HNP-18-045, Submittal of Relief Request I4R-18, Reactor Vessel Closure Head Nozzle Repair Technique, Inservice Inspection Program, Fourth Ten-Year Interval

Submittal of Relief Request I4R-18, Reactor Vessel Closure Head Nozzle Repair Technique, Inservice Inspection Program, Fourth Ten-Year Interval

ML18108A094
Person / Time
Site:	Harris
Issue date:	04/18/2018
From:	Hamilton T Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML18108A093	List: HNP-18-045, Submittal of Relief Request I4R-18, Reactor Vessel Closure Head Nozzle Repair Technique, Inservice Inspection Program, Fourth Ten-Year Interval
References
HNP-18-045
Download: ML18108A094 (35)

Text

Tanya M. Hamilton Vice President Harris Nuclear Plant 5413 Shearon Harris Rd New Hill NC 27562-9300 919-362-2502 PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ENCLOSURE 3 THIS LETTER IS UNCONTROLLED 10 CFR 50.55a April 18, 2018 Serial: HNP-18-045 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400/Renewed License No. NPF-63

Subject:

Relief Request I4R-18, Reactor Vessel Closure Head Nozzle Repair Technique, Inservice Inspection Program, Fourth Ten-Year Interval Ladies and Gentlemen:

Pursuant to 10 CFR 50.55a(z)(1), Duke Energy Progress, LLC (Duke Energy), hereby requests Nuclear Regulatory Commission (NRC) approval of the attached relief request for the Shearon Harris Nuclear Power Plant, Unit 1 (HNP) inservice inspection program, fourth ten-year interval.

The request is similar to relief requests previously approved by the NRC staff (Agencywide Documents Access and Management System (ADAMS) Accession Numbers ML12270A258, ML13238A154, ML14093A075, ML15203A702, ML15342A043, and ML16343A220).

Duke Energy has identified one nozzle with flaws that requires repair as a result of examinations of the reactor head performed in the current refueling outage. Relief is requested in accordance with 10 CFR 50.55a(z)(1). The provisions of this relief are applicable to the fourth ten-year inservice inspection interval at HNP, which commenced on September 9, 2017 and is currently scheduled to end on September 8, 2027, as identified in the Fourth Interval Inservice Inspection Plan, submitted to the NRC on October 23, 2017 (ADAMS Accession No. ML17296A323).

While this relief request identifies the same code cases as the previous relief requests identified above, updated versions of the applicable code cases, as approved by the NRC and identified in the Fourth Interval Inservice Inspection Plan, are referenced in this submittal.

The calculations to support this request are provided in Enclosure 3 and contain information considered proprietary to Framatome, Inc. On behalf of Framatome, 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 information. Upon removal of the proprietary information in Enclosure 3, the balance of this letter is uncontrolled.

Duke Energy requests approval of this request by April 26, 2018, to support startup from the current refueling outage.

PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ENCLOSURE 3 THIS LETTER IS UNCONTROLLED

PROPRIETA RY INFORMATIO N -WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ENCLOSURE 3 THIS LETTER IS UNCONTROLLED U.S. Nuclear Regulatory Commission Page 2 HNP-18-045 This letter contains a regulatory commitment as identified in Enclosure 4.

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

Sincerely,

~~fh- ct-1~

Tanya M. Hamilton

Enclosures:

1. Relief Request I4R-18

2. Affidavit Supporting Withholding of Proprietary Information

3. Calculation 32-9176350-006, Shearon Harris Unit 1 CRDM/CET Nozzle As-Left J-groove Weld Analysis (PROPRIETARY)

4. Regulatory Commitment cc: J. Zeiler, NRC Senior Resident Inspector, HNP M. Barillas, NRC Project Manager, HNP NRC Regional Administrator, Region II PROPRIETA RY INFORMATIO N -WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ENCLOSURE 3 THIS LETTER IS UNCONTROLLED

PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ENCLOSURE 3 THIS LETTER IS UNCONTROLLED U.S. Nuclear Regulatory Commission Page 2 HNP-18-045 This letter contains a regulatory commitment as identified in Enclosure 4.

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

Sincerely, Tanya M. Hamilton

Enclosures:

1. Relief Request I4R-18

2. Affidavit Supporting Withholding of Proprietary Information

3. Calculation 32-9176350-006, Shearon Harris Unit 1 CRDM/CET Nozzle As-Left J-groove Weld Analysis (PROPRIETARY)

4. Regulatory Commitment cc: J. Zeiler, NRC Senior Resident Inspector, HNP M. Barillas, NRC Project Manager, HNP NRC Regional Administrator, Region II PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ENCLOSURE 3 THIS LETTER IS UNCONTROLLED

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 1 HNP-18-045 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400/Renewed License No. NPF-63 Relief Request I4R-18, Reactor Vessel Closure Head Nozzle Repair Technique, Inservice Inspection Program, Fourth Ten-Year Interval Enclosure 1 Relief Request I4R-18

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 1 Page 1 of 25 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400/Renewed License No. NPF-63 Relief Request I4R-18, Reactor Vessel Closure Head Nozzle Repair Technique, Inservice Inspection Program, Fourth 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 Code Item Number: B4.20 (Code Case N-729-4, Alternative Examination Requirements for PWR Reactor Vessel Upper Heads with Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1)

==

Description:==

RVCH Penetration Nozzle 33 Size: 4 Inch Nominal Outside Diameter Material: SB-167, Alloy 600 (UNS N06600)

2. Applicable Code Edition and Addenda Shearon Harris Nuclear Power Plant, American Society of Mechanical Engineers Unit 1 (HNP), Inservice Inspection Boiler and Pressure Vessel Code Program (ISI) - Fourth Interval (ASME Code)Section XI, 2007 Edition through 2008 Addenda Shearon Harris Nuclear Power Plant, American Society of Mechanical Engineers Unit 1, RVCH Code of Construction Boiler and Pressure Vessel Code Section III, 1971 Edition through Winter 1971 Addenda

3. Applicable Code Requirements ASME Code, Section Xl, 2007 Edition through 2008 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.

ASME Code, Section Xl, 2007 Edition through 2008 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..

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 1 Page 2 of 25 ASME Code, Section Xl, 2007 Edition through 2008 Addenda IWA-4411 states:

Welding, brazing, fabrication, 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, 2007 Edition through 2008 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, 2007 Edition through 2008 Addenda IWA-4610(a) states in part:

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

ASME Code,Section III, 2001 Edition through 2003 Addenda:

NB-5245, Partial Penetration Welded Joints, requires progressive surface examination of partial penetration welds NB-5331(b) states that indications characterized as cracks, lack of fusion, or incomplete penetration are unacceptable regardless of length.

Code Case N-638-6, 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-729-4, Alternative Examination Requirements for PWR Reactor Vessel Upper Heads with Nozzles Having Pressure-Retaining Partial-Penetration WeldsSection XI, Division 1, Figure 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, 2007 Edition through 2008 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.

ASME Code, Section Xl, 2007 Edition through 2008 Addenda, IWA-3300 specifies requirements for characterization of flaws detected by inservice examination.

ASME Code, Section Xl, 2007 Edition through 2008 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, 2007 Edition through 2008 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

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 1 Page 3 of 25 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 Flaws requiring repair were detected during the ISI program ultrasonic testing (UT) examination of the HNP RVCH nozzle penetration 33. 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. Nozzle (33) will be repaired under this request. Figure 10 shows the location of the axial indications and Figure 11 shows the relative location of the nozzle on the RVCH. Table 1 provides sizing and characterization information on the flaws leading to the repair activities.

The repair technique is intended to be the same as was used previously for nozzles 5, 14, 17, 18, 23, 30, 37, 38, 40, 49, 51, 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 flaws, 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.

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 Code Section III, 1971 Edition including Addenda through Winter 1971, Duke Energy Progress, LLC (Duke Energy), 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 2007 Edition through the 2008 Addenda of ASME Code Section XI, Code Case N-638-6, Code Case N-729-4, and the alternatives discussed below.

Basic steps for the IDTB repair are:

1. Removal of lower portion of existing Thermal Sleeve Assembly to provide access for IDTB weld repair (note that the Thermal Sleeve Assembly is not shown in Figures 1-10).

2. Roll expansion of the Control Rod Drive Mechanism (CRDM) housing tube above the area of repair. This stabilizes the nozzle to prevent any movement when the lower 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 indications. 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

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 1 Page 4 of 25 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. Welding of new Lower Thermal Sleeve Assembly.

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 as-found dimensions.

Stresses introduced during the controlled roll expansion process implemented per design and fabrication controls will not create regions that would be more susceptible to primary water stress corrosion cracking (PWSCC) than other regions that have been previously evaluated and found acceptable. Two fabrication parameters are controlled to ensure the nozzle roll expansion is effective in performing its design function of mechanical support for the nozzle prior to the application of the inside diameter temper bead weld, and does not introduce stresses sufficient to initiate PWSCC in the region above the roll expansion zone. The parameters of interest are tool insertion depth and torque on the assembly tool.

Tool insertion depth, based on tooling setup height, will be controlled so that the rolled region is contained within the RVCH penetration bore. Tool insertion depth will be adjusted based on the as-found head thickness. The torque applied to the roll expander is controlled so that the desired amount of plastic deformation occurs. The torque limiter assembly will be set and independently verified with a calibrated torque wrench prior to use.

As noted above, the roll expansion process will be completed for nozzle 33 and the two parameters of interest that could impact the susceptibility to PWSCC will be validated to be within process specifications. As a result, there is high confidence that adequate measures will be applied in the nozzle 33 repair such that PWSCC in the region above the roll expansion zone is not expected to initiate.

Duke Energy has determined that repair of the RVCH nozzle penetration 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. Examinations 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-729-4, 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. Figure 3 also shows the post weld PT scope.

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 1 Page 5 of 25 ASME Code Section III, 2001 Edition including Addenda through 2003, NB-5245, requires progressive surface examination of partial penetration welds. The Construction Code requirement for progressive PT examination, in lieu of volumetric examination, is specified because volumetric examination is not practical for the conventional partial penetration weld configurations. Therefore, the following combination of UT and PT examinations are proposed.

For this modification, the weld, except for the taper transition, is suitable for UT examination, and a final surface PT examination shall 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 reasonable assurance of structural integrity.

UT volumetric examination of the repaired/modified configuration will be performed as specified in ASME Code Case N-638-6, 4(a)(2) and 4(a)(4). Regulatory Guide 1.147, Revision 18, has conditionally approved Code Case N-638-6 with the condition that the demonstration for UT examination of the repaired volume be required using representative samples which contain construction type flaws. The acceptance criteria of NB-5330, in ASME Code Section III, 2001 Edition through 2003 Addenda, will apply to all flaws identified within the repaired volume.

ASME Code 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 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 scanned 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.

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 1 Page 6 of 25 PT examination will be performed on the entire surface area as indicated by Figure 3. In addition, the volume in question will be examined to the extent practical using the 45L and 70L (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. The final examination of the new weld and immediate surrounding region will be sufficient to verify that defects have not been induced in the ferritic low alloy steel RVCH base material, due to welding, to the extent practical.

The combination of performing the PT and UT examinations depicted in Figure 3 during the IDTB repair provides reasonable assurance of structural integrity and that all unacceptable flaws associated with the weld repair area have been removed. Thus, Duke Energy requests relief from the progressive surface examination requirements specified in NB-5245.

b. Triple Point Anomaly ASME Code Section III, 2001 Edition including Addenda through 2003, NB-5331(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.

The outermost penetration was modeled due to the applied loading conditions being representative and bounding relative to all other locations in the RVCH. The initial flaw size for the triple point anomaly analysis is 0.10 inches. Crack growth analysis determines the future flaw size and concludes that it is acceptable for the stated life. The outermost hillside nozzle is explicitly modeled, meaning that both extremes of interaction between the IDTB weld and the original J-groove weld are considered (i.e., these welds are very close to each other on the uphill side, and are relatively far away from each other on the downhill side).

A fracture mechanics analysis has been performed for the design configuration to provide justification, in accordance with ASME Code 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.

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

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 1 Page 7 of 25 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 ASME Code,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.

At the end of 40 years, the growth of the axial flaw, and the circumferential flaw, is calculated to be less than 0.001 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, St =mc/SFm, where mc 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.

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 1 Page 8 of 25 The fracture margin for the limit load calculations include 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 Code 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 Code Section XI criteria. 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.

Duke Energy requests relief from the acceptance criteria specified in NB-5331(b) to permit anomalies, as described herein, at the triple point area to remain in service.

c. 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 by analytical evaluation as required by 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 butter material, the HNP RVCH nozzle repair design configuration is considered to be acceptable for a minimum of 30 years of operation, based on EPFM analysis failure criteria, following an IDTB weld repair.

The outermost penetration was modeled due to the applied loading conditions being the same or worse than all other locations in the RVCH. The initial flaw size for the J-groove weld is conservatively assumed to include all of the weld and buttering. This is highly conservative since the buttering sees post weld heat treatment, which would tend to reduce welding residual stresses, making it less susceptible to PWSCC. While the analysis considers crack growth on both uphill and downhill sides, the weld on the downhill side of the outermost nozzle has the largest area. Therefore, the largest possible initial flaw size on the downhill side is considered.

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 stable for a minimum

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 1 Page 9 of 25 of 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 based on the detailed EPFM Analysis that was performed. Following the detailed EPFM analysis, a primary stress limit analysis (PSLA) as required per IWB-3610(d)(2) was subsequently performed to demonstrate the service life of the RVCH considering all thirteen repairs (twelve previously completed repairs and one planned repair) performed through April 2018. The results from the PSLA show that the overall service life of the RVCH is 5 years from the April 2018 outage.

Therefore, the RVCH nozzle repairs performed during the April 2018 outage are acceptable for 5 years from the time of repair. Measurements will be taken during the repair activity to confirm assumed local conditions (head thickness and J-groove weld size) in the PSLA are valid. Duke Energy commits to take measurements during the repair process to confirm the actual local head thickness bounds the PSLA assumptions for a limiting life greater than or equal to 5 years as a condition for this request. This commitment is identified in Enclosure 4 of this submittal.

Successive examinations required by IWB-3132.3 will not be performed on the nozzle (33) that requires repair this outage for the duration of the life of the repairs 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.

In summary, the acceptable fatigue crack growth life is based on primary stress limits as specified in NB-3000. The analysis shows acceptability of the RVCH nozzle repairs for 5 years from the time of repair and documents acceptability beyond the planned RVCH replacement in the fall of 2019.

Relief is requested from flaw characterization and subsequent examination requirements.

The potential for debris from a cracking J-groove partial penetration weld was considered. The evaluation is generic with respect to penetration location in the RVCH. Radial cracks were

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 1 Page 10 of 25 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.

d. Inservice Inspections Successive examinations required by Code Case N-729-4 shall be performed on nozzle 33 during each subsequent refueling outage or until the RVCH is replaced. Code Case N-729-4 provides requirements for the inservice inspection of RVCHs with nozzles having partial penetration welds. Code Case N-729-4 Table 1, Item B4.20, permits either volumetric or surface examination. Item B4.20 examination requirements are specified in Figure 2 of Code Case N-729-4. 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 future inservice inspections. This examination area is equivalent to that required by Figure 2 in Code Case N-729-4, 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.

Non-repaired RVCH nozzles will continue to be examined in accordance with Code Case N-729-4 and 10 CFR 50.55a(g)(6)(ii)(D), using a qualified ultrasonic examination procedure with demonstrated leak path assessment capability.

Therefore, preservice inspection following repair and future inservice inspections will comply with Code Case N-729-4 as modified by 10 CFR 50.55a(g)(6)(ii)(D) and as depicted in Figure 9.

e. 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 in Appendix C, Section C.3.2, of . 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).

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 1 Page 11 of 25

6. Duration of Proposed Alternatives 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. For the weld anomaly and as-left J-groove weld analyses, the design life starts at the time of the repair. For the most limiting PWSCC evaluation, the analyses described above and others in the modification that will be implemented under 10 CFR 50.59 support a maximum exam interval of 2.2 effective full power years (EFPY) or 803 effective full power days for continued plant operation. The 2.2 EFPY exam interval starts at the time of the repair and is reset to an additional 2.2 EFPY upon demonstration of acceptable results from examinations performed during each refueling outage. Duke Energy examines all repaired RVCH penetration nozzles every refueling outage (every 18 months) in accordance with ASME Code Case N-729-4 as conditioned by 10 CFR 50.55a(g)(6)(ii)(D). The next refueling outage is currently scheduled to start on or about October 12, 2019. There are 537 calendar days from April 23, 2018, to October 12, 2019. Therefore, the periodic examinations each refueling outage will provide reasonable assurance of the structural integrity of RVCH nozzles.

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.

As described in section 5.c of this Enclosure, analysis shows acceptability of the RVCH nozzle repairs for a minimum life of 5 years is predicted based on the as-left J-groove flaw evaluation.

The provisions of this relief request are applicable to the fourth ten-year inservice inspection interval for HNP which commenced on September 9, 2017, and is currently scheduled to end on September 8, 2027. The repair installed in accordance with the provisions of this relief shall remain in place for the 5-year design life of the repair, until another alternative is approved by the Nuclear Regulatory Commission (NRC), or until the RVCH is replaced.

7. Additional Information

a. Mockup In support of over 130 similar repairs, Framatome, Inc. 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.

A NiCrFe alloy 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

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 1 Page 12 of 25 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-638-6 Duke Energy intends to implement Code Case N-638-6 for the inservice inspection program during the fourth ten-year interval as submitted to the NRC on October 23, 2017 (ADAMS Accession No. ML17296A323).

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 1 Page 13 of 25

8. Precedents

1. Davis-Besse Nuclear Power Station Relief Request RR-A34, September 20, 2010, ADAMS Accession Number ML102571569

2. Calvert Cliffs Nuclear Power Plant Relief Request RR-PZR-01, December 9, 2011, ADAMS Accession Number ML113360526

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, September 13, 2013, ADAMS Accession Number ML13238A154

5. Shearon Harris Nuclear Power Plant, Unit 1, Relief Request I3R-13, April 11, 2014, ADAMS Accession Number ML14093A075

6. Shearon Harris Nuclear Power Plant, Unit 1, Relief Request I3R-15, September 18, 2015, and January 6, 2016, ADAMS Accession Numbers ML15203A702 and ML15342A043

7. Shearon Harris Nuclear Power Plant, Unit 1, Relief Request I3R-16, December 27, 2016, ADAMS Accession Number ML16343A220

9. References

1. ASME Code Case N-638-6 Similar and Dissimilar Metal Welding Using Ambient Temperature Machine GTAW Temper Bead Technique,Section XI, Division 1

2. NRC Regulatory Guide 1.147, Revision 18, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, ADAMS Accession Number ML16321A336

3. ASME Code Case N-729-4 Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1

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

5. Shearon Harris Nuclear Power Plant, Unit 1, Relief Request I3R-15, April 2, 2015, ADAMS Accession Numbers ML15092A236, ML15105A521, ML15114A480, and ML15120A406

6. Shearon Harris Nuclear Power Plant, Unit 1, Relief Request I3R-16, October 19, 2016, ADAMS Accession Numbers ML16294A218, ML16295A159 and ML16298A133

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 1 Page 14 of 25

.. -_l,,.__ _

I r

r I

I II I I I

LL+JJ I I I Figure 1. Machining

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 1 Page 15 of 25 "TRIPLE PCNNT Figure 2. Welding

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 1 Page 16 of 25 r-..__

"x" is 1.5" for all nozzles.=:. 30° and 1" for all nozzles> 30° relative to the head surface Figure 3. Examination Areas Pre-Weld PT I-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 Relief Request I4R-18 HNP-18-045 Enclosure 1 Page 17 of 25 CRDMNozzle Head Figure 4. UT 0° and 45° L-wave Beam Coverage Looking Clockwise and Counter-clockwise

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 1 Page 18 of 25 CRDMNozzle Hea d Figure 5. UT 45° L-wave Beam Coverage Looking Down

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 1 Page 19 of 25 CRDMNozzle Head Figure 6. UT 45° L-wave Beam Coverage Looking Up

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 1 Page 20 of 25 CRDMNozzle Head Figure 7. UT 70° L-wave Beam Coverage Looking Down

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 1 Page 21 of 25 CRDMNozzle Head Figure 8. UT 70° L-wave Beam Coverage Looking Up

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 1 Page 22 of 25 A ___ T "a" is 1.5" for all nozzles~ 30° and 1" for all nozzles > 30° relative to the head surface Figure 9. PSI and ISi Weld and Nozzle Base Metal Surface Examination Area (A-B-C-D)

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 1 Page 23 of 25 SA-533 GR. B, CL. 1 (5 3/4" MIN THICK)

CLADDING (3/16")

~

ALLOY 82/182 J-GROOVE WELD CROM NOZZLE (ALLOY 600)

LOCATION OF AXIAL INDICATION Figure 10. Location of Axial Indications

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 1 Page 24 of 25

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<t> T><EAN"'- COlJPLE POF'f Figure 11. Reactor Vessel Head Penetration Locations

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 1 Page 25 of 25 Table 1. Flaw Characteristics Ind. ID/O Depth to Thru Azimuth Orientation Nozzle No. D Ind. Wall Length (degrees) Ax/Circ Type 33 1 OD 0.526 0.100 0.223 8.2 AXIAL PWSCC 33 2 OD 0.531 0.095 0.260 352.5 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". Tube 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 Relief Request I4R-18 HNP-18-045 HNP-18-045 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400/Renewed License No. NPF-63 Relief Request I4R-18, Reactor Vessel Closure Head Nozzle Repair Technique, Inservice Inspection Program, Fourth Ten-Year Interval Enclosure 2 Affidavit Supporting Withholding of Proprietary Information

AFFIDAVIT COMMONWEALTH OF VIRGINIA )

) ss.

CITY OF LYNCHBURG )

1. My name is Philip A Opsal. I am Manager, Product Licensing, for Framatome Inc., (formally known as AREVA Inc.), and as such I am authorized to execute this Affidavit.

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

3. I am familiar with the Framatome Inc. information contained in the following document: Framatome Inc. Calculation Summary Sheet, 32-9176350-006, "Shearon Harris Unit 1 CRDM/CET Nozzle As-Left J-groove Weld Analysis," referred to herein as "Document."

Information contained in this Document has been classified by Framatome Inc. as proprietary in accordance with the policies established by Framatome Inc. for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by Framatome Inc. and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Documentas proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document 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 Framatome Inc. to determine whether information should be classified as proprietary:

(a) The information reveals details of Framatome lnc.'s 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 Framatome Inc.

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

(e) The information is vital to a competitive advantage held by Framatome Inc.,

would be helpful to competitors to Framatome Inc., and would likely cause substantial harm to the competitive position of Framatome Inc.

The information in this Document is considered proprietary for the reasons set forth in paragraphs 6(b), 6(c) and 6(d) above.

7. In accordance with Framatome lnc.'s policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside Framatome Inc. only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. Framatome lnc.'s 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.

SUBSCRIBED before me this ~l\ptb

~--

day of Ajvt;l , 2018.

Sherry L. McFaden NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 10/31/18 Reg.# 7079129 SHERRYL. MCFADEN Notary Public Commqnwealth of Virginia 7079129 My Commission Expires Oct 31, 2018

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 HNP-18-045 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400/Renewed License No. NPF-63 Relief Request I4R-18, Reactor Vessel Closure Head Nozzle Repair Technique, Inservice Inspection Program, Fourth Ten-Year Interval Enclosure 4 Regulatory Commitment

U.S. Nuclear Regulatory Commission Relief Request I4R-18 HNP-18-045 Enclosure 4 Page 1 of 1 Regulatory Commitment The following table identifies an action committed to in this letter. Statements in this submittal with the exception of those in the table below are provided for information purposes and are not considered commitments.

Commitment Expected Completion Dates Duke Energy commits to take measurements This commitment will be completed prior during the repair process to confirm that the to entering Mode 5 for plant startup.

actual local head thickness bounds the primary stress limit analysis assumptions for a limiting design life greater than or equal to 5 years as described in Enclosure 1 of this submittal.