ML18330A143

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Relief Request Number RR 5-7, Proposed Alternative to ASME Section XI Code Requirements for Repair of Reactor Pressure Vessel Head Penetrations
ML18330A143
Person / Time
Site: Palisades Entergy icon.png
Issue date: 11/26/2018
From:
Entergy Nuclear Operations
To:
Office of Nuclear Reactor Regulation
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ML18330A141 List:
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PNP 2018-054
Download: ML18330A143 (32)
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ATTACHMENT 1 PNP 2018-054 Relief Request Number RR 5-7, Proposed Alternative to ASME Section XI Code Requirements for Repair of Reactor Pressure Vessel Head Penetrations 31 pages follow

I ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1) 1.0 ASME Code Component Affected I Applicable Code Edition

_______________ H ** _____________ ****.. ** _____________.... """"................... _----_.. _..... _._............ -............ _.......... _..... __.... _.. _---_.

Component:

! Reactor Vessel Closure Head (RVCH)

        • _._... __ ** ___ M ** _______ ** _____________ ** ___ ** _.... __ ********** __ ** __ ** ____ ******** _ ****** ___ **** _...... __ **** __.. _ ******** ___ ** _.. __ ** _____ ** _ ** __ ** ___ ** ____ ** ___._ *** ___

==

Description:==

Alternative Requirements for the Repair of Reactor Vessel l

Head Penetrations (VHPs) with Nozzles Having Pressure-Retaining Partial-Penetration J-groove Welds i ! Code Class:

Class 1

        • ____........ ______________________________________ * *.. ______ *** *.. _M ** ___ ** _ **** ______________ ** ___
        • M.............. _--_. __.... __.... _--_.... __........ _------_.. _----_.. _-------------------

Examination Category:

ASME Code Case N-729-4 Code Item:

84.20 Identification:

VHP Numbers 25, 33 and 36 Reference Drawing:

232-122-11 Closure Head Assembly

... M.......... _...

Material:

Inconel' Alloy 600 (S8-167) UNS N06600 Intervallnservice Inspection (lSI) and Repair/Replacement Programs: American Society of Mechanical Engineers (ASME) 80iler and Pressure Vessel (8&PV) Code,Section XI, 2007 Edition through the 2008 Addenda. Examinations of the VHPs are performed in accordance with 10CFR50.55a(g)(6)(ii)(D), which specifies the use of Code Case N-729-4, with conditions.

Code of Construction [Reactor Pressure Vessel (RPV)]: ASME Section III, 1965 Edition through Winter 1965 Addenda.

2.0 APPLICABLE CODE REQUIREMENTS The applicable requirements of the following ASME 8&PV Code and Code Cases from which relief is requested are as follows:

ASME Code,Section XI, 2007 Edition through 2008 Addenda IW8-3420 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 IW8-3500.

1 of 31

ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1)

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, 2001 Edition through 2003 Addenda NB-5245, Partial Penetration Welded Joints, specifies progressive surface examination of partial penetration welds.

NB-5331 (b) states:

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 post weld heat treatment.

Paragraph 1 (a) states in part:

This Case shall not be used to repair SA-302, Grade B material, unless the material has been modified to include 0.4% to 1.0% nickel, quenching and tempering, and application of a fine gain practice.

Paragraph 2.1 (a) states:

The materials that are used for procedure qualification testing shall receive a heat treatment that is at least equivalent to the time and temperature already applied to the materials being welded.

Paragraph 1 (g) states:

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

2 of 31

ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1)

The applicable requirements of the following ASME B&PV Code and Code Case(s) from which relief is not specifically requested are as follows:

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

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.

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 1I1... provided the requirements of IWA-4222 through IWA-4226, as applicable, are met.....

IWA-4224.1, Identical material Procured to a Later Edition or Addenda of the Construction Code,Section III or Material Specification.

(a) Materials, including welding materials may meet the requirements of later dates...

(b) Differences in the specified material tensile stress...

IWA-4400 provides welding, brazing, metal removal, fabrication, and installation requirements related to repair/replacement activities.

IW A-4411 states:

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

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 3 of 31

ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1) used, provided the substitution is as listed in IWA-4221 (c). Filler metal requirements shall be reconciled, as required, in accordance with IWA-4224.

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.

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, Fig. 2, "Examination Volume for Nozzle Base Metal and Examination Area for Weld and Nozzle Base Metal," is applicable to the VHPs.

3.0 REASON FOR REQUEST Flaws requiring repair were detected during the lSI program ultrasonic (UT) examination of the Palisades Nuclear Plant (PNP) VHP numbers 25, 33 and 36 during R026. The flaws are axially oriented, in the tube wall, near the low hill side, and approximately adjacent to the J-groove weld root. Three (3) nozzles will be repaired under this request. Figure 10 shows the relative location of the nozzles in the RVCH and Figure 11 shows the location of the axial indications.

The repair technique, sometimes referred to as the half-nozzle repair, is intended to be the same as was used previously for nozzles 29 and 30 in 2004 with the exception that surface stress improvement will be performed using rotary peening in place of abrasive water jet machining (AWJM). 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. The new weld also attaches a replacement lower nozzle that provides a means for reattaching the CEDM extension and grid structure. This technique requires 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 is 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 Nuclear Operations, Inc. (ENO) proposes to perform the modification of the VHPs utilizing the Inside Diameter Temper Bead (IDTB) welding method to restore the pressure boundary of the degraded nozzle penetrations. 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 4 of 31

ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1) minimum preheat temperature and no PWHT. The modification described below will be performed in accordance with the 2007 Edition through 2008 Addenda of ASME Section XI, Code Case N-638-6, Code Case N-729-4, and the alternatives discussed in 4.0.

Basic steps for the IDTB repair are:

1. Cut grid structure adjoining the target nozzle and surrounding extensions.
2. Cut the nozzle close to the underside of the head and remove the nozzle extension.
3. 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.
4. Machining to remove the lower nozzle to above the J-groove weld eliminating the portions of the nozzle containing the unacceptable indication(s). This machining operation also establishes the weld preparation area (Refer to Figure 1).
5. Liquid penetrant (PT) examination of the machined area (Refer to Figure 3).
6. Welding the remaining portion of the nozzle and the new replacement lower nozzle using Alloy 52M weld material (Refer to Figure 2).
7. Machining the weld and nozzle to provide a surface suitable for nondestructive examination (NDE).
8. PT and UT examination of the weld and adjacent region (Refer to Figure 3).
9. Rotary peening of the repair region most susceptible to PWSCC.

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

The location of the VHP nozzle welds relative to the inner and outer spherical radii of the RVCH, and the existing J-groove weld will vary depending upon the location of the VHP nozzle and 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 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. The parameters of interest are tool insertion depth and the torque setting 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. The torque applied to the roll expander is controlled so that the desired amount of plastic deformation occurs. The 5 of 31

ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PAUSADESNUCLEARPLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1) 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 nozzles 25,33 and 36 and the two parameters of interest that could impact the susceptibility to PWSCC will be validated to be within process specifications. Additionally, rotary peening will be applied to remediate the tensile surface stresses in the roll expanded region. As a result, there is high confidence that adequate measures will be applied in the modification of nozzles 25,33 and 36 such that PWSCC in the region above the roll expansion zone is not expected to initiate.

ENO has determined that modification of the VHPs utilizing the alternatives specified in this request will provide an acceptable level of quality and safety. Relief is requested in accordance with 1 OCFRSO.5Sa(z)(1).

4.0 PROPOSED ALTERNATIVE AND BASIS FOR USE 4.1 Welding Requirements Code Case N-638-6 paragraph 1 (a) states in part:

This Case shall not be used to repair SA-302, Grade B material, unless the material has been modified to include 0.4% to 1.0% nickel, quenching and tempering, and application of a fine gain practice.

The RVCH material is SA-302 Grade B -Modified, quenched and tempered plate. The CMTRs from Lukens Steel Company support the SA-302, Grade B material as having been modified to include 0.4% to 1.0% nickel and also that the material was quenched and tempered. Aluminum content is not reported on the CMTRs and the CMTRs do not identify that a fine grain practice was applied during the steelmaking process.

Therefore, it is unknown if Aluminum-Nitride (AIN) pinning of the prior-austenite grain boundaries occurred that would have resulted in fine grains. It is also unknown if carbide formers such as Nb or V were intentionally added to the "modified" formulation to promote fine grains, as these elements are not reported on the CMTR.

EPRI Report 1014351 provides a comparison of the chemical and mechanical properties, heat treatment, and grain refinement practices of SA-302, Grade B Modified to SA-533, Grade B Class 1 materials. The chemical composition and the mechanical properties of SA-302, Grade B Modified materials are essentially identical to SA-S33, Grade B Class 1, especially in the case when both materials have been Quenched and Tempered (which is the case at PNP). Prior to 1987, the prescriptive Quench and 6 of 31

ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1)

Temper was the primary difference between SA-302, Grade B plate and SA-533, Grade B Class 1 plate specifications. The SA 533 specification in ASME Section II did not include a fine grain practice requirement until 1987. Code Case N-638-6 does not prohibit its use on SA-533, Grade B Class 1 plate manufactured prior to 1987.

The GTAW ambient temperature temper bead welding process is designed to develop a tough, ductile microstructure in the weld heat affected zone (HAl) that is equivalent or superior to the surrounding base material. When performing GTAW ambient temperature temper bead welding in accordance with Code Case N-638-6, cooling rates are sufficiently high to obtain a very high percentage martensitic microstructure in the HAl. Tempering of the HAl is accomplished by the heat introduced from adjacent weld beads and successive weld layers. The degree of tempering is ideal for developing excellent notch toughness. Thus, two beneficial steps necessary to achieve an optimum HAl microstructure occur during temper bead welding - a very high cooling rate step and tempering step(s). Finally, assurance of adequate notch toughness in the HAl is obtained by the performance of impact testing (Charpy V-notch testing) of the HAl in accordance with Code Case N-638-6, paragraph 2.1 (e)(4). The Framatome welding procedure, which will be used for performing the IOTB welding on the PNP RVCH, meets these requirements as the average lateral expansion value of the HAl Charpy V-notch specimens from the procedure qualification was greater than that of the unaffected base material.

The acceptable UT examination results reported at each RFD since installation in 2004 provide evidence for the two previous IOTB repairs at PNP in 2004 that the plate material was not adversely affected by the temper bead weld process. Therefore, based on the discussion provided END requests relief from the fine grain practice requirement specified in Code Case N-638-6 paragraph 1 (a).

Code Case N-638-6 paragraph 2.1 (a) states:

The materials that are used for procedure qualification testing shall receive a heat treatment that is at least equivalent to the time and temperature already applied to the materials being welded.

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 HAl for acceptance.

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ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1)

The temper bead weld procedure qualification is required to demonstrate that the Charpy V-notch test results from the weld heat affected zone are no less than the Charpy V-notch test results for the unaffected base material. EPRI Report 1025169 documents that simulated PWHT on procedure qualification test plates degrades the base material notch toughness which increases the difference between the base material and the test weld heat affected zone, thereby making it less difficult to meet the temper bead qualification requirements. Therefore, simulated PWHT does not provide conservative results for temper bead weld qualifications.

The RVCH material at PNP has 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of PWHT and the weld procedure qualification test plate has 37Y2 hours of simulated PWHT. Code Case N-638-6 requires prior simulated PWHT on the temper bead qualification test plate to be equivalent to or greater than the total aggregate time applied to the component to be welded. There is no maximum limit on the simulated PWHT time. This has been recognized by the ASME Code as a non-conservative requirement that was changed in Code Case N-638-

7. Paragraph 2.1 (a) of N-638-7 was revised to neither require nor prohibit prior simulated PWHT on the test plate used for temper bead procedure qualification.

Paragraph 2.1 (a) of N-638-7 states, "Prior simulated postweld heat treatment on the procedure qualification test assembly is neither required nor prohibited. However, if used, the simulated postweld heat treatment shall not exceed the time or temperature already applied to the base material to be welded."

Code Case N-638-7 is listed in Table 2, "Conditionally Accepted Section XI Code Cases," of Draft Regulatory Guide DG-1342. As stated in the Proposed Rule published in Vol. 83, No. 159 of the Federal Register, dated Thursday, August 16, 2018, the condition applied to N-638-7 is identical to the condition that was applied to N-638-6.

Therefore, the revised simulated PWHT requirements specified in paragraph 2.1 (a) have been accepted by the NRC.

The simulated PWHT time requirement for at least equivalent time and temperature for the temper bead qualification test plate is a non-conservative approach for temper bead welding. Therefore, END requests relief from the procedure qualification heat treatment requirement specified in N-638-6, paragraph 2.1 (a).

Code Case N-638-6 Paragraph 1 (g) states:

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

Rotary peening is performed on the final layer to provide further assurance of the modified configuration being resistant to PWSCC. However, peening on the final layer 8 of 31

ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1) of a temper bead weld is prohibited by ASME Code Case N-638-6, paragraph 1 (g).

This prohibition is referring to the high cold-work peening that is traditionally used for configuration distortion control during welding, as was interpreted by ASME XI-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 heat affected zone and 20 mils at the base metal). The uniform compressive stress layer created by the rotary peening process does not inhibit subsequent NOE. Furthermore, this residual compressive stress layer has been shown to greatly reduce PWSCC initiation.

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 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.

However, the increased resistance to PWSCC initiation provided by rotary peening will not influence the inspection frequency for the modified nozzles as required by ASME Code Case N-729-4 as conditioned by 10CFR50.55a(g)(6)(ii)(O). Therefore, END requests relief from Code Case N-638-4, paragraph 1 (g).

4.2 IDTS Modification Acceptance Examinations ASME Section III, 2001 Edition including Addenda through 2003, 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 are proposed.

For a modified VHP, the 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 will be performed as specified in ASME Code Case N-638-6, 4(a)(2) and 4(a)(4). Regulatory Guide 1.147, Rev. 18, has conditionally approved Code Case N-638-6 with the condition that UT volumetric examinations are demonstrated using representative samples which contain construction type flaws (see 6.1). ASME Section III, NB-5112 requires the use of 9 of 31

ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1) procedures "that have been proven by actual demonstration to the satisfaction of the inspector". The acceptance criteria of NB-5331, in ASME Section III, 2001 Edition including Addenda through 2003, 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 2.79 in. The scanning is 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. Additionally, the low alloy steel extending to % in.

beneath the weld into the low alloy steel base material (see Figure 3) will be examined using the 0° L-wave transducer searching for evidence of under bead cracking and lack of fusion in the heat-affected zone. The repair volume receives essentially 100% UT examination coverage as shown in Figure 4 through Figure 8.

In addition to the UT examinations, a surface PT examination will be performed on the entire weld as shown in Figure 3. 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 acceptance criteria of NB-5350 in ASME Section 111,2001 Edition including Addenda through 2003 shall apply.

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

4.3 Triple Point Anomaly ASME Section III, 2001 Edition including Addenda through 2003, NB-5331 (b) states:

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.

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ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1)

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 in. in through wall extent 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 was performed for the design configuration to provide justification, in accordance with Section XI, for operating with the postulated triple point anomaly. The anomaly is modeled as a 0.10 in. deep crack-like defect, initiating 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 shown in Figure 12 and discussed below.

Circumferential and Axial Flaws: Flaw propagation is across the nozzle wall thickness from the 00 to the 10 of the nozzle housing (analysis paths 13 & 16 and paths 11 & 14).

The shortest path through the new Alloy 52M weld material is through paths 13 and 16. By using a fatigue crack growth rate twice that of the rate of in-air austenitic stainless steel material, that is used to bound the Alloy 600/690 nozzle and Alloy 52M weld materials, it is ensured that another potential path through the heat affected zone between the new repair weld and the Alloy 600 nozzle material is also bounded.

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

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ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1)

Cylindrical Flaw: Flaw propagation extends down the outside surface of the repair weld between the upper and lower triple points (analysis paths 12 & 15).

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 analysis demonstrate that a 0.10 inch weld anomaly is acceptable for 27 years of operation following a VHP nozzle 10 temper bead weld repair. Acceptable design margins have been demonstrated for all flaw propagation paths considered in the analysis. The minimum fracture toughness margin has been shown to be 4.34 for circumferential and axial flaw propagation Paths 11, 13, 14, and 16 and 3.58 for cylindrical flaw propagation Paths 12 and 15, as compared to the required margin of

--J10 (3.16) for normal operating conditions per Section XI, IW8-3612. Fatigue crack growth is minimal. The maximum final flaw size is determined to be less than twice the initially assumed flaw size considering all flaw propagation paths. A limit load analysis was also performed considering the ductile Alloy 600/Ailoy 690 materials along flaw propagation Paths 11, 13, 14, and 16. This analysis showed a limit load margin of 7.96 for normal operating conditions, as compared to the required margin of 2.7 per Section XI, IW8-3644.

Since the postulated 00 flaw in the weld anomaly at the upper triple point 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.

The crack-like defects due to the weld anomaly at the lower triple point are exposed to primary coolant however, the materials at the lower triple point are Alloy 52M, Alloy 690, and low alloy steel, therefore are only subject to fatigue crack growth.

This evaluation is prepared in accordance with ASME Code Section XI, 2007 Edition including Addenda through 2008, 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.

ENO requests relief from the acceptance criteria specified in N8-5331 (b) to permit anomalies, as described herein, at the triple point area to remain in service.

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ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1) 4.4 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. 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 is conservatively assumed 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 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/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-leW flaw into the low alloy steel would be by fatigue crack growth under cyclic loading conditions.

The J-groove flaws were evaluated using worst-case CEDM outermost nozzle penetration configuration with postulated flaw sizes on uphill and downhill sides of the J-groove weld. Fatigue crack growth for cyclic loading conditions using operational stresses from pressure and thermal loads and crack growth rates from Section XI, Nonmandatory Appendix A, Subarticle A-4300 for ferritic material in a primary water environment were calculated. The results of this evaluation show that, based on a combination of linear elastic fracture mechanics (LEFM) analysis and elastic-plastic fracture mechanics (EPFM) analysis of a postulated remaining flaw in the original Alloy 821 Alloy 182 J-groove weld and buttering for the modified RVCH nozzle is acceptable for 27 years of operation.

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 27 years of service. Although the postulated flaw did not satisfy ASME Code Section XIIWB-3612 for all transient loading conditions, LEFM analysis 13 of 31

ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1) determined that the uphill side of the reactor head penetration was the worst case position for the postulated flaw, while EPFM determined that the downhill side was the worst case position, calculated the final flaw size by fatigue crack growth, and identified the controlling service conditions for evaluation by EPFM.

The transients applicable for the "as-left" J-groove weld are those due to normal and upset conditions only. The controlling loading condition was identified to be during normal cooldown, 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 (0.597 kips/in) was less than the J-integral of the low alloy steel head material (1.711 kips/in) at a crack extension of 0.1 inch. Flaw stability during ductile flaw growth was easily demonstrated using safety factors of 3.0 for primary stress intensity factors and 1.5 for secondary stress intensity factors. The applied tearing modulus (15.18 kips/in) was less than the material tearing modulus of the low allow steel head material (31.28 kips/in). Considering LEFM condition only, the controlling transient was the end of the cooldown (uphill side flaw),

with a safety margin on the applied stress intensity factor of 1.80 compared to the required safety margin of --./2.

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 27 years of service based on the detailed EPFM analysis that was performed. Following the detailed EPFM analysis, a primary stress limits analysis (PSLA) as required per IWB-361 0(d)(2) was subsequently performed to demonstrate the service life of 27 years for the RVCH.

Successive examinations required by IWB-3132.3 will not be performed on the nozzles repaired this outage (VHPs 25, 33 and 36) for the duration of the life of the repairs because analytical evaluation of the worst-case postulated flaw is performed to demonstrate the acceptability for 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 27 years from the time of repair and documents acceptability beyond the planned plant shutdown in 2022.

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ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PAUSADESNUCLEARPLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1)

Relief is requested from flaw characterization specified in IWB-3420 and subsequent examination requirements specified in IWB-2420(b) and IWB-2420(c).

4.5 Preservice Inspection (PSI) and Inservice Inspection (151) of VHPs Repaired Successive examinations required by Code Case N-729-4 shall be performed on nozzles 25,33 and 36 during each subsequent refueling outage. 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. The post-weld surface examination shown in Figure 3 will be used for the PSI examinations required by Code Case N-729-4, paragraph -2220.

Volumetric examination will be used for lSI for modified VHPs. 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 volume depicted in this figure at several locations. Figure 9 of this relief request will be used to establish the examination volume for future inservice inspections. This examination volume exceeds that required by Figure 2 in Code Case N-729-4, as it examines the nozzle weld and the volume above the nozzle weld that includes the rotary peened surfaces.

Non-repaired RVCH CEOM and ICI nozzles will continue to be examined in accordance with Code Case N-729-4 as modified by 10 CFR 50.55a(g)(6)(ii)(0), using a qualified ultrasonic examination procedure with technical justification demonstrating leak path assessment capability.

Therefore, future inservice inspections will comply with Code Case N-729-4 as modified by 10 CFR 50.55a(g)(6)(ii)(0) and as depicted in Figure 9.

4.6 General Corrosion Impact on Exposed Low Alloy Steel The 10TB nozzle modification 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 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 10TB 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 corrosion products pack the annulus between the lower nozzle extension and the RVCH bore, the long-term corrosion rate and overall release of Fe into the PCS is expected to be negligible.

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ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1) 4.7 Conclusions Implementation of an IDTS repair to the RVCH nozzle penetrations will produce an effective repair that will restore and maintain the pressure boundary integrity of the PNP VHPs. Similar modifications have been performed successfully and have been in service for several years without any known degradation [e.g., Shearon Harris (2012, 2013,2015,2016 and 2018) and PNP (2004)]. This 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).

5.0 DURATION OF PROPOSED ALTERNATIVE The overall acceptable life of the repair design is based on the most limiting life predicted by three evaluations: 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 27 -year design life starts at the time of the repair. A PWSCC evaluation of the IDTS repair is included in the Life Assessment Summary, Ref. [4]. The compressive stress imparted by the rotary peening process is expected to mitigate the residual tensile surface stresses in the roll expanded transition area, the heat affected zone between the Alloy 52M IDTS weld and the Alloy 600 CEDM nozzle.

Therefore, the PWSCC evaluation concluded that due to the compressive stresses achieved by rotary peening, PWSCC initiation is not expected during the 27-year design life of the repair.

ENO will examine RVCH penetrations 25, 33, and 36 every refueling outage in accordance with ASME Code Case N-729-4 as conditioned by 10CFR50.55a(g)(6)(ii)(D) and as depicted in Figure 9.

The provisions of this relief are applicable to the fifth ten-year inservice inspection interval for PNP which began on December 13, 2015, and is currently scheduled to end on December 12, 2025. PNP is currently scheduled to operate for approximately 3%

years and then close in the spring of 2022. The repair installed in accordance with the provisions of this relief shall remain in place for the remaining life of the plant/repair.

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ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PAUSADESNUCLEARPLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1) 6.0 ADDITIONAL INFORMATION 6.1 VHP Weld Qualification Mockup UT Acceptance Volumetric examination is required by Code Case N-638-6. NRC Reg. Guide 1.147, Rev. 18 imposes a condition for this code case that requires UT demonstration on representative samples which contain construction type flaws. Framatome, in support of many similar modifications, has 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 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 7.0.

The demonstration was conducted for ANI! and NRC, Region III Inspectors using an earlier revision of the same UT procedure to be used at PNP. Subsequent revisions were non-technical changes.

To satisfy this requirement, an IDTB weld half-nozzle repair mockup containing reflectors to simulate construction type flaws applicable to this weld process has been used. It contains a series of 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 outside diameter so that the hole face is 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 has demonstrated the ability to detect a linear weld fabrication triple point anomaly extending 0.05 in. and greater into the weld.

A NiCrFe alloy calibration block is used and contains a series of electrical-discharge machining (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 contains 1/4T, 1/2T, and 3/4T deep end-drilled holes and side-drilled holes that are used for calibration.

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. Prior to examination of the repair welds at PNP, UT personnel will practice using the data files from the demonstration described above.

17 of 31

ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1) 7.0 PRECEDENTS

1. Palisades Nuclear Plant, Relief Request #1 and Relief Request #2, October 11, 2005 (ADAMS Accession Number ML052870321)
2. Davis-Besse Nuclear Power Station Relief Request RR-A34, September 20, 2010 (ADAMS Accession Number ML102571569)
3. Calvert Cliffs Nuclear Power Plant Relief Request RR-PZR-01, December 9, 2011 (ADAMS Accession Number ML113360526)
4. Shearon Harris Nuclear Power Plant, Unit 1, Relief Request 13R-09, October 2, 2012 (ADAMS Accession Number ML12270A258)
5. Shearon Harris Nuclear Power Plant, Unit 1, Relief Request 13R-11, September 13,2013 (ADAMS Accession Number ML13238A154)
6. Shearon Harris Nuclear Power Plant, Unit 1, Relief Request 13R-13, April 11, 2014 (ADAMS Accession Number ML14093A075)
7. Shearon Harris Nuclear Power Plant, Unit 1, Relief Request 13R-15, September 18,2015, and January 6, 2016 (ADAMS Accession Numbers ML15203A702 and ML15342A043)
8. Shearon Harris Nuclear Power Plant, Unit 1, Relief Request 13R-16, December 27,2016 (ADAMS Accession Number ML16343A220)
9. Shearon Harris Nuclear Power Plant, Unit 1, Relief Request 14R-18, April 18, 2018, (ADAMS Accession Number ML18108A094) 18 of 31

ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PAUSADESNUCLEARPLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1)

8.0 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. Framatome Evaluation #51-5047343 Revision -007, "Palisades CEDM Nozzle IDTB Repair - Life Assessment Summary" (Proprietary)(See Attachment 2)
5. EPRI Report 1014351, "Repair and Replacement Applications Center: Topical Report Supporting Expedited NRC Review of Code Cases for Dissimilar Metal Weld Overlay Repairs," December 2006
6. EPRI Report 1025169, "Welding and Repair Technology Center: Welding and Repair Technical Issues in ASME Section XI," December 2012 19 of 31

ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR SO.SSa(z)(1)

Figure 1 Nozzle Machining 20 of 31

ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PAUSADES NUCLEAR PLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1)

Figure 2 Nozzle Weld UPPER TRIPLE POINT ALLOY 52M WELD LOWER TRIPLE POINT LOWER REPLACEMENT NOZZLE 21 of 31

ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PAUSADESNUCLEARPLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR SO.SSa(z)(1)

Figure 3 Nozzle Examination Pre-Weld PT k-I-o-p Post - Weld PT m-n-o-p-q Post - Weld UT a-b-c-d-e-f-g-h-j-a

22. of 31

ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PAUSADESNUCLEARPLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR SO.SSa(z)(1)

Figure 4 Nozzle UT 0° and 45°L Beam Coverage Looking Clockwise and Counter-clockwise NOZZLE HEAD 23 of 31

ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PAUSADES NUCLEAR PLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR SO.SSa(z)(1)

Figure 5 Nozzle 45°L UT Beam Coverage Looking Down NOZZLE HEAD 24 of 31

ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1)

Figure 6 Nozzle 45°L UT Beam Coverage Looking Up NOZZLE HEAD 25 of 31

ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1)

Figure 7 Nozzle 700L UT Beam Coverage Looking Down NOZZLE HEAD 26 of 31

ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1)

Figure 8 Nozzle 700L UT Beam Coverage Looking Up NOZZLE HEAD 27 of 31

ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1)

Figure 9 Nozzle lSI UT Examination 1"

UT I a-b-c-d-a UT e-f (leak path) 28 of 31 Note: Examination volume includes the rolled transition and a portion of the nozzle material above the rolled transition.

ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PAUSADESNUCLEARPLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1) 1-45 CEDM Nozzles 46-53 ICI Nozzles Figure 10 Reactor Vessel Head Penetration Locations (Plan View Looking Down)

Note: Penetrations 29 and 30 were repaired during the Fall 2004 refueling outage 29 of 31

ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR SO.SSa(z)(1)

CEDM NOZZLE Figure 11 Indication Location 30 of 31 J-GROOVE WELD INDICATION LOCATION

ATTACHMENT 1 ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT Relief Request Number RR 5-7 Proposed Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1)

Figure 12 Crack Propagation Paths Uphill Side Downhill Side 31 of 31

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