NMP1L3447, Constellation Energy Generation, LLC - Response to Request for Additional Information - Proposed Relief Request Associated with Reactor Pressure Vessel Water Level Instrumentation Partial Penetration Nozzle Repairs

Constellation Energy Generation, LLC - Response to Request for Additional Information - Proposed Relief Request Associated with Reactor Pressure Vessel Water Level Instrumentation Partial Penetration Nozzle Repairs

ML22033A134
Person / Time
Site:	Dresden, Peach Bottom, Nine Mile Point, Limerick, Clinton, Quad Cities, FitzPatrick, LaSalle
Issue date:	02/02/2022
From:	David Gudger Constellation Energy Generation
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
[[::JAF-22-0015|JAF-22-0015]], NMP1L3447, RS-22-016
Download: ML22033A134 (12)

Text

200 Exelon Way Kennett Square, PA 19348 www.exeloncorp.com 10 CFR 50.55a RS-22-016 [[::JAF-22-0015|JAF-22-0015]] NMP1L3447 February 2, 2022 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249 James A. FitzPatrick Nuclear Power Plant Renewed Facility Operating License No. DPR-59 NRC Docket No. 50-333 LaSalle County Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374 Limerick Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-39 and NPF-85 NRC Docket Nos. 50-352 and 50-353 Nine Mile Point Nuclear Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-63 and NPF-69 NRC Docket Nos. 50-220 and 50-410 Peach Bottom Atomic Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-44 and DPR-56 NRC Docket Nos. 50-277 and 50-278 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265

Subject:

Response to Request for Additional Information - Proposed Relief Request Associated with Reactor Pressure Vessel Water Level Instrumentation Partial Penetration Nozzle Repairs

Response to Request for Additional Information - Proposed Relief Request Associated with Reactor Pressure Vessel Water Level Instrumentation Partial Penetration Nozzle Repairs February 2, 2022 Page 2

References:

1) Letter from D. Gudger (Exelon Generation Company, LLC) to U.S.

Nuclear Regulatory Commission, Proposed Relief Request Associated with Reactor Pressure Vessel Water Level Instrumentation Partial Penetration Nozzle Repairs, dated August 12, 2021 (ML21224A123)

2) Email from B. Purnell (U.S. Nuclear Regulatory Commission) to T.

Loomis (Exelon Generation Company, LLC), Exelon Generation Company, LLC - Request for Additional Information Regarding Proposed Fleet Alternative for Repair of Water Level Instrumentation Partial Penetration Nozzles, dated January 13, 2022 (ML22020A064)

In the Reference 1 letter, Constellation Energy Generation, LLC (CEG) requested approval of a relief request associated with the repair of water level instrumentation (WLI) partial penetration nozzles on the Reactor Pressure Vessel (RPV). In the Reference 2 email, the U.S. Nuclear Regulatory Commission Staff requested additional information. Attached is our response.

There are no commitments contained in this response.

If you have any questions or require additional information, please contact Tom Loomis at 610-765-5510.

Respectfully, David T. Gudger Sr. Manager - Licensing & Regulatory Affairs Constellation Energy Generation, LLC Attachments: Response to Request for Additional Information

Response to Request for Additional Information - Proposed Relief Request Associated with Reactor Pressure Vessel Water Level Instrumentation Partial Penetration Nozzle Repairs February 2, 2022 Page 3 cc: Regional Administrator - NRC Region I Regional Administrator - NRC Region III NRC Senior Resident Inspector - Clinton Power Station NRC Senior Resident Inspector - Dresden Nuclear Power Station NRC Senior Resident Inspector - James A. FitzPatrick Nuclear Power Plant NRC Senior Resident Inspector - LaSalle County Station NRC Senior Resident Inspector - Limerick Generating Station NRC Senior Resident Inspector - Nine Mile Point Nuclear Station NRC Senior Resident Inspector - Peach Bottom Atomic Power Station NRC Senior Resident Inspector - Quad Cities Nuclear Power Station NRC Project Manager - Clinton Power Station NRC Project Manager - Dresden Nuclear Power Station NRC Project Manager - James A. FitzPatrick Nuclear Power Plant NRC Project Manager - LaSalle County Station NRC Project Manager - Limerick Generating Station NRC Project Manager - Nine Mile Point Nuclear Station NRC Project Manager - Peach Bottom Atomic Power Station NRC Project Manager - Quad Cities Nuclear Power Station W. DeHass, Pennsylvania Bureau of Radiation Protection A. L. Peterson, NYSERDA Illinois Emergency Management Agency - Division of Nuclear Safety

ATTACHMENT Response to Request for Additional Information

Response to Request for Additional Information Page 1 of 8 Request for Additional Information (RAI) 1:

Section 4 of the proposed alternative states, in part, that: The original partial penetration J-groove weld and a remnant of the original nozzle will remain in place. A flaw evaluation will demonstrate the acceptability of leaving the original partial penetration J-groove weld and remnant nozzle, with a maximum postulated flaw, in place for one cycle. Section 5.C of the proposed alternative states, in part, that: This evaluation will be used to demonstrate compliance with a combination of Subarticle IWB-3610 [of the ASME Code,Section XI,] and ASME Code Case N-749-x or similar code case , as applicable. The NRC staff understands that this evaluation will confirm qualitative information provided in this alternative request.

However, the application does not provide detailed information on how this flaw evaluation will be performed.

Discuss in detail how the confirmatory flaw evaluation will be performed. Specifically, discuss the scope, assumptions, input parameters, methodology, and acceptance criteria. For example, the scope of the flaw evaluation may include the components such as the existing J-groove weld and the new attachment welds. The input parameters may include flaw size, flaw locations, and crack growth rates. The methodology may include growth due to fatigue and stress corrosion cracking, weld residual stresses, the analytical model, and applied loadings.

Response

A confirmatory flaw evaluation will be performed similar to the evaluation approved for the Limerick Generating Station, Unit 2 (Safety Evaluation Report dated March 5, 2019 (ML19009A002)) and the one-cycle approval for Peach Bottom Atomic Power Station, Unit 2 (Safety Evaluation Report dated April 23, 2021 (ML21110A680)).

The purpose of this evaluation is to determine the suitability of leaving a degraded J-groove weld at the instrument nozzle on the reactor vessel following the repair of the leaking nozzle through the end of licensed plant operation. Since a potential flaw in the J-groove weld cannot be sized by currently available nondestructive examination techniques, it is conservatively assumed that the as-left condition of the remaining J-groove weld includes degraded or cracked weld material extending through the entire J-groove weld and Alloy 600 remnant nozzle material.

The scope of the flaw evaluation is to determine the suitability of leaving the existing J-groove weld in place following the repair. Many details related to assumptions, input parameters, methodology, and acceptance criteria will be proprietary; however, assumptions will be consistent with and similar to those in the proprietary analyses developed for the May 2018 Limerick Generating Station, Unit 2 repair (as referenced in Safety Evaluation Report dated March 5, 2019, ML19009A002), and in the November 2020 Peach Bottom Atomic Power Station, Unit 2 repair (as referenced in Safety Evaluation Report dated April 23, 2021 ML21110A680).

In general, design inputs would include the geometry of the original and repair instrument nozzles, the material properties of new and existing components, the fracture material properties, design and steady state operating conditions, and operating condition transients.

In general, the methodology will postulate a conservative initial flaw. A finite element crack model will be used to obtain stress intensity factors. Steady state and transient loads will then be applied to calculate crack growth. ASME Code Case N-749-x or similar code case as

Response to Request for Additional Information Page 2 of 8 modified by the Nuclear Regulatory Commission may then be used to determine the appropriate method of fracture toughness analysis.

The acceptance criteria will be based on the applicable ASME Section XI Code Edition and, if necessary, ASME Section XI Code Case N-749-x or similar code case as modified by the Nuclear Regulatory Commission. Acceptance of the postulated flaw will be based on available fracture toughness or ductile tearing resistance using the safety factors consistent with those used in the proprietary analyses developed for the Limerick Generating Station, Unit 2 May 2018 repair (as referenced in Safety Evaluation Report dated March 5, 2019, ML19009A002),

and in the November 2020 Peach Bottom Atomic Power Station, Unit 2 November 2020 repair (as referenced in Safety Evaluation Report dated April 23, 2021 ML21110A680).

As discussed in the relief request, the final analysis will be submitted to the U.S. Nuclear Regulatory Commission (NRC).

RAI-2

Section 5.E of the proposed alternative states, in part, that: A corrosion evaluation will be performed to consider potential material degradation due to the repair of the RPV WLI partial penetration nozzle. The repair will result in the RPV [low alloy steel] being exposed to the reactor coolant. However, the application does not provide detailed information on how this corrosion evaluation will be performed.

Describe in detail how the corrosion evaluation will be performed. Specifically, discuss the scope, assumptions, input parameters, methodology, and acceptance criteria.

Response

A corrosion evaluation will be performed similar to the evaluation approved for the Limerick Generating Station, Unit 2 (Safety Evaluation Report dated March 5, 2019 (ML19009A002))

and the one-cycle approval for Peach Bottom Atomic Power Station, Unit 2 (Safety Evaluation Report dated April 23, 2021 (ML21110A680)).

The repair of the reactor vessel nozzle will change the penetration configuration in the following ways: 1) the repair exposes the low alloy steel (LAS) reactor vessel to water conditions, 2) the repair includes a new Alloy 690 nozzle as part of the pressure boundary, and 3) the repair includes a new Alloy 52M weld pad and partial penetration J-groove weld as part of the pressure boundary. Also, the pipe coupling and weld to the nozzle may be dissimilar metals.

The corrosion evaluation considers potential material degradation due to each of these changes.

The scope of the corrosion evaluation is to evaluate potential corrosion material degradation due to each of these changes. Many details related to assumptions, input parameters, methodology, and acceptance criteria will be proprietary. However, assumptions will be consistent with and similar to those in the proprietary analyses developed for the May 2018 Limerick Generating Station, Unit 2 repair (as referenced in Safety Evaluation Report dated March 5, 2019, ML19009A002), and in the November 2020 Peach Bottom Atomic Power Station, Unit 2 repair (as referenced in Safety Evaluation Report dated April 23, 2021 ML21110A680). The corrosion mechanisms considered with be the general corrosion, galvanic corrosion, crevice corrosion, and stress corrosion cracking. Assumed corrosion rates will be conservative and based on established industry experience and testing.

Response to Request for Additional Information Page 3 of 8 Design inputs will include the geometry of the repair, the material properties of new and existing components, the water chemistry of wetted components, galvanic interactions, and operating conditions.

The evaluation methodology will assess the potential material degradation due to postulated corrosion mechanisms.

Acceptance for potential material loss due the postulated corrosion mechanisms will be based on ASME Section XI and Section III analysis As discussed in the relief request, the final analysis will be submitted to the NRC.

RAI-3

Section 5.F of the proposed alternative states that a lost parts evaluation will be performed to assess the potential for nozzle segments to enter the RPV during power operation. However, the application does not provide detailed information on how this evaluation will be performed.

Describe in detail how the lost parts evaluation will be performed. Specifically, discuss the scope, assumptions, input parameters, methodology, and acceptance criteria. Confirm that this evaluation would also assess the potential for the original J-groove weld material to enter the RPV during power operation.

Response

A loose parts evaluation will be performed consistent with existing loose part evaluation procedures to assess the potential for nozzle segments to enter the Reactor Pressure Vessel (RPV) during power operations. This evaluation will assess the potential impact and safety concerns on the fuel and internal RPV components. This evaluation will also consider interfacing systems, flow blockage, and adverse chemical reactions.

This evaluation will assess the potential and impact for the original J-groove weld material and remnant nozzle to enter the RPV during power operation.

RAI-4

Section 7 of the proposed alternative discusses precedents for the repair of reactor vessel head penetration and J-groove welds. In the half-nozzle repair of a reactor vessel head penetration nozzle for a control rod drive mechanism, a corner location (called a triple point) exists where the Alloy 52M weld metal is in contact with low alloy steel of the RPV closure head and the Alloy 600 of the original nozzle or Alloy 690 of the replacement nozzle. In the triple point location, a welding anomaly (e.g., lack of fusion) could exist. For similar precedents, licensees have performed a flaw evaluation to demonstrate that the anomaly at the triple point location will not affect the structural integrity of the repair. The proposed alternative does not discuss whether a triple point location will exist after the repair.

Discuss whether a triple point location will exist in the repair of a WLI nozzle. If a triple point exists, confirm that the flaw evaluation described in Section 5.C of the proposed alternative will demonstrate that the welding anomaly does not affect structural integrity of the WLI nozzle for one cycle.

Response to Request for Additional Information Page 4 of 8

Response

The proposed repair design joins two base materials with a J-groove weld to eliminate the triple point as seen in previous half-nozzle repair designs on Control Rod Drive Mechanism (CRDM) nozzle replacements. In the case of the configuration below, materials and dimensions may be different depending on the specific design. The materials shown are the Alloy 52M weld pad and the Alloy 690 nozzle. The J-groove weld is attaching the installed weld pad to the replacement nozzle. There is no triple point joining three separate base materials together in this design.

RAI-5

The configuration of a typical repaired WLI nozzle, a discussion of the repair process, and the specific nondestructive examinations that will be performed (on the pad or the replacement J-groove weld) are not described in the proposed alternative.

A. Provide a diagram of a typical installed replacement WLI nozzle, including the new weld(s),

original J-groove weld, the remnant nozzle, and the replacement nozzle. The installed new WLI nozzle diagram should show how the new nozzle is attached to the new weld pad (i.e.,

whether the new nozzle is inserted into the bore of the RPV wall). If a portion of the new nozzle is inserted into the bore of the RPV wall, discuss whether a weld is installed at the junction of the new nozzle and the bore of the RPV wall.

B. Provide a diagram showing the examination coverage of the installed replacement WLI nozzle, the new weld pad, and the new partial penetration weld that attached the new nozzle to the new weld pad. Discuss the nondestructive examination and associated acceptance criteria for the new welds and the replacement WLI nozzle.

C. Discuss and/or list the step-by-step repair process.

Response

A. Enclosure 1 provides a diagram of a typical installed replacement WLI nozzle. For this repair design, there is no weld installed at the junction of the new nozzle and bore of the RPV wall. Enclosure 1 shows a temporary plug that may be used during the repair but will be removed prior to completion of the repair. This diagram shows a reducing coupling; however, depending on the plant specific piping configurations, different couplings will be utilized.

B. Enclosure 2 provides a typical diagram. The specific NDE is dependent on the final design of the installed weld pad and J-groove weld. NDE is discussed in part C. The acceptance criteria for the weld pad is in accordance with the applicable ASME Code Case utilized, ASME Section XI, the construction code or ASME Section III. The acceptance criteria for the J-groove weld is in accordance with the applicable editions of ASME Section XI and the construction code or ASME Section III. The NDE exams and coverage are in accordance with the ASME Code Case utilized, ASME Section III, the construction code, and ASME Section V.

C. A summary of the repair and NDE performed are discussed as follow:

a. Install foreign material exclusion sealing plug, detach piping near coupling, cut the existing Alloy 600 nozzle outboard of the RPV, grind the nozzle flush with the RPV

Response to Request for Additional Information Page 5 of 8 shell OD surface, and attach the capacitor discharge studs (welding and boring tools, if necessary) to RPV.

b. Then, perform surface and volumetric examinations of the RPV shell OD surface in preparation for installing Alloy 52M weld pad.

c. Install a weld dam to accommodate for depositing weld pad, deposit the Alloy 52M weld pad in accordance with ASME Code Case N-638-x, N-839-x, or similar code case, as approved or conditionally approved by the NRC in the latest revision of Regulatory Guide 1.147, and in accordance with the construction code, perform post weld grinding of the weld pad, and conduct dimensional inspection of weld pad.

Then, perform surface and ultrasonic examinations of the weld pad upon completion of a 48-hour hold time as applicable.

d. Remove weld dam, perform final machining of the weld pad bore, perform a dimensional measurement of the final bore. Then, perform surface examination of the final bore.

e. Machine replacement Alloy 690 nozzle. Then, perform visual and surface examinations of the replacement Alloy 690 nozzle.

f. Weld new coupling to nozzle. Then, perform visual and surface examinations of the coupling-to-nozzle weld.

g. Machine J-groove bevel in the weld pad. Then, perform visual and surface examinations of the J-groove bevel.

h. Perform installation and welding of the replacement Alloy 690 nozzle. Then, perform a progressive surface examination of J-groove weld joining the replacement Alloy 690 nozzle to the weld pad.

i. Remove capacitor discharge studs attached to RPV, if installed. Then, perform surface examination of RPV at the capacitor discharge stud attachment locations.

j. Attach piping and remove foreign material exclusion sealing plug.

RAI-6

The proposed alternative does not discuss whether an extent of condition evaluation will be performed prior to the repair of a WLI nozzle.

Discuss the extent of condition evaluation that will be performed on the reactor vessels remaining WLI nozzles before a WLI nozzle is repaired using the proposed alternative.

Response

An extent of condition will be performed. Accessible nozzles with the same design as the repaired nozzle will receive a bare metal VT-2 performed at the normal operating pressure. This exam will look for evidence of through-wall leakage, degradation due to corrosion of a pressure retaining boundary and evidence of pressure/flow loss or flow impairment.

Response to Request for Additional Information Page 6 of 8 Errata:

In addition, the Reactor Pressure Vessel Code of Construction for James A. FitzPatrick Nuclear Power Plant has been corrected to be Section III 1965 Edition, through Winter 1966 Addenda from Section III 1967 Edition, through Winter 1969 Addenda. The updated table is shown below:

ASME SECTION ASME CONSTRUCTION PLANT INTERVAL START END XI EDITION CODE (RPV)

Section III 1971 Edition, Clinton Power Station, Fourth 2013 Edition July 1, 2020 June 30, 2030 through Summer 1973 Unit 1 Addenda Dresden Nuclear 2007 Edition, January 20, Power Station, Units 2 Fifth through 2008 January 19, 2023 2013 Section III 1963 Edition, and 3 Addenda through Summer 1964 Dresden Nuclear Addenda January 20, Power Station, Units 2 Sixth 2017 Edition January 19, 2033 2023 and 3 2007 Edition, Section III 1965 Edition, James A. FitzPatrick Fifth through 2008 August 1, 2017 June 15, 2027 through Winter 1966 Nuclear Power Plant Addenda Addenda Section III 1968 Edition, 2007 Edition, LaSalle County September 30, through Summer 1970 Fourth through 2008 October 1, 2017 Stations, Units 1 and 2 2027 Addenda, except Addenda Paragraph N-355 Section III 1968 Edition, 2007 Edition, through Summer 1969 Limerick Generating February 1, Fourth through 2008 January 31, 2027 Addenda, except that Station, Units 1 and 2 2017 Addenda Article 4 of the Winter 1969 Addenda applies Nine Mile Point Nuclear Fifth 2013 Edition August 23, 2019 August 22, 2029 Section I 1962 Edition Station, Unit 1 Section III 1971 Edition, Nine Mile Point Nuclear Fourth 2013 Edition October 6, 2018 August 22, 2028 through Winter 1972 Station, Unit 2 Addenda Peach Bottom Atomic Section III 1965 Edition, December 31, Power Station, Units 2 Fifth 2013 Edition January 1, 2019 through Winter 1965 2028 and 3 Addenda Quad Cities Nuclear 2007 Edition, Power Station, Units 1 Fifth through 2008 April 2, 2013 April 1, 2023 and 2 Addenda Section III 1965 Edition, through Summer 1965 Quad Cities Nuclear Addenda Power Station, Units 1 Sixth 2017 Edition April 2, 2023 April 1, 2033 and 2

Response to Request for Additional Information Page 7 of 8 Enclosure 1 - Typical Reactor Pressure Vessel Level Instrument Nozzle Repair Temporary Plug Remnant Nozzle RPV Shell Gap between nozzle remnant and new nozzle Alloy 52M weld pad Alloy 52M J-groove weld Alloy 690 nozzle Fillet weld Reducing coupling

Response to Request for Additional Information Page 8 of 8 - NDE Examination Example