Text

Response to Requests for Additional Information for Relief for Alternate Requirements for Repair of Reactor Vessel Head Penetrations with Nozzles Having Pressure-Retaining Partial-Penetration J-Groove Welds
	ML16043A148
Person / Time
Site:	Byron, Braidwood
Issue date:	02/11/2016
From:	Gullott D; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML16043A357	List: RS-16-045, Response to Requests for Additional Information for Relief for Alternate Requirements for Repair of Reactor Vessel Head Penetrations with Nozzles Having Pressure-Retaining Partial-Penetration J-Groove Welds;
References
	RS-16-045
	Download: ML16043A148 (17)
	v • d • e

4300 Winfield Road Warrenville, IL 60555 ExeLon Generation

630 657 2000 Office Proprietary information contained in Attachments 2 through 5 Withhold from public disclosure under 10 CFR 2.390 When separated, the cover letter, Attachment 1 and Attachment 6 are non-proprietary RS-16-045 10 CFR 50.55a February 11, 2016 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Braidwood Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457 Byron Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455

Subject:

Response to Requests for Additional Information for Relief for Alternate Requirements for Repair of Reactor Vessel Head Penetrations with Nozzles Having Pressure-Retaining Partial-Penetration J-Groove Welds

References:

1) Letter from D. M. Gullott (Exelon Generation Company, LLC) to U. S. Nuclear Regulatory Commission, "Requests for Relief for Alternate Requirements for Repair of Reactor Vessel Head Penetrations with Nozzles Having Pressure-Retaining Partial-Penetration J-Groove Welds," dated September 11, 2015

2) Email from J. Wiebe (NRC) to J. Krejcie (Exelon Generation Company, LLC),

Preliminary Request for Additional Information Regarding the Braidwood and Byron Requests for Relief Regarding Repair of Reactor Vessel Head Penetration J-Groove Welds, dated January 6, 2016 In accordance with 10 CFR 50.55a, "Codes and standards," paragraph (z)(1), Exelon Generation Company, LLC (EGC), requested relief from the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (B&PV Code) specifically related to the repair of the degraded reactor vessel closure head (RVCH) penetration nozzles and their associated partial penetration J-groove attachment welds.

Specifically, the relief request proposed to perform an alternative repair technique using the Areva Inside Diameter Temper Bead (IDTB) welding method to restore the pressure boundary of a degraded nozzle. EGC submitted this request to the NRC in Reference 1.

Subsequent to submittal of Reference 1, the NRC requested additional information to support the review of the subject relief request (Reference 2).

Note, the subject relief request was submitted as a contingency effort to support repair of potential indications requiring repair as detected during an Inservice Inspection (ISI) Program

February 11, 2016 U.S. Nuclear Regulatory Commission Page 2 ultrasonic (UT) examination of the affected vessel head penetrations (VHPs). The relief request was submitted on a preemptive basis to ensure a timely repair and minimize impact on NRC resource's on an emergent basis should Braidwood Station or Byron Station discover a flaw that requires a repair in the upcoming outages. The need for a repair could be determined as early as during the Byron Station refueling outage in spring 2016 (132R19). includes the response to the NRC's requested additional information.

Attachments 2 through 5 provide supporting documents containing proprietary information as defined by 10 CFR 2.390, "Public inspection, exemption, requests for withholding." Areva, Inc,(Areva) as the owner of the proprietary information has executed the enclosed affidavit, which identifies that the enclosed proprietary information has been handled and classified as proprietary, is customarily held in confidence, and has been withheld from public disclosure.

The proprietary information was provided to EGC by Areva as referenced by the affidavit. The proprietary information has been faithfully reproduced in the attached information such that the affidavit remains applicable. Areva hereby requests that the attached proprietary information be withheld, in its entirety, from public disclosure in accordance with the provisions of 10 CFR 2.390 and 10 CFR 9.17. The affidavit supporting the proprietary nature of the information is contained in Attachment 6. EGC will provide a follow-up supplement to this response containing the non-proprietary versions of the Attachment 2 through 5 documents.

There are no regulatory commitments contained within this letter.

Should you have any questions concerning this letter, please contact Ms. Jessica Krejcie at (630) 657-2816.

Respectfully, David M. Gullott Manager Licensing Exelon Generation Company, LLC : Response to Request for Additional Information : AREVA Document #51-9234023-000, "Corrosion Evaluation of Byron Units 1 and 2 and Braidwood Units 1 and 2 IDTB Weld Repairs" (PROPRIETARY) : AREVA Document #51-9240805-001, "Byron Units 1 & 2, and Braidwood Units 1

& 2 IDTB Reactor Vessel Head Penetration Nozzle Weld Repair-Life Assessment Summary" (PROPRIETARY) : AREVA Document #51-9233902-000, "PWSCC Evaluation for Contingency RVCH Nozzle Repairs at Byron Units 1 and 2 and Braidwood Units 1 and 2" (PROPRIETARY) : AREVA Document #51-9245035-001, "Ambient IDTB Weld Interpass Temperature Evaluation" (PROPRIETARY) : AREVA Inc., Affidavit for AREVA documents, February 10, 2016 Response to Request for Additional Information Response to Request for Additional Information In accordance with 10 CFR 50.55a, "Codes and standards," paragraph (z)(1), Exelon Generation Company, LLC (EGC), requested relief from the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (B&PV Code) specifically related to the repair of the degraded reactor vessel closure head (RVCH) penetration nozzles and their associated partial penetration J-groove attachment welds.

Specifically, the relief request proposed to perform an alternative repair technique using the Areva Inside Diameter Temper Bead (IDTB) welding method to restore the pressure boundary of a degraded nozzle. EGC submitted this request to the NRC in Reference 1.

Subsequent to submittal of Reference 1, the NRC requested additional information to support the review of the subject relief request (Reference 2). The below information provides the response to the NRC's requested additional information.

NRC RAI-1 The NRC staff notes that Section 1.0 of Attachment 1 to the relief request, ASME Code Components Affected, indicates that this request applies to vessel head penetrations 1-78 except those with previous repairs but does not state that these include the core exit thermocouple (CFTC), control rod drive mechanisms (CRDM), and reactor vessel level indication systems (RVLIS) nozzles. Confirm that all types of nozzles discussed in the request are included in section 1 of the request.

EGC Response to RAI-1:

All types of nozzles discussed in RAI-1 are included in Section 1.0 of Reference 1. With exception to those nozzles previously repaired; the 78 nozzles identified in Section 1.0 of Reference 1 encompass the 5 core exit thermocouple (CETC) nozzles, 53 control rod drive mechanism (CRDM) nozzles, and 2 reactor vessel level indication systems (RVLIS) nozzles and 18 spare CRDM nozzles.

Previous repairs at Byron consist of:

Unit 1, P-31, P-43, P-64 and P-76 Unit 2, P-6 and P-68 Previous repairs at Braidwood consist of:

- Unit 1, P-69 Unit 2, no repairs NRC RAI-2 In Section 4.0 of Attachment 1 to relief requests, the licensee stated, in part, that, "The modification described in Section 4.0 will be performed in accordance with ASME Code Case N-638-4 and the alternatives discussed in Section 5.0. H Page 1 of 10 Response to Request for Additional Information

a. Clarify whether all requirements of ASME Code Case N-638-4, including NRC imposed conditions, will be met, except the requirement in 1(g) for which relief is requested.

b. In determining interpass temperatures, Code Case N-638-4 requires the use of thermocouples (paragraph 3(e)(1)) unless an impracticality exists. Will thermocouples be used to determine interpass temperatures? If not please describe the method that will be used and justify the impracticality of using thermocouples.

c. Has the temper bead welding procedure proposed for use for this request been previously qualified? If so, did the Charpy V-notch tests meet the requirements of the code case or was it necessary to adjust the temperature as permitted by option 2 of the code case? If a temperature adjustment was required, please provide the adjusted temperature and describe how the adjusted temperature will be used in the evaluation of reactor vessel integrity analyses, e.g., pressure-temperature curves.

EGC Response to RAI-2a:

Yes, all requirements of ASME Code Case N-638-4, including NRC imposed conditions contained in Regulatory Guide 1.147, will. be met. Additional clarification is provided in the response to NRC RAI-2b below.

EGC Response to RAI-2b:

Code Case N-638-4, Section 3(e) requires the interpass temperature to be determined by one of the following methods i.e., 3(e)1, 3(e)2 or 3(e)3. Although the Code Case does not specifically prefer one method over another method, NRC imposed conditions per Regulatory Guide 1.147 must be considered, endorsing the use of thermocouples per code case method 3(e)1. Thermocouple use is deemed impractical for this application since the weldment area is inaccessible (internal bore welding) and does not allow for direct monitoring of the interpass temperature (physically restrained). For this application, a heat flow calculation was prepared, option 3(e)2, to determine the interpass temperature (51-9245035-001). The total temperature increase is conservatively estimated to be 183.6°F over the course of the entire weld process.

Thus, the maximum interpass temperature of 350°F will not be exceeded in accordance with the code case. The Interpass Temperature evaluation is included as Attachment 5.

EGC Response to RAI-2c:

Yes, the Welding Procedure Specification (WPS) provides qualified instructions and is in accordance with ASME Section XI, Code Cases N-638-1 and N-638-4 for Ambient Temperature Temper Bead. The Charpy V-notch tests meet the requirements of Code Cases N-638-1 and N-638-4 including impact testing without any temperature adjustment.

NRC RAI-3:

In Section 5.1 of Attachment 1 to relief requests, the licensee stated, in part, Page 2 of 10 Response to Request for Additional Information "Therefore, the following combination of UT and penetrant testing (PT) examinations are proposed. "

"PT examination will also be repeated on the remediated surfaces. "

a. Provide the percent coverage of the required examination volume and surface of weld repairs that the UT and PT will achieve. If the ultrasonic and surface exam coverage is less than essentially 100 percent as required, justify the acceptability of a lesser amount of examination coverage.

b. Provide the acceptance criteria to be used for flaws detected by the PT.

c. Describe the "remediated surfaces."

EGC Response to RAI-3a:

Approximately 70% of the modified CRDM and RVLIS nozzle weld surface is expected to be scanned by UT considering two-directional coverage. 100% of the modified CETC nozzle weld surface will be scanned by UT meeting coverage requirements.

There is no portion of the CRDM or RVLIS nozzle weld volume that does not receive at least single direction UT coverage.

100% of the modified CRDM, RVLIS, and CETC nozzle weld surfaces will be examined post IDTB welding via surface PT. PT examination will also be repeated on the remediated (peened) surfaces post application of the IDTB weld.

There is no portion of the repair that does not receive surface liquid penetrant examination.

Considering single direction UT, the combination of performing PT and UT examinations during the IDTB repair provides 100% coverage (volumetric and surface) of the modified CRDM, RVLIS, and CFTC nozzle surfaces.

EGC Response to RAI-3b:

The acceptance criteria that will be used for flaws detected by PT for weld metal, adjacent base materials and base materials are per ASME Section III, 2001 through summer of 2003 addenda, paragraph NB-2546.

EGC Response to RAI-3c:

Rotary peening (surface remediation) is performed on the final IDTB weld surface and roll expanded surface to alleviate tensile stress introduced by the new IDTB weld and mechanical roll expansion process.

The remediated surfaces or peened surfaces consist of the new IDTB weld surface, Alloy 600 (PWSCC susceptible) heat affected zone and the roll expanded ID surface. The upper portion of the nozzle inside the RVCH, prior to cutting the lower portion of the nozzle out, is slightly roll expanded to secure the upper portion of the nozzle to the RVCH and keep it in place. The roll expanded surface is above the IDTB weld. The remediation surface length on the ID of the nozzle or peened surface length is 3.72 inches minimum. Specifically, the ID surface to be Page 3 of 10 Response to Request for Additional Information peened (3.72 inch minimum) starts at the lower edge of the IDTB weld and continues upward to include the roll expanded (or machining extent) surface; this includes a 1/4" surface above the upper extent of the roll expanded surface, which is also the extent of the PT examination but does not include the entire ID of the remnant alloy 600 nozzle.

NRC RAI-4:

Clarify the precise areas which will undergo rotary peening.

EGC Response RAI-4:

The precise areas that will undergo rotary peening are described above in the EGC Response to RAI-3c.

NRC-RAI-5:

(a) Clarify whether the entire area (e.g., new weld, heat affected'zone (HAZ), and entire inside diameter (ID) of the remnant of the Alloy 600 nozzle) following rotary peening will be inspected by the UT and PT; (b) Provide percent coverage that is anticipated to be obtained from the UT and PT following peening; (c) If the UT and PT are not performed after peening, provide justifications that rotary peening process would not cause potential adverse effects on structural integrity of the components (e.g., potential surface microcracking and growth of subsurface fabrication anomalies).

EGC Response RAI-5a:

The entire area as described in EGC response to NRC-RAI-3c will be inspected by PT examination following rotary peening. UT examination will not be performed following rotary peening.

EGC Resoonse RAI-5b:

100% of the modified CRDM, RVLIS, and CETC nozzle weld surface, alloy 600 heat affected zone and the roll expanded ID surface will be examined post rotary peening via surface PT.

Refer to EGC Response RAI-3a. UT examination will not be performed following rotary peening. Refer to EGC Response RAI-5c.

EGC Resoonse RAI-5c:

LIT examination will not be performed following rotary peening. Rotary peening will only affect the surface condition and alleviates residual stresses associated with PWSCC. Rotary peening produces a compressive stress depth at the repair surface areas on the ID of the alloy 600 nozzle, namely the IDTB weld, alloy 600 HAZ and rolled transition surface where elevated tensile stresses may be present. Corrosion testing was performed for rotary peened samples that underwent massive bend and EPRI type U-bend testing. The results concluded that, after 2000 hrs of primary water exposure at 360°C, none of the rotary peened samples showed cracking at the surface, concluding that rotary peening mitigates the effects of PWSCC.

The rotary peening process is a qualified process that does not interfere or adversely impact the ability to perform PT examination. Furthermore, the rotary peening process was concluded to Page 4 of 10 Response to Request for Additional Information not induce a surface condition that would accelerate PWSCC behavior with evidence demonstrating a preventive effect towards PWSCC. Thus, providing assurance of structural integrity (no adverse effect) and precluding any influence on potential surface micro-cracking or growth of subsurface fabrication anomalies.

NRC RAI-6:

For the subject repaired RVCH penetration nozzle(s), clarify whether the licensee will meet the ASME Code Case N-729-1 required preservice and subsequent inservice inspections as well as the required frequency of inspections, as mandated by 10 CFR 50.55a(g)(6)(ii)(D).

EGC Response to RAI-6:

EGC will meet the ASME Code Case N-729-1 required preservice (PSI) and subsequent inservice (ISI) inspections, as well as the required frequency of inspections, as mandated by 10 CFR 50.55a(g)(6)(ii)(D) to the extent practical since the IDTB weld repair method removes much of the examination area depicted in figure 2 of Code Case N-729-1. Any IDTB repaired nozzles will continue to be examined each cycle in accordance with 10 CFR 50.55a(g)(6)(ii)(D)5.

NRC RAI-7:

Section 1.0 of Attachment 3 to the relief request stated, in part, that, "The CETC Nozzle is considered to be the controlling component, and therefore, this analysis will consider the CETC Nozzle only. "

Discuss the reason for considering the CETC nozzle to be the controlling component among the three types of nozzles (CRDM, RVLIS, and CETC).

EGC Response RAI-7:

This analysis is a bounding analysis covering three different groups of nozzles (CRDM, RVLIS, and CETC) at multiple locations on the RVCH, and across 4 units. The physical RVCH design and load conditions are identical across all units.

An evaluation of the CETC Nozzle is considered to envelope the CRDM/RVLIS Nozzles for the following reasons:

The CETC Nozzles are located further from the vessel centerline than the CRDM/RVLIS Nozzles resulting in a greater hillside angle which tends to produce larger stresses.

The CETC Nozzles have a smaller nominal weld length fused to the RVCH base metal.

The CETC Nozzles incorporate a lower nozzle which imparts additional loading to the IDTB weld as compared to the CRDM/RVLIS which does not incorporate a lower nozzle.

NRC RAI-8:

Section 5.3 in Attachment 1 to the relief requests discusses worst-case postulated flaw sizes in the J-groove weld.

Page 5 of 10 Response to Request for Additional Information

a. Describe the worst-case flaw sizes.

b. Section 6.0 refers to "40 years of plant life after a repair," "33 plant life," and "33 extended full power years (EFPY)." Explain clearly the difference between 40 years plant life, 33 years plant life, and 33 EFPY, how each number of year (calendar year or EFPY) is obtained, and why the 33 EFPY is bounding.

EGC Response RAI-8a:

The nozzle with the largest hillside angle is the worst case location, which for the current configuration is the CETC nozzle. The radial-axial flaws were postulated since the hoop stresses are dominant. Linear Elastic Fracture Mechanics evaluations were performed for the final flaw size from the fatigue crack growth evaluation. The postulated initial and final worst case flaw size is included as ai and of in Table 6-1 of Attachment 4 of Reference 1 for the uphill position and Table 6-2 of Attachment 4 of Reference 1 for the downhill position.

EGC Response to RAI-8b:

The IDTB weld repair is intended to support the remaining life of the original 40 year plant design basis (40 year operating license) plus license renewal extension (20 year additional operating license) or 60 year plant design basis life overall. At the time an IDTB weld repair would be performed (if applicable) for each individual Byron and Braidwood Unit, there would be approximately 10 to 11 years of life remaining from the original 40 year plant life on any of the 4 Units.

EFPY refers to "Effective Full Power Years" a duration which takes into account plant down time due to refueling outages that typically occur over a 26 day period when the plant is not at full power and considered a conservative estimate. For example, 33 years of plant life is approximately 12,045 days, whereas considering EFPY years for this same duration is approximately 11,473 days (22 outages at approximately 572 days not at full power in a 33 year span). Thus, 33 EFPY was considered as a conservative bounding life duration given the most limiting evaluation, i.e., as-left J-Groove flaw evaluation; evaluated to be acceptable for 33 years of plant life.

The calculations (e.g., as-left J-Groove flaw evaluation or weld anomaly flaw evaluation) provide an individual assessment of life expectancy associated with the IDTB weld repair. A repair life assessment evaluation was specifically prepared and included as Attachment 3.

NRC RAI-9:

Discuss whether qualified welding procedures were used to perform the weld repairs, discuss whether the welding procedures and qualification satisfy the industry code, and reference the industry code that the welding procedure and qualification satisfies.

EGC Response RAI-9:

See EGC Response RAI-2c.

N RC RAI-1 0:

The NRC staff notes that the licensee discussed weld residual stresses in Section 5.3 of, Attachment 3, and Attachment 5 to the relief requests.

Page 6 of 10 Response to Request for Additional Information

a. Clarify whether at least fourth order polynomial approximation was used to represent the through wall welding residual stress distributions for new repair weld and original J-groove weld;

b. If less than fourth order polynomial approximation for residual stress distributions was used, discuss the reason and justify its adequacy and validity.

EGC Response NRC RAI-10a:

Various fracture mechanics calculation methods were applied for different parts of the weld stress model as follows:

Original J-Groove Weld:

An explicit flaw finite element model was used to calculate the stress intensity factors. The weld residual stress distribution is used directly, with no polynomial curve fitting required. Refer to

'Analytical Methodology' provided in Section 2.0.3 (AREVA Doc #32-9244434, Attachment 5 of Reference 1).

New Repair Weld (IDTB):

The weight function method, which can account for highly nonlinear stress distributions, was used to calculate the stress intensity factors for evaluating 3600 circumferential and axial surface flaws for the new repair weld. The reported path line weld residual stress distributions are used directly, with no polynomial fitting required. Refer to 'Stress Intensity Factor (SI F)

Solutions' provided in Section 2.2 under 'Methodology' (AREVA Doc #32-9244389, Attachment 3 of Reference 1). The postulated 3600 circumferential and axial surface flaws are represented in Figure 2-1 (Refer to crack propagation paths 1 a/1 b/1 c and 2a/2b/2c).

For evaluating cylindrical flaws in the new repair weld, a third order polynomial fit was only used to represent the welding residual stress distributions over the flaw depth as required for the solution utilized (see Section XI, Appendix A, A-3200(b)). Refer to 'Stress Intensity Factor (SIF)

Solutions' provided in Section 2.2 under 'Methodology' and Appendix A (AREVA Doc #32-9244389, Attachment 3 of Reference 1). The postulated cylindrical flaws are represented by the vertical paths (interface between the I DTB repair weld and the RV Closure Head) in Figure 2-1 (Refer to crack propagation paths 3a/3b/3c and 4a/4b/4c).

EGC Response to NRC RAI-10b:

The third order polynomial fit was only used to represent the welding residual stress distributions over the flaw depth for evaluating cylindrical flaws for the new repair weld (see response to 10a). Although through wall welding residual stress distributions can be highly nonlinear in general, the polynomial fit only needs to describe the stress distribution over the much shorter crack depth (see Section XI, Appendix A, A-3200(b)). This short segment (refer to section 4.1 of Referencel Attachment 3 for segment length (Areva document #32-9244389),

representing only 9% of the overall through wall stress distribution considering the final flaw size, can be adequately represented with a third order fit as used in the analysis.

Page 7 of 10 Response to Request for Additional Information Note:

The initial flaw size depth is 0.1 inch, refer to Section 2.1 (AREVA Doc #32-9244389, of Reference 1); and the bounding final crack depth for a cylindrical flaw for the new repair weld is small, refer to Table 5-24, Path 4b (AREVA Doc #32-9237284, Attachment 2 of Reference 1).

NRC RAI-11:

The NRC staff notes that the request briefly discussed the general corrosion of exposed low alloy steel (LAS) in Section 5.6 of Attachment 1 to the relief requests and stated that the general corrosion rate is conservatively estimated to be 0.0036 inch/year. The corrosion of the exposed base metal has negligible impact on the RVCH and is acceptable for 40 years from the time the modification is installed.

a. Discuss in detail and provide justifications on why the galvanic corrosion, hydrogen embrittlement, stress corrosion cracking (SCC), and crevice corrosion are not expected to be a concern for the exposed RVCH LAS base metal;

b. Provide detailed discussions on how the licensee determined that the general corrosion rate will be 0.0036 inch/year, and why this estimation of general corrosion rate is conservative.

c. Provide detailed discussions on why the corrosion of the exposed base metal is acceptable.

EGC Response RAI-1 1 a:

A corrosion evaluation was performed of the exposed low alloy steel for each of the Byron and Braidwood Units and included as Attachment 2. This document evaluated the potential corrosion concerns of the components and weld that establish the vessel head penetration pressure boundary following the contingency IDTB weld repair. The associated impacts due to galvanic corrosion, hydrogen embrittlement, stress corrosion cracking, and crevice corrosion are discussed below:

Galvanic Corrosion:

The results of the NRC's boric acid corrosion test program have shown that the galvanic difference between ASTM A533 Grade B Steel, Alloy 600, and Type 308 stainless steel (used in reactor vessel cladding) is not significant enough to consider galvanic corrosion as a strong contributor to the overall boric acid corrosion process as stated in Section 5.3 of Attachment 2.

Thus, it was judged that galvanic corrosion between the exposed RVCH low alloy steel and Alloy 600, Alloy 690, or their weld metals is not a concern for this repaired configuration. This is supported by many cases of low alloy steel coupled to stainless steel or nickel-base alloys being exposed to primary coolant without known galvanic corrosion observed.

Hydrogen Embrittlement:

The hydrogen concentration in the RVCH will be greatest at the exposed low alloy steel surface and will decrease across the thickness of the RVCH to the trace concentration of hydrogen in the steel. Thus, the average hydrogen concentration in the RVCH is less than the maximum Page 8 of 10 Response to Request for Additional Information hydrogen concentrations summarized in Section 5.5 of Attachment 2. Thus, it was judged that hydrogen embrittlement is not a concern for the exposed RVCH low alloy steel in this repaired configuration. This is supported by many cases of low alloy steel being exposed to primary coolant without cracking attributed to hydrogen embrittlement observed.

Stress Corrosion Cracking (SCC):

Extensive PWR and BWR operating experience related to low alloy steel being exposed to reactor coolant has resulted in no known occurrences of SCC of low alloy steel reactor vessel material to any significant depth as stated in Section 5.4 of Attachment 2. Thus, it was judged that SCC of the exposed low alloy steel RVCH is not a concern for this repaired configuration.

Crevice Corrosion:

Crevice corrosion is not expected to be an issue for the CRIDM, RVLIS and spare nozzles due to the open geometry of the final repair configuration. The final repair configuration for the thermocouple column will create crevice conditions between the low alloy steel in the RVCH and portions of the new nozzle. However, crevice corrosion is not expected to be more severe than general corrosion for the low alloy steel. Furthermore, any corrosion deposits will likely plug the crevice path causing corrosion in the crevice to cease as stated in Section 5.2 of Attachment 2.

Thus, it was judged that crevice corrosion is not a concern for this repaired configuration.

EGC Response RAI-1 1 b:

The general corrosion rate was conservatively determined for the exposed low alloy steel of the repair corrosion evaluation and included as Attachment 2 with details provided below.

Based on the evaluation under Section 5.1.2 of Attachment 2, laboratory test data at low and high velocity was considered. The trend in the data indicated that the general corrosion rate (deareted, high temperature) increased with increasing flow velocity. The highest general corrosion rate from the test data (i.e., at the highest flow rate) was used in the evaluation. It was confirmed that the high flow velocity from the test data bounds the actual plant-specific flow velocity at the repair location. The corrosion rate for shutdown conditions (aerated, low temperature) from the test data was also considered. The overall general corrosion rate is based on a 90% capacity factor, which is estimated to be equivalent to 18 months of operation followed by 2 months shutdown, which resulted in an overall general corrosion rate of 0.0036 ipy.

EGC Response RAI-1 1 c:

The repair life assessment evaluation was performed for the exposed low alloy steel for each of the Byron and Braidwood Units and included as Attachment 3.

Per Section 3.1 of the repair life assessment evaluation, the analysis determined that the total surface (radial) corrosion in the penetration bore for 40 years would be small and acceptable.

Thus, this small amount of corrosion volume lost was determined to not have a significant impact.

Page 9 of 10 Response to Request for Additional Information

References:

1)

Letter from D. M. Gullott (Exelon Generation Company, LLC) to U. S. Nuclear Regulatory Commission, "Requests for Relief for Alternate requirements for Repair of Reactor Vessel Head Penetrations with Nozzles Having Pressure-Retaining Partial-Penetration J-Groove Welds," dated September 11, 2015

2)

Email from J. Wiebe (NRC) to J. Krejcie (Exelon Generation Company, LLC),

Preliminary Request for Additional Information Regarding the Braidwood and Byron Requests for Relief Regarding Repair of Reactor Vessel Head Penetration J-Groove Welds, dated January 6, 2016 Page 10 of 10

ATTACHMENT AREVA Inc., Affidavit for AREVA documents, February 10, 2016

AFFIDAVIT COMMONWEALTH OF VIRGINIA ss.

CITY OF LYNCHBURG

1.

My name is Morris Byram. I am Manager, Product Licensing, for AREVA Inc.

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

2.

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

3.

1 am familiar with the AREVA information contained in the following AREVA Engineering Information Records: 51-9233902-000, "PWSCC Evaluation for Contingency RVCH Nozzle Repairs at Byron Units 1 and 2 and Braidwood Units 1 and 2," dated June 24, 2015; 51-9234023-000, "Corrosion Evaluation of Byron Units I and 2 and Braidwood Units I and 2 IDTB Weld Repairs," dated May 12, 2015; 51-9240805-001, "Byron Units 1 & 2, and Braidwood Units 1 & 2 IDTB Reactor Vessel Head Penetration Nozzle Weld Repair-Life Assessment Summary," dated October 16, 2015; and 51-9245035-001, "Ambient IDTB Weld Interpass Temperature Evaluation," dated September 28, 2015, and referred to herein as "Documents." Information contained in these Documents has been classified by AREVA as proprietary, in accordance with the policies established by AREVA Inc. for the control and protection of proprietary and confidential information.

4.

These Documents contain information of a proprietary and confidential nature and are of the type customarily held in confidence by AREVA and not made available to the

public. Based on my experience, I am aware that other companies regard information of the kind contained in these Documents as proprietary and confidential.

5.

These Documents have been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in these Documents be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6.

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

(a)

The information reveals details of AREVA's research and development plans and programs or their results.

(b)

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

(c)

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

(d)

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

(e)

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

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

7.

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

8.

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

9.

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

SUBSCRIBED before me this to*

day of &&~

2016.

Sherry L. McFaden NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 10/31/18 Reg. # 7079129 F-MY6 m=slWn foxopirst Oct 31,2018 14PRRPMN