Text

Relief Request for Half-Nozzle Repair of Reactor Vessel Closure Head Penetration 46 - Technical Report, ANP-4023NP, Revision 0, December 2022
	ML23088A217
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Site:	Arkansas Nuclear
Issue date:	12/31/2022
From:	; Framatome
To:	; Office of Nuclear Reactor Regulation
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ML23088A214	List: 2CAN032304, Supplement to the Request for Alternative Regarding the Half-Nozzle Repair of Reactor Vessel Closure Head Penetration 46; ML23088A217;
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ENCLOSURE 2 2CAN032304 ARKANSAS NUCLEAR ONE, UNIT 2 RELIEF REQUEST FOR HALF-NOZZLE REPAIR OF RVCH PENETRATION #46 - TECHNICAL REPORT, ANP-4023NP, REVISION 0, DECEMBER 2022 (NON-PROPRIETARY)

0414-12-F04 (Rev. 004, 04/27/2020)

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 3 Nature of Changes Section(s)

Item or Page(s) Description and Justification 1 All Initial Issue

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 4 Contents Page

1. ASME CODE COMPONENT AFFECTED ............................................................ 7

2. APPLICABLE CODE EDITION AND ADDENDA .................................................. 7

3. APPLICABLE CODE REQUIREMENTS .............................................................. 7

4. REASON FOR REQUEST.................................................................................... 8

5. PROPOSED ALTERNATIVE AND BASIS FOR USE ......................................... 11 5.1. Welding Requirements............................................................................. 11 5.2. IDTB Modification Acceptance Examinations .......................................... 13 5.3. Triple Point Anomaly ................................................................................ 13 5.4. Flaw Characterization and Successive Exams - RVCH Original J-Groove Weld............................................................................ 16 5.5. lnservice Inspection (ISI) of VHPs ........................................................... 18 5.6. General Corrosion Impact on Exposed Low Alloy Steel .......................... 19 5.7. Conclusions ............................................................................................. 20

6. DURATION OF PROPOSED ALTERNATIVE .................................................... 21

7. ADDITIONAL INFORMATION ............................................................................ 21 7.1. VHP Weld Qualification Mockup UT Acceptance ..................................... 21

8. PRECEDENTS ................................................................................................... 22

9. REFERENCES ................................................................................................... 23

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 5 List of Figures Figure 1 Nozzle Machining ....................................................................................... 25 Figure 2 Nozzle Weld ............................................................................................... 26 Figure 3 Nozzle Examination .................................................................................... 27 Figure 4 Nozzle 0° and 45°L UT Beam Coverage Looking Clockwise and Counter-Clockwise ..................................................................................... 28 Figure 5 Nozzle 45°L UT Beam Coverage Looking Down ........................................ 29 Figure 6 Nozzle 45°L UT Beam Coverage Looking Up ............................................ 30 Figure 7: Nozzle 70°L UT Beam Coverage Looking Down ........................................ 31 Figure 8: Nozzle 70°L UT Beam Coverage Looking Up ............................................ 32 Figure 9: Nozzle ISI UT Examination ......................................................................... 33 Figure 10: Reactor Vessel Head Penetration Locations .............................................. 34 Figure 11: Indication Location...................................................................................... 35

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 6 Nomenclature Acronym Definition ADAMS Agencywide Documents Access and Management System ANO-2 Arkansas Nuclear One, Unit 2 ASME American Society of Mechanical Engineers B&PV Boiler and Pressure Vessel Code CEDM Control Element Drive Mechanism CMTR Certified Material Test Report Cr Chromium EPFM Elastic Plastic Fracture Mechanics Fe Iron GTAW Gas Tungsten Arc Welding HAZ Heat Affected Zone ICI In-core Instrumentation IDTB Inside Diameter Temper Bead ISI Inservice Inspection LEFM Linear Elastic Fracture Mechanics NDE Nondestructive Examination Ni Nickel NRC Nuclear Regulatory Commission PCS Primary Coolant System PT Dye Penetrant Examination PWHT Post Weld Heat Treatment PWSCC Primary Water Stress Corrosion Cracking RFO Refueling Outage RPV Reactor Pressure Vessel RVCH Reactor Vessel Closure Head UT Ultrasonic Examination VHP Vessel Head Penetration

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 7 ENTERGY NUCLEAR OPERATIONS, INC.

ARKANSAS NUCLEAR ONE, UNIT 2 Relief Request Number ANO2-RR-22-001

1. ASME CODE COMPONENT AFFECTED Component: Reactor Vessel Closure Head (RVCH) Penetration #46 Code Class: 1 Exam. Cat.: American Society of Mechanical Engineers (ASME)

Code Case N-729-6 Item No. B4.20 Unit: Arkansas Nuclear One, Unit 2 (ANO-2)

Interval: Fifth (5th) (March 26, 2020 to March 25, 2030)

2. APPLICABLE CODE EDITION AND ADDENDA ASME Section XI, 2007 Edition through 2008 Addenda ASME Section XI Code Case N-729-6, as amended in 10 CFR 50.55a(g)(6)(ii)(D)

ASME Section III, 1968 Edition through Summer 1970 Addenda (Original Construction Code)

ASME Section III, Subsection NB, 1992 Edition

3. APPLICABLE CODE REQUIREMENTS The applicable requirements of the following ASME Boiler & Pressure Vessel (B&PV) Code and Code Cases from which relief is requested are as follows:

ASME Code,Section XI, 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.

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 8

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).

ASME Code,Section III, 1992 Edition

NB-5245, Partial Penetration Welded Joints, specifies progressive surface examination of partial penetration welds.

NB-5330(b) states:

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

Code Case N-638-7, Similar and Dissimilar Metal Welding Using Ambient Temperature Machine GTAW Temper Bead Technique This Code Case provided requirements for automatic or machine gas tungsten arc welding (GTAW) of Class 1 components without the use of preheat or post-weld heat treatment. The condition imposed on this Code Case by Regulatory Guide (RG) 1.147, Revision 19 (effective at the time of implementation) is addressed below in paragraph 7.1.

Paragraph 1(g) states:

Peening may be used, except on the initial and final layers.

Paragraph 2(b) permits use of existing welding procedures qualified in accordance with previous revisions of the Code Case. When the existing welding procedure was qualified in accordance with N-638-4, the test coupon base material was post-weld heat treated to comply with paragraph 2.1(a) of the Code Case (N-638-4) which states:

The materials shall be post-weld heat treated to at least the time and temperature that was applied to the materials being welded.

4. REASON FOR REQUEST Entergy Operations, Inc. (Entergy) ANO-2 performed ultrasonic (UT) examinations of RVCH penetration nozzles in Refueling Outage 2R28 (Fall 2021), in accordance with ASME Code Case N-729-6 (Item No. B4.20)1, and detected an axial, planar indication in Control Element Drive Mechanism (CEDM) Nozzle #46. The indication was located along the outside diameter, downhill side of the nozzle in the J-groove weld fillet region (see Figure 11). Eddy 1

Code Case N-729-6 as amended in 10 CFR 50.55a(g)(6)(ii)(D) and supplemented by Relief Request ANO2-ISI-022 (Reference 4)

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 9 current (ECT) examination along the outside diameter of the nozzle and J-groove weld confirmed that the indication was surface connected. A supplemental liquid penetrant (PT) examination of the flaw was also performed to confirm the indication's location. The UT leak path assessment on CEDM Nozzle #46 did not provide any evidence of base material degradation along the RVCH nozzle bore. Additionally, the bare metal visual examination (Item B4.10) of the RVCH did not identify any evidence of reactor coolant leakage such as boric acid deposits or base material wastage.

The UT examination indicated that the flaw in CEDM Nozzle #46 was 0.204-inch deep and 0.400-inch long. The total length of the flaw was estimated to be 1.03-inches based on the UT and supplemental PT examinations. The PT examination also indicated that the flaw extended into the thread relief region of the nozzle while the lower tip of the flaw was approximately 0.7-inch above the bottom of the nozzle. Figure 10 shows the relative location of the nozzles in the RVCH and Figure 11 shows the general location of the axial indication.

The repair technique sometimes referred to as the half-nozzle repair, was performed to correct the identified condition on CEDM Nozzle #46. The half-nozzle repair involved 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. The new weld also attached a replacement lower nozzle that provided a means for re-attaching the guide cone. This technique required relief from certain aspects of the ASME B&PV Code as described below.

Because of the risk of damage to the RVCH material properties or dimensions, it was not feasible to apply the post weld heat treatment (PWHT) requirements of the original Construction Code. As an alternative to the requirements of the RVCH Code of Construction, Entergy performed a modification of CEDM Nozzle #46 utilizing the Inside Diameter Temper Bead (IDTB) welding method to restore the pressure boundary of the degraded nozzle penetration. The IDTB welding method was 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 PWHT. The modification described below was performed in accordance with the 2007 Edition through 2008 Addenda of ASME Section XI, Code Case N-638-7, Code Case N-729-6, and the alternatives discussed in Section 5.

Basic steps for the IDTB repair are:

1. Roll expansion of the nozzle above the area to be modified to stabilize the nozzle and prevent any movement when the nozzle is separated from the nozzle-to-RVCH J-groove weld.

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

3. PT examination of the machined area (Refer to Figure 1).

4. Welding the remaining portion of the nozzle and the new replacement Alloy 690 nozzle

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 10 using Alloy 52M weld material (Refer to Figure 2).

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

6. PT and UT examination of the weld and adjacent region (Refer to Figure 3).

Note the figures in this request are provided to assist in clarifying the above description.

They are not intended to provide design information such as the location of the CEDM nozzle weld relative to the inner and outer spherical radii of the RVCH.

Stresses introduced during the controlled roll expansion process implemented per design and fabrication controls did not create regions that would be more susceptible to Primary Water Stress Corrosion Cracking (PWSCC) than other regions that had been previously evaluated and found acceptable. Two fabrication parameters were controlled to ensure that the nozzle roll expansion is effective in performing its design function of mechanical support for the nozzle prior to the application of the IDTB weld. The parameters of interest are tool insertion depth and the torque setting on the assembly tool.

Tool insertion depth, based on tooling setup height, was controlled so that the rolled region was contained within the RVCH penetration bore. The torque applied to the roll expander was controlled so that the desired amount of plastic deformation occurred. The torque limiter assembly was set and independently verified with a calibrated torque wrench prior to use.

As noted above, the roll expansion process was completed for CEDM Nozzle #46 and the two parameters of interest (tool insertion depth and applied torque) that could impact the susceptibility to PWSCC were validated to be within process specifications. Additionally, rotary peening was applied to remediate the tensile surface stresses in the roll expanded region. As a result, adequate measures were applied during the modification of Nozzle #46 with respect to PWSCC for the life of the repair.

Entergy submitted a relief request for one cycle of operation beginning after 2R28 (Reference 11). The NRC Safety Evaluation approving the one cycle relief request was issued on April 7, 2022 (ML22073A095). This relief request contains the analyses and justification through the remainder of the current renewed license (July 17, 2038). Entergy has determined that modification of CEDM Nozzle #46 utilizing the alternatives specified in this request will provide an acceptable level of quality and safety for the remainder of the current renewed license. Relief is requested in accordance with 10 CFR 50.55a(z)(1).

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 11

5. PROPOSED ALTERNATIVE AND BASIS FOR USE 5.1. Welding Requirements The half nozzle repair on CEDM Nozzle #46 was performed using the ambient temperature temper bead process of Code Case N-638-7. Paragraph 2(b) of the Code Case permitted use of welding procedures qualified in accordance with previous Code Case revisions.

Accordingly, the welding procedure used on Nozzle #46 was qualified in accordance with N-638-4 (an earlier revision). Code Case N-638-4, paragraph 2.1(a) states:

The materials shall be post-weld heat treated to at least the time and temperature that was applied to the materials being welded.

Post-weld heat treatment (PWHT) can slightly degrade the fracture (notch) toughness of low alloy steels. Therefore, it is both reasonable and conservative to perform a simulated PWHT of test samples that will be used to evaluate base materials that have received PWHT during fabrication and placed into reactor service. However, it is not conservative to perform a simulated PWHT of welding qualification test plate material that will be compared to the temper bead heat affected zone (HAZ) for acceptance.

The temper bead weld procedure qualification is required to demonstrate that the Charpy V-notch test results from the weld HAZ are no less than the Charpy V-notch test results for the unaffected base material. EPRI Report 1025169, Section 3.0 (Reference 5) documents that simulated PWHT on procedure qualification test plates degrades the notch toughness of the test plate increasing the contrast between the impact properties of the base material test plate and the temper bead weld HAZ. In other words, the simulated PWHT makes passing the impact testing requirements of the temper bead procedure qualification less difficult.

Therefore, simulated PWHT on the temper bead test coupon does not provide conservative results when the simulated PWHT time exceeds the actual PWHT time applied to the component during construction.

The RVCH material at ANO-2 had at least 50 hours5.787037e-4 days 0.0139 hours 8.267196e-5 weeks 1.9025e-5 months of PWHT but the weld procedure qualification test plate had 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months of simulated PWHT. This condition did not comply with Code Case N-638-4, paragraph 2.1(a) which required simulated PWHT on the temper bead qualification test plate to be equivalent to or exceed the total aggregate time applied to the component to be welded. There is no maximum limit on the simulated PWHT time.

The simulated PWHT requirement of Code Case N-638 was recognized by the ASME Code Committee as non-conservative and was changed in Revision 7. Code Case N-638-7, paragraph 2.1(a) states that simulated PWHT of the test assembly is neither required nor prohibited. However, if used, the simulated PWHT shall not exceed the time or temperature already applied to the base material to be welded. The welding procedure used to implement the half nozzle repair on CEDM Nozzle #46 complies with this requirement.

Code Case N-638-7 was conditionally approved by the NRC in RG 1.147, Revision 19 at the time of implementation. The NRC condition, was unrelated to simulated PWHT, and stated:

Demonstration of ultrasonic examination of the repaired volume is required using representative samples that contain construction-type flaws. Therefore, the enhanced and more conservative simulated PWHT requirements in Code Case N-638-7 have also been

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 12 approved by the NRC. The NRC condition that was applied to Code Case N-638-7 was incorporated into subsequent versions of the Case; therefore, the most recent issuance of RG 1.147 (Revision 20, issued in December 2021) lists Code Case N-638-10 as an acceptable Section XI Code Case in Table 1.

In summary, ambient temperature temper bead welding was performed on CEDM Nozzle

46 in accordance with Code Case N-638-7 while the welding procedure was qualified in accordance with Code Case N-638-4. The qualified welding procedure did not comply with the simulated PWHT requirements of Revision 4 of the Code Case but did comply with the enhanced and more conservative simulated PWHT requirements in Revision 7 (i.e.,

N-638-7).

Therefore, Entergy requests approval to apply the simulated PWHT requirements of Case N-638-7, paragraph 2.1(a) when using the temper bead welding procedure on CEDM Nozzle #46.

Code Case N-638-7, paragraph 1(g) states:

Peening may be used, except on the initial and final layers.

Rotary peening was performed on the final layer to provide further assurance of the modified configuration being resistant to PWSCC. However, peening on the final layer of a temper bead weld is prohibited by ASME Code Case N-638-7, paragraph 1(g). This prohibition refers to the high cold-work peening that is traditionally used for configuration distortion control during welding, as was interpreted by ASME Xl-1-13-19 for Code Case N-606-1.

This is not considered applicable to the rotary peening process, which is highly controlled, uniform, and only influences a shallow surface layer (approximately 10 mils at the HAZ and 20 mils at the base metal). The uniform compressive stress layer created by the rotary peening process does not inhibit subsequent NDE. Furthermore, this residual compressive stress layer has been shown to greatly reduce PWSCC initiation. Recognizing these benefits, the ASME Code Committee revised Code Case N-638 (i.e., N-638-8) to allow use of peening processes designed to reduce residual surface tensile stresses on the final layer or surface of the weld.

Upon completion of peening, visual and surface examinations were performed on the peened surface. However, while the peening operation provides increased resistance to PWSCC initiation, the inspection frequency of ISI examinations on CEDM Nozzle #46 will comply with that specified in Item B4.20 of Code Case N-729-6 as approved by the NRC in 10 CFR 50.55a(g)(6)(ii)(D).

ASME Code Section III, Nonmandatory Appendix W, W-2140, clearly describes the beneficial nature of compressive stresses for the mitigation of stress corrosion cracking (SCC) susceptibility. It states that shot peening, as a form of stress improvement, can be used to place the inside diameter of piping in a compressive residual stress state to resist SCC. Extensive laboratory testing performed as part of MRP-61, "An Assessment of the Control Rod Drive Mechanism (CRDM) Alloy 600 Reactor Vessel Head Penetration PWSCC Remedial Techniques," indicates that shot peening successfully inhibits PWSCC initiation.

With rotary peening, the shot is captured in a flap and regularly spaced such that it uniformly imparts compressive stresses on metal surfaces.

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 13 Therefore, Entergy requests relief from Code Case N-638-7, paragraph 1(g).

5.2. IDTB Modification Acceptance Examinations ASME Section Ill, 1992 Edition, NB-5245, specifies progressive surface examination of partial penetration welds. The Construction Code requirement for progressive surface examination, in lieu of volumetric examination, was because volumetric examination is not practical for the conventional partial penetration weld configurations. Therefore, the following combination of UT and PT examinations were performed.

The modified Vessel Head Penetration (VHP) weld is suitable for UT examination and the weld is accessible from both the top and bottom sides (Refer to Figure 4 through Figure 8).

UT volumetric examination of the modified configuration was performed as specified in ASME Code Case N-638-7, 4(a)(2) and 4(a)(4). RG 1.147, Revision 19, conditionally approved Code Case N-638-7 with the condition that UT volumetric examinations be demonstrated using representative samples which contain construction type flaws. See Section 7.1 for details. The acceptance criteria of NB-5330 of the 1992 Edition of the ASME Section Ill Code apply to all flaws identified within the examined volume.

The UT examination system is capable of scanning from cylindrical surfaces with inside diameters of approximately 2.82-inch. The scanning was performed using a 0° L-wave transducer, 45° L-wave transducers in two opposed axial directions, and 70° L-wave transducers in two opposed axial directions as well as 45° L-wave transducers in two opposed circumferential directions. The weld received 100% examination coverage.

Additionally, the low alloy steel extending to 0.25-inch beneath the weld into the low alloy steel base material (see Figure 3) was examined using the 0° L-wave transducer searching for evidence of under bead cracking and lack of fusion in the HAZ. These examinations satisfy ASME Section III, NB-5330 requirements. The repair volume was extended to include 1-inch of Alloy 600 nozzle material above the weld and 1-inch of Alloy 690 material below the weld. UT examination coverage is as shown in Figure 4 through Figure 8 of this submittal.

In addition to the UT examinations, a surface PT examination was performed on the entire weld as shown in Figure 3. The final examination of the new weld and immediate surrounding region was 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 acceptance criteria of NB-5350 in ASME Section III, 1992 Edition were applicable.

The combination of performing PT and UT examinations depicted in Figure 3 during the IDTB repair provided assurance of structural integrity. Thus, Entergy requests relief from the progressive surface examination requirements specified in NB-5245.

5.3. Triple Point Anomaly ASME Section Ill, NB-5330(b) states:

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 14 Indications characterized as cracks, lack of fusion, or incomplete penetration are unacceptable regardless of length.

An artifact of ambient temperature temper bead welding is an anomaly in the weld at the triple point. There are two triple points in the modification. The upper 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 lower triple point is the point in the repair weld where the low alloy steel RVCH base material, the Alloy 690 replacement nozzle, and the Alloy 52M weld intersect. The locations of the upper and lower triple points for the VHP modification are shown in Figure 2.

The 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 through-wall extent and are assumed to exist, for purposes of analysis, around the entire bore circumference at the triple point location.

Linear elastic fracture mechanics (LEFM) and limit load Life of Repair analyses (Reference 6) were performed for this repair considering flaws at the weld anomaly triple point locations. The analyses resulted in acceptable crack growth for the life of the repair, from the time of repair to the end of 60 year plant operating life, for a total of 17 years. The process for the Life of Repair analyses was as follows:

1. The initial flaw size for the postulated flaws in the triple point anomaly analysis was 0.100-inches. Crack growth analysis determined the future flaw size and concluded that it is acceptable for the stated life.

2. A fracture mechanics analysis was performed for the design configuration to provide justification, in accordance with ASME Section XI, for operating with the postulated triple point anomaly. The anomaly was modeled as a 0.100-inch deep crack-like defect, initiating at the triple point locations, considering the most susceptible material for propagation. Postulated flaws could be oriented within the anomaly such that there are three possible flaw propagation paths, as discussed in Items 3 and 4 below.

3. Circumferential and Axial Flaws: Flaw propagation was across the nozzle wall thickness from the outside diameter to the inside diameter of the nozzle housing for the upper and lower triple points.

a. The shortest paths were through the nozzle thickness at the upper and lower triple points (see Figure 2). By using a fatigue crack growth rate of twice that of the rate of Alloy 600 material to bound the Alloy 600/690 nozzle and Alloy 52M weld materials either in-air (for upper triple point flaws) or exposed to the reactor water environment (lower triple point flaws), it was ensured that all potential paths through the HAZ between the new repair weld and the Alloy 600 nozzle material was bounded.

b. For completeness, two types of flaws were postulated at the outside surface of the nozzle IDTB repair weld. A 360-degree continuous circumferential flaw, lying in a horizontal plane, was considered to be a conservative representation of crack-like defects that would exist in the weld triple point anomaly. This flaw was

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 15 subjected to axial stresses in the nozzle. An axially oriented semi-circular outside surface flaw was 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.

4. Cylindrical Flaw: Flaw propagation extended down from the upper triple point or up from the lower triple point along the outside surface of the repair weld between the upper and lower triple points.

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

5. The results of the LEFM and limit load analysis have demonstrated that a 0.100-inch weld anomaly is acceptable for life of repair, which is 17-years of operation following the VHP nozzle inside diameter temper bead weld repair. Acceptable design margins were demonstrated for all flaw propagation paths considered in the analysis. For the low alloy steel RVCH base material, the minimum fracture toughness margin has been shown to be greater than the required margin of 10 (3.16) for normal/upset/test conditions where pressure exceeds 20% of the design pressure, and 2 (1.41) for emergency/faulted conditions and normal/upset/test conditions where pressurization does not exceed 20%

of the design pressure and during which the minimum temperature is not less than RTNDT per ASME Section XI, IWB-3613. A limit load analysis was also performed considering the ductile Alloy 600/Alloy 690 materials along flaw propagation path lines.

This analysis showed a limit load margin greater than the required margin per ASME Section XI, IWB-3642.

For qualification of the IWB-3613 criteria, the analysis credits the following analytical limitations after IDTB repair of CEDM Nozzle #46 (in recognition of the LTOP lift setting)*:

a. The minimum fluid temperature for performing [

]

b. The maximum [ ] transient pressure when the fluid temperature is less or equal to [ ] is [ ].

c. The maximum [ ] transient pressure when the fluid temperature is less or equal to [ ] is [ ]

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 16

(*) Note that, per Section 5.2.2.4 of Reference [12], the low temperature overpressure protection (LTOP) lift setting is limited to 430 psig for reactor coolant fluid temperature less or equal to 220°F and therefore, [

] operations are protected from exceeding this pressure at the

[ ] temperatures identified in the analysis limitations above.

6. Since the postulated outside diameter flaw in the weld anomaly at the upper triple point is 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. [

]

7. [

] satisfying the ASME Section XI criteria.

Entergy requests relief from the acceptance criteria specified in NB-5330(b) of ASME Section III to permit anomalies, as described herein, at the triple point area to remain in service.

5.4. Flaw Characterization and Successive Exams - RVCH Original J-Groove Weld The assumptions of IWB-3600 of ASME Section XI are that cracks are fully characterized in accordance with IWB-3420 in order to compare the calculated parameters to the acceptable parameters addressed in IWB-3500. There are no qualified UT examination techniques for examining the original nozzle-to-RVCH J-groove welds. Therefore, since it is impractical to characterize the flaw geometry that may exist therein, it was conservatively assumed that the "as-left" condition of the remaining J-groove weld includes flaws extending through the entire Alloy 82/Alloy 182 J-groove weld and buttering. It was 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/Alloy 182 weld would propagate by PWSCC through the weld and buttering to the interface with the low alloy steel RVCH material. Any growth of the postulated "as-left" flaw into the low alloy steel would be by fatigue crack growth under cyclic loading conditions.

Based on a combination of linear elastic and elastic-plastic fracture mechanics the "Life of Repair" analyses (Reference 7) resulted in a fatigue crack growth life for the "as-left" J-groove flaw of at least 17 years after the IDTB repair recognizing the following analytical limitations ensured by the LTOP lift setting*:

1. The minimum fluid temperature for performing [

]

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 17

2. The maximum [ ] transient pressure when the fluid temperature is less or equal to [ ]

3. The maximum [ ] transient pressure when the fluid temperature is less or equal to [ ]

(*) Note that, per Section 5.2.2.4 of Reference [12], the low temperature overpressure protection (LTOP) lift setting is limited to 430 psig for reactor coolant fluid temperature less or equal to 220°F and therefore, [ ] operations are protected from exceeding this pressure at the [ ] temperatures identified in the analysis limitations above.

The process for the Life of Repair analyses was as follows:

Relief is requested from flaw characterization specified in IWB-3420.

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 18 In addition, the potential for debris from a cracked J-groove weld remnant was considered.

Radial cracks (relative to the nozzle) were postulated to occur in the J-groove weld due to the dominance of higher hoop stresses relative to axial stresses. The possibility of transverse cracks occurring that could subsequently intersect the radial cracks is considered remote as there are minimal driving forces for cracks in the transverse direction. The radial cracks would relieve the driving forces for any potential transverse cracks. There are no known service conditions that could drive radial cracks and transverse cracks to intersect to produce a loose part. There is extensive operating experience with remnant J-groove welds for which there are no known cases of debris generation (loose parts) due to PWSCC of the remnant J-groove weld. Therefore, cracking of the J-groove weld resulting in debris (loose parts) is not expected.

5.5. lnservice Inspection (ISI) of VHPs Code Case N-729-6 as approved by the NRC in 10 CFR 50.55a specifies requirements for performing ISI examinations of RVCHs with nozzles having partial penetration welds. Code Case N-729-6 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-729-6.

ISI examination of CEDM Nozzle #46 will be performed using a volumetric examination method. In lieu of the volumetric examination region that extends 1.5-inch above and 1.5-inch below the J-groove weld shown in Figure 2 of Case N-729-6, an alternative examination region will be interrogated for the repair weld. The examination volume will extend up to the outer surface of the head (greater than 1.5-inch above the repair weld),

including the rotary peened surfaces (including the roll transition region), and 1-inch below the repair weld as shown in Figure 9. Examination coverage below the weld will be less than the 1.5-inch requirement due to geometric limitations; however the coverage will extend a minimum of 1-inch below the weld and will obtain the maximum volume practical.

Examination coverage of 1-inch below the repair weld is considered sufficient due to the following:

The replacement nozzle material (Alloy 690) is resistant to PWSCC

The replacement nozzle is non-pressure boundary material

The new pressure boundary weld (Alloy 52M) is resistant to PWSCC The repair performed during RFO 2R28 modified the examination volume depicted in Figure 2 of Code Case N-729-6. Figure 9 of this submittal establishes the examination volume for ISI examinations. The examination volume also includes the rotary peened surfaces. Successive examinations required by Code Case N-729-6 will be performed on CEDM Nozzle #46 during each refueling outage.

All other ANO-2 RVCH CEDM and ICI nozzles will continue to be examined in accordance with Code Case N-729-6 as modified by 10 CFR 50.55a(g)(6)(ii)(D) and other NRC approved alternatives.

Therefore, future ISI examinations will comply with Code Case N-729-6 as modified by 10 CFR 50.55a(g)(6)(ii)(D) and as depicted in Figure 9.

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 19 5.6. General Corrosion Impact on Exposed Low Alloy Steel The IDTB nozzle modification left 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 wetted surface. Galvanic corrosion, hydrogen embrittlement, SCC, and crevice corrosion are not expected to be a concern for the exposed low alloy steel base metal. General corrosion of the exposed low alloy steel base metal will occur in the area between the IDTB weld and the original J-groove weld. Due to the depletion of oxygen, tight geometry, and lack of Primary Coolant System (PCS) flow at the exposed low alloy steel, general corrosion will significantly decrease after a period of time. As the surface of the low alloy steel passivates, the long-term corrosion rate is expected to be negligible. However, a conservative, sustained, corrosion rate was applied and the resultant increase in bore diameter was considered in the reinforcement calculation (per NB-3330) as part of the ASME Section III analysis. The corrosion evaluation (Reference 8) and the ASME Section III analysis (Reference 10) are attached to this submittal.

Galvanic Corrosion The results of the NRCs boric acid corrosion program have shown that the galvanic difference between SA-533 Grade B, Alloy 600, and Type 308 stainless steel (nominal chemistry of RVCH cladding) is not significant enough to consider galvanic corrosion as a strong contributor to the overall boric acid corrosion process (NUREG-1823). Therefore, it was judged that galvanic corrosion between the exposed RVCH low alloy steel, Alloy 600, Alloy 690, or their weld metals is not a concern for this repair configuration. This is supported by studies documented in EPRI Report 1000975 in which low alloy steel specimens were coupled and uncoupled to stainless steel exposed to a borated water environment at various temperatures. The corrosion rates for the coupled and uncoupled conditions were determined to be similar. Additionally, galvanic corrosion of carbon steel coupled to stainless steel in boric acid solution in the absence of oxygen should be quite low. The results of this study are also applicable to nickel-based alloys as austenitic stainless steels have approximately the same corrosion potential as nickel-based alloys such as Alloy 600 and Alloy 690.

Hydrogen Embrittlement Hydrogen embrittlement occurs when a material property is degraded due to the presence of hydrogen. This type of damage usually occurs in combination with an acting stress. The hydrogen concentration in the RVCH will be greatest at the exposed surface and decreases across the thickness of the RVCH to the trace concentration of hydrogen in the low alloy steel. Hydrogen concentrations in the reactor coolant system are deemed insufficient to induce hydrogen cracking in the low alloy steel of the RVCH. Therefore, it was determined that hydrogen embrittlement is not a concern for the exposed RVCH low alloy steel in the repaired configuration. This conclusion is supported by many cases of low alloy steels being exposed to primary coolant without any observed cracking due to hydrogen embrittlement.

Stress Corrosion Cracking

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 20 There is extensive Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) operating experience related to low alloy steels being exposed to the reactor coolant environment. This operating experience has not identified any known occurrence of stress corrosion cracking of the low alloy steel of RVCHs. Likewise, there are no existing ASME Section XI Code rules or NRC regulations addressing this issue in RVCH low alloy steels in PWR reactor coolant environment. Therefore, it has been determined that stress corrosion cracking of the low alloy steel of the RVCH is not a concern for this repair configuration.

Crevice Corrosion The geometry of the gap between the RVCH and replacement nozzle could create conditions for crevice corrosion. However, operating experience for PWRs shows that crevice corrosion of low alloy steels associated with these half nozzle repairs is not a problem in PWR systems due to expected low oxygen contents. Furthermore, the surface of the low alloy steel material will passivate with time, decreasing the rate of corrosion within the crevice. Therefore, it was determined that crevice corrosion of the low alloy is not a concern.

General Corrosion Corrosion of the exposed low alloy steel is not expected to be a concern based on existing operating experience. The surface of the low alloy steel material will passivate with time, decreasing the rate of general corrosion. As corrosion products fill this gap, they will isolate the low alloy steel surface from the reactor coolant system, thereby, impeding the transport of oxygen which is necessary to sustain continued general corrosion. Due to the depleted oxygen, passivated surface, tight geometry, and lack of appreciable reactor coolant flow at the exposed low alloy steel, general corrosion will decrease significantly after a period of time.

5.7. Conclusions The IDTB repair to RVCH CEDM Nozzle #46 produced an effective repair that restored and maintained the pressure boundary integrity of the VHP. Other IDTB modifications have been performed successfully (see Section 8) and were in service for several years without any known degradation [e.g., Shearon Harris (2012, 2013, 2015, 2016 and 2018) and Palisades (2004, 2018, and 2020)]. This alternative provides improved structural integrity and reduced likelihood of leakage for the primary system. Detailed finite element based Life of Repair analyses (Reference 6) resulted in a crack growth life for the triple point anomaly flaw of at least 17 years after the repair (in 2038 after 60 years of plant operation).

Likewise, "Life of Repair" analyses performed on the "as-left" J-groove flaw (Reference 7) resulted in a fatigue crack growth life of at least 17 years after the repair (in 2038 after 60 years of plant operation). Corrosion of the exposed low alloy steel base material is not a concern due to lack of oxygen, tight geometry, and lack of reactor coolant system flow in the exposed region. The analyses and evaluations discussed herein justify continued use of the nozzle repair for the current operating life of the plant. Accordingly, the use of the alternative provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1).

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 21

6. DURATION OF PROPOSED ALTERNATIVE The overall acceptable life of the repair is based on the most limiting life predicted by three evaluations: the weld anomaly analysis (see 5.3) , the as-left J-groove weld analysis (see 5.4), and the PWSCC evaluation of the original Alloy 600 nozzle. A PWSCC evaluation of the IDTB repair is included in the (Life) Assessment Summary (Reference 9). For the weld anomaly and as-left J-groove weld analyses, the demonstrated 17-year life starts at the time of the repair. The compressive stress imparted by the rotary peening process is expected to mitigate the residual tensile surface stresses in the Alloy 600 CEDM nozzle material at the roll expanded transition area and adjacent to the Alloy 52M IDTB weld. Therefore, the PWSCC evaluation concluded that due to the compressive stresses achieved by rotary peening, PWSCC initiation is not expected during the 17-year life of the repair.

The duration of the proposed alternative is until the end of the current renewed license (July 17, 2038).

7. ADDITIONAL INFORMATION 7.1. VHP Weld Qualification Mockup UT Acceptance Volumetric examination was required by Code Case N-638-7. NRC RG 1.147, Revision 19 imposed a condition for this code case that required UT demonstration on representative samples which contained construction type flaws. Framatome, in support of many similar modifications, performed demonstrations using IDTB weld repair mockups since VHP modifications at Oconee Nuclear Station in 2001. The most recent procedure demonstration took place during the 2010 Davis Besse control rod drive mechanism (CRDM) repair campaign which included review of recorded automated data showing UT responses obtained from an IDTB weld mockup for the half-nozzle repair. This is the same mockup used for the procedure demonstration for Shearon Harris VHP nozzle modifications listed in Section 8.

To satisfy this requirement, an IDTB weld half-nozzle repair mockup containing reflectors to simulate construction type flaws applicable to this weld process was used. It contained 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 contained flat bottom holes drilled from the mockup outside diameter so that the hole face was normal to the nozzle surface to simulate under-bead cracking, and lack of bond, or lack of fusion throughout the weld volume. The examination procedure demonstrated the ability to detect a linear weld fabrication triple point anomaly extending 0.05-inch and greater into the weld.

A Nickle-Chromium-Iron (NiCrFe) alloy calibration block was used and contained a series of EDM notches at nominal depths of 10%, 25%, 50%, and 75% deep from both inside diameter and outside diameter surfaces in both the axial and circumferential orientation.

The block also contained 1/4T, 1/2T, and 3/4T deep end-drilled holes and side-drilled holes that were used for calibration.

During the repair at ANO-2, the site crew performed training on mockups for each of their respective specialties, i.e., machinists train on machining mockups, welders train on welding

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 22 mockups, and NDE personnel train on NDE mockups. Prior to examination of the repair weld at ANO-2, UT personnel practiced using the data files from the demonstration described above.

8. PRECEDENTS

1. Nuclear Management Company (NMC) letter to the NRC, "Request for Relief from ASME Section XI Code Requirements for Repair of Reactor Pressure Vessel Head Penetrations," dated October 11, 2005, ADAMS Accession Number ML052870321.

2. FirstEnergy Nuclear Operating Company (FENOC) letter to the NRC, "10 CFR 50.55a Request for Alternate Repair Methods for Reactor Pressure Vessel Head Penetration Nozzles," dated April 1, 2010, ADAMS Accession Number ML100960276.

3. Constellation Energy letter to the NRC, "Relief Request for Modifications to Pressurizer Heater Sleeves and Lower Level Nozzle Penetrations (RR-PZR-01),"

dated January 31, 2011, ADAMS Accession Number ML110340059.

4. Progress Energy letter to the NRC, "Relief Request I3R-09 Reactor Vessel Closure Head Nozzles Inservice Inspection Program - Third Interval," dated May 3, 2012, ADAMS Accession Number ML12131A663.

5. Progress Energy letter to the NRC, "Relief Request I3R-11 Reactor Vessel Closure Head Nozzles Inservice Inspection Program - Third Interval," dated May 22, 2013, ADAMS Accession Number ML13143A167.

6. Progress Energy letter to the NRC, "Relief Request I3R-13 Reactor Vessel Closure Head Nozzle 37, Inservice Inspection Program - Third Ten-Year Interval," dated November 22, 2013, ADAMS Accession Number ML13329A354.

7. Progress Energy letter to the NRC, "Relief Request I3R-15, Reactor Vessel Closure Head Nozzle Repair Technique, Inservice Inspection Program - Third Ten-Year Interval," dated April 2, 2015, ADAMS Accession Number ML15092A236.

8. Progress Energy letter to the NRC, "Relief Request I3R-16, Reactor Vessel Closure Head Nozzle Repair Technique, Inservice Inspection Program - Third Ten-Year Interval," dated October 19, 2016, ADAMS Accession Number ML16294A218.

9. Progress Energy letter to the NRC, "Relief Request I3R-18, Reactor Vessel Closure Head Nozzle Repair Technique, Inservice Inspection Program - Fourth Ten-Year Interval," dated April 18, 2018, ADAMS Accession Number ML18108A094.

10. Entergy letter to the NRC, "Relief Request Number RR 5-7 Proposed Alternative to ASME Section XI Code Requirements for Repair of Reactor Pressure Vessel Head Penetrations," dated November 26, 2018, ADAMS Accession Number ML18330A142.

11. Entergy letter to the NRC, "Relief Request Number RR 5-8 Proposed Alternative to ASME Section XI Code Requirements for Repair of Reactor Pressure Vessel Head

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 23 Penetrations," dated September 23. 2020, ADAMS Accession Number ML20267A387.

12. Florida Power & Light Company letter to the NRC, "Unit 1 Relief Request 21 and Unit 2 Relief Request 31 Request for Additional Information Response," dated April 14, 2003, ADAMS Accession Number ML031060268.

9. REFERENCES

1. ASME Code Case N-638-7, "Similar and Dissimilar Metal Welding Using Ambient Temperature Machine GTAW Temper Bead Technique,Section XI, Division 1".

2. NRC Regulatory Guide 1.147, Revision 19, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1", ADAMS Accession Number ML19128A244.

3. ASME Code Case N-729-6 "Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial- Penetration Welds,Section XI, Division 1".

4. Entergy letter to the NRC, "Request for Alternative to 10 CFR 50.55a(g)(6)(ii)(D)

Examination Requirements - Relief Request ANO2-ISI-022," dated November 24, 2020, ADAMS Accession Number ML20329A202.

5. EPRI Report 1025169, Welding and Repair Technology Center: Welding and Repair Technical Issues in ASME Section XI.

6. [ ] Document Number 32-9352239-001.

7. [

] Document Number 32-9352384-001.

8. [

] Document Number 51-9338948-001.

9. [ ] Document Number 51-9352242-000.

10. [

] Document Number 32-9348826-002.

11. Relief Request ANO2-RR-21-002, Half Nozzle Repair of Reactor Vessel Closure Head Penetration #46, November 5, 2021, ADAMS Accession No. ML21309A007.

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 24

12. Arkansas Nuclear One, Unit 2, Safety Analysis Report Amendment 30 (Redacted),

April 2022, ADAMS Accession No. ML22124A153.

13. Response to Request for Additional Information Concerning Relief Request ANO2-RR-21-002 Support the Repair of the Reactor Vessel Closure Head Penetration #46, November 7, 2021, ADAMS Accession No. ML21312A017.

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 25 Figure 1 Nozzle Machining

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 26 Figure 2 Nozzle Weld

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 27 Figure 3 Nozzle Examination Pre - Weld PT k-l-o-p Post - Weld PT m-n-o-p-q Post - Weld UT a-b-c-d-e-f-g-h-j-a NOTE: For Post - Weld PT, extent of examination above and below the weld was 1-inch for Nozzle #46. In addition, the examination included a minimum of 0.81-inch above the rolled transition area.

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 28 Figure 4 Nozzle 0° and 45°L UT Beam Coverage Looking Clockwise and Counter-Clockwise

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 29 Figure 5 Nozzle 45°L UT Beam Coverage Looking Down

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 30 Figure 6 Nozzle 45°L UT Beam Coverage Looking Up

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 31 Figure 7: Nozzle 70°L UT Beam Coverage Looking Down

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 32 Figure 8: Nozzle 70°L UT Beam Coverage Looking Up

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 33 Figure 9: Nozzle ISI UT Examination UT a-b-c-d-a UT e-c (leak path)

Note: Extent of examination above the weld extends up to the outer surface of the head and below the weld is 1-inch for Nozzle #46.

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 34 Figure 10: Reactor Vessel Head Penetration Locations Notes:

1. Figure 10 shows the locations of the RVCH penetrations. There are 81 CEDM penetrations, eight In-core Instrument penetrations, and one Vent line.

2. CEDM Penetration #46 is highlighted.

Framatome Inc. ANP-4023NP Revision 0 Arkansas Nuclear One, Unit 2 Relief Request for Half-Nozzle Repair of RVCH Penetration #46 Technical Report Page 35 Figure 11: Indication Location

ML23088A217

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