking requires the presence of hydrogen and low toughness materials. It is manifested by intergranular cracking of susceptible materials and normally occurs within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of welding.

The machine GTAW process is inherently free of hydrogen. Unlike the shielded metal arc welding process, GTAW filler metals do not rely on flux coverings that may be susceptible to moisture absorption from the environment. Conversely, the GTAW process utilizes dry inert shielding gases that cover the molten weld pool from oxidizing atmospheres. Any moisture on the surface of the component being welded is vaporized ahead of the welding torch. The vapor is prevented from being mixed with the molten weld pool by the inert shielding gas that blows the vapor away before it can be mixed. Furthermore, modern filler metal manufacturers produce wires having very low residual hydrogen. This is important because filler metals and base materials are the most realistic sources of hydrogen for the automatic or machine GTAW temper bead welding. Therefore, the potential for hydrogen-induced cracking is greatly reduced by using the machine GTAW process. Extensive research has been performed by EPRI. EPRI Report 1013558, Temperbead Welding Applications, 48 Hour Hold Requirements for Ambient Temperature Temperbead Welding (ADAMS Accession Number ML070670060), provides justification for starting the 48-hour hold after completing the third temper bead weld layer rather than waiting for the weld to cool to ambient temperature.

HNP requests relief from commencing the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> hold period when the weld reaches ambient temperature. The 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> hold period will commence upon completion of the third weld layer.

This approach has been previously considered by the NRC staff in the conditional approval of N-638-4 in Rev. 16 of Regulatory Guide 1.147 when using austenitic materials and for dissimilar metal weld overlays in the approval of HNP Relief Request I3R-1, ADAMS Accession Number ML072760737.

d. Triple Point Anomaly ASME Section III, 2001 Edition including Addenda through 2003, NB-5330(b) states:

Indications characterized as cracks, lack of fusion, or incomplete penetrations are unacceptable regardless of length.

An artifact of ambient temperature temper bead welding is an anomaly in the weld at the triple point. The triple point is the point in the repair weld where the low alloy steel RVCH base material, the Alloy 600 nozzle, and the Alloy 52M weld intersect. The location of the triple point anomaly is shown in Figure 2. This anomaly consists of an irregularly shaped very small void.

Mock-up testing has verified that the anomalies are common and do not exceed 0.10 inches in length and are assumed to exist, for purposes of analysis, around the entire bore circumference at the triple point elevation.

A fracture mechanics analysis has been performed for the design configuration to provide justification, in accordance with Section XI, for operating with the postulated triple point anomaly. The anomaly is modeled as a 0.10 inch, circular crack-like defect, extending 360 degrees around the circumference at the triple point location, considering the most susceptible material for propagation. Postulated flaws could be oriented within the anomaly such that there are two possible flaw propagation paths, as discussed below.

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 10 of 27 Path 1: Flaw propagation is across the nozzle wall thickness from the OD to the ID of the nozzle housing (analysis paths 1 & 2).

This is the shortest path through the new Alloy 52M weld material. By using a fatigue crack growth rate twice that of the rate of Alloy 600 material, it is ensured that another potential path through the heat affected zone between the new repair weld and the Alloy 600 nozzle material is also bounded.

For completeness, two types of flaws are postulated at the outside surface of the nozzle IDTB repair weld. A 360 degree continuous circumferential flaw, lying in a horizontal plane, is considered to be a conservative representation of crack-like defects that may exist in the weld triple point anomaly. This flaw is subjected to axial stresses in the nozzle. An axially oriented semi-circular outside surface flaw is also considered since it would lie in a plane normal to the higher circumferential stresses. Both of these flaws would propagate toward the inside surface of the nozzle.

Path 2: Flaw propagation extends down the outside surface of the repair weld between the weld and the RVCH (analysis paths 3 through 6).

A cylindrically oriented flaw is postulated to lie along this interface, subjected to radial stresses with respect to the nozzle. This flaw may propagate through either the new Alloy 52M weld material or the low alloy steel RVCH base material.

The results of the analyses demonstrate that the 0.10 inch weld anomaly is acceptable for a 40 year design life of the HNP nozzle repair. The minimum fracture toughness margins for flaw propagation Paths 3 through 6 have been shown to be acceptable compared to the required margins of 10 for normal/upset conditions and 2 for emergency/faulted (and test) conditions per Section XI, IWB-3612. A limit load analysis was performed considering the ductile weld repair material along flaw propagation Path 1 & 2. The analysis showed that for the postulated circumferential flaw the minimum margin on allowable stress is 1.43. For the axial flaw the minimum margin on allowable flaw depth is 3.9. Fracture toughness margins have also been demonstrated for the postulated cylindrical flaws. For the cylindrical flaws, it is shown that the applied shear stress at the remaining ligament is less than the allowable shear stress per NB-3227.2.

The final crack size (length and depth) in the axial and circumferential direction at the end of 40 years is:

Axial flaw: final depth (af) is 0.1008 inch, since length/depth is 2, length = 0.202 inch.

Circumferential flaw: the final flaw depth of the 360° circumferential flaw is 0.1002 inch.

The final crack sizes are acceptable based on ASME Code,Section XI, IWB-3640 flaw evaluations which demonstrate that the final flaw sizes satisfy the applicable Code acceptance criteria, as discussed below.

For flaws in the IDTB weld, the applicable section is IWB-3640. Following the procedures in IWB-3641 and acceptance criteria of IWB-3642 the flaw evaluation based on Appendix C is performed.

For the circumferential flaw, the stress margin is calculated per Article C-5000 of ASME Code Section XI.

The stress margin:

St/m = 1.43 where m is the membrane stress,

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 11 of 27 St =m c/SFm, where m c is the critical membrane stress, and SFm is the safety factor of 2.7 per C-2620 For axial flaws, the calculated stress ratio (SFm h/f) is 0.519 and the nondimensional flaw length is 0.211. Thus the allowable flaw size (a/t) determined from Table C-5410-1 of ASME Code Section XI is 0.75 and allowable flaw depth is 0.395 inch. Thus the allowable flaw size margin, aallow/af= 3.9.

The margins of 1.43 for circumferential and 3.9 for axial flaws exceed the required margins of the ASME Code; therefore, the flaw evaluations demonstrate that the required margins of IWB-3600 are satisfied.

The fracture margin calculation includes the required safety factors and hence the required margin is only 1.0. Thus the calculated margins, 1.43 for circumferential flaws and 3.9 for axial flaws, are acceptable.

This evaluation is prepared in accordance with ASME Section XI and demonstrates that for the intended service life of the repair, the fatigue crack growth is acceptable and the crack-like indications remain stable. This satisfies the ASME Section XI criteria but does not include considerations of stress corrosion cracking such as PWSCC. Since the crack-like defects due to the weld anomaly are not exposed to the primary coolant and the air environment is benign for the materials at the triple point, the time-dependent crack growth rates from PWSCC are not applicable.

Relief is requested to permit anomalies, as described herein, at the triple point area to remain in service.

e. Flaw Characterization and Successive Examinations - RVCH Original J-Groove Weld The assumptions of IWB-3600 are that cracks are fully characterized in order to compare the calculated parameters to the acceptable parameters addressed in IWB-3500. The original nozzle-to-RVCH J-groove weld is extremely difficult to examine with UT due to the compound curvature and fillet radius around the nozzle circumference. These conditions preclude UT coupling and control of the sound beam needed to perform flaw sizing with reasonable confidence in the measured flaw dimensions. Therefore, it is impractical to characterize the flaw geometry that may exist therein. As these J-groove welds have not been fully examined with qualified techniques, they are assumed to have unacceptable flaws.

A flaw in the J-groove weld cannot be sized by currently available nondestructive examination techniques. It is conservatively assumed that the as-left condition of the remaining J-groove weld includes flaws extending through the entire Alloy 82/182 J-groove weld and butter material.

It is further postulated that the dominant hoop stresses in the J-groove weld would create a situation where the preferential direction for cracking would be radial. A radial crack in the Alloy 82/182 weld metal would propagate by PWSCC, through the weld and butter, to the interface with the low alloy steel head material, where it would blunt, or arrest. Any growth of the postulated as-left flaw into the low alloy steel head would be by fatigue crack growth under cyclic loading conditions.

The J-groove flaws have been evaluated for acceptance in accordance with the analytical evaluation requirements of IWB-3132.3 using worst-case postulated flaw sizes. The results of this evaluation show that, based on a combination of linear elastic and elastic-plastic fracture mechanics analysis of a postulated remaining flaw in the original Alloy 182 J-groove weld and

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 12 of 27 butter material, the HNP RVCH nozzle repair design configuration is considered to be acceptable for 30 years of operation following an IDTB weld repair.

Linear-elastic (LEFM) and elastic-plastic (EPFM) fracture mechanics analyses were used to demonstrate that the remaining worst-case as-left J-groove flaw would be acceptable for 30 years of service. Although the postulated flaw did not satisfy ASME Code Section XI IWB-3612 for all transient loading conditions, LEFM analysis determined that the uphill side of the reactor head penetration was the worst case position for the postulated flaw, calculated the final flaw size by fatigue crack growth, and identified the controlling service conditions for evaluation by EPFM.

For normal and upset conditions, the controlling loading condition was identified to be a reactor trip, for which it was shown, using safety factors of 1.5 on primary loads and 1.0 on secondary loads, that the applied J-integral (0.785 kips/in) was less than the J-integral of the low alloy steel head material (2.473 kips/in) at a crack extension of 0.1 inch. For emergency and faulted conditions, the controlling loading condition was a large loss of coolant accident, for which it was shown that with safety factors of 1.5 on primary loads and 1.0 on secondary loads that the applied J-integral (2.359 kips/in) was less than the J-integral of the low alloy steel head material (2.474 kips/in) at a crack extension of 0.1 inch. Flaw stability during ductile flaw growth was easily demonstrated for both loading conditions using safety factors of 3.0 and 1.5 for the reactor trip and 1.5 and 1.0 for the large loss of coolant accident.

It is likely that the flaws detected by UT examination would be removed when the lower portion of the nozzle is machined away from the J-groove weld. However, as discussed above, flaws are postulated to exist in the remaining portion of the J-groove weld and shown in the evaluation to be acceptable for 30 years of service.

Successive examinations required by IWB-3132.3 will not be performed because analytical evaluation of the worst-case postulated flaw is performed to demonstrate the acceptability of continued operation. A reasonable assurance of the RVCH structural integrity is maintained without the successive examination by the fact that evaluation has shown the worst case flaw to be acceptable for continued operation.

Relief is requested from flaw characterization and subsequent examination requirements.

The potential for debris from a cracking J-groove partial penetration weld was considered.

Radial cracks were postulated to occur in the weld due to the dominance of hoop stresses at this location. This possibility of occurrence of transverse cracks that could intersect the radial cracks is considered remote. There are no forces that would drive a transverse crack. The radial cracks would relieve the potential transverse crack driving forces. Hence it is unlikely that a series of transverse cracks could intersect a series of radial cracks resulting in any fragments becoming dislodged.

f. Inservice Inspections Code Case N-729-1 provides requirements for the inservice inspection of RVCHs with nozzles having partial penetration welds. Code Case N-729-1 Table 1, Item 4.20, permits either volumetric or surface examination. Item 4.20 examination requirements are specified in Figure 2 of Code Case N-729-1. The repair proposed by this relief request removes much of the examination area depicted in this figure at several locations. Figure 9 of this relief request will be used to establish the examination area for the preservice inspection following repair and for

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 13 of 27 future inservice inspections. This examination area is equivalent to that required by Figure 2 in Code Case N-729-1, as it examines the nozzle weld and the same area above the nozzle weld as would be required by Figure 2 in the Code Case.

Therefore, preservice inspection following repair and future inservice inspections will comply with Code Case N-729-1 as modified by 10 CFR 50.55a(g)(6)(ii)(D) and as depicted in Figure 9.

g. General Corrosion Impact on Exposed Low Alloy Steel The IDTB nozzle repair leaves a small portion of low alloy steel in the RVCH exposed to primary coolant. An evaluation was performed for the potential corrosion concerns at the RVCH low alloy steel (LAS) wetted surface. Galvanic corrosion, hydrogen embrittlement, SCC, and crevice corrosion are not expected to be a concern for the exposed LAS base metal. General corrosion of the exposed LAS base metal will occur in the area between the IDTB weld and the J-groove weld. The general corrosion rate is conservatively estimated to be 0.0036 inch/year. The corrosion of the exposed base metal has negligible impact on the RVCH and is acceptable for 40 years from the time the modification is installed.

CONCLUSIONS Implementation of an IDTB repair to the RVCH nozzle penetration will produce an effective repair that will restore and maintain the pressure boundary integrity of the HNP RVCH. Similar repairs have been performed successfully and have been in service for several years without any known degradation. Any repairs to RVCH nozzles using the subject techniques will occur as design change plant modifications in accordance with the HNP Quality Assurance Program.

This will ensure that the assumptions of the calculations supporting this request and any conditions identified in the Safety Evaluation are satisfied. The alternative provides improved structural integrity and reduced likelihood of leakage for the primary system. Accordingly, the use of the alternative provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1).

6.

Duration of Proposed Alternatives The analyses described above and others in the modification that will be implemented under 10 CFR 50.59 support a design life expectancy of 2.2 effective full power years (EFPY). The structural and fracture mechanics analyses results are based upon expected repair parameters which may vary during implementation. The design lifetime is sensitive to the length of the Alloy 52M weld ligament, and the actual limiting ligament length may vary depending upon the as-found and as-left conditions. The design life will be re-evaluated if necessary using as-built data and incorporated into the modification, future NDE inspection schedules, and asset management plans. HNP will examine all repaired RVCH penetration nozzles every refueling outage in accordance with ASME Code Case N-729-1 as conditioned by 10 CFR 50.55a(g)(6)(ii)(D). The periodic examinations will provide reasonable assurance of the structural integrity of RVCH nozzles prior to exceeding the design life of the repair.

The 2.2 EFPY life takes no credit for a delay in PWSCC initiation time, and is based on PWSCC crack growth of an undetected flaw in the remaining Alloy 600 nozzle. It is estimated to take 2.2 EFPY for the crack to propagate from an undetected flaw of 10% of the Alloy 600 nozzle wall thickness to 75% of the original Alloy 600 nozzle wall thickness. Thus, ISI examinations at

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 14 of 27 subsequent refueling outages showing no PWSCC flaw initiation will support an additional 2.2 EFPY of operation.

The 30 year life is predicted based on the as-left J-groove flaw evaluation. The 2.2 EFPY is based on a separate PWSCC evaluation in the exposed original Alloy 600 nozzle. The overall acceptable life of the repair design is based on the most limiting life predicted amongst the weld anomaly analysis, the as-left J-groove analysis and the PWSCC evaluation of the original Alloy 600 nozzle, which is 2.2 EFPY.

The provisions of this relief are applicable to the third ten-year inservice inspection interval for HNP which commenced on May 2, 2007 and will end on May 1, 2017. The repairs installed in accordance with the provisions of this relief shall remain in place for the design life of the repair, until another alternative is approved by the NRC, or until the RVCH is replaced.

7.

Additional Information

a. Mockup AREVA, in support of over 128 similar repairs, has performed many qualifications using mockups since the IDTB control rod drive mechanism nozzle repairs at Oconee Nuclear Station in 2001. During these repair evolutions, the site crew performs training on mockups for each of their respective specialties, i.e., machinists train on machining mockups, welders train on welding mockups, and NDE personnel train on NDE mockups.

An IDTB weld repair NDE mockup was fabricated to replicate the expected configuration. It contains a series of electrical-discharge machining (EDM) notches at the triple point to simulate the triple point anomaly at various depths into the nozzle wall and cracking at the IDTB weld to low alloy steel interface. It also contains flat bottom holes drilled from the mockup outer diameter so that the hole is normal to the surface to simulate under bead cracking, lack of bond, and lack of fusion.

An Inconel calibration block is used and contains a series of EDM notches at nominal depths of 10%, 25%, 50%, and 75% deep from both ID and OD surfaces in both the axial and circumferential orientation. The block also contains 1/4T, 1/2T, and 3/4T deep end holes and side drilled holes that are used for calibration.

This is the same mockup used for the procedure qualification for the Davis Besse CRDM nozzle repairs in 2010.

b. ASME Code Case N-638-1 HNP adopted ASME Code Case N-638-1 in the Third Interval Inservice Inspection Program submittal to the NRC as HNP-08-038 (ADAMS Accession No. ML081330463). Later revisions of the code case have not been adopted.

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 15 of 27

8.

Precedents

1. Davis-Besse Nuclear Power Station Relief Request RR-A34, April 1, 2010, ADAMS Accession Number ML100960276.

2. Calvert Cliffs Nuclear Power Plant Relief Request RR-PZR-0 1, January 31, 2011, ADAMS Accession Number ML110340059.

3. Shearon Harris Nuclear Power Plant, Unit 1, Relief Request I3R-09, October 2, 2012, ADAMS Accession Number ML12270A258.

4. Shearon Harris Nuclear Power Plant, Unit 1, Relief Request I3R-11, May 22, 2013, ADAMS Accession Number ML13238A154.

5. Shearon Harris Nuclear Power Plant, Unit 1, Relief Request I3R-13, November 22, 2013, ADAMS Accession Number ML13329A354.

9.

References

1. EPRI Report 1013558, Temperbead Welding Applications, 48 Hour Hold Requirements for Ambient Temperature Temperbead Welding, EPRI, Palo Alto, CA and Hermann &

Associates, Key Largo, FL, December 2006.

2. ASME Code Case N-638-1 Similar and Dissimilar Metal Welding Using Ambient Temperature Machine GTAW Temper Bead Technique,Section XI, Division 1.

3. NRC Regulatory Guide 1.147, Revision 15, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1.

4. NRC Regulatory Guide 1.147, Revision 16, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1

5. ASME Code Case N-729-1 Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1.

6. Shearon Harris Nuclear Power Plant, Unit 1, Relief Request I3R-13, November 22, 2013, ADAMS Accession Numbers ML13329A354 and ML13330A996.

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 16 of 27 Figure 1. Machining

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 17 of 27 Figure 2. Welding

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 18 of 27 Figure 3. Examination Areas Pre-Weld PT l-m-n-o-p-q Post-Weld PT m-n-s-p-q-r Post-Weld UT (Weld) a-b-c-d-e-h Post Weld UT (Nozzle Material) e-f-g-h

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 19 of 27 Figure 4. UT 0° and 45° L-wave Beam Coverage Looking Clockwise and Counter-clockwise

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 20 of 27 Figure 5. UT 45° L-wave Beam Coverage Looking Down

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 21 of 27 Figure 6. UT 45° L-wave Beam Coverage Looking Up

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 22 of 27 Figure 7. UT 70° L-wave Beam Coverage Looking Down

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 23 of 27 Figure 8. UT 70° L-wave Beam Coverage Looking Up

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 24 of 27 Figure 9. PSI and ISI Weld and Nozzle Base Metal Surface Examination Area (A-B-C-D)

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 25 of 27 Figure 10. Location of Axial Indication

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 26 of 27 Figure 11. Reactor Vessel Head Penetration Locations

U.S. Nuclear Regulatory Commission Revised Relief Request I3R-15 HNP-15-034 Enclosure 1 Page 27 of 27 Table 1. Flaw Characteristics Nozzle Ind.

No.

ID/OD Depth to Ind.

Thru Wall Length Azimuth (degrees)

Orientation Ax/Circ Type 14 1

OD 0.526 0.100 0.223 41.1 AXIAL PWSCC 18 1

OD 0.428 0.198 0.260 21.5 AXIAL PWSCC 23 1

OD 0.485 0.141 0.297 27.3 AXIAL PWSCC Notes:

1. Flaws are in the tube outside diameter (OD) extending inward toward the tube inside diameter (ID) and approximately parallel with the nozzle axis (axially oriented) at the lower toe side of the weld.

2. 0° Azimuth is the lowest point (downhill) on the nozzle. Progression is CCW looking up.

3. Tube diameter, OD 4.002", ID 2.750". Thickness, 0.626" Nom.

4. Dimensions are in inches.

5. Scans performed from the tube ID. Flaws are located at the OD.

U.S. Nuclear Regulatory Commission HNP-15-034 HNP-15-034 Shearon Harris Nuclear Power Plant, Unit No. 1 Docket No. 50-400/Renewed License No. NPF-63 Relief Request I3R-15 Revision and Supplement Reactor Vessel Closure Head Nozzle Repair Technique Inservice Inspection Program - Third Ten-Year Interval Affidavit Supporting Withholding of Proprietary Information

A F F I D A V I T COMMONWEALTH OF VIRGINIA

)

) ss.

CITY OF LYNCHBURG

)

1.

My name is Gayle Elliott. I am Manager, Product Licensing, for AREVA Inc.

(AREVA) and as such I am authorized to execute this Affidavit.

2.

I am familiar with the criteria applied by AREVA to determine whether certain AREVA information is proprietary. I am familiar with the policies established by AREVA to ensure the proper application of these criteria.

3.

I am familiar with the AREVA information contained in Engineering Information Records 51-9176114-002 entitled, Corrosion Evaluation of Shearon Harris RV Head Penetration IDTB Weld Repair, dated April 2015, and 51-9176115-003, entitled, PWSCC Assessment of the Alloy 600 Nozzle in the Shearon Harris RV Head Penetration IDTB Weld Repair, dated April 2015 and referred to herein as Documents. Information contained in these Documents has been classified by AREVA as proprietary in accordance with the policies established by AREVA Inc. for the control and protection of proprietary and confidential information.

4.

These Documents contain information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in these Documents as proprietary and confidential.

5.

These Documents have been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in these Documents be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure

is requested qualifies under 10 CFR 2.390(a)(4) Trade secrets and commercial or financial information.

6.

The following criteria are customarily applied by AREVA to determine whether information should be classified as proprietary:

(a)

The information reveals details of AREVAs research and development plans and programs or their results.

(b)

Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c)

The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA.

(d)

The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA in product optimization or marketability.

(e)

The information is vital to a competitive advantage held by AREVA, would be helpful to competitors to AREVA, and would likely cause substantial harm to the competitive position of AREVA.

The information in these Documents is considered proprietary for the reasons set forth in paragraphs 6(c), 6(d) and 6(e) above.

7.

In accordance with AREVAs policies governing the protection and control of information, proprietary information contained in these Documents has been made available, on a limited basis, to others outside AREVA only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8.

AREVA policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9.

The foregoing statements are true and correct to the best of my knowledge, information, and belief.

68%6&5,%('before me this day of 2015.

Danita R. Kidd NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 12/31/16 Reg. # 205569 0° *UI:A 0 0

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U.S. Nuclear Regulatory Commission HNP-15-034 Proprietary Information Withhold from Public Disclosure Under 10 CFR 2.390 Trade Secrets and Commercial or Financial Information HNP-15-034 Shearon Harris Nuclear Power Plant, Unit No. 1 Docket No. 50-400/Renewed License No. NPF-63 Relief Request I3R-15 Revision and Supplement Reactor Vessel Closure Head Nozzle Repair Technique Inservice Inspection Program - Third Ten-Year Interval Proprietary Calculations Calculation 51-9176114-002 Corrosion Evaluation of Shearon Harris RV Head Penetration IDTB Weld Repair (Proprietary)

Calculation 51-9176115-003 PWSCC Assessment of the Alloy 600 Nozzle in the Shearon Harris RV Head Penetration IDTB Weld Repair (Proprietary)

U.S. Nuclear Regulatory Commission HNP-15-034 HNP-15-034 Shearon Harris Nuclear Power Plant, Unit No. 1 Docket No. 50-400/Renewed License No. NPF-63 Relief Request I3R-15 Revision and Supplement Reactor Vessel Closure Head Nozzle Repair Technique Inservice Inspection Program - Third Ten-Year Interval Redacted/Non-Proprietary Calculations Calculation 51-9215673-001 Corrosion Evaluation of Shearon Harris RV Head Penetration IDTB Weld Repair (Non-Proprietary)

Calculation 51-9215674-001 PWSCC Assessment of the Alloy 600 Nozzle in the Shearon Harris RV Head Penetration IDTB Weld Repair (Non-Proprietary)

20004-021 (01/30/2014)

Page 1 of 22 AREVA Inc.

Engineering Information Record Document No.:

51

- 9215673 -

001 Corrosion Evaluation of Shearon Harris RV Head Penetration IDTB Weld Repair (NonProprietary)

AREVA INC. PROPRIETARY This document and any information contained herein is the property of AREVA Inc. (AREVA) and is to be considered proprietary and may not be reproduced or copied in whole or in part. This document shall not be furnished to others without the express written consent of AREVA and is not to be used in any way which is or may be detrimental to AREVA. This document and any copies that may have been made must be returned to AREVA upon request.

Controlled Document

Controlled Document

20004-021 (01/30/2014)

Document No.: 51-9215673-001 Corrosion Evaluation of Shearon Harris RV Head Penetration IDTB Weld Repair (NonProprietary)

Page 3 Record of Revision Revision No.

Pages/Sections/

Paragraphs Changed Brief Description / Change Authorization 000 All Original Issue. The corresponding proprietary version is in AREVA document 51-9176114-001.

001 All The corresponding proprietary document is 51-9176114-002.

001 Throughout Updated form to most recent revision.

001 Throughout This revision includes incorporation of document amendment 159-9216054-000.

001 Section 1.0 Included a sentence about previously repaired nozzles.

Included the purpose of Revision 1.

001 Section 2.0 Clarification about AWJM being optional.

001 Section 2.0 Changed Progress Energy to Duke Energy.

001 Figure 2-1 Added word unrepaired to title of figure.

001 Figure 2-3 Removed word no from title of figure.

001 Figure 2-2 Included new Reference 2 and the word optional in the title of figure.

001 Section 4.0 Included sentence about shutdown conditions.

001 Section 4.1.2 Included discussion of and reference to cross flow velocities.

001 Section 4.4 Included reference for first sentence.

001 Section 5.1 Included clarification of results with respect to AWJM.

Included note about Dmax.

001 Table 5-1 Included results to reflect new revision of drawing.

001 Section 7.0 Included discussion of AWJM at the end of the second paragraph.

001 Section 8.0 Updated Reference 1 to Rev. 3 001 Section 8.0 New Reference 2, further references renumbered 001 Section 8.0 Updated Reference 3 to Rev. 6.

001 Section 8.0 Fixed error in date of Reference 7 001 Section 8.0 New Reference 21 Controlled Document

Document No.: 51-9215673-001 Corrosion Evaluation of Shearon Harris RV Head Penetration IDTB Weld Repair (NonProprietary)

Page 4 Table of Contents Page SIGNATURE BLOCK................................................................................................................................ 2 RECORD OF REVISION.......................................................................................................................... 3 LIST OF TABLES..................................................................................................................................... 5 LIST OF FIGURES................................................................................................................................... 6 1.0 PURPOSE..................................................................................................................................... 7

2.0 BACKGROUND

............................................................................................................................ 7 3.0 KNOWN OCCURANCES OF EXPOSED CARBON/LOW ALLOY STEEL BASE METAL......... 12 4.0 CORROSION OF EXPOSED LOW ALLOY STEEL AT LOCATION A....................................... 13 4.1 General Corrosion of Exposed Base Metal..................................................................... 13 4.1.1 Oxygen Level in the Repaired Area.................................................................. 13 4.1.2 General Corrosion Rate.................................................................................... 13 4.1.3 General Corrosion Rate in the HAZ.................................................................. 14 4.2 Crevice Corrosion of Exposed Base Metal...................................................................... 14 4.3 Galvanic Corrosion of Exposed Base Metal.................................................................... 14 4.4 Stress Corrosion Cracking of Exposed Base Metal........................................................ 15 4.5 Hydrogen Embrittlement of Exposed Base Metal............................................................ 15 5.0 ESTIMATE OF FE RELEASE RATE FROM EXPOSED LOW ALLOY STEEL AT LOCATION A.............................................................................................................................. 16 5.1 CRDM, CET, and Spare Nozzle Repairs........................................................................ 16 6.0 CORROSION OF ALLOY 52M................................................................................................... 19

7.0 CONCLUSION

S.......................................................................................................................... 20

8.0 REFERENCES

............................................................................................................................ 21 Controlled Document

Document No.: 51-9215673-001 Corrosion Evaluation of Shearon Harris RV Head Penetration IDTB Weld Repair (NonProprietary)

Page 5 List of Tables Page TABLE 5-1: ESTIMATED EXPOSED AREA OF LAS AND FE RELEASE RATE................................. 18 Controlled Document

Document No.: 51-9215673-001 Corrosion Evaluation of Shearon Harris RV Head Penetration IDTB Weld Repair (NonProprietary)

Page 6 List of Figures Page FIGURE 2-1: CURRENT CONFIGURATION OF CRDM/CET/SPARE NOZZLE AT HNP (REFERENCE 3, STEP 1)............................................................................................................................... 8 FIGURE 2-2: IDTB CRDM NOZZLE REPAIR CONFIGURATION WITH NO OPTIONAL REMOVAL OF WELD OVERLAP, INCLUDING REMEDIATION (SHOWN BEFORE REPLACEMENT THERMAL SLEEVE ATTACHMENT FOR CLARITY) (REFERENCE 3, STEP 6)................. 9 FIGURE 2-3: IDTB CRDM NOZZLE REPAIR CONFIGURATION WITH OPTIONAL REMOVAL OF WELD OVERLAP, NOT INCLUDING REMEDIATION (SHOWN BEFORE REPLACEMENT THERMAL SLEEVE ATTACHMENT FOR CLARITY) (REFERENCE 3, STEP 5A)............ 10 FIGURE 2-4: IDTB CRDM NOZZLE REPAIR CONFIGURATION [REFERENCE 3, STEP 7].............. 11 Controlled Document

Document No.: 51-9215673-001 Corrosion Evaluation of Shearon Harris RV Head Penetration IDTB Weld Repair (NonProprietary)

Page 7 1.0 PURPOSE The purpose of this document is to evaluate potential corrosion concerns arising from the final geometrical configuration of the proposed inner diameter temper bead (IDTB) weld repair of the control rod drive mechanism (CRDM), core exit thermocouple (CET), and/or spare nozzle penetrations due to potential degradation at these potentially repaired locations. Sixty-five (65) nozzle reactor vessel closure head (RVCH) penetrations, including fifty-two (52) CRDM nozzles with thermal sleeves, four (4) CET nozzles with no thermal sleeves, and nine (9) spare locations with no thermal sleeves are potential locations for this repair to be performed [1]. Six of these sixty-five nozzles (four (4) CRDM, one (1) CET, one (1) spare) have been repaired to-date using abrasive water jet machining (AWJM) remediation techniques [3]. The materials with potential corrosion concerns evaluated within this document include the low alloy steel RVCH exposed as well as the new Alloy 52M used to create a new weld during the proposed repair.

The purpose of Revision 1 is to reflect changes to the design specification [1] and drawing [2][3] making the use of the AWJM surface remediation technique in this repair optional.

2.0 BACKGROUND

In December 2000, inspections of the Alloy 600 CRDM nozzle penetrations in the RVCH at a domestic pressurized water reactor (PWR) identified leakage in the region of the partial penetration weld between the RVCH and the CRDM nozzle. This leakage, identified as the result of primary water stress corrosion cracking (PWSCC), was repaired using manual grinding and welding. In February 2001, the manual repair of several CRDM nozzles at another domestic PWR with similar defects resulted in extensive radiation dose to the personnel due to the location and access limitations. Therefore, the Babcock & Wilcox (B&W) Owners Group (BWOG) commissioned Framatome (now AREVA) to provide an automated repair process that was ultimately implemented at a third domestic PWR [1].

Due to concerns that similar degradation at CRDM, CET, and spare nozzle locations may have occurred at Shearon Harris Unit 1, Duke Energy has contracted AREVA to adapt this repair for Shearon Harris as a contingency. In the current unrepaired configuration (Figure 2-1 [3]), a leak of primary water through a PWSCC crack in a potentially susceptible Alloy 600 penetration or Alloy 82/182 J-groove weld could cause the buildup of boric acid on the outer portion of the RVCH. If required, the IDTB nozzle repair method shown in Figure 2-2, Figure 2-3, and Figure 2-4 will be used [3]. Figure 2-2 depicts the proposed repair method [

] and after the optional AWJM remediation [2] is performed. Figure 2-3 depicts the proposed repair method [

]

Note that Figure 2-3 does not show the AWJM remediation. Both Figure 2-2 and Figure 2-3 are shown before replacement thermal sleeve attachment for clarity purposes. Figure 2-4 depicts the final repair, including the replacement thermal sleeve (where applicable). [

] Therefore this drawing is applicable for the CRDM, CET, and spare locations described in Section 1.0 of this document. As stated in Section 1.0, the materials with potential corrosion concerns evaluated within this document include the low alloy steel RVCH exposed as well as the new Alloy 52M used to create a new weld during the proposed repair.

Controlled Document

Document No.: 51-9215673-001 Corrosion Evaluation of Shearon Harris RV Head Penetration IDTB Weld Repair (NonProprietary)

Page 8 Figure 2-1: Current Unrepaired Configuration of CRDM/CET/Spare Nozzle at HNP (Reference 3, Step 1)

Original Alloy 600 housing Low Alloy Steel RVCH Original J-Groove Weld Original Thermal Sleeve Controlled Document

Document No.: 51-9215673-001 Corrosion Evaluation of Shearon Harris RV Head Penetration IDTB Weld Repair (NonProprietary)

Page 9 Figure 2-2: IDTB CRDM Nozzle Repair Configuration with no Optional Removal of Weld Overlap, including Optional Remediation (Shown before Replacement Thermal Sleeve Attachment for Clarity) (Reference 2,3, Step 6)

Controlled Document

Document No.: 51-9215673-001 Corrosion Evaluation of Shearon Harris RV Head Penetration IDTB Weld Repair (NonProprietary)

Page 10 Figure 2-3: IDTB CRDM Nozzle Repair Configuration with Optional Removal of Weld Overlap, not including Remediation (Shown before Replacement Thermal Sleeve Attachment for Clarity)

(Reference 3, Step 5A)

Controlled Document

Document No.: 51-9215673-001 Corrosion Evaluation of Shearon Harris RV Head Penetration IDTB Weld Repair (NonProprietary)

Page 11 Figure 2-4: IDTB CRDM Nozzle Repair Configuration [Reference 3, Step 7]

Controlled Document

Document No.: 51-9215673-001 Corrosion Evaluation of Shearon Harris RV Head Penetration IDTB Weld Repair (NonProprietary)

Page 12 3.0 KNOWN OCCURANCES OF EXPOSED CARBON/LOW ALLOY STEEL BASE METAL The primary reactor coolant system (RCS), the pressurizer, reactor vessel, and the steam generator are clad with either a stainless steel or nickel-base alloy in order to prevent corrosion of the carbon or LAS base metal.

Throughout the operating history of domestic PWRs, there have been many cases where a localized area of the carbon or LAS base metal has been exposed to the primary coolant. Several such instances are listed below:

1960s Yankee-Rowe reactor vessel - Surveillance capsules fell from holder assemblies to the bottom of the vessel, releasing test specimens and other debris, leading to perforations in the cladding.

1990 Three Mile Island Unit 1 steam generator - Several tubes have separated within the tubesheet area exposing the tubesheet material to primary coolant. (LER 289-1990-005) 1990 ANO Unit 1 pressurizer - A leak was detected at the pressurizer upper level tap nozzle within the steam space in December 1990. The repair consisted of removing the outer section of the nozzle followed by welding a new section of nozzle to the OD of the pressurizer. (LER 313-1990-021) 1991 Oconee-Unit 1 steam generator - A misdrilled tubesheet hole in the upper tubesheet of one of the steam generators, during plugging operation in 1991, led to exposure of a small area of unclad tubesheet to primary coolant. [Note: This area of the tubesheet has since been patched and is no longer exposed to coolant.]

1993 McGuire-Unit 2 reactor vessel - A defect in the vessel cladding was discovered during an inspection in July 1993; the defect is believed to have occurred as a result of a pipe dropped in the vessel during construction (1975).

1993 SONGS-Unit 2 hot leg nozzle - A repair to a hot leg nozzle was completed during the 1993 outage at the SONGS Unit 2. This repair consisted of replacing a section of the existing Alloy 600 nozzle with a new nozzle section fabrication from Alloy 690. A gap approximately [

] wide exists between the two nozzle sections where the carbon steel base metal is exposed to the primary coolant. Verbal communication with SONGS personnel indicated that the hot leg nozzle containing this repair was removed and the exposed carbon steel examined.

1994 Calvert Cliffs-Unit 1 pressurizer - Two leaking heater nozzles in the lower head of the pressurizer were partially removed and the penetrations were plugged in 1994. (LER 317-1994-003) 1997 Oconee-Unit 1 OTSG manway - During the end-of-cycle (EOC) 17 refueling outage, a degraded area was observed in the bore of the 1B once through steam generator (OTSG). Subsequent inspection revealed a

[

] long circumferential damaged area to the cladding surface of the manway opening. The exposure of the base metal was confirmed by etching.

2001 CRDM repairs at Oconee Unit 2, Oconee Unit 3, Crystal River Unit 3, Three Mile Island Unit 1, and Surry Unit 1. (LER 270-2001-002, 287-2001-003, 302-2001-004, 289-2001-002, 280-2001-003) 2002 CRDM repairs at Oconee Unit 1 and Oconee Unit 2. (LER 269-2002-003, 270-2002-002) 2003 CRDM/CEDM repairs at St. Lucie Unit 2 and Millstone Unit 2, half nozzle repairs of STP-1 bottom mounted instrument nozzles, half nozzle repairs of pressurizer instrument nozzles at Crystal River Unit 3. (LER 389-2003-002, 498-2003-003, 302-2003-003) 2005 Half-nozzle modification for the TMI-1 pressurizer vent nozzle.

In each of these instances, carbon or LAS base metal was exposed to primary coolant in a localized area. Each plant returned to normal operation with the base metal exposed; in the case of Yankee-Rowe, the vessel operated for roughly 30 years with the base metal exposed.

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Page 13 4.0 CORROSION OF EXPOSED LOW ALLOY STEEL AT LOCATION A Several types of corrosion can occur when carbon and LAS base metal are exposed to primary coolant. During operating conditions, the primary coolant is deaerated at high temperatures (650°F design temperature [1])

depending on the location within the RCS. During shutdown conditions, the primary coolant temperature approaches 70°F and may become aerated and/or stagnant depending on the location within the RCS. The following sections discuss the possible corrosion mechanisms for the exposed LAS base metal at Location A (see Figure 2-3 for depiction of Location A).

4.1 General Corrosion of Exposed Base Metal General corrosion is defined as a type of corrosion attack (deterioration) uniformly distributed over a metal surface and corrosion that proceeds at approximately the same rate over a metal surface [4]. Stainless steels and nickel-base alloys (e.g., wrought Type 304, Type 316, Alloy 600, and Alloy 690 and their equivalent weld metals) are essentially unsusceptible to general corrosion in a PWR environment due to their passive protective surface layer. Carbon and LASs, however, may be susceptible to general corrosion depending on the service environment. The major factors affecting the general corrosion susceptibility of LAS are temperature, fluid velocity, boric acid concentration, and time. The general corrosion rates of carbon and LASs in aerated and deaerated conditions are discussed below.

4.1.1 Oxygen Level in the Repaired Area 4.1.2 General Corrosion Rate Many investigators have reported corrosion rates of carbon and LAS in various environments [7, 8, 9, 10, 11, 12, 13, 14, 15, 16]. In several instances, the corrosion rates for carbon and LASs have been observed to be similar in PWR environments; this data is applicable to carbon and LAS materials such as A-302, SA-533 (the material of the Shearon Harris RVCH [1]), and SA-516 [7, 9, 10, 14]. The Electric Power Research Institute (EPRI) has published a handbook on boric acid corrosion [17]. This handbook summarizes the industry field experience with boric acid corrosion incidents, a discussion of boric acid corrosion mechanisms, and a compilation of prior boric acid corrosion testing and results. In one evaluation, ASTM A302 Grade B was exposed to primary coolant in aerated and deaerated conditions [18]. It was shown that under deaerated conditions (i.e., during operation), the corrosion rate depended on temperature, fluid velocity, boric acid concentration, and time [18]. At the maximum velocity tested (36 ft/sec), [

] the corrosion rate was determined to be 0.003 inch/year (ipy); this was the maximum corrosion rate reported in the report.

Under static conditions (i.e., stagnant) at 650°F, a maximum corrosion rate of 0.0009 ipy was reported. In this Controlled Document

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Page 14 same study at shutdown conditions (aerated, at low temperature), the maximum corrosion rate was determined to be 0.0015 inch for a two month shutdown, or 0.009 ipy [18].

4.1.3 General Corrosion Rate in the HAZ 4.2 Crevice Corrosion of Exposed Base Metal The environmental conditions in a crevice can become aggressive with time and can cause accelerated local corrosion. The proposed as-repaired geometry shown in Figure 2-3 is open and does not contain any crevices.

Experiments were conducted (not as a part of this work scope) to determine the crevice corrosion rate of LAS.

The results indicate that the crevice corrosion rate for both aerated and deaerated conditions is less than the respective general corrosion rate [12, 18]. Operating experience from PWRs shows that crevice corrosion is not normally a problem in PWR systems with expected low oxygen contents [14].

[

]

4.3 Galvanic Corrosion of Exposed Base Metal Galvanic corrosion may occur when two dissimilar metals in contact are exposed to a conductive solution or coupled together. The three essential components to galvanic corrosion are 1) materials possessing different surface potential, 2) a common electrolyte, and 3) a common electrical path. The larger the potential difference between the metals, the greater the likelihood of galvanic corrosion. Low alloy and carbon steel are more anodic than stainless steels and nickel-base alloys [19] and could therefore be subject to galvanic attack when coupled and exposed to reactor coolant.

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Page 15 Several corrosion tests were performed (not as a part of this work scope) to determine the influence of coupling.

In one test, carbon steel specimens were coupled and uncoupled to stainless steel and exposed to simulated reactor shutdown conditions. The corrosion rates while coupled and uncoupled were determined to be similar [18].

Additionally, galvanic corrosion of carbon steel coupled to stainless steel in boric acid solution in the absence of oxygen should be quite low, therefore the galvanic corrosion rate should be low and the total corrosion is expected to be about equal to the general corrosion rate [18]. Austenitic stainless steels, such as Type 304, have approximately the same corrosion potential as nickel-base alloys such as Alloy 690. Therefore, galvanic corrosion studies of LAS and stainless steel give insight into the galvanic corrosion of LAS and nickel-base alloys. Specimens made from 5% chromium steel coupled to Type 304 stainless steel were exposed to aerated water at 500°F for 85 days (~2000 hours) with no evidence of galvanic corrosion. In the test above, the corrosion rates was note affected by coupling [16]. Additionally, results of the NRCs boric acid corrosion test program have shown that the galvanic difference between ASTM A533 Grade B, Alloy 600, and 308 stainless steel is not significant enough to consider galvanic corrosion as a strong contributor to the overall boric acid corrosion process [20].

[

]

4.4 Stress Corrosion Cracking of Exposed Base Metal Stress corrosion cracking (SCC) can only occur when the following three conditions are present:

A susceptible material A tensile stress An aggressive environment Under normal PWR conditions (deaerated), primary water is not a particularly aggressive environment for LAS unless an (unexpected) departure from normal operating conditions occurs [21]. This service environment does not generally support localized corrosion of LAS therefore the likelihood of a pit or notch forming which would contribute a stress concentrator or SCC initiation site is negligible. Extensive experience with exposed LAS in PWRs [

] has not resulted in any reported SCC. [

]

4.5 Hydrogen Embrittlement of Exposed Base Metal Hydrogen embrittlement occurs when a materials properties are degraded due to the presence of hydrogen. This type of damage usually occurs in combination with a stress, residual, applied, or otherwise. Hydrogen embrittlement is typically observed in high pressure hydrogen environments and in deformed metals and is characterized by ductility losses and lowering of the fracture toughness [4]. High pressure hydrogen environments are not typical of PWR systems and are defined as an environment with approximately 5,000-10,000 psi [22]. Although hydrogen is added to PWR water to scavenge oxygen (see Section 4.1.1 of this report),

the primary contributor of hydrogen diffusion into the LAS is the corrosion process. Corrosion tests on LAS in deaerated boric acid solutions indicated that the maximum concentration of the hydrogen in the steel from the corrosion process was less than 2 ppm and did not increase with time [18]. The quantity of hydrogen that may accumulate at locations within the coolant system is not expected to induce hydrogen embrittlement in materials at these locations. [

]

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Page 16 5.0 ESTIMATE OF FE RELEASE RATE FROM EXPOSED LOW ALLOY STEEL AT LOCATION A 5.1 CRDM, CET, and Spare Nozzle Repairs Based on Figure 2-3 of this report, the limiting dimensions after boring, with or without AWJM, for D and W are:

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Page 17 The estimated exposed LAS surface area and the corresponding Fe release rate are summarized in Table 5-1.

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Page 18 Table 5-1: Estimated Exposed Area of LAS and Fe Release Rate

[

] Note that since limiting dimensional values were used, this conservative estimate will not change based on the use of AWJM surface remediation techniques.

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Page 19 6.0 CORROSION OF ALLOY 52M Alloy 52M (Alloy 52 modified) is the specified IDTB weld repair for the CRDM/CET/spare nozzle penetrations. To better understand the potential corrosion concerns of Alloy 52M, information regarding the corrosion of Alloy 690 (the associated base metal to Alloy 52M) and Alloy 52 (the base Alloy 52 not containing some of the alloying element of Alloy 52M) will also be presented. The corrosion resistance of Alloy 52M is expected to be similar to that of Alloy 52 and Alloy 690. The difference between these Alloy 52 and 52M is only minor alloying elements for enhanced weldability. The chromium content, which provides the corrosion resistance of the material, is similar. The corrosion resistance of Alloy 690 has been extensively studied as a result of numerous PWSCC failures in mill annealed Alloy 600 in primary water environments. As a result of Alloy 600 failures, Alloy 690 has been chosen by the nuclear industry as the replacement material of choice for Alloy 600 components.

A comprehensive review for the use of Alloy 690 in PWR systems cites numerous investigations and test results under a wide array of conditions, including both primary (high temperature deoxygenated water) and secondary coolant environments. The first Alloy 690 steam generator went online in May 1989 with no reported failures as to the date of that publication due to environmental degradation (August 1997) [25]. No environmental degradation of Alloy 690 or its related weld metals has been reported since the August 1997. Crevice and general corrosion of austenitic nickel-base materials is not expected to be of great concerns in typical PWR conditions

[26].

Alloy 52M was tested (not as a part of this work scope) in accelerated corrosion conditions by testing a weld mockup that simulated nozzle safe end repairs. The testing consisted of 400°C (752°F) steam plus hydrogen doped with 30 ppm each of fluoride, chloride, and sulfate anions. The hydrogen partial pressure was controlled at approximately 75kPa with a total steam pressure of 20 MPa; this environment has been previously used to accelerate the simulated PWSCC of nickel-base alloys. After a cumulative exposure of 2051 hours0.0237 days <br />0.57 hours <br />0.00339 weeks <br />7.804055e-4 months <br /> (equivalent to 45.6 EFPY), no environmental degradation was detected on the surface of the Alloy 52M welds. Small microfissures on the surface of the Alloy 52M welds, stressed in tension, did not serve as initiation sites for environmental degradation, nor did they propagate during the tests. Stress corrosion cracks initiated in the also-tested Alloy 182 welds in exposure times less than one-fifth the total exposure time of the Alloy 52M specimens

[27]. This study, along with many others [examples in References 28, 29, 30], indicate that the Alloy 52M weld metal in the proposed repair [

].

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Page 20

7.0 CONCLUSION

S The information presented above describes the potential corrosion mechanisms that may affect the exposed LAS in the proposed Shearon Harris RVCH penetration repair configuration in Figure 2-3. Based on this evaluation, the modification is found acceptable (as detailed below).

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Page 21

8.0 REFERENCES

1.

AREVA Document 08-9172870-003, Design Specification, Shearon Harris RVCH CRDM and CET Nozzle Penetration Modification.

2.

AREVA Document 39-9238279-000, SI for 02-9175500E-006.

3.

AREVA Drawing 02-9175500E-006, Shearon Harris CRDM ID Temberbead Weld Repair.

4.

ASM Metals Handbook, Eleventh Edition, Volume 13A Corrosion, 2003.

5.

P. Pastina, et. al., The Influence of Water Chemistry on the Radiolysis of the Primary Coolant Water in Pressurized Water Reactors, Journal of Nuclear Materials, 264 (1999) 309-318.

6.

P. Scott, A Review of Irradiation Assisted Stress Corrosion Cracking, Journal of Nuclear Materials, 211 (1994) 101-122.

7.

Whitman, G. D. et. al., A Review of Current Practice in Design, Analysis, Materials, Fabrication, Inspection, and Test, ORNL-NSIC-21, ORNL, December 1967.

8.

Vreeland, D. C. et. al., Corrosion of Carbon and Low-Alloy Steels in Out-of-Pile Boiling Water Reactor Environment, Corrosion, v17, June 1961, p. 269.

9.

Vreeland, D. C. et. al., Corrosion of Carbon Steel and Other Steels in Simulated Boiling-Water Reactor Environment: Phase II, Corrosion, v18, October 1962, p. 368.

10.

Uhlig, H. H., Corrosion and Corrosion Control, John Wiley & Sons, New York, 1963.

11.

Copson, H. R., Effects of Velocity on Corrosion by Water, Industrial and Engineering Chemistry, v44, No. 8, p. 1745, August 1952.

12.

Vreeland, D. C., Corrosion of Carbon Steel and Low Alloy Steels in Primary Systems of Water-Cooled Nuclear Reactors, Presented at Netherlands-Norwegian Reactor School, Kjeller, Norway, August 1963.

13.

Pearl, W. L. and Wozadlo, G. P., Corrosion of Carbon Steel in Simulated Boiling Water and Superheated Reactor Environments, Corrosion, v21, August 1965, p. 260.

14.

DePaul, E. J., Corrosion and Wear Handbook for Water-Cooled Reactors, McGraw-Hill Book Company, Inc. 1957.

15.

Tackett, D. E. et. al., Review of Carbon Steel Corrosion Data in High-Temperature Water, High-Purity Water in Dynamic Systems, USAEC Report, WAPD-LSR(C)-134, Westinghouse Electric Corporation, October 14, 1955.

16.

Ruther, W. E. and Hart, R. K., Influence of Oxygen on High Temperature Aqueous Corrosion of Iron, Corrosion, v19, April 1963, p. 127.

17.

Boric Acid Corrosion Guidebook, Revision 1: Managing the Boric Acid Corrosion Issues at PWR Power Stations, EPRI, Palo Alto, CA: 2001. 1000975.

18.

Evaluation of Yankee Vessel Cladding Penetrations, Yankee Atomic Electric Company to the U. S.

Atomic Energy Commission, WCAP-2855, License No. DPR-3, Docket No. 50-29, October 15, 1965.

19.

ASM Metals Handbook, Ninth Edition, Volume 13 Corrosion, 1987.

20.

U. S. NRC publication NUREG-1823, U.S. Plant Experience with Alloy 600 Cracking and Boric Acid Corrosion of Light-Water Reactor Pressure Vessel Materials, NRC Accession No. ML051390139.

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Page 22

21.

P.M. Scott and D.R. Tice, Stress Corrosion in Low Alloy Steels, Nuclear Engineering and Design, Volume 119, 1990.

22.

Gray, H.H., Hydrogen Embrittlement Testing (STP 543), Opening Remarks, 1974, ASTM.

23.

AREVA 38-2200979-00, Submittal of Information Requested by Progress Energy to be Provided to AREVA NP for Shearon Harris Unit 1 Reactor Vessel Head Repair.

24.

AREVA 51-1177703-00, Estimated Cobalt Release Rates.

25.

Crum, J.R., Nagashima, T., Review of Alloy 690 Steam Generator Studies, Eighth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, August 10-14, 1997, ANS.

26.

Fyfitch, S. (2012) Corrosion and Stress Corrosion Cracking of Ni-Base Alloys. In: Konings R.J.M., (ed.)

Comprehensive Nuclear Materials, volume 5, pp. 69-92, Amsterdam: Elsevier.

27.

Jacko, R.J., et. al., Accelerated Corrosion Testing of Alloy 52M and Alloy 182 Weldments, Eleventh International Conference on Environmental Degradation of Materials in Nuclear System, August 10-14, 2003, ANS.

28.

Sedricks, A.J., et. al., Inconel Alloy 690 - A New Corrosion Resistant Materials, Boshoku Gijutsu, Japan Society of Corrosion Engineering, v28, No. 2, pp. 82-95, 1979.

29.

Brown, C.M., and Mills, W.J., Effect of Water on Mechanical Properties and Stress Corrosion Behavior of Alloy 600, Alloy 690, EN82H Welds, and EN52 Welds, Corrosion, v55(2), February 1999.

30.

Mills, W.J., and Brown, C.M., Fracture Behavior of Nickel-Based Alloys in Water, Ninth International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, August 1-5, 1999, TMS.

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20004-021 (01/30/2014)

Page 1 of 9 AREVA Inc.

Engineering Information Record Document No.:

51

- 9215674 -

001 PWSCC Assessment of the Alloy 600 Nozzle in the SH RV Head Penetration IDTB Weld Repair (NonProprietary)

AREVA INC. PROPRIETARY This document and any information contained herein is the property of AREVA Inc. (AREVA) and is to be considered proprietary and may not be reproduced or copied in whole or in part. This document shall not be furnished to others without the express written consent of AREVA and is not to be used in any way which is or may be detrimental to AREVA. This document and any copies that may have been made must be returned to AREVA upon request.

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Document No.: 51-9215674-001 PWSCC Assessment of the Alloy 600 Nozzle in the SH RV Head Penetration IDTB Weld Repair (NonProprietary)

Page 3 Record of Revision Revision No.

Pages/Sections/

Paragraphs Changed Brief Description / Change Authorization 000 Initial Issue Original Issue. The corresponding proprietary version is in AREVA document 51-9176115-002 001 All Corresponding proprietary version is 51-9176115-003.

001 Throughout Updated form to most recent revision.

001 Throughout This revision includes incorporation of document amendment 159-9216055-000.

001 Section 1.0 Included a sentence about previously repaired nozzles.

Included the purpose of Revision 1.

Scope clarification 001 Section 2.0 Changed Progress Energy to Duke Energy 001 Section 3.1 Renamed heading 001 Section 3.2 Included two new justified assumptions.

001 Section 5.0 Clarification about AWJM being optional.

001 Section 5.1 New section to differentiate estimated repair life with and without AWJM 001 Section 5.1 Included reference 4 to add additional justification for crack growth rate used for PWSCC 001 Section 5.1 Clarified that this evaluation is for axial flaws only 001 Section 5.2 New section to differentiate estimated repair life with and without AWJM 001 Section 6.0 Included estimated repair life if no AWJM surface remediation is used 001 Section 7.0 Updated Reference 1 to Revision 3 001 Section 7.0 Updated Reference 2 to Revision 6 001 Section 7.0 New Reference 3, 4, 6, 15 001 Section 7.0 Updated Reference 7 Controlled Document

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Page 4 Table of Contents Page SIGNATURE BLOCK................................................................................................................................ 2 RECORD OF REVISION.......................................................................................................................... 3 1.0 PURPOSE..................................................................................................................................... 5

2.0 BACKGROUND

............................................................................................................................ 5 3.0 ASSUMPTIONS............................................................................................................................ 5 3.1 Assumptions Requiring Verification................................................................................... 5 3.2 Justified Assumptions........................................................................................................ 5 4.0 DESIGN INPUTS.......................................................................................................................... 5 5.0 EVALUATION............................................................................................................................... 5 5.1 Scenario #1: Weld Repair with Abrasive Waterjet Machining........................................... 6 5.2 Scenario #2: Weld Repair without Surface Remediation.................................................. 7

6.0 CONCLUSION

S............................................................................................................................ 8

7.0 REFERENCES

.............................................................................................................................. 9 Controlled Document

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Page 5 1.0 PURPOSE The purpose of this document is to estimate a lifetime for the inside diameter temperbead (IDTB) weld repairs on the Shearon Harris reactor vessel closure head (RVCH) control rod drive mechanism (CRDM) type nozzles (this includes 52 CRDM nozzles, 4 core exit thermocouple (CET) nozzles, and 9 spare locations [1]). This evaluation considers nozzles in the as-repaired condition. This evaluation is limited in scope to primary water stress corrosion cracking (PWSCC) concerns of the original Alloy 600 nozzles, which is specifically focused on the nozzle material directly adjacent to the IDTB weld and the roll transition region due to the elevated surface tensile stress.

The purpose of Revision 1 is to reflect changes to the design specification [1] and drawing [2][3] making the use of the AWJM surface remediation technique in this repair optional.

2.0 BACKGROUND

Operating experience has shown that Alloy 600 CRDM nozzles and their welds are susceptible to primary water stress corrosion cracking (PWSCC) [1]. Consequently, the B&W Owners Group (BWOG) commissioned AREVA to provide an automated repair process that was ultimately implemented at Oconee Unit 2. Duke Energy has contracted AREVA to adapt this repair for Shearon Harris Unit 1 as a contingency [1].

3.0 ASSUMPTIONS 3.1 Assumptions Requiring Verification None.

3.2 Justified Assumptions

1. [

].

2. [

].

3. Other assumptions made in this evaluation are discussed in Section 5.0.

4.0 DESIGN INPUTS Design inputs and sources are discussed in Section 5.0.

5.0 EVALUATION The IDTB weld repair process includes: 1) removing the thermal sleeve (not applicable to CET nozzles and spare locations), 2) roll-expanding the original nozzle above the flaw, 3) removal of the lower portion of the nozzle, 4) ambient temperbead welding of the remaining nozzle material to the RV head with Alloy 52M, 5) AWJM remediation of the repaired area [2] 1 (this remediation is optional [3]), and 6) replacement of the thermal sleeve (where applicable). For the repair configuration, the final pressure boundary is comprised of the remaining Alloy 1 The referenced repair drawing is applicable to all 65 CRDM penetrations. This includes 52 CRDM nozzles, 9 spares, and 4 CET nozzles.

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Page 6 600 nozzle material and the Alloy 52M weld. The purpose of this evaluation is to approximate the service life for the remaining weld repaired Alloy 600 nozzles. Since AWJM is an optional surface remediation process, two scenarios will be evaluated: with and without the use of the AWJM surface remediation process.

5.1 Scenario #1: Weld Repair with Abrasive Waterjet Machining The U.S. nuclear industry has developed a conservative 75% through-wall (axial) acceptance criterion for the depth of axial flaws that initiate and propagate from the ID of CRDM nozzles at or above the J-groove weld [7].

Note that a version of this 75% criterion has also been added to newer versions of the ASME Code [8].

The AWJM process uses high-pressure water and an abrasive to remove material and flaws. The process leaves a compressive residual surface stress, thereby mitigating PWSCC [9]. Similarly, shot peening produces a layer of compressive residual surface stress and is used in many industries for surface modification of components in aggressive environments. Shot peening of Alloy 600 has been shown to reduce the occurrence of PWSCC and delay initiation, but does not mitigate pre-existing non-detectable cracks [10,11,12]. AWJM remediation removes pre-existing non-detectable cracks within the remediation depth and leaves a new surface which has had no prior exposure to primary water. A quantitative lifetime has not been established for AWJM remediation, although there is significant data in the literature to show that the use of shot peening of new components significantly extends their lifetime. Shot peened surfaces, (i.e. surfaces with compressive residual stress) are not expected to be susceptible to PWSCC for extended periods of time.

The Electric Power Research Institute (EPRI) Materials Reliability Program (MRP) has conducted a test program of various Alloy 600 RV head penetration PWSCC mitigation techniques including the AWJM process [13].

Other mitigation techniques tested in the program include: excavation weld repair, laser cladding, laser weld repair, EDM, flapper wheel, shot peening, brush nickel plating, and electroless nickel plating. Alloy 600 samples mitigated by each technique were autoclave tested under accelerated SCC conditions at 750°F in doped steam with hydrogen. AWJM was shown to be the best mitigation technique with no SCC being observed after exposures of up to 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />. These results provide qualitative support to the claim that AWJM mitigates PWSCC initiation.

Since PWSCC has not been shown to initiate in the compressive stress region induced by the AWJM process, the dominant degradation mechanism is assumed to be general corrosion. The average corrosion rate for Alloy 600 in primary water is less than or equal to [

] [14].

The compressive stress layer was measured to be 0.003 inches thick in an AWJM remediated CRDM nozzle

[

].

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Page 7 The assumptions used in this scenario are as follows:

1. [

] an undetected flaw 0.002 inches deep

[

] is assumed present. This results in a 0.001 inch layer of compressive stress which must be breached before PWSCC can initiate.

2. [

]

3. [

]

4. [

].

It is estimated that it will take 12.5 EFPY for the remaining compressive stress layer to be breached [

]. Thereafter, it is estimated that a crack could propagate to 75% of the original wall thickness in 2.3 EFPY [

]. The total estimated lifetime of the repair is 14.8 EFPY if AWJM surface remediation techniques are used.

5.2 Scenario #2: Weld Repair without Surface Remediation By not performing the remediation process, the remnant Alloy 600 nozzle will be more susceptible to PWSCC adjacent to the IDTB weld and at the roll expansion transition. As described previously, initiation of a PWSCC crack can sometimes take decades, but laboratory testing and operating experience also indicates that initiation can occur almost immediately in some cases.

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Page 8 The assumptions used in this scenario are as follows:

It is estimated that a crack could propagate to 75% of the original wall thickness in 2.2 EFPY [

]. The total estimated lifetime of the repair is 2.2 EFPY if AWJM surface remediation techniques are not used.

6.0 CONCLUSION

S An evaluation of PWSCC initiation and growth was performed for the IDTB weld repair process with and without the use of AWJM remediation. Conservative assumptions were used for the flaw initiation time and crack growth rate. The industry adopted 75% though-wall flaw acceptance criterion was used.

The estimated time to breach the compressive stress layer and reach a flaw through 75% of the original wall thickness is 14.8 EFPY if AWJM surface remediation techniques are used. If AWJM surface remediation techniques are not used, the estimated time to reach a flaw through 75% of the original wall thickness is 2.2 EFPY. Recall that this estimate is expected to be conservative based on the assumptions listed in Section 5.2.

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7.0 REFERENCES

1.

AREVA Document 08-9172870-003, Design Specification - Shearon Harris RVCH CRDM and CET Nozzle Penetration Modification.

2.

AREVA Drawing 02-9175500E-006, Shearon Harris CRDM ID Temper Bead Weld Repair.

3.

AREVA Document 39-9238279-000, SI for 02-9175500E-006.

4.

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