Text

Proposed Inservice Inspection Alternative FNP-ISI-ALT-05-03, Version 1.0
	ML18108A070
Person / Time
Site:	Farley
Issue date:	04/18/2018
From:	Wheat J; Southern Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML18108A148	List: NL-18-0193, Proposed Inservice Inspection Alternative FNP-ISI-ALT-05-03, Version 1.0;
References
	NL-18-0193 FNP-ISI-ALT-05-03, Ver. 1.0
	Download: ML18108A070 (41)
	v • d • e

Enclosure 4 to this letter contains Proprietary Information to be withheld fro m public disclosure per 10 CFR 2.390. When separated from Enclosure 4 this transmittal document is decontrolled.

t. Southern Nuclear Regulatory Affairs 40 Inverness Center Parkway Post Office Box 1295 Birmingham, AL 35242 APR 1 8 2018 205 992 5000 tel 205 992 7601 fax Docket Nos.: 50-364 NL-18-0193 U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant- Unit 2 Proposed lnservice Inspection Alternative FNP-ISI-ALT-05-03, Version 1.0 Ladies and Gentlemen:

In accordance with the provisions of 10 CFR 50.55a(z)(2), Southern Nuclear Operating Company (SNC) hereby requests Nuclear Regulatory Commission (NRC) approval of proposed inservice inspection (lSI) alternative FNP-ISI-ALT-05-03, Version 1. This request proposes extending the inspection period by one operating cycle beyond the interval currently authorized under alternative FNP-ISI-ALT-15, Version 1 (ML14084A203). The staff previously authorized an inspection interval of six operating cycles (approximately 9 calendar years or 8.6 effective full power years (EFPY)) in accordance with 10 CFR 50.55a(3)(i) for both Farley Units (ML14262A317). The proposed alternative would allow the ASME Code Case N-770-2, Inspection Item B for Reactor Pressure Vessel (RPV) Cold Leg (CL) Dissimilar Metal (OM) welds to be examined after 7 operating cycles (approximately 10.5 calendar years or 9.7 EFPY) for Farley Unit 2, and is being submitted in accordance with 10 CFR 50.55a(z)(2), "hardship without a compensating increase in the level of quality and safety, to align planned mitigation activities of the hot and cold leg OM welds and avoid multiple core barrel removal and installation evolutions over consecutive refueling outages.

As previously stated in FNP-ISI-ALT-15, limitations in access and outer surface geometry prohibits compliance with ASME Code and 10 CFR 50.55a examination coverage requirements from the outer surface. Therefore, core barrel removal is required to access the inner surface of the subject OM welds.

By performing the CL weld inspections in conjunction with an approved stress improvement process during refueling outage 2R27, SNC would reduce unnecessary radiation exposure to personnel, and reduce outage risk from multiple infrequently performed heavy lift evolutions.

SNC requests NRC review and approval of this alternative by February 15, 2019 to facilitate the use of proposed inspection date extension when authorized in accordance with 10 CFR 50.55a(z).

This letter contains no NRC commitments. If you have any questions, please contact Jamie Coleman at 205.992.6611.

U.S. Nuclear Regulatory Commission NL-18-0193 Page 2 Respectfully submitted,

~~-r~

Justin T. Wheat Nuclear Licensing Manager JTW/NDJ Enclosures 1. Proposed Alternative FNP-ISI-ALT-05-03, Version 1.0 in accordance with 10 CFR 50.55a(z)(2)

2. Applicability of the generic circumferential flaw tolerance evaluation

3. Westinghouse Letter CAW-18-4695 Application for Withholding Proprietary Information from Public Disclosure

4. PROPRIETARY Westinghouse Letter LTR-SDA-17-035 Technical Justification to Support Extended Volumetric Examination Interval for Joseph M. Farley Unit 2 Reactor Vessel Inlet Nozzle to Safe End Dissimiliar Metal Welds

5. NON-PROPRIETARY Westinghouse Letter LTR-SDA-17-035 Technical Justification to Support Extended Volumetric Examination Interval for Joseph M. Farley Unit 2 Reactor Vessel Inlet Nozzle to Safe End Dissimiliar Metal Welds Cc: Regional Administrator, Region II NRR Project Manager- Farley Nuclear Plant Senior Resident Inspector- Farley Nuclear Plant RTYPE: CFA04.054

Joseph M. Farley Nuclear Plant Unit 2 Enclosure 1 Proposed Alternative FNP-ISI-ALT-05-03, Version 1.0 in accordance with 10 CFR 50.55a(z)(2) to NL-18-0193 Proposed Alternative FNP-ISI-ALT-05-03, Version 1.0 in accordance with 10 CFR 50.55a(z)(2)

Plant Site - Unit:

Farley Nuclear Plant (FNP) - Unit 2 Interval-Interval Dates:

Fifth Inspection Interval, December 1, 2017 through November 30, 2027 Requested Date for Approval and Basis:

Approval is requested by February 15, 2019 for an extension of the interval for examination of the Reactor Pressure Vessel (RPV) Inlet I Cold Leg (CL) nozzle to safe-end dissimilar metal (OM) welds. The extension is requested to allow deferral of the examinations from the FNP -

Unit 2 26th Refueling Outage (2R26), which is currently scheduled to start spring 2019, to the following Refueling Outage. The requested approval date is needed to allow the finalization of the work activities and associated outage planning and preparation needed to prepare for Refueling Outage 2R26. The proposed alternative would allow these examinations to be performed during the FNP Unit 2, 27th Refueling Outage (2R27), which is scheduled to start fall 2020.

ASME Code Components Affected:

The affected components are the RPV CL OM welds at FNP Unit 2. The identified examinations are Examination Category ASME Code Case N-770-2, Inspection Item B.

FNP - Unit 2 Components:

Component 10 Description APR 1-41 00-14DM Inlet nozzle OM weld at 335 degrees APR 1-4200-14DM Inlet nozzle OM weld at 215 degrees APR 1-4300-14DM Inlet nozzle OM weld at 95 degrees

Applicable Code Edition and Addenda

The applicable Code edition and addenda is ASME Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," 2007 Edition with Addenda through 2008.

Applicable Code Requirements:

10CFR50.55a(g)(6)(ii)(F) requires licensees of existing, operating pressurized-water reactors as of August 17, 2017 to implement the requirements of ASME Code Case N-770-2 instead of N-770-1, subject to conditions specified in paragraphs (g)(6)(ii)(F)(2) through (13) of that section, by the first refueling outage starting after August 17, 2017. Code Case N-770-2, Inspection Item 8 requires unmitigated butt weld at Cold Leg operating temperature ;;:: 525°F and < 580°F to be volumetrically examined every second inspection period not to exceed 7 years. FNP Unit 2 previously received authorization to defer these examinations until the 2019 refueling outage (2R26) as documented in ML14262A317.

E1-1 to NL-18-0193 Proposed Alternative FNP-ISI-ALT-05-03, Version 1.0 in accordance with 10 CFR 50.55a(z)(2)

Background and Reason for Request:

Relief is being requested at this time due to the scheduling aspects of the inspection requirement conflicting with the current DM weld mitigation plans at FNP Unit 2. Due to this conflict, SNC is requesting an alternative based on a hardship without a compensating increase in quality or safety as compared to the requirements of Code Case N-770-2, as conditioned by 10CFR50.55a.

Examination of Code Case Item A-2 (hot leg) and Code Case Item B (cold leg) welds are performed from the inside surface (I D) of the pipe at FNP Unit 2 due to extremely limited access from the outside surface (OD) of the pipe and undesirable weld geometry not conducive to achieving the necessary examination coverage when scanning from the OD.

The FNP Unit 2 Item A-2 and Item B welds are located inside a "sandbox, .. which was installed during original plant construction after all welding was completed. The inspection of the Item A-2 (hot leg) welds from the ID does not require removal of the reactor vessel (RV) lower internals (core barrel), only the upper internals, while the inspection of the Item B (cold leg) welds from the ID requires that both the upper internals and core barrel be removed for access.

The work scope for the FNP Unit 2 Refueling Outage 2R26 currently includes examination of the hot leg DM welds in accordance with ASME Code Case N-770-2 item A-2 and examination and wear measurements of upper internals control rod guide tube guide cards in accordance with MRP-227-A and WCAP-17451-P. In addition, a 9-year inspection interval is currently authorized by staff evaluation of FNP-ISI-ALT-15 Version 1 (ML14262A317), with the result that the CL DM weld examinations are also required no later than 2R26.

Limitations in access and outer surface geometry prohibits examination coverage compliance with ASME Code and 10CFR50.55a conditions CL DM weld examinations from the outer surface, therefore require removal of the core barrel from the reactor vessel to gain access to the reactor vessel nozzle interior surfaces. To remove and reinstall the core barrel requires implementation of detailed planning and precision lifts to ensure that the core barrel and/or reactor vessel are not damaged. In addition, the core barrel is extremely radioactive which adds to the complexity when lifting the core barrel in and out of the reactor vessel. Because of this complexity utilities typically include all applicable examinations requiring core barrel removal to be included in the same refueling outage. Thus, other examinations that would be scheduled with the core barrel removed are:

(1) ASME Code Section XI Category B-A RV shell weld volumetric examinations (2) ASME Code Section XI Category B-D RV nozzle-to-shell volumetric examinations (3) ASME Code Section XI Category B-N-3 core support structure visual examinations (4) MRP-227-A Core Barrel Girth Weld Enhanced Visual (EVT-1) examinations The four examinations listed above can all be completed during the fall 2020 refuel outage without deviating from the NRC authorized alternative in the case of RV volumetric examinations (ML101750402) or the industry "needed" requirements, as the required intervals are based on ASME Code Section XI 10-year intervals or in the case of (4), within two cycles of entering the period of extended operation. Consequently, only the inlet nozzle DM welds require NRC authorization to delay completion until the fall 2020 (2R27) refuel outage.

The previously submitted alternative authorized by the staff allowed SNC to align the hot leg (HL) and CL DM weld examinations to optimize inspection tooling deployment; i.e. every other E1-2

Enclosure 1 to NL-18-0193 Proposed Alternative FNP-ISI-ALT-05-03, Version 1.0 in accordance with 10 CFR 50.55a(z)(2)

HL OM weld examination would also include a CL OM weld examination. SNC has subsequently finalized a mitigation strategy for both the HL and CL OM Welds for FNP Unit 2 that will utilize the Mechanical Stress Improvement Process (MSIP) during the fall 2020 refueling outage. The HL OM Weld examination is required by N-770-2 inspection item A-2 during the spring 2019 outage. Performing the HL OM Weld volumetric ultrasonic testing (UT) and surface eddy current testing (ET) examination one outage prior to MSI P allows SNC to simplify the operator burden for water management and the maintenance burden for heavy lifts during the subsequent outage. As allowed by N-770-2, a HL examination no more than one cycle in advance of MSIP satisfies the pre-MSIP examination requirement and subsequently allows MSIP to occur prior to initial flood up of the cavity. This eliminates the need to flood up the cavity, remove the RPV head and upper internals, defuel, perform the pre-MSIP HL UT and ET and then replace the upper internals and RPV head and drain the cavity to allow personnel access to the cavity "sandboxes" for MSIP implementation, followed by a reflood of the cavity, removal of the core barrel to allow the post-MSIP UT and ET of the CL OM Welds.

A post-MSIP examination is required for all mitigated welds. Therefore, under the current authorization, the CL examination would be required during the 2019 outage and then again in 2020 to satisfy the post-MSIP examination requirement. This results in the core barrel removal during two consecutive outages. This is an undesirable hardship due to the additional complicated lift involved and the additional dose and personnel safety risk. Core barrel removal requires it to be raised above the refueling cavity water level during transfer from the reactor vessel to the cavity storage stand. The radiation exposure levels for this activity can be high and necessitate unrelated work to stop and unnecessary personnel to leave containment.

During the previous FNP Unit 2 core barrel lift in 2010, personnel accumulated 152.7 millirem (mR) from the removal and reinstallation activity. More recently in 2016 FNP Unit 1 core barrel removal and installation resulted in 495.5 mR dose accumulation due to difficulties during reinstallation.

The NRC has recognized this hardship in recent staff evaluations (ML16074A001 for Comanche Peak 1, ML16313A042forVogtle 1 and2, ML16190A133forBeaverValley2, ML16174A091 for South Texas Project 1, ML15232A543 for McGuire Unit 1)

Moving mitigation into the 2019 refueling outage for the sole purpose of avoiding multiple core barrel removal evolutions also results in unnecessary hardship due to the multiple other heavy lifts, dose and refueling cavity complications and burden as discussed above. For FNP Unit 2, moving MSI P and the entire scope of examinations associated with a core barrel removal into the 2019 refueling outage results in additional burden and complication because of the need to perform the upper internals Control Rod Guide Tube (CRGT) guide card wear measurements as required by MRP-227-A as submitted and approved for FNP Unit 2 in the FNP Reactor Internals Aging Management Program Plan (ML15226A225 and ML17135A252). While MRP-227-A allows as few as 10 CRGTs to be measured, WCAP-17451-P recommends measurement of a minimum of 41 (of 48) CRGTs for FNP Unit 2 to attain a 95°/o confidence that the wear sample did not overlook a high wear location. This activity is required based on accumulated EFPY of between 32 and 34, which FNP Unit 2 is projected to exceed if deferred beyond the spring 2019 outage. SNC has no desire to deviate from the industry requirements and recent operating experience regarding guide card wear does not support a strong technical basis for a deviation.

Performing the CRGT wear measurements involves removal and temporary storage of the control rod drives from the CRGTs to access the guide cards. CRGT wear measurements in conjunction with the examinations involved with core barrel removal listed above thus would further complicate rigging, lifting, and the congestion associated with movement of inspection tooling, and reactor equipment.

E1-3

Enclosure 1 to NL-18-0193 Proposed Alternative FNP-ISI-ALT-05-03, Version 1.0 in accordance with 10 CFR 50.55a(z)(2)

Proposed Alternative and Basis for Use:

SNC proposes to extend the inspection interval for code case N-770-2, category B welds for an additional operating cycle to allow examination during Refueling Outage 2R27 which is scheduled to begin in fall 2020. The subject examinations would need to be performed before the end of the Refueling Outage 2R26 in spring 2019, pending approval of this proposed alternative.

The volumetric examinations of the CL DM welds were last performed in 2010 along with eddy current surface examinations. The examinations were performed from the inner surface of the DM welds. No relevant indications were noted during this examination.

The ET procedure is qualified to detect fatigue and IGSCC/IDSCC cracks having a depth of 0.04" (1 mm) and greater and a length of 0.24" (6mm) and greater. As such an ID surface-connected flaw with a depth and length up to 0.04" (1 mm) in depth and 0.24" (6mm) in length are not qualified for detection.

It is important to note that ET is primarily used in detecting very small surface breaking flaws and not for depth sizing. The vendor procedure requires an indication with a length of 0.25" or greater to be recorded. There is no minimum depth requirement. In relation to the postulated flaw evaluation previously provided, an axial flaw aspect ratio (flaw length/flaw depth) of 2 is conservatively assumed in the crack growth analysis, since axial flaw growth due to PWSCC is limited to the width of the DM weld. Based on the assumed aspect ratio and minimum flaw reporting length a 0.125" deep initial flaw assumption would be consistent and reasonable for the initial flaw depth.

SNC's use of a postulated initial flaw size of 7.5% or 0.245" is thus considered reasonable and conservative for the RPV Cold Leg examinations as this initial depth assumption corresponds to a 1:1 aspect ratio when compared with the minimum recordable length using supplemental ET in combination with the volumetric examination required by ASME Code Case N-770-1 (now N-770-2). This rationale was previously evaluated in the staffs report authorizing FNP-ISI-ALT-15 Version 1. As discussed in Enclosure 4, the analysis results demonstrate that the time limiting case for leak tightness (axially oriented postulated flaws) would not progress through to an unacceptable depth in less than 10.5 EFPY. Currently, the EFPY for FNP Unit 2 at the requested inspection interval of 10.5 years corresponds to a projected EFPY of 9. 7 (6.8 actual through 2R25 plus 2.9 projected to the start of 2R27).

Therefore, in accordance with 10CFR50.55a(z)(2) "hardship without a compensating increase in the level of quality and safety", SNC requests approval to extend the Inspection Interval to include FNP Unit 2 Refueling Outage 2R27 currently scheduled to being fall 2020.

Duration of Proposed Alternative:

The proposed alternative would extend the volumetric Inspection Interval for Examination Category B, of Code Case N-770-2 up to and including Refueling Outage 2R27, currently scheduled to start in fall 2020.

E1-4 to NL-18-0193 Proposed Alternative FNP-ISI-ALT 03, Version 1.0 in accordance with 10 CFR 50.55a(z)(2)

Precedents:

NRC Safety Evaluation dated August 27, 2015, for McGuire Nuclear Station Unit 1. Request for Relief Nos. 15-M-001 for the Fourth 10-Year lnservice Inspection Program Interval (TAC NO.

MF5817) (ML15232A543).

NRC Safety Evaluation Dated for Comanche Peak Nuclear Power Plant, Unit 1- Relief Request 183-3, Alternative to ASME Code,Section XI, Examination Requirements for Reactor Pressure Vessel Cold Leg Inspection Frequency (CAC NO. MF6125) (ML16074A001).

NRC Safety Evaluation Dated June 30, 2016 for South Texas Project, Unit 2 Request for Relief No. RR-ENG-3-20 for Extension of the Inspection Frequency of the Reactor Vessel Cold-Leg Nozzle To Safe-end Welds Flaw Analysis (CAC NO. MF7428), (ML16174A091).

Status:

Pending NRC approval.

E1-5

Joseph M. Farley Nuclear Plant Unit 2 Enclosure 2 Applicability of the Generic Circumferential Flaw Tolerance Evaluation to NL-18-0193 Proposed Alternative FNP-181-ALT-05-03, Version 1.0 in accordance with 10 CFR 50.55a(z)(2)

Applicability of the generic circumferential flaw tolerance evaluation Because of the potential for pipe rupture, the most limiting flaw from the aspect of safety significance is a circumferentially oriented flaw which propagates to the point of structural failure. In order to provide reasonable assurance that a postulated circumferentially oriented flaw would not propagate sufficiently through the thickness and around the circumference, the Materials Reliability Program (MAP) performed bounding flaw tolerance analyses for various cold leg locations in the PWR fleet which are documented in MRP-349.

The figure directly relevant to FNP-ISI-ALT-05-03 from MRP-349 is Figure 5-4, which depicts four safe-end length and RV inlet temperature combinations. Based on the circumferential crack growth results shown in Figure 5-4, the bounding combined case of a 25°/o inner diameter repair, higher cold leg temperature (565°F), and longer safe end, a circumferential flaw will not propagate from an initial depth/thickness ratio of 15o/o to an unacceptable depth/thickness ratio (57°/o) per IWB-3600 of Section XI in ten years of continued operation. For the temperatures (535 °F) most closely related to FNP (537°F, 538°F nominal), in neither the short or long safe end case does a circumferential flaw propagate from 15°/o to an unacceptable depth/thickness ratio in less than 20 years. The technical basis developed for FNP Unit 2 is based on the generic analysis in MRP-349 which was further discussed in the response to RAI #1 of FNP-ISI-ALT-13 (ML13130A119).

The information necessary for the staff to conduct confirmatory analyses pertinent to FNP Unit 2 is as follows:

The inner diameter of the weld is 27.47 inches, the weld thickness is 3.27 inches.

The FNP Unit 2 safe-end is assumed to be 4.7 inches based on design drawings. The NOE data indicates them to be 4.54 inches thus the design length is conservative.

The RPV low alloy steel nozzles were buttered with alloy 182 and subsequently stress relieved with the entire reactor vessel. The stainless steel safe-ends were then welded onto the nozzles at the RPV fabricator shop (Combustion Engineering) with Alloy 82/182 filler material. The 00 and 10 of the single V-groove dissimilar metal (OM) welds were machined to the final weld configuration. The safe-ends were then machined with the piping side weld preparation and field welded to the stainless-steel RCS loop elbow with stainless steel filler material. There were no 10 repairs documented in the weld traveler for any of the cold leg OM welds thus the residual stresses for the FNP Unit 2OM welds are expected to be lower than the generically assumed 25o/o 10 or 50o/o 10 repair cases. The bounding residual axial stress profiles applicable to the FNP Unit 2 is Figure 5-4 of MRP-349. Enclosure 4provides the FNP Unit 2 specific axial flaw tolerance evaluation. The summary conclusions of that Enclosure 4 provide reasonable assurance that a postulated axially oriented flaw would not propagate to an unacceptable depth prior to the fall 2020 refueling outage.

Based on the PWSCC crack growth curve shown in Figure 7-1 of Enclosure 4, an undetected flaw in a baseline inspection with a flaw depth of 0.24 inch, which is 7.5°/o of the original weld thickness, would not reach the maximum end-of-evaluation period allowable flaw depth of 75o/o of the original wall thickness in less than 10.5 EFPY. Since the last volumetric examinations of the CL OM welds, FNP Unit 2 accumulated 6.8 actual EFPY through 2R25 and with an additional projected EFPY of 2.9 until the start of 2R27. FNP Unit 2 would therefore project to accumulate 9.7 EFPY in the period between the 2010 examinations and the outage proposed for deferral, thus a difference of 0.8 EFPY between operating time required for the and the operating time postulated to reach the allowable flaw depth.

to NL-18-0193 Proposed Alternative FNP-ISI-ALT-05-03, Version 1.0 in accordance with 10 CFR 50.55a(z)(2)

References:

1. Dominion Engineering Calculation C-8869-00-01, Rev. 0. 'Welding Residual Stress Calculation for Farley Units 1 and 2 RPV Inlet Nozzle DMW ." (Dominion Engineering Inc.

Proprietary)

2. Combustion Engineering Drawing E-233-896, "Pressure Vessel Final Machining For:

Westinghouse Electric Corp., 157 I.D. P.W.R.," Revision 7. (Westinghouse Proprietary)

3. Combustion Engineering Drawing E-233-897, "Nozzle Details For: Westinghouse Electric Corp., 157 I.D. P.W.R.," Revision 2. (Westinghouse Proprietary)

4. Combustion Engineering Drawing E-233-943, "Pressure Vessel Final Machining -Sections, Westinghouse Electric Corporation, 157" I.D. P.W.R.," Revision 3. (Westinghouse Proprietary)

5. Combustion Engineering Drawing E-233-926, "Nozzle Details For: Westinghouse Electric Corp., 157 I.D. P.W.R.," Revision 4. (Westinghouse Proprietary)

6. Materials Reliability Program: Primary Water Stress Corrosion Cracking (PWSCC) Flaw Evaluation Guidance (MRP-287). EPRI, Palo Alto, CA: 2010. 1021023.

7. Materials Reliability Program: Welding Residual Stress Dissimilar Metal Butt-Weld Finite Element Modeling Handbook (MRP-317). EPRI, Palo Alto, CA: 2011. 1022862.

8. American Petroleum Institute, API 579-1/ASME FFS-1 (API 579 Second Edition), "Fitness-For-Service," June 2007.

9. Materials Reliability Program: Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds (MRP-115), EPRI, Palo Alto, CA: 2004. 1006696. (EPRI Proprietary)

10. Rules for lnservice Inspection of Nuclear Power Plant Components, ASME Boiler &

Pressure Vessel Code,Section XI, 2001 Edition through 2003 Addenda.

E2-2

Joseph M. Farley Nuclear Plant Unit 2 Enclosure 3 Westinghouse Letter CAW-18-4695 Application for Withholding Proprietary Information from Public Disclosure

Westinghouse Non-Proprietary Class 3

@Westinghouse Westinghouse Electric Company 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA U.S. Nuclear Regulatory Commission Direct tel: (412) 374-4643 Document Control Desk Direct fax: (724) 940-8560 115 55 Rockville Pike e-mail: greshaja@westinghouse.com Rockville, MD 20852 CAW-18-4695 January 10, 2018 APPLICATION FOR WITilliOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

LTR-SDA-17-035, Revision 0, Attachment A, "Technical Justification to Support the Extended Volumetric Examination Interval for Joseph M. Farley Unit 2 Reactor Vessel Inlet Nozzle to Safe End Dissimilar Metal Welds" (Proprietary)

The Application for Withholding Proprietary Information from Public Disclosure is submitted by Westinghouse Electric Company LLC ("Westinghouse"), pursuant to the provisions of paragraph (b)( 1) of Section 2.390 of the Nuclear Regulatory Commission's ("Commission's") regulations. It contains commercial strategic information proprietary to Westinghouse and customarily held in confidence.

The propri~tary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-18-4695 signed by the owner of the proprietary information, Westinghouse. The Affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.

Accordingly, this letter authorizes the utilization of the accompanying Affidavit by Southern Nuclear Operating Company.

Correspondence with respect to the proprietary aspects of the Application for Withholding or the Westinghouse Affidavit should reference CAW-18-469 5, and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 2 Suite 259, Cranberry Township, Pennsylvania 16066.

~~

1-ames A. Gresham, Manager Regulatory Compliance

CAW-18-4695 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

ss COUNTY OF BUTLER:

I, James A. Gresham, am authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC ("Westinghouse") and declare that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief.

Executed on: , [TJf rK ~~uk- --

James A. Gresham, Manager Regulatory Compliance

3 CAW-18-4695 (1) I am Manager, Regulatory Compliance, Westinghouse Electric Company LLC ("Westinghouse"),

and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.

(2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.3 90 of the Nuclear Regulatory Commission's ("Commission's") regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit.

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.

(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from pu~lic disclosure is owned and has been held in confidence by Westinghouse.

(ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitute Westinghouse policy and provide the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of

4 CAW-18-4695 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage (e.g., by optimization or improved marketability).

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It ((ontains patentable ideas, for which patent protection may be desirable.

(iii) There are sound policy reasons behind the Westinghouse system which include the following:

(a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.

(b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.

(c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

5 CAW-18-4695 (d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.

(f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iv) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, is to be received in confidence by the Commission.

(v) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manp.er or method to the best of our knowledge and belief.

(vi) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in LTR-SDA-17-035 Attachment A, "Technical Justification to Support the Extended Volumetric Examination Interval for Joseph M. Farley Unit 2 Reactor Vessel Inlet Nozzle to Safe End Dissimilar Metal Welds" (Proprietary), for submittal to the Commission, being transmitted by Southern Nuclear Operating Company letter. The proprietary information as submitted by Westinghouse is that associated with the technical justification to support the extended volumetric examination interval for Joseph M. Farley Unit 2 reactor vessel inlet nozzle to safe end dissimilar metal welds, and may be used only for that purpose.

(a) This information is part of that which will enable Westinghouse to provide technical justification to support the extended volumetric examination interval

6 CAW-18-4695 for Joseph M. Farley Unit 2 reactor vessel inlet nozzle to safe end dissimilar metal welds.

(b) Further, this information has substantial commercial value as follows:

(i) Westinghouse plans to sell the use of similar information to its customers for the purpose of providing technical justification to support the extended volumetric examination interval for reactor vessel to safe end dissimilar metal welds.

(ii) Westinghouse can sell support and defense of industry guidelines and acceptance criteria for plant-specific applications.

(iii) The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors ofWestinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprietary and non-proprietary versions of a document, furnished to the NRC in connection with requests for generic and/or plant-specific review and approval.

In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the Affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(l).

COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

Southern Nuclear Operating Company Letter for Transmittal to the NRC The following paragraphs should be included in your letter to the NRC Document Control Desk:

Enclosed are:

1. LTR-SDA-17-035, Revision 0, Attachment A, "Technical Justification to Support the Extended Volumetric Examination Interval for Joseph M. Farley Unit 2 Reactor Vessel Inlet Nozzle to Safe End Dissimilar Metal Welds." (Proprietary)

2. LTR-SDA-17-035, Revision 0, Attachment B, "Technical Justification to Support the Extended Volumetric Examination Interval for Joseph M. Farley Unit 2 Reactor Vessel Inlet Nozzle to Safe End Dissimilar Metal Welds." (Non-Proprietary)

Also enclosed are the Westinghouse Application for Withholding Proprietary Information from Public Disclosure CAW-18-4695, accompanying Mfidavit, Proprietary Information Notice, and Copyright Notice.

As Item 1 contains information proprietary to Westinghouse Electric Company LLC ("Westinghouse"), it is supported by an Affidavit signed by Westinghouse, the owner of the information. The Affidavit sets forth the basis on which the information may be withheld from public disclosure by the Nuclear Regulatory Commission ("Commission") and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.390 of the Commission's regulations.

Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10*CFR Section 2.390 of the Commission's regulations.

Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Westinghouse Mfidavit should reference CAW-18-4695 and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 2 Suite 259, Cranberry Township, Pennsylvania 16066.

Joseph M. Farley Nuclear Plant Unit 2 Enclosure 5 NON-PROPRIETARY Westinghouse Letter LTR-SDA-17-035 Technical Justification to Support Extended Volumetric Examination Interval for Joseph M. Farley Unit 2 Reactor Vessel Inlet Nozzle to Safe End Dissimiliar Metal Welds

Westinghouse Non-Proprietary Class 3 Page 1 of20 LTR-SDA-17-035 Revision 0 Attachment B January 10, 2018 Attachment B (Non-Proprietary)

Technical Justification to Sup~ort the Extended Volumetric Examination Interval for Joseph M. Farley Unit 2 Reactor Vessel Inlet Nozzle to Safe End Dissimilar Metal Welds

- "This record was final approved on 1/10/2018 5:15:19 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 2 of20 LTR-SDA-17-035 Revision 0 Attachment B January 10, 2018 Technical Justification to Support the Extended Volumetric Examination Interval for Joseph M. Farley Unit 2 Reactor Vessel Inlet Nozzle to Safe End Dissimilar Metal Welds 1.0 Introduction Service induced cracking of the nickel-base alloy components and weldments have been occurring more and more frequently in recent years, resulting in the need to repair and/or replace these components. Such cracking and leakage have been observed in the reactor vessel upper and bottom head penetration nozzles as well as the dissimilar metal (DM) butt welds of the pressurizer and reactor vessel nozzles exposed to the high reactor coolant temperatures. These Pressurized Water Reactor (PWR) power plant field experiences and the potential for Primary Water Stress Corrosion Cracking (PWSCC) require reassessment of the examination frequency as well as the overall examination strategy for nickel-base alloy components and weldments. ASME Code Case N-770-2 [I] provides the visual and volumetric inspection guidelines for the primary system piping DM butt welds to augment the current inspection requirements.

In accordance with ASME Code Case N-770-2 guidelines, volumetric examinations are required for the unmitigated DM butt welds at the Reactor Vessel (RV) inlet nozzles every second inspection period not exceeding 7 years. A volumetric examination was previously performed for the Joseph M. Farley Unit 2 RV inlet nozzle to safe end DM butt welds during the Spring 20 I 0 Refueling Outage (RFO). The next volumetric examination for the RV inlet nozzle DM welds was planned during the Spring 2019 RFO in accordance with a previously accepted relief request [2, 3].

The fracture mechanics evaluation in this letter report will determine the impact of performing the volumetric examination on Joseph M. Farley Unit 2 during the Fall 2020 RFO. The time interval between the previous Unit 2 examination during the Spring 20 I 0 RFO and the planned examination during the Fall 2020 RFO is I 0.5 calendar years, rather than the 7 calendar years allowed by Code Case N-770-2 or the 9 calendar years allowed by the previously accepted relief request [2, 3]. Therefore, Joseph M. Farley Unit 2 is seeking additional relaxation from the ASME Code Case N-770-2 examination requirement and the previous relief request to be able to defer the volumetric examination to the Fall 2020.

The technical justification to support this relief request is developed in this letter report based on a flaw tolerance analysis. The objective of the flaw tolerance analysis is to determine the largest initial axial and circumferential flaw sizes that could be left behind in service and remain acceptable until the next planned inspection. This maximum allowable initial flaw size can then be compared to a flaw size which would have ~een detected during the Spring 20 I 0 RFO inlet nozzle D¥ weld examination based on the inspection detection capability.

The following sections provide a discussion of the methodology, geometry, loading and the flaw tolerance analyses perfonned to develop the technical justification for deviating from the volumetric examination requirements of ASME Code Case N-770-2.

- - This record was final approved on 1/10/2018 5:15:19 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 3 of20 LTR-SDA-17-035 Revision 0 Attachment B January 10, 2018 2.0 Methodology In order to support the technical justification for deferring the volumetric examination from the previously accepted relief request extension of Spring 2019 RFO to the Fall 2020 RFO for Joseph M. Farley Unit 2, it is necessary to demonstrate the structural integrity of the RV inlet nozzle DM welds subjected to the PWSCC growth mechanism. To demonstrate the structural integrity of the DM welds, it is essential to determine the maximum allowable initial flaw size that would be acceptable in the DM welds for the duration between examinations. This maximum allowable initial flaw size would be the largest flaw size that would remain acceptable until the Fall 2020 RFO. The maximum allowable initial flaw size for a given plant operation duration can be determined by subtracting the PWSCC growth for that plant operation duration from the maximum allowable end-of-evaluation period flaw size, which is determined in accordance with ASME Code Section XI [4].

To determine the maximum allowable end-of-evaluation period flaw sizes and the crack tip stress intensity factors used for the PWSCC analysis, it is necessary to establish the stresses, crack geometry and the material properties at the locations of interest. The applicable loadings which must be considered consist of piping reaction loads acting at the DM weld regions and the welding residual stresses which exist in the region of interest.

The piping loads at the RV inlet nozzle DM weld locations are used to determine PWSCC growth. In addition to the piping loads, the effects of welding residual stresses are also considered. For PWSCC, the crack growth model for the DM weld material is based on that given in MRP-115 for Alloy 182 weld material [5]; this PWSCC growth model is also documented in ASME Section XI [4]. The nozzle geometry and piping loads used in the fracture mechanics analysis are shown in Section 3.0. A discussion of the welding residual stress distributions used for the DM welds is provided in Section 4.0. The determination of the maximum allowable end-of-evaluation period flaw sizes is discussed in Section 5.0.

The maximum allowable initial flaw size will be determined based on the crack growth due to the PWSCC growth mechanism at the R V inlet nozzle DM weld. The PWSCC growth is calculated based on the normal operating temperature and the crack tip stress intensity factors resulting from the normal operating steady state piping loads and welding residual stresses as discussed in Section 6.0. Section 7.0 provides the crack growth curves used in developing the technical justification to deviate from the ASME Code Case N-770-2 guidelines by deferring the volumetric inspection of the R V inlet nozzle DM welds from the Spring 2019 to Fall 2020 RFO.

- - This record was final approved on 1/10/2018 5:15:19 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 4 of20 LTR-SDA-17-035 Revision 0 Attachment B January 10, 2018 3.0 Nozzle Geometry and Loads The DM weld geometry for the Joseph M. Farley Unit 2 RV inlet nozzles is based on the nozzle detail drawings [6]. The operating temperature of the reactor vessel inlet nozzles is based on the Joseph M.

Farley relief request [3]. The RV inlet nozzle geometry and normal operating temperature used in the analysis are summarized in Table 3-1.

The piping reaction loads at the RV inlet nozzle DM weld locations are summarized in Table 3-2. These loads are used in determining the maximum allowable end-of-evaluation period flaw sizes and the PWSCC growth.

Table 3-1 Joseph M. Farley Unit 2 Reactor Vessel Inlet Nozzle Geometry and Normal Operating Temperature Dimension Outside Diameter (in) 34.00 Inside Diameter (in) 27.47 Thickness (in) 3.27 RV Inlet Nozzle Normal Operating Temperature = 541 op Table 3-2 Joseph M. Farley Unit 2 Reactor Vessel Inlet Nozzle Piping Loads a,c,e

'*'*'*This record was final approved on 1/10/2018 5:15:19 PM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 5 of20 L TR-SDA-17-035 Revision 0 Attachment B January 10, 2018 4.0 Dissimilar Metal Weld Residual Stress Distribution The plant specific welding residual stresses used in the PWSCC growth analysis are determined from the finite element stress analysis (FEA) in [7] based on the Joseph M. Farley Unit 1 and 2 RV inlet nozzle DM weld specific configuration. Figure 4-1 shows a sketch of the Joseph M. Farley Unit 2 inlet nozzle DM weld configuration. The FEA in [7] is based on a two-dimensional axisymmetric model of the inlet nozzle DM weld region. The FEA model geometry includes a portion of the low alloy steel nozzle, the stainless steel safe end, a portion of the stainless steel piping, the DM weld attaching the nozzle to the safe end, and the stainless steel weld attaching the safe end to the piping. The FEA model also assumes a 360° inside surface weld repair with a repair depth of 50% through the DM weld thickness, which is consistent with MRP-287 guidance [8]. As documented in the previous Joseph M. Farley relief request

[3], the RV inlet nozzle DM weld did not undergo any inside diameter surface weld repairs during fabrication of the DM welds; therefore, the 50% weld repair through the DM weld thickness is conservative. The following fabrication sequence was simulated in the FEA and matches the information provided in the reactor vessel nozzle details drawings [6]:

The inlet nozzle was buttered with weld-deposited Alloy 82/182 material. Nozzle and buttering are post weld heat treated (PWHT) at 1,1 00°F.

The inlet nozzle was welded to the safe end ring forging using an Alloy 82/182 weld. The inner diameter of the dissimilar metal weld is machined to finished size.

An assumed 50% inside surface weld repair 360° around the circumference was conservatively simulated in the Alloy 82/182 weld, which is consistent with MRP-287 [8].

Shop hydrostatic test was then performed at a pressure of3110 psig and a temperature of300°F.

The safe end was then machined for the piping side weld preparation.

The machined safe end was welded to a long segment of stainless steel piping using a stainless steel weld.

A plant hydrostatic test was performed at 2485 psig pressure with a temperature of 300°F.

After the plant hydrostatic test, normal operating temperature and pressure was uniformly applied three times to consider any shakedown effects, after which the model was set to normal operating conditions.

Based on the FEA model, residual stresses at three different paths (centerline of the DM weld, nozzle side of the DM weld, and safe-end side of the DM weld) in the DM weld were obtained. Additionally, a recommended stress distribution path was also provided, which is a representation of the limiting stress from all three paths through the DM weld. The recommended axial and hoop stress profiles were used in the generation of the crack growth charts to determine the maximum allowable initial flaw sizes (Section 7.0). The hoop and axial welding residual stresses for the recommended stress profiles at 100% normal operating conditions (operating pressure and temperature) are shown in Figure 4-2.

- - This record was final approved on 1/10/2018 5:15:19 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 6 of20 L TR-SDA-17-035 Revision 0 Attachment B January 10, 2018 Dissimilar Metal Weld Nozzle Forging Safe-End Butter Figure 4-1: Joseph M. Farley Unit 2 Reactor Vessel Inlet Nozzle DM Weld Configuration

- - This record was final approved on 1/10/2018 5:15:19 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 7 of20 LTR-SDA-17-035 Revision 0 Attachment B January 10, 2018 a,c,e Figure 4-2: Joseph M. Farley Unit 2 Reactor Vessel Inlet Nozzle DM Weld 100°/o Normal Operating Recommended Residual Stress Profiles Through DM Weld with 50%. Inside Surface Weld Repair

- - This record was final approved on 1/10/2018 5:15:19 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 8 of20 L TR-SDA-17-035 Revision 0 Attachment B January 10, 2018 5.0 Maximum Allowable End-of-Evaluation Period Flaw Size Determination In order to develop the technical justification to defer the volumetric examination of the RV inlet nozzle DM welds from the previously accepted relief request extension of Spring 2019 RFO to Fall 2020 RFO, the first step is the determination of the maximum allowable end-of-evaluation period flaw sizes. The maximum allowable end-of-evaluation period flaw size is the size to which an indication is allowed to grow until the next inspection or evaluation period. This particular flaw size is determined based on the piping loads, geometry and the material properties of the component. The evaluation guidelines and procedures for calculating the maximum allowable end-of-evaluation period flaw sizes are described in paragraph IWB-3640 and Appendix C of the ASME Section XI Code [4].

Rapid, nonductile failure is possible for ferritic materials at low temperatures, but is not applicable to the nickel-base alloy material. In nickel-base alloy material, the higher ductility leads to two possible modes of failure, plastic collapse or unstable ductile tearing. The second mechanism can occur when the applied J integral exceeds the J 1c fracture toughness, and some stable tearing occurs prior to failure. If this mode of failure is dominant, then the load-carrying capacity is less than that predicted by the plastic collapse mechanism. The maximum allowable end-of-evaluation period flaw sizes of paragraph IWB-3640 for the high toughness materials are determined based on the assumption that plastic collapse would be achieved and would be the dominant mode of failure. However, due to the reduced toughness of the DM welds, it is possible that crack extension and unstable ductile tearing could occur and be the dominant mode of failure. To account for this effect, penalty factors called "Z factors" were developed in ASME Code Section XI, which are to be multiplied by the loadings at these welds. In the current analysis for Joseph M. Farley Unit 2, Z factors based on MRP-216 [9] are used in the analysis to provide a more representative approximation of the effects of the DM welds; these Z factors are also. documented in ASME Code Section XI [4]. The use of Z factors in effect reduces the maximum allowable end-of-evaluation period flaw sizes for flux welds and thus has been incorporated directly into the evaluation performed in accordance with the procedure and acceptance criteria given in IWB-3640 and Appendix C of ASME Code Section XI. It should be noted that the maximum allowable end-of-evaluation period flaw sizes are limited to only 75% of the wall thickness in accordance with the requirements of ASME Section XI paragraph IWB-3640 [4].

The maximum allowable end-of-evaluation period flaw sizes determined for both axial and circumferential flaws have incorporated the relevant material properties, pipe loadings and geometry.

Loadings under normal, upset, emergency and faulted conditions are considered in conjunction with the applicable safety factors for the corresponding service conditions required in the ASME Section XI Code.

For circumferential flaws, axial stress due to the pressure, deadweight, thermal expansion, seismic and pipe break. loads are considered in the evaluation. As for the axial flCi;WS, hoop stress resulting from pressure loading is used.

The maximum allowable end-of-evaluation period flaw sizes for the axial and circumferential flaws at the RV inlet nozzle DM welds are provided in Table 5-1. The maximum allowable end-of-evaluation period axial flaw size was calculated with an assumed aspect ratio (flaw length/flaw depth) of 2. The aspect ratio of 2 is reasonable because the axial flaw growth due to PWSCC is limited to the width of the DM weld configuration. For the circumferential flaw, a conservative aspect ratio of I 0 is used.

- - This record was final approved on 1/10/2018 5:15:19 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 9 of20 LTR-SDA-17-035 Revision 0 Attachment B January 10, 2018 It should be noted that the resulting maximum allowable end-of-evaluation period flaw sizes were limited by the ASME Code limit of75% of the weld thickness for both flaw configurations.

Table 5-l Maximum End-of-Evaluation Period Allowable Flaw Sizes (Flaw Depth/Wall Thickness Ratio- aft)

Axial Flaw Circumferential Flaw (Aspect Ratio= 2) (Aspect Ratio= 10) 0.75 0.75

- - This record was final approved on 1/10/2018 5:15:19 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 10 of20 L TR-SDA-17-035 Revision 0 Attachment B January 10, 2018 6.0 PWSCC Growth Analysis A PWSCC growth analysis was perform~d to determine the maximum allowable initial flaw size that would be acceptable based on ASME Section XI acceptance criteria [4] for the operating duration from the Spring 2010 to the Fall 2020 RFO. The maximum allowable initial flaw size for the given plant operation duration is determined by subtracting the crack growth due to PWSCC for the specific plant operation duration from the maximum allowable end-of-evaluation period flaw size shown in Table 5-1.

Crack growth due to PWSCC is calculated for both axial and circumferential flaws using the normal operating condition steady-state stresses. For axial flaws, the stresses included pressure and residual stresses, while for circumferential flaws, the stresses considered are pressure, 1OOo/o power normal thermal expansion, deadweight and residual stresses. The input required for the crack growth analysis is basically the information necessary to calculate the crack tip stress intensity factor (K1), which depends on the geometry of the crack, its surrounding structure and the applied stresses. The geometry and loadings for the nozzles of interest are discussed in Section 3.0 and the applicable residual stresses used are discussed in Section 4.0. Once K1 is calculated, PWSCC growth can be calculated using the applicable crack growth rate for the nickel-base alloy material (Alloy 182) from MRP-115 [5], which is also documented in ASME Section XI [4]. For all inside surface flaws, the governing crack growth mechanism for the R V inlet nozzle is PWSCC.

Using the applicable stresses at the DM welds, the crack tip stress intensity factors can be determined based on the stress intensity factor expressions from API-579 [10]. The through-wall stress distribution profile is represented by a 4th order polynomial:

Where:

ao, a~. a 2 , a 3, and a 4 are the stress profile curve fitting coefficients; xis the distance from the wall surface where the crack initiates to the crack tip; t is the wall thickness; and a is the stress perpendicular to the plane of the crack.

The stress intensity factor calculations for semi-elliptical inside surface axial and circumferential flaws

+/-

are expressed in the general form as follows:

K1 ~ ~ Gj(a/c, aft, t/R,<l>)~j (~Y

~Q J=O Where:

a Crack depth c Half crack length along surface t Thickness of cylinder

- - This record was final approved on 1/10/2018 5:15:19 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 11 of20 LTR-SDA-17-035 Revision 0 Attachment B January 10, 2018 R Inside radius

<l> Angular position of a point on the crack front Gj Gj is influence coefficient for fh stress distribution on crack surface (i.e., G0, a~.

G2, G3, G4)

Q The shape factor of an elliptical crack is approximated by:

Q = I + 1.464(alc)L65 for ale~ I or Q = I + 1.464(c/a)L65 for ale> I The influence coefficients at various points on the crack front can be obtained by using an interpolation method. Once the crack tip stress intensity factors are determined, PWS~C growth calculations can be performed using the crack growth rate below with the applicable normal operating temperature.

The PWSCC growth rate used in the crack growth analysis is based on the Electric Power Research Institute (EPRI) recommended crack growth curve for Alloy I82 material [5]:

-da =exp [-Qg - (I-- - I )] a(K)P dt R T Tref Where:

da Crack growth rate in m/sec (in/hr) dt Qg Thermal activation energy for crack growth= I30 kJ/mole (3I.O kcal/mole)

R Universal gas constant= 8.3I4 x I0- 3 kJ/mole-K CI.I03 x I0-3 kcal/mole- 0 R)

T = Absolute operating temperature at the location of crack, K ( 0 R)

Tref Absolute reference temperature used to normalize data= 598.I5 K (1 076.67°R) a Crack growth amplitude 1.50 x I0- 12 at 325°C (2.47 x I0-7 at 6I7°F)

~ Exponent = I.6 K Crack tip stress intensity factor MPav'm (ksiv'in)

The normal operating temperature used in the crack growth analysis is 54 I °F at the R V inlet nozzle as reported in the previous relief request [3]. It should be noted that the fatigue crack growth mechanism is not considered in the crack growth analysis as it is considered to be small when compared to the crack growth due to the PWSCC growth mechanism at the reactor vessel inlet nozzle for the duration of interest. This is demonstrated by the low fatigue usage factor of 0.002I at the inlet nozzle location of interest in the reactor vessel analytical report CENC-II9I [II]. Therefore, it is not necessary to consider fatigue crack growth in the evaluation.

'**"This record was final approved on 1/10/2018 5:15:19 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 12 of20 LTR-SDA-17-035 Revision 0 Attachment B January 10, 2018 7.0 Technical Justification for Deferring the Volumetric Examination In accordance with ASME Code Case N-770-2 [1], the volumetric examination interval for the unmitigated RV inlet nozzle to safe end DM welds must not exceed 7 years. Joseph M. Farley Unit 2 is seeking relaxation from the ASME Code Case N-770-2 requirement in order to defer the volumetric examination of the reactor vessel inlet nozzle to safe end DM welds from the Spring 2019 [2, 3] to Fall 2020 RFO. Technical justification can be developed to support deferring the volumetric examination by calculating the maximum allowable initial flaw size that could be left behind in service and remain acceptable between the inspections. This maximum allowable initial flaw size can then be compared to a flaw size which would have been detected during the Spring 2010 RFO inlet nozzle DM weld examination based on the inspection detection capability.

The maximum allowable initial flaw depth is determined by subtracting the amount of PWSCC growth for a conservative duration of 10.5 effective full power years (EFPY) from the maximum allowable end-of-evaluation period flaw depth shown in Table 5-1. The end-of-evaluation period flaw depth is calculated based on the guidelines given in paragraph IWB-3640 and Appendix C of the ASME Section XI Code [4]. The PWSCC growth at the Alloy 82/182 weld is calculated based on the normal operating condition, piping loads, and the welding residual stresses at the DM weld as well as the crack growth model in MRP-115 [5]. The maximum allowable initial flaw depth was calculated for an axial flaw with an assumed aspect ratio of 2. An aspect ratio of 2 is reasonable for the axial flaw due to the DM weld configuration since any PWSCC axial flaw growth is limited to the width of the weld. For the circumferential flaw, a conservative aspect ratio of 10 is used in the crack growth analysis.

The PWSCC growth analysis of the circumferential flaws considered two cases: normal operating piping loads (deadweight and normal thermal loads) with residual stresses from the recommended profile (shown in Figure 4-1) and normal operating piping loads without residual stresses in order to obtain the most limiting crack growth results since a portion of the axial residual stress profile is compressive. It was determined that the case which included piping loads and welding residual stresses was limiting for circumferential flaws. The inclusion of welding residual stresses in the evaluation is conservative for the circumferential flaw evaluation since the RV inlet nozzle DM weld did not undergo any weld repairs.

The PWSCC growth curves and the maximum allowable initial flaw sizes for an axial flaw and a circumferential flaw are shown in Figures 7-1 and 7-2, respectively. The horizontal axis displays service life in EFPY, and the vertical axis shows the flaw depth to wall thickness ratio (aft). The maximum allowable end-of-evaluation period flaw sizes are also shown in these figures for the respective flaw configurations. Based on the crack growth results from Figures 7-1 and 7-2, the maximum allowable initial flaw sizes for the axial and circumferenth~.l flaws are tabulated in Table 7-1 for 10.5 EFPY.

- - This record was final approved on 1/10/2018 5:15:19 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 13 of20 LTR-SDA-17-035 Revision 0 Attachment B January 10, 2018 Table 7-1 Joseph M. Farley Unit 2 Maximum Allowable Initial Flaw Sizes for 10.5 EFPY Axial Flaw Circumferential Flaw (Aspect Ratio = 2) (Aspect Ratio = 10)

Maximum Allowable Initial 0.075 0.646 Flaw Size (aft)

Flaw Depth (in) 0.245 2.11 Flaw Length (in) 0.490 21.1 Note: Aspect ratio= flaw length/flaw depth Wall thickness (t) = 3.27 inches The flaw sizes shown in Table 7-1 are the largest axial and circumferential flaw sizes that could be left behind in service and remain acceptable from the Spring 2010 to Fall 2020 RFO (10.5 EFPY) for Joseph M. Farley Unit 2. In accordance with the Ultrasonic Testing (UT) detection and sizing requirements in ASME Section XI Appendix VIII, Supplement 10 [4], the minimum required detectable flaw depth is 10% of the wall thickness. Therefore, the maximum allowable initial circumferential flaw size is above the minimum flaw depth requirement per the UT detection capabilities, and thus would have been reasonably detected at the previous inspection of the DM welds.

In addition to the required baseline volumetric UT examination of the RV inlet nozzle DM weld, Joseph M. Farley Unit 2 also conducted Eddy Current Testing (ET) on the RV inlet nozzle DM welds as noted in

[3]. The ET inspection determined that were no unacceptable indications detected in the Spring 2010 RFO for Joseph M. Farley Unit 2.

The ET examination is an additional means to detecting surface breaking indications on the inside surface of the DM weld. The Joseph M. Farley qualification process for the ET procedure is discussed in [12],

which is a Nuclear Regulatory Commission (NRC) request for additional information response provided by Joseph M. Farley for justification of their ET procedure. Per [12], the qualification process and practical trial for the ET procedure is in accordance with European Network for Inspection Qualification (ENIQ) guidelines. Based on the qualification guidelines as discussed in [12], the Eddy Current examination is capable of detecting fatigue and intergranular stress corrosion cracking (IGSCC) I interdendritic stress corrosion cracking (IDSCC) cracks as small as 0.04" deep by 0.24" long.

The Joseph M. Farley ET inspection procedure from the Spring 2010 RFO was capable of detecting surface connected flaws of 0.04" in depth and 0.24" long per [12]. For Joseph M. Farley Unit 2, the maximum initial axial flaw depth is 0.245" and flaw length is 0.490" from Table 7-1; these postulated flaw dimension are greater than the Joseph M. Farley ET flaw depth of 0.04" and flaw length of 0.24".

As a result, the calculated maximum allowable initial axial flaw size is large enough, if present, to have been detected during the last Spring 2010 RFO examination of the RV inlet nozzle DM welds at Joseph M. Farley Unit 2. Similar justification was used in the previous Joseph M. Farley relief request [3], as well as for Comanche Peak Unit 1 [13] and South Texas Unit 1 [14] RV inlet nozzle DM weld alternate inservice inspection relief request for axial initial flaw depth less than 10% of the through-wall thickness.

Furthermore, the NRC staff safety evaluation report response [3] for the Farley relief request [2] and the safety evaluation report responses for the Comanche Peak relief request [13] and the South Texas relief

- - This record was final approved on 1/10/2018 5:15:19 PM. (This statement was added by the PRIME system upon its validation)

/

Westinghouse Non-Proprietary Class 3 Page 14 of20 LTR-SDA-17-035 Revision 0 Attachment B January 10, 2018 request [14] accepted the use of the licensee's ET qualification process to justify the acceptability for initial flaw sizes less than 10% of the through-wall thickness when supplemented with volumetric examinations performed by UT as required by the ASME Code Case N-770-2.

Therefore, the maximum allowable initial axial and circumferential flaw sizes in Table 7-1 would have been detected during the Spring 2010 RFO inlet nozzle DM weld examination. Since, there were no indications found during the Spring 2010 RFO for the inlet nozzle DM weld as documented in [3 ], the technical justification developed in this letter report can be used to defer the volumetric examination for the Joseph M. Farley Unit 2 RV inlet nozzle DM welds from the previously accepted relief request extension of Spring 2019 RFO to the Fall 2020 RFO.

Please note that between Spring 2010 RFO and Fall 2020, Farley Unit 2 will have operated for 9.67 EFPY at 100% power based on plant data (considering both actual and projected cycles). Thus, Figure 7-1, also shows that an axial flaw size of 9.2% of the wall thickness (flaw depth = 0.3" and flaw length =

0.6") could be left behind in service and remain acceptable from the Spring 2010 to Fall 2020 RFO for Farley Unit 2. These postulated flaw dimensions (flaw depth = 0.3" and flaw length = 0.6") for 9.67 EFPY are still greater than the Joseph M. Farley ET flaw depth of 0.04" and flaw length of 0.24". Thus, there is more margin between the postulated flaw size and the ET detection capability with the consideration of9.67 EFPY based on projected plant operational data.

...... This record was final approved on 1/10/2018 5:15:19 PM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 15 of20 LTR-SDA-17-035 Revision 0 Attachment B

-i

r January 10, 2018 f

r.;r

~ 0.8 a

0

T---,----r--- ---~----~---~--- ~ ----~---1---~---- ---T---,----r--- ---+---f---+--~ ---: --+---~---

I I I I I I I I

---~----r---T---I 1 I I 1 I I I I I

~ ---"'1----~---T--- ----r---~---~----

~ =~ =~E =~ =~

---~---~----r--- ---~----r---T--- ----~---~---~---- ~---"'1----r---

en ---~-- - -~---+--- ~-

~---*---~---- -~ ~zt --* *--~ ~~ - --~-~ -~ -~ -~ ~ -~ - ~ !~ -:: -~ !---~----~---

!l

J I I

'-----4----i----

1 I I

I 1 I

---~----~---

1 I I

+I ---

I

~--

I I I

- ~---~----

I I

-- I I I

---~---

I I

---~----~---+---

I

i---i---;----

1 I I

- 1t ~---i ____ i __ _

I I I 0.7

' I I I I I

~

I I I

l I I

~

Ill

"'C ---t---~----~--- - --- ---~----~---

a -- ---:---- ~---

"'C

~~~~~===~===~=== ====t===1===~==== ===~===~===~~~=~ ~~~ ~~~~ ~~~l~~~~~~~~

I I I

~ I I I

-r--- -

lYE

- - - T--- -~---

I I I I I I I I I I I I I I 0

J

__. ~0.6 ---T---,----r--- ---,----r---T--- ----r---T---,----

I I I I I I I I I I I I I I I I f I I I

---T---~----,---- ---~----r---T-- -T---~---- T---~----r---

__. ---T---.,----r--- ---,----r---T---

1 I I 1 I I 1 I I

---,----.----r--- ---~----~---~--

1 I I I

~---~-------- -f---~---- ---~---~----~---

.---"'1----~--- ----t----r----T--- --- .. ----t----t----

1 I I 1 I I 1 I I 1 I I 0

~ ; ___ ___ ...,. ____ ... __ _ ----~---~---~----

- --*----4____ ._ __ _ ---~----~---~-- -~---~---- - ~ ~---~----~---

=

1 I I 1 I I 1 I I 1 I I ' I I I I I I I

~ ---~----~---*--- ----~---~---~---- ---i----~---~-- -~---~---- - 1t ~---~----~---

0.5 I I I 1 I I 1 I I I I I CXl ~

~

I'll

J I I I I I I ~ I I I

- J_ f--- ~--- -~---

=

~===~ =~=I===~==== ===~====~~~~t=== ====~===t===~=

(11 - -t t----:----~---

.:.. -5 t----:----

---l----~----'----

c.o

-u :.= I

---~----

I

---r---~----

I

=

--- *r---7---

I I I I

r---r---~---- -j_

I I

~---

~ E-o 0.4 ' 1 I I I I I I I I I

I 1

I

---T---,----r---

I I I

-, -i--- ~- -- -~- --

I I I

~ ---T---,----r--- ----r---T---,---- ---T---,----r---

---~

-i ' I I I I I

--1-f--- ~--- -~---

r r.;r 1 I I I I I

,s

-r--- ~--- -:----

---T---~----r--- ----r---~---,----

1 I I I I I

~3 -sc. 0.3 ---~---~----~--- ----i---i---~----

-J_f--- ~- -- -~- --

1 I I I I I I I I I

~ ---4---~----L---

1 I I --~---

I ---~----~---~---

I I I ----~---*---~----

I I I

(!)

~ ---~---J----L--- ~----L--- ---~----~---~--- ----~---i---~----

3. 1 I t I I I I I I I I - -a- ~---~----~---.

____ L __ _

=

~ ---1---~----~--- ---~----~--- ---i----~---t--- ----~---1---i----

---~---J I 1 I

~

en ---t---~----~---* ---~----~---+---- --t---~----~--- ---~----~---t--------~---t---~---- _ j_f---~----~--- *

~ 0.2 I :---:

---f--- ~--- -~---~-- --~~--- -~ -- _--:--=_-

1 I I I I I I I I I I I I I I I I I I

~ ---~--- -~- -- ~--- ---- ~- -- ~--- ~---- i ---;--- -;---

I c.

~ ---T---,----r-- ---,----r-- T---

I I J I I ,

---,----r---,--- ----r---T---,---- ---T---,----r---

I I I I I I 0"

' I J

---,----r --T---

I I I I I I I I I f---~----~---

---~---,----r-- ---,----r---~--- ----r---~---~----

'< 1

---~---~----~--

I I I

--- ~- --

I I I

---4----~---~---

I I I I

~---T---~---- ;---~----~---

T

(!) 0.1 1 I t I I I I I I I I 1 ----- - - -~-f---~----~---

1 I I

-u ! ! Service Life= 9.67 EFPY I I JJ 1 I I ----L---i---~----

I I I

- ~t ~--- ~--- -~---

___ J. ___ J ____ L __ _

~

~---+----:---- ~-l1 ___ j I ____ I~---

---~--- ----L---~---J----

m en ----~---1----:--- Service Life= 10.5 EFPY ---+---

'< 0 1 l 1 t I t I I I I 1 3 0 2 4 6 8 10 12 14 c: Service Life (EFPY)

"'C 0

J

~

~ Figure 7-1: PWSCC Growth Curve for Joseph M. Farley Unit 2 Inlet Nozzle Axial Flaw (DM weld), Aspect Ratio= 2 c:

a(5" 2.

Westinghouse Non-Proprietary Class 3 Page 16 of20 L TR-SDA-17-035 Revision 0 Attachment B

~

T January 10, 2018 c;r 0.8

.ttA, (i)

- - .. - - - -* - - - .- -- - - -* - - - .- - -- -~

~.,.-

0 0

a. 1- r- ~ *I ...:~

i- 1- *I - ~ ~

~

en /

I ..,ftt"' I 0.7

!l

J ASME Maximum Allowable End-of-Evaluation Period Flaw Size 1

~ I I>> """"'

"C I a<

"C 1- 1- 1- ~

I

~

,...- ..,~

.-0.6

([) I a.

0

J

...... ~ -'"'"~" I

...... = I Maximum Initial Flaw Size Allowed ~ I

~ ~ ~1 During Spring 2010 RFO (a/t) = 0.646 ~

I 0.5

~

CXI fll fll

~ I

=

~

...... ~

c v

~ / I

-~

co ~

-u ~ 0.4 ~

I I

~ 1/ I

! ~

....=

I c;r g ;c. 0.3 ~ I

~ I

([) ~

3 Q ~

I a

([)

~

/ I i

I>>

~ 0.2 L

/ I

a. l..;" I a.

([)

a. 1/ I

~-""'"

0"

0.1 ""- I

([) I I

-u ~ Service Life= 10.5 EFPY

n ..... """" I I

~ I m I en 0 I 0 5 10 15 20 25 30 35 40 45 50 3

c: Service Life (EFPY)

"C 0

J

~

[ Figure 7-2: PWSCC Growth Curve for Joseph M. Farley Unit 2 Inlet Nozzle Circumferential Flaw (DM weld), Aspect Ratio= 10 a:

ac:r 2-

Westinghouse Non-Proprietary Class 3 Page 17 of20 LTR-SDA-17-035 Revision 0 Attachment B January 10, 2018 8.0 Summary and Conclusions A volumetric examination of the reactor vessel inlet nozzle to safe end DM butt welds was performed during the Spring 2010 RFO at Joseph M. Farley Unit 2. The next required volumetric examination is planned during the Spring 2019 RFO in accordance with the previously approved relief request [2, 3].

However, the volumetric examination will be deferred to the Fall 2020 RFO for the reactor vessel inlet nozzle DM welds. Since the time interval between the previous examination and the planned examination exceeds 7 years, which deviates from the Code Case N-770-2 inspection interval requirements, a flaw tolerance evaluation was completed to defer the volumetric examination of the inlet nozzle DM welds.

This letter report provides technical justification to support the relaxation request by performing a flaw tolerance analysis to determine the largest initial axial and circumferential flaws that could be left behind in service and remain acceptable between the planned examinations. This maximum allowable initial flaw size can then be compared to any flaw size which would have been detected during the previous inlet nozzle DM weld examinations.

Based on the PWSCC growth analysis results from Section 7.0 which is for a conservative duration of 10.5 EFPY, the maximum allowable initial flaw sizes for the reactor vessel inlet nozzle DM welds are tabulated in Table 8-1 for Joseph M. Farley Unit 2. These allowable initial axial and circumferential flaw sizes have been shown to be acceptable in accordance with the ASME Section XI IWB-3640 acceptance criteria through the Fall 2020 RFO for Joseph M. Farley Unit 2 taking into account of potential PWSCC growth since the last volumetric and surface examinations.

In accordance with the Ultrasonic Tes.ting (UT) detection and sizing requirements in ASME Section XI Appendix VIII, Supplement 10 [4], the minimum required detectable flaw depth is 10% of the wall thickness. In addition to the UT examination of the RV inlet nozzle DM weld, supplemental Eddy Current Testing (ET) was performed on the RV inlet nozzle DM welds for Joseph M. Farley Unit

2. Based on the qualification guidelines as discussed in [12], the Eddy Current examination is capable of detecting fatigue and IGSCCIIDSCC cracks as small as 0.04" deep by 0.24" long.

Based on the Joseph M. Farley Unit 2 results in Table 8-1, the calculated maximum allowable initial axial flaw size (flaw depth = 0.245" and flaw length = 0.490") for 10.5 EFPY is large enough to have been detected during the last Spring 2010 RFO examination of the RV inlet nozzle DM welds. Furthermore, the postulated flaw dimensions for 9.67 EPFY (flaw depth= 0.3" and flaw length= 0.6") are greater than the Joseph M. Farley ET flaw depth of 0.04" and flaw length of 0.24". Thus, there is more margin between the postulated flaw size and the ET detection capability with the consideration of 9.67 EFPY based on projected pla~t operational data. Similar justification was used in the previo1:1s Joseph M. Farley Units 1 and 2 relief request [3], as well as for Comanche Peak Unit 1 and South Texas Unit 1 RV inlet nozzle DM weld alternate inservice inspection relief request for axial initial flaw sizes less than 10% of the through-wall thickness. Furthermore, the NRC staff safety evaluation report response [3] to the Farley relief request [2] and the safety evaluation report responses for the Comanche Peak relief request

[13] and the South Texas relief request [14] accepted the use of the licensee's ET qualification process to justify the acceptability for initial flaw sizes less than 10% of the through-wall thickness when supplemented with volumetric examinations performed by UT as required by the ASME Code Case N-770-2. Therefore, deferring the volumetric examination for the Joseph M. Farley Unit 2 RV inlet nozzle

'**This record was final approved on 1/10/2018 5:15:19 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 18 of20 LTR-SDA-17-035 Revision 0 Attachment B January 10, 2018 DM welds from the Spring 2019 RFO allowed by Code Case N-770-2 to the Fall2020 RFO is technically justified.

Table 8-1 Joseph M. Farley Unit 2 Maximum Allowable Initial Flaw Sizes based on 10.5 EFPY Axial Flaw Circumferential Flaw (Aspect Ratio = 2) (Aspect Ratio= 10)

Maximum Allowable Initial 0.075 0.646 Flaw Size (aft)

Flaw Depth (in) 0.245 2.11 Flaw Length (in) 0.490 21.1 Note: Aspect ratio= flaw length/flaw depth Wall thickness (t) = 3.27 inches

- - This record was final approved on 1/10/2018 5:15:19 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 19 of20 LTR-SDA-17-035 Revision 0 Attachment B January 10, 2018 9.0 References

1. ASME Code Case N-770-2,Section XI Division 1, "Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities," Approval Date June 9, 2011.

2. United States Nuclear Regulatory Commission Letter Dated December 5, 2014, "Joseph M. Farley, Units I and 2, (FNP-ISI-ALT-15, Version I) Alternative to Inservice Inspection Regarding Reactor Pressure Vessel Cold-Leg Nozzle Dissimilar Metal Welds (TAC Nos. MF3687 and MF3688)."

(ADAMS Accession Number ML14262A317)

3. Southern Company, NL-14-0295, "Joseph M. Farley Nuclear Plant, Proposed Inservice Inspection Alternative FNP-ISI-ALT-15, Version 1.0," March 24, 2014, Docket Nos.: 50-348, 50-364.

4. ASME Boiler & Pressure Vessel Code, 2007 Edition with 2008 Addenda,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components."

5. Materials Reliability Program: Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds (MRP-115), EPRI, Palo Alto, CA: 2004.

I006696.

6. Drawings for Joseph M. Farley Unit 2 RV inlet nozzles:

a. Combustion Engineering Drawing E-233-943 Revision 3, "Pressure Vessel Final Machining

-Sections, Westinghouse Electric Corporation, 157" I.D. P.W.R."

b. Combustion Engine~ring Drawing E-233-926 Revision 4, "Nozzle Details For: We~tinghouse Electric Corp., 157 J.D. P.W.R."

7. Dominion Engineering, Inc. Document C-8869-00-0 I, Revision 0, "Welding Residual Stress Calculation for Farley Units 1 and 2 RPV Inlet Nozzle DMW," May 2013.

8. Materials Reliability Program: Primary Water Stress Corrosion Cracking (PWSCC) Flaw Evaluation Guidance (MRP-287), EPRI, Palo Alto, CA: 2010, 1021023.

9. Materials Reliability Program: Advanced FEA Evaluation of Growth of Postulated Circumferential PWSCC Flaws in Pressurizer Nozzle Dissimilar Metal Welds (MRP-216, Rev. 1): Evaluations Specific to Nine Subject Plants. EPRI, Palo Alto, CA: 2007. 1015400.

10. American Petroleum Institute, API 579-1/ASME FFS-1 (API 579 Second Edition), "Fitness-For-Service," June 2007.

11. Combustion. Engineering, Inc. Report CENC-1191, "Analytical Report for Alabama Power and Light Company, J. M. Farley Station Unit No.2 Reactor Vessel," May 1974.

12. Southern Nuclear Company, Inc. Letter NL-14-1193, "Joseph M. Farley Nuclear Plant Response to Request for Additional Information Regarding Proposed Alternative to lnservice Inspection Requirements of ASME Code Case N-770-1," Docket Nos. 50-348 and 50-364, Dated August 1, 2014. (ADAMS Accession Number ML14213A484)

- - This record was final approved on 1/10/2018 5:15:19 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 20 of20 LTR-SDA-17-035 Revision 0 Attachment B January 10, 2018 I3. United States Nuclear Regulatory Commission Letter Dated March I4, 20I6, "Comanche Peak Nuclear Power Plant, Unit I - Relief Request IB3-3, Alternative to the ASME Code,Section XI, Examination Requirements for Reactor Pressure Vessel Cold-Leg Weld Inspection Frequency (CAC No. MF6I25)." (ADAMS Accession Number ML16074AOOJ)

I4. United States Nuclear Regulatory Commission Letter Dated August 2I, 20I5, "South Texas Project, Unit I - Request for Relief No. RR-ENG-3-I7 for Extension of the Inspection Frequency of the Reactor Vessel Cold-Leg Nozzle to Safe-End Welds with Flaw Analysis (Tac No. MF6I74)."

(ADAMS Accession Number ML15218A367)

- - This record was final approved on 1/10/2018 5:15:19 PM. (This statement was added by the PRIME system upon its validation)

v t e Letter Sequence Request
EPID:L-2018-LLR-0062, Proposed Inservice Inspection Alternative FNP-ISI-ALT-05-03, Version 1.0 (Open)
Initiation Request Acceptance Administration Withholding Request Acceptance Questions Answers Results Approval... Other: ML18334A228
MONTHYEAR2018-04-18 [Table View]

NL-18-0193, Proposed Inservice Inspection Alternative FNP-ISI-ALT-05-03, Version 1.0

Text

Applicable Code Edition and Addenda

References:

Subject:

~===~ =~=I===~==== ===~====~~~~t=== ====~===t===~=

=

Navigation menu

Search