RA-21-0132, Application for Subsequent Renewed Operating Licenses. Part 7 of 8

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Application for Subsequent Renewed Operating Licenses. Part 7 of 8
ML21158A200
Person / Time
Site: Oconee  Duke energy icon.png
Issue date: 06/07/2021
From:
Duke Energy Carolinas
To:
Office of Nuclear Reactor Regulation
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References
RA-21-0132
Download: ML21158A200 (417)


Text

Oconee Nuclear Station Units 1, 2, and 3 Application for Subsequent License Renewal Appendix E - Applicants Environmental Report Attachment F: Other Consultation Letters

41 O S. Wilmington St

( -, DUKE NCRH15

<(; ENERGY Raleigh , NC 27601 o* 919.546.5285 Nov. 4, 2020 Fran Marshall Environmental Affairs Administrator SCDHEC 2600 Bull Street Columbia, SC 29201

Subject:

Duke Energy - Oconee Nuclear Station Units 1, 2 & 3 Subsequent License Renewal - Thermophilic Organisms

Dear Ms. Marshall:

As per your recommendation regarding DHEC's assessment of thermophilic organisms for Oconee Nuclear Station, Duke Energy had Charles W. Calmbacher, Certified Industrial Hygienist, review the information and offer his professional opinion on this matter. This summary, and Mr. Calmbacher's findings, are provided in the attached letter. We request that DHEC consider this information in support of your review and response to our request concerning any potential public health concerns associated with thermophilic organisms. The original request dated Nov 11, 2019 for Oconee Nuclear Station is also attached.

If you have any questions, please feel free to contact me at 919-793-4220 or arun .kapur@duke-energy.com .

Sincerely, Arun Kapur Lead Environmental Specialist Attachments Charles W. Calmbacher Letter Thermophilic Letter to DHEC cc: John Estridge, ONS BUILDING ASMARTER ENERGY FUTURE"'

.::I ENERCON Excel/ence- Eve,y p1ojec1. Every doy.

October 21 , 2020 Mr. Arun Kapur Duke Energy 410 S. Wilmington St, NCRH 15 Raleigh, NC 27601 RE: Impact of Thermophilic Organisms - Oconee Nuclear Station

Dear Mr. Kapur:

An updated review of the potential impact of Naegleria fowleri on Lake Keowee, and associate waterways, was performed relative to the renewal application for the Duke Energy Oconee Nuclear Power Station. A review performed during the previous licensing cycle for the plant by the State of South Carolina (Dr. Brown) concluded that the potential for increased exposure to N.

fowleri from ONS' s thermal discharge is not a concern. The conditions leading to that conclusion have not changed, and are not proposed to change, since that determination. The cooling water is discharged at temperatures below levels that promote growth of the organism. Cooling towers are not used to cool discharge water. This reduces the potential for growth of the ameba before discharge. The deep discharge of cooling water in the lake keeps water temperature below optimal levels for propagation of the ameba.

In performing this updated review, I utilized the following information and documents:

  • Hains, Dr. John, Practical Limnology, " What Lurks in that Water?", The Sentinel, September-October 2016. Pp. 10-11.
  • Application for Renewed Operating Licenses Oconee Nuclear Station, Units 1, 2, and 3 Volume IV, June 1998. pp. 4-37-38 & Attachment I.

Based on current permit requirements, NPDES permits, discharge location, and discharge parameters, it is reasonable to assume that the potential for increasing the exposure to N. fowleri to the general public is negligible. I see no impediment to the license renewal, or continued operation, of the Oconee Nuclear Power Station based on the potential of impacting public health and safety relative to increased potential for the growth ofthermophilic microorganisms. Infection by N. fowleri is considered "rare" by the CDC. There is little reason to believe that the actions of the ONS should promote or increase the potential for elevated exposure or risk to the public at large.

Respectfully submitted, Charles W. Calmbacher, Ph.D., CIH Certified Industrial Hygienist cc: Ed Asbury Tom Slavonic 2790 Mosside Boulevard, Suite 70S Monroeville, PA 15146 Phone: 724.213.7702 enercon.com

el_~ DUKE J. Ed Burchfield, Jr.

~ ENERGY. V.ICB Pres/dBnt Occnee Nuclear Station Dub Energy ON01VP / 7800 Roch8ster Hwy Seneca, SC 29672

~ 864.873.3478

, 864.873.5791 Ed.Burchfield OdukHlle,py.com BY U.S. MAIL - RETURN RECEIPT REQUESTED November 11, 2019 Fran Marshall Environmental Affairs Administration SCDHEC 2600 Bull Street Columbia, SC 29201 RE: Duke Energy - Oconee Nuclear Station Units 1, 2, and 3 Subsequent License Renewal

Dear Ms. Marshall:

Duke Energy is seeking a response from DHEC concerning the potential existence and Perceived public health risks associated with thermophilic organisms that may be present in the portion of Lake Keowee that receives the cooling water discharge from our Oconee Nuclear Station (ONS). Information concerning the reason for this request and specific microorganisms of concern is presented below. Figures depicting the station site and the vicinity within a 6-mile radius of the station are attached.

Reason for this Request and Microorganisms of Concern Duke Energy Carolinas, LLC (Duke Energy) is preparing an application for renewing the operating licenses for Oconee Nuclear Station Units 1, 2, and 3 (ONS) for an additional 20 years (see Table 1). Duke Energy is contacting you for assistance in assessing the impacts from continued operation during this renewed license period.

Table 1. ONS Licensing Dates License Expiration Extended License Expiration ONS Unit Date Date Unit 1 February 6, 2033 February 6, 2053 Unit2 October 6, 2033 October 6, 2053 Unit3 July 19, 2034 July 19, 2054 As part of the renewal process, the U.S. Nuclear Regulatory Commission (NRC) requires that the license renewal application include an environmental report (ER) that assesses the impacts from continued operation and any refurbishment undertaken to enable the continued operation

of the units. One area of potential environmental impact concerns potential public health risks associated with microorganisms.

Information to Support Consultation on Thermophlllc Microorganisms In Regulatory Guide 1437, Supplement 1, Generic Environmental Impact Statement for License Renewal of Nuclear Plants (GEIS), the NRC considered health impacts from thermophilic organisms posed to both the public and plant workers because ideal conditions for thennophilic bacteria can result from nuclear facility operations and discharges. The NRC designated public health impacts resulting from thermophilic organisms as a Category 2 issue requiring plant-specific analysis. Information to be considered in evaluating impacts includes thermal discharge temperature; thermal characteristics of the receiving water bodies; thermal conditions tor the enhancement of Naegleria fowleri and other pathogens; and potential impacts to public health.

The GEIS discussion of microbiological hazards focuses on the thermophilic microorganisms Legionella spp. (which can be a hazard in cooling towers) and the pathogenic amoeba, N. fowleri (which can be a hazard resulting from cooling water discharges). ONS's cooling system does not use cooling towers but does have a thermal discharge to publicly accessible water.

Naegleria spp. is ubiquitous in nature and thrives in heated water bodies at temperatures ranging from 95-106°F or higher is rarely found in water cooler than 95°F, and infection rarely occurs in water temperatures of 95°F or less (NRC 2013, Section 3.9.3). SCDHEC, South Carolina's state public health agency, characterized the risk of infection from N. fowleri 's as very rare, but warns that "recreational water users should assume that N. fowler! is present in warm freshwater across the United States and be aware that there is always a low-level risk of infection." There have been only eight cases of Primary Amebic Meningoencephalitis, the infection caused by N. fowleri, in South Carolina from 1962 to 2018.

ONS utilizes an open-cycle cooling system in which cooling water is withdrawn from Lake Keowee from its intake channel on the south side of the ONS plant, heated in the condensers, and returned to Lake Keowee through the discharge point on the northeast side of the ONS plant. ONS discharges heated cooling water at a depth of approximately 20 feet. The lake waters near the discharge area are open to the public. Activities in the area include recreational boating, fishing, and scuba diving. Lake Keowee has residential housing and public swimming areas as well.

The current NPDES permit for ONS establishes both a maximum allowable discharge temperature, and a limit for increases of water temperature between the intake and discharge.

The maximum discharge temperature is100°F as a daily average, unless critical hydrological, meteorological, and electrical demand conditions apply. In such situations, the discharge temperature shall not be allowed to exceed 103°F. The maximum temperature rise above the intake temperature is limited to 22°F when the intake temperature Is greater than 68°F. In the 2013 permit renewal application, Duke Energy requested the daily maximum value of 100°F to a 7-day average not to exceed 100°F.

As part of a CWA Section 316(a) demonstration monitoring, Duke Energy monitors water temperatures at several Lake Keowee stations. The most recent report submitted to SCOH EC is 2 of 6 I P a t'. 1'

from 2013 and covers the years 2006 - 2011. The closest station to the plant's discharge is Location 508, which is approximately 200 meters from the discharge. The annual maximum measured surface temperatures at Location 508 in the years 2006 to 2011 ranged from 92.5 °F in 2009 to 94.8°F in 2008 . The annual maximum temperatures were similar to values reported in 1995 and 2007 reports. The report also noted that no exceedances of permit thermal limits occurred over the 2006 - 2011 period.

As noted above, N. fowleri is rarely found in water cooler than 95°F, and infection rarely occurs in water temperatures of 95°F or less. While the immediate discharge area could have temperatures in the summer above 95°F, the maximum temperatures recorded 200 meters from the discharge were below 95°F, indicating lower risk. In addition, the discharge point and this monitoring point is located in an area of deep water, approximately 23 meters. The N. fowleri infection risk is higher in shallow, warm water.

The Friends of Lake Keowee published an article in their newsletter in 2016 from Dr. J. Hains of Clemson University that addressed the risk posed by ONS's heated discharge for promoting N.

fowleri. Dr. Hains wrote, "The temperature at which this organism grows best is reported to be far greater (approximately 112°F) than the ONS discharge. Moreover, that water originates in the coldest, deepest depths of Lake Keowee, not an optimal habitat. To my knowledge there have been no studies of the distribution of this organism in Lake Keowee (or in other nearby lakes in recent times)."

Additionally, for the first license renewal of ONS, Duke Energy consulted with SCDHEC to determine if the continued operation of Oconee will have public health impacts due to the enhancement of thermophilic organisms. By letter dated October 25, 1996, Dr. John F. Brown, State Toxicologist at SCDHEC, summarized the agency's position and opinion regarding the public health implications of continued operation of Oconee. Regarding the potential public health hazard from pathogenic microorganisms whose abundance might be promoted by ONS's artificial warming of recreational waters, Dr. Brown indicated that there seems to be no significant threat to off-site persons near such heated recreational waters.

As stated earlier, this letter seeks your input on any potential public health concerns associated with our proposed continued operation of ONS. We appreciate your notifying us of your comments and any information you believe Duke Energy should consider in the preparation of the ER. Duke Energy plans to include this letter and any response you provide in the ER.

3 of 6 IPa g c

Should you or your staff have any questions or comments, please contact Mike Ruhe at (980) 373-3231 / Mike.Ruhe@duke-energy.com.

Sincerely,

>eifJ-¢111 J. Ed Burchfield, Jr Site Vice President Oconee Nuclear Station Attachments:

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cc: Myra Reece SCDHEC 2600 Bull Street Columbia, SC 29201

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January 8, 2021 Mr. Arun Kapur Duke Energy Lead Environmental Specialist 410 S. Wilmington Street NCRH 15 Raleigh, North Carolina 27601 Re: Oconee Nuclear Station Units 1,2 and 3 Subsequent License Renewal - Consultation on Thermophilic Microorganisms

Dear Mr. Kapur:

After a review of the report on the impact of thermophilic organisms associated with discharge from the Oconee Nuclear Station (ONS) prepared by Enercon, DHEC technical staff does not take any exception with the findings, including that "there is little reason to believe that the actions of the ONS should promote or increase the potential for elevated exposure or risk to the public at large."

As stated in the Duke-Energy letter dated November 11 , 2019, Naegleria spp. is ubiquitous in nature and thrives in water bodies at temperatures ranging from 95 to 106 degrees F. SCDHEC routinely warns recreational water users across the state that there is always a low risk of infection from naturally occurring organisms like N. fowleri in warm freshwater.

In our view, the Enercon consultant has met the NRC requirement of plant-specific analysis of this issue. The discharge temperature, location and demonstration monitoring all support the conclusion that an increased risk to the public to exposure to thermophilic microorganisms is not expected.

We agree with your consultant that the conditions that led the former DHEC State Toxicologist who evaluated this issue and memorialized his findings in the October 1996 letter have not substantially changed and are not expected to change. As Dr. Brown referenced, there is no ongoing monitoring program for these microorganisms in this location largely because no increased risk has been identified. That remains true today.

For all of the reasons listed in the Enercon letter/report, we concur that there is "no impediment to the license renewal or continued operation of the ONS based on the potential of impacting public health and safety relative to increased potential for growth of thermophilic microorganisms."

S.C Department of Hea lth and Environmen tal Control 2600 Bull Stree t Co lumbia. SC 29201 (803) 898 3432 www scdhec gov

If you have questions or need any additional information, please don't hesitate to contact me by phone (803.422.1805) or e-mail (marshaf2@dhec.sc.gov) .

Thank you, s/Fran W. Marshall Fran W. Marshall, JD, MSPH Director of the Office of Applied Science cc: Mike Ruhe, Duke Energy, Director, Environmental Policy and Affairs Rounette Nader, Duke Energy, Director Nuclear Engineering Myra C. Reece, DHEC Director of Environmental Affairs Bryan Rabon, DHEC Bureau of Water, Manager Aquatic Science Programs

(_~ DUKE J. Ed Burdlfleld, J r.

ENERGY. YIC8 President Oconee Nuclesr Station Duke Energy ONOIVP / 7800 Rochester Hwy Seneca, SC 29672

- 864.873.3478

' 864.873.5791 Ed.Bufthfteld Oduks-ene,gy com BY U.S. MAIL RETURN RECEIPT REQUESTED November 11, 2019 Ms. Kimberly Bose, Secretary Federal Energy Regulatory Commission 888 First Street N.E.

Washington, DC 20426 RE: Duke Energy - Oconee Nuclear Station Units 1, 2, and 3 Subsequent License Renewal

Dear Ms. Bose:

Duke Energy Carolinas, LLC (Duke Energy) is seeking to renew the operating license for Oconee Nuclear Station Units 1, 2, and 3 (ONS) for an additional 20 years (see Table 1). Duke Energy and ONS have safely and reliably provided electricity to our Carolinas customers for decades. ONS has generated clean and cost-effective power, provided thousands of well-paying jobs, and produced substantial economic benefits for the Carolinas. Renewing the licenses of ONS is important for our customers, communities and environment.

Table 1. ONS Licensing Dates Extended License Expiration ONS Unit License Expiration Date Date Unit 1 February 6, 2033 February 6, 2053 Unit2 October 6, 2033 October 6, 2053 Unit 3 July 19, 2034 July 19, 2054 Duke Energy's nuclear fleet plays an important role in the company's efforts to lower carbon emissions. In 2018, the Duke Energy nuclear fleet generated more than 72 billion kilowatt-hours of electricity and avoided the release of about 54 million tons of carbon dioxide- equivalent to keeping more than 1o million passenger cars off the road. The company has set aggressive carbon reduction goals of at least 50% by 2030 and net-zero by 2050 and keeping its nuclear fleet operating is key to achieving these goals.

Renewing the nuclear licenses will provide significant value to Duke Energy customers, as well as continue to support Carolinas communities through jobs, tax revenues and partnerships.

Duke Energy employs about 5,000 workers in its nuclear group, with additional contract workers supporting refueling outages and project work. In 2018, the Duke Energy also paid more than

$300 million in property and payroll taxes associated with the nuclear stations, benefiting local governments and school districts. In addition, nuclear employees support the communities where they live and work by donating time and funds through sponsorships and volunteer activities.

The ONS site is situated on 510 acres in eastern Oconee County, South Carolina (SC),

approximately eight miles northeast of Seneca, SC, on the southern shore of Lake Keowee.

During the license renewal term, Duke Energy proposes to continue operating the units as currently operated. There are currently no ground-disturbing activities or refurbishment anticipated at the ONS site during the subsequent license renewal period. Figures depicting the station site and the vicinity within a 6-mile radius of the station are enclosed.

Should you or your team have any questions or comments about ONS or the license renewal process, please contact Alan Stuart at (980) 373-2079 / Alan.Stuart@duke-energy.com.

Sincerely, J Cf ILJff)I/

J. Ed Burchfield, Jr Site Vice President Oconee Nuclear Station Attachments:

Figure 1. ONS Site Figure 2. ONS 6-mile Vicinity 2 of 4 I f' .i ~. ,*

Figure 1. ONS Site Legend c::IONSSite


====iMiles 0 0.25 0.5 3 of 41 I' :1 g L'

Figure 2. ONS 6-mile Vicinity Legend a Airport . . Surface Water D Heliport 0NSS1te U.S. Route i_-_i 8-Mile Radius

- - State Highway Place Local Road c:Jeounty ---==:::::,M~s 0 2

- Railroad 4of4j Pa~l'

(_~ DUKE J. Ed Burchfield, Jr.

ENERGY. \r,ce President Oconee Nuclear Station Duke Energy ONOI VP f 7800 Rochester Hwy Seneca, SC 29672 864.873.3478 f 864.873.5791 Ed.Bu11:hfie/d Oduke-ene,gy.oom evu.s. MAIL RETURN RECEIPT REQUESTED November 11, 2019 Doug Hof, President Friends of Lake Keowee Society 4065 Keowee School Rd Seneca, SC 29672 RE: Duke Energy - Oconee Nuclear Station Units 1, 2, and 3 Subsequent License Renewal

Dear Mr. Hof:

Duke Energy Carolinas, LLC (Duke Energy) is seeking to renew the operating license for Oconee Nuclear Station Units 1, 2, and 3 (ONS) for an additional 20 years (see Table 1). Duke Energy and ONS have safely and reliably provided electricity to our Carolinas customers for decades. ONS has generated clean and cost-effective power, provided thousands of well-paying jobs, and produced substantial economic benefits for the Carolinas. Renewing the licenses of ONS is important for our customers, communities and environment.

Table 1. ONS Licensing Dates Extended License Expiration ONS Unit License Expiration Date Date Unit 1 February 6, 2033 February 6, 2053 Unit2 October 6, 2033 October 6, 2053 Unit3 July 19, 2034 July 19, 2054 Duke Energy's nuclear fleet plays an important role in the company's efforts to lower carbon emissions. In 2018, the Duke Energy nuclear fleet generated more than 72 billion kilowatt-hours of electricity and avoided the release of about 54 million tons of carbon dioxide - equivalent to keeping more than 1o million passenger cars off the road. The company has set aggressive carbon reduction goals of at least 50% by 2030 and net-zero by 2050 and keeping its nuclear fleet operating is key to achieving these goals.

Renewing the nuclear licenses will provide significant value to Duke Energy customers, as well as continue to support Carolinas communities through jobs, tax revenues and partnerships.

Duke Energy employs about 5,000 workers in its nuclear group, with additional contract workers supporting refueling outages and project work. In 2018, the Duke Energy also paid more than

$300 million in property and payroll taxes associated with the nuclear stations, benefiting local governments and school districts. In addition, nuclear employees support the communities where they live and work by donating time and funds through sponsorships and volunteer activities.

The ONS site is situated on 510 acres in eastern Oconee County, South Carolina (SC),

approximately eight miles northeast of Seneca, SC, on the southern shore of Lake Keowee.

During the license renewal term, Duke Energy proposes to continue operating the units as currently operated. There are currently no ground-disturbing activities or refurbishment anticipated at the ONS site during the subsequent license renewal period. Figures depicting the station site and the vicinity within a 6-mile radius of the station are enclosed.

Should you or your team have any questions or comments about ONS or the license renewal process, please contact Mlkayla Kreuzberger at (864) 873-4204 / Mikayla.Kreuzberger@duke-energy.com or Alan Stuart at (980) 373-2079 / Alan.Stuart@duke-energy.com.

Sincerely,

)Cl~

J. Ed Burchfield, Jr Site Vice President Oconee Nuclear Station Attachments:

Figure 1. ONS Site Figure 2. ONS 6-mile Vicinity

o~ - - ~====iMles 0.25 0.5 3 Of 41 fl ,l g t'

Figure 2. ONS 6-mile Vicinity Legend a Airport . . Surface Water D Heliport 0NSS1te U.S. Route i_-_i 8-Mile Radius

- - State Highway Place Local Road c:Jeounty ---==:::::,M~s 0 2

- Railroad 4of4j Pa~l'

(_~ DUKE J. Ed Burchfield, Jr.

ENERGY. \rice President Oconee Nuclsar Station Duke Energy ONO 1VP / 78()() Rochester Hwy Seneca, SC 29672

" 864.873.3478

864.873.5791 Ed.Burchfield @dul<e-ene,r,y.com BYU.S. MAIL RETURN RECEIPT REQUESTED November 11, 2019 Rob Auhlebach Advocates for Quality Development P.O. Box 802 Seneca, SC 29679 RE
Duke Energy - Oconee Nuclear Station Units 1, 2, and 3 Subsequent Ucense Renewal

Dear Mr. Auhlebach:

Duke Energy Carolinas, LLC (Duke Energy) is seeking to renew the operating license for Oconee Nuclear Station Units 1, 2, and 3 (ONS) for an additional 20 years (see Table 1). Duke Energy and ONS have safely and reliably provided electricity to our Carolinas customers for decades. ONS has generated clean and cost-effective power, provided thousands of well-paying jobs, and produced substantial economic benefits for the Carolinas. Renewing the licenses of ONS is important for our customers, communities and environment.

Table 1. ONS Licensing Dates Extended License Expiration ONS Unit License Expiration Date Date Unit 1 February 6, 2033 February 6, 2053 Unit 2 October 6, 2033 October 6, 2053 Unit3 July 19, 2034 July 19, 2054 Duke Energy's nuclear fleet plays an Important role in the company's efforts to lower carbon emissions. In 2018, the Duke Energy nuclear fleet generated more than 72 billion kilowatt-hours of electricity and avoided the release of about 54 million tons of carbon dioxide - equivalent to keeping more than 10 million passenger cars off the road. The company has set aggressive carbon reduction goals of at least 50% by 2030 and net-zero by 2050 and keeping its nuclear fleet operating is key to achieving these goals.

Renewing the nuclear licenses will provide significant value to Duke Energy customers, as well as continue to support Carolinas communities through jobs, tax revenues and partnerships.

Duke Energy employs about 5,000 workers in its nuclear group, with additional contract workers supporting refueling outages and project work. In 2018, the Duke Energy also paid more than

$300 million in property and payroll taxes associated with the nuclear stations, benefiting local governments and school districts. In addition, nuclear employees support the communities where they live and work by donating time and funds through sponsorships and volunteer activities.

The ONS site is situated on 510 acres in eastern Oconee County, South Carolina (SC),

approximately eight miles northeast of Seneca, SC, on the southern shore of Lake Keowee.

During the license renewal term, Duke Energy proposes to continue operating the units as currently operated. There are currently no ground-disturbing activities or refurbishment anticipated at the ONS site during the subsequent license renewal period. Figures depicting the station site and the vicinity within a 6-mile radius of the station are enclosed.

Should you or your team have any questions or comments about ONS or the license renewal process, please contact Mikayla Kreuzberger at (864) 873-4204 / Mikayla.Kreuzberger@duke-energy.com or Alan Stuart at (980) 373-2079 / Alan.Stuart@duke-energy.com.

Sincerely,

)Cl~~

J. Ed Burchfield, Jr Site Vice President Oconee Nuclear Station Attachments:

Figure 1. ONS Site Figure 2. ONS 6-mile Vicinity

Legend c::J ONS Site

--~=~

0 0.25 M~es 05 3 Of 4 I J' i l )! l '

Figure 2. ONS 6-mlle Vicinity Pickens Countv Central Clemsofi Legend D Airport . . Surface Water c::J ONS Site D Heliport U.S. Route

- - State Highway

  • - _18-Mile Radius Place Local Road t::'J eounty

__.... Railroad 4of 4 I I' ;1 g c

/ -.. DUKE J. Ed Burchfield, Jr.

  • ' ENERGY. Vice Pres/d6nt Oconee Nuclear Station Duke Energy ONOt VP / 7800 Roch8ster Hwy Seneca. SC 296'12 o 864.873.3478 I 884.873.5791 Ed.Burchfield OcJuke.energy.com BY U.S. MAIL RETURN RECEIPT REQUESTED November 11, 2019 Chris Starker, Land Conservation Manager Upstate Forever 507 Pettigru St Greenville, SC 29601 RE: Duke Energy - Oconee Nuclear Station Units 1, 2, and 3 Subsequent License Renewal

Dear Mr. Starker:

Duke Energy Carolinas, LLC (Duke Energy) is seeking to renew the operating license for Oconee Nuclear Station Units 1, 2, and 3 (ONS) for an additional 20 years (see Table 1). Duke Energy and ONS have safely and reliably provided electricity to our Carolinas customers for decades. ONS has generated clean and cost-effective power, provided thousands of well-paying jobs, and produced substantial economic benefits for the Carolinas. Renewing the licenses of ONS is important for our customers, communities and environment.

Table 1. ONS Licensing Dates Extended License Expiration ONS Unit License Expiration Date Date Unit 1 February 6, 2033 February 6, 2053 Unit2 October 6, 2033 October 6, 2053 Unit3 July 19, 2034 July 19, 2054 Duke Energy's nuclear fleet plays an important role In the company's efforts to lower carbon emissions. In 2018, the Duke Energy nuclear fleet generated more than 72 billion kilowatt-hours of electricity and avoided the release of about 54 million tons of carbon dioxide - equivalent to keeping more than 1O million passenger cars off the road. The company has set aggressive carbon reduction goals of at least 50% by 2030 and net-zero by 2050 and keeping its nuclear fleet operating is key to achieving these goals.

Renewing the nuclear licenses will provide significant value to Duke Energy customers, as well as continue to support Carolinas communities through jobs, tax revenues and partnerships.

Duke Energy employs about 5,000 workers in its nuclear group, with additional contract workers supporting refueling outages and project work. In 2018, the Duke Energy also paid more than

$300 million in property and payroll taxes associated with the nuclear stations, benefiting local governments and school districts. In addition, nuclear employees support the communities where they live and work by donating time and funds through sponsorships and volunteer activities.

The ONS site is situated on 510 acres In eastern Oconee County, South Carolina (SC),

approximately eight miles northeast of Seneca, SC, on the southern shore of Lake Keowee.

During the license renewal term, Duke Energy proposes to continue operating the units as currently operated. There are currently no ground-disturbing activities or refurbishment anticipated at the ONS site during the subsequent license renewal period. Figures depicting the station site and the vicinity within a 6-mile radius of the station are enclosed.

Should you or your team have any questions or comments about ONS or the license renewal process, please contact Mikayla Kreuzberger at (864) 873-4204 I Mikayla.Kreuzberger@duke-energy.com or Alan Stuart at (980) 373-2079 / Alan.Stuart@duke-enerqy.com.

Sincerely,

)UIJ..PfJJ/

J. Ed Burchfield, Jr Site Vice President Oconee Nuclear Station Attachments:

Figure 1. ONS Site Figure 2. ONS 6-mile Vicinity 2 of 4 I I'., ):.:

Figure 1. ONS Site Legend c:::JoNSSite


====::iMll 0 0.25 0.5 ea 3 of 4 I 1' :1 ~ e

Figure 2. ONS 6-mlle Vicinity legend r:I Airport Surface Water D Heliport US, Route

.. - State Highway c::JoNSSrte

  • - . I B-M*e Radius Place Local Road Q eounty 0

==::::,MHea 2

~ Railroad 4 Of 41 p d )' ~*

( ~ DUKE

~ ENERGY.

J. Ed Burchfltld, Jr.

V,ce Pl'9Sid9nt Oconee Nuclear Station Duke Energy ONOtVP / 7800 Rochester Hwy Sen8ca, SC 29672 Il 864.873.3478 I 864.873.5791 Ed.Burchfield @cJuke.energy.com BYU.S. MAIL RETURN RECEIPT REQUESTED November 11, 2019 Sara Green, Executive Director SC Wildlife Federation 215 Pickens Street Columbia, SC, 29205 RE: Duke Energy - Oconee Nuclear Station Units 1, 2, and 3 Subsequent License Renewal

Dear Ms. Green:

Duke Energy Carolinas, LLC (Duke Energy) is seeking to renew the operating license for Oconee Nuclear Station Units 1, 2, and 3 (ONS) for an additional 20 years (see Table 1). Duke Energy and ONS have safely and reliably provided electricity to our Carolinas customers for decades. ONS has generated clean and cost-effective power, provided thousands of well-paying jobs, and produced substantial economic benefits for the Carolinas. Renewing the licenses of ONS is important for our customers, communities and environment.

Table 1. ONS Ucensing Dates Extended License Expiration ONS Unit License Expiration Date Date Unit 1 February 6, 2033 February 6, 2053 Unit2 October 6, 2033 October 6, 2053 Unit3 July 19, 2034 July 19, 2054 Duke Energy's nuclear fleet plays an important role in the company's efforts to lower carbon emissions. In 2018, the Duke Energy nuclear fleet generated more than 72 billion kilowatt-hours of electricity and avoided the release of about 54 million tons of carbon dioxide - equivalent to keeping more than 10 million passenger cars off the road. The company has set aggressive carbon reduction goals of at least 50% by 2030 and net-zero by 2050 and keeping its nuclear fleet operating is key to achieving these goals.

Renewing the nuclear licenses will provide significant value to Duke Energy customers, as well as continue to support Carolinas communities through jobs, tax revenues and partnerships.

Duke Energy employs about 5,000 workers in its nuclear group, with additional contract workers supporting refueling outages and project work. In 2018, the Duke Energy also paid more than

$300 million in property and payroll taxes associated with the nuclear stations, benefiting local governments and school districts. In addition, nuclear employees support the communities where they live and work by donating time and funds through sponsorships and volunteer activities.

The ONS site is situated on 510 acres in eastern Oconee County, South Carolina (SC),

approximately eight miles northeast of Seneca, SC, on the southern shore of Lake Keowee.

During the license renewal term, Duke Energy proposes to continue operating the units as currently operated. There are currently no ground-disturbing activities or refurbishment anticipated at the ONS site during the subsequent license renewal period. Figures depicting the station site and the vicinity within a 6-rnile radius of the station are enclosed.

Should you or your team have any questions or comments about ONS or the license renewal process, please contact Mikayla Kreuzberger at (864) 873-4204 / Mikayla.Kreuzberger@duke-energy.com or Alan Stuart at (980) 373-2079 / Alan.Stuart@duke-energy.com .

Sincerely,

) ~ aJ/id; J. Ed Burchfield, Jr Site Vice President Oconee Nuclear Station Attachments:

Figure 1. ON$ Site Figure 2. ONS 6-mile Vicinity

Figure 1. ONS Site legend c:::JoNSSrte

~ - - ~:;====;tMiln 0 0.25 05 3 Of 41 i' ,l ,!_! l'

Figure 2. ONS 6-mlle Vicinity Salem

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Central Legend a Airport Surface Water D

Heliport U.S. Route State Highway c:JoNSSite 1_ _ , &-Mlle Radius Place Local Road O eounty

  • - Railroad 4 or 4 I 1, ,t g ~*

el_, DUKE J. Ed Burchfield, Jr.

~ ; ENERGY. Vk:e President Oconee Nuclear Station Duke Energy ONOIVP I7800 Rochester Hwy Sen8cs, SC 29672 864.873.3478 f 864.873.5791 Ed.Burchfield Oduk8-en8Jr,y.com BY U.S. MAIL RETURN RECEIPT REQUESTED November 11, 2019 Douglas Spencer Southeastern Power Administration 1166 Athens Tech Rd.

Elberton, GA 30635 RE: Duke Energy - Oconee Nuclear Station Units 1, 2, and 3 Subsequent License Renewal

Dear Mr. Spencer:

Duke Energy Carolinas, LLC (Duke Energy) is seeking to renew the operating license for Oconee Nuclear Station Units 1, 2, and 3 (ONS) for an additional 20 years (see Table 1). Duke Energy and ONS have safely and reliably provided electricity to our Carolinas customers for decades. ONS has generated clean and cost-effective power, provided thousands of well-paying jobs, and produced substantial economic benefits for the Carolinas. Renewing the licenses of ONS is important for our customers, communities and environment.

Table 1. ONS Licensing Dates Extended License Expiration ONS Unit License Expiration Date Date Unit 1 February 6, 2033 February 6, 2053 Unit2 October 6, 2033 October 6, 2053 Unit3 July 19, 2034 July 19, 2054 Duke Energy's nuclear fleet plays an important role in the company's efforts to lower carbon emissions. In 2018, the Duke Energy nuclear fleet generated more than 72 billion kilowatt-hours of electricity and avoided the release of about 54 million tons of carbon dioxide - equivalent to keeping more than 1O million passenger cars off the road. The company has set aggressive carbon reduction goals of at least 50% by 2030 and net-zero by 2050 and keeping its nuclear fleet operating is key to achieving these goals.

Renewing the nuclear licenses will provide significant value to Duke Energy customers, as well as continue to support Carolinas communities through jobs, tax revenues and partnerships.

Duke Energy employs about 5,000 workers in its nuclear group, with additional contract workers supporting refueling outages and project work. In 2018, the Duke Energy also paid more than

$300 million in property and payroll taxes associated with the nuclear stations, benefiting local govemments and school districts. In addition, nuclear employees support the communities where they live and work by donating time and funds through sponsorships and volunteer activities.

The ONS site is situated on 510 acres in eastern Oconee County, South Carolina (SC),

approximately eight miles northeast of Seneca, SC, on the southern shore of Lake Keowee.

During the license renewal term, Duke Energy proposes to continue operating the units as currently operated. There are currently no ground-disturbing activities or refurbishment anticipated at the ONS site during the subsequent license renewal period. Figures depicting the station site and the vicinity within a 6-mile radius of the station are enclosed.

Should you or your team have any questions or comments about ONS or the license renewal process, please contact Ed Bruce at (704) 382-5239 / Ed.Bruce@duke-enerqy.com.

Sincerely, J. Ed Burchfield, Jr Site Vice President Oconee Nuclear Station Attachments:

Figure 1. ONS Site Figure 2. ONS 6-mile Vicinity 2 of 4 I P a ~* l'

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Figure 2. ONS 6-mile Vicinity Central Clemson ,

Legend a Airport Surface Water a Heliport U.S. Route

.. -~

c:]oNSSite 1_ _ , 6-Mile Radius

,___ Slate Highway Place Local Road c:J eounty ~--===>Mtloa 0 2

- Railroad 4 of 41 P ar l'

/_~ DUKE J. Ed Burchfield, Jr.

<{; ENERGY. V-IC9 President Oconee Nuctear Station Duke Energy ONOtVP f 1800 Roch9Slflf Hwy Seneca, SC 29672

  • 864.873 3478 I 864.873.5791 Ed.Burchfield @dukHnergy.com BY U.S. MAIL RETURN RECEIPT REQUESTED November 11, 2019 Amanda Brock, County Administrator Oconee County 415 S Pine St Walhalla, SC 29691 RE: Duke Energy - Oconee Nuclear Station Units 1, 2, and 3 Subsequent License Renewal

Dear Ms. Brock:

Duke Energy Carolinas, LLC (Duke Energy) is seeking to renew the operating license for Oconee Nuclear Station Units 1, 2, and 3 (ONS) for an additional 20 years (see Table 1). Duke Energy and ONS have safely and reliably provided electricity to our Carolinas customers for decades. ONS has generated clean and cost-effective power, provided thousands of well-paying jobs, and produced substantial economic benefits for the Carolinas. Renewing the licenses of ONS is important for our customers, communities and environment.

Table 1. ONS Licensing Dates Extended License Expiration ONS Unit License Expiration Date Date Unit 1 February 6, 2033 February 6, 2053 Unit2 October 6, 2033 October 6, 2053 Unit3 July 19, 2034 July 19, 2054 Duke Energy's nuclear fleet plays an important role in the company's efforts to lower carbon emissions. In 2018, the Duke Energy nuclear fleet generated more than 72 billion kilowatt-hours of electricity and avoided the release of about 54 million tons of carbon dioxide - equivalent to keeping more than 1o million passenger cars off the road. The company has set aggressive carbon reduction goals of at least 50% by 2030 and net-zero by 2050 and keeping its nuclear fleet operating is key to achieving these goals.

Renewing the nuclear licenses will provide significant value to Duke Energy customers, as well as continue to support Carolinas communities through jobs, tax revenues and partnerships.

Duke Energy employs about 5,000 workers in its nuclear group, with additional contract workers supporting refueling outages and project work. In 2018, the Duke Energy also paid more than

$300 million in property and payroll taxes associated with the nuclear stations, benefiting local governments and school districts. In addition, nuclear employees support the communities where they live and work by donating time and funds through sponsorships and volunteer activities.

The ONS site is situated on 510 acres in eastern Oconee County, South Carolina (SC),

approximately eight miles northeast of Seneca, SC, on the southern shore of Lake Keowee.

During the license renewal term, Duke Energy proposes to continue operating the units as currently operated. There are currently no ground-disturbing activities or refurbishment anticipated at the ONS site during the subsequent license renewal period. Figures depicting the station site and the vicinity within a 6-mile radius of the station are enclosed.

Should you or your team have any questions or comments about ONS or the license renewal process, please contact Mikayla Kreuzberger at (864) 873-4204 / Mikayla.Kreuzberger@duke-energy.com.

Sincerely, Ja~ji J. Ed Burchfield, Jr Site Vice President Oconee Nuclear Station Attachments:

Figure 1. ONS Site Figure 2. ONS 6-mile Vicinity cc: Phil Shirley, Director Oconee County Parks, Recreation and Tourism 1031 South Dover Rd, Seneca, SC 29672 2 of 4 I l' a ~ ~*

Figure 1. ONS Site Legend c::JoNSSite


===::,Mi 0 0.25 05 les 3 of 4 I I' ,1 g ~

Figure 2. ONS 6-mlle Vicinity Pickens County Central Clemson Legend II Airport Surface water D Heliport U.S. Route 0NSS1te

1. _, &-Mile Radius

-- State Highway Place Local Road t=J county

- - - Railroad

£ 'J DUKE J. Ed Burchfield, Jr.

~ ENERGY. V-,ce President Oconee Nuclear Station DuhEnergy ONOtVP / 78()() Rochflster Hwy Seneca, SC 29672 c 864.873.3478 I 864.873.5791 Ed.Burchf181d OdukHne,gy.oom BY U.S. MAIL RETURN RECEIPT REQUESTED November 11, 2019 Ken Roper, County Administrator Pickens County 222 McDaniel Ave Ste B2, Pickens, SC 29671 RE: Duke Energy - Oconee Nuclear Station Units 1, 2, and 3 Subsequent License Renewal

Dear Mr. Roper:

Duke Energy Carolinas, LLC (Duke Energy) is seeking to renew the operating license for Oconee Nuclear Station Units 1, 2, and 3 (ONS) for an additional 20 years (see Table 1). Duke Energy and ONS have safely and reliably provided electricity to our Carolinas customers for decades. ONS has generated clean and cost-effective power, provided thousands of well-paying jobs, and produced substantial economic benefits for the Carolinas. Renewing the licenses of ONS is important for our customers, communities and environment.

Table 1. ONS Licensing Dates Extended license Expiration ONS Unit License Expiration Date Date Unit 1 February 6, 2033 February 6, 2053 Unit2 October 6, 2033 October 6, 2053 Unit 3 July 19, 2034 July 19, 2054

~

Duke Energy's nuclear fleet plays an important role in the company's efforts to lower carbon emissions. In 2018, the Duke Energy nuclear fleet generated more than 72 billion kilowatt-hours of electricity and avoided the release of about 54 million tons of carbon dioxide - equivalent to keeping more than 1O million passenger cars off the road. The company has set aggressive carbon reduction goals of at least 50% by 2030 and net-zero by 2050 and keeping its nuclear fleet operating is key to achieving these goals.

Renewing the nuclear licenses will provide significant value to Duke Energy customers, as well as continue to support Carolinas communities through jobs, tax revenues and partnerships.

Duke Energy employs about 5,000 workers in its nuclear group, with additional contract workers supporting refueling outages and project work. In 2018, the Duke Energy also paid more than

$300 million in property and payroll taxes associated with the nuclear stations, benefiting local governments and school districts. In addition, nuclear employees support the communities where they live and work by donating time and funds through sponsorships and volunteer activities.

The ONS site is situated on 510 acres in eastern Oconee County, South Carolina (SC),

approximately eight miles northeast of Seneca, SC, on the southern shore of Lake Keowee.

During the license renewal term, Duke Energy proposes to continue operating the units as currently operated. There are currently no ground-disturbing activities or refurbishment anticipated at the ONS site during the subsequent license renewal period. Figures depicting the station site and the vicinity within a 6-mile radius of the station are enclosed.

Should you or your team have any questions or comments about ONS or the license renewal process, please contact Mikayla Kreuzberger at (864) 873-4204 / Mikayla.Kreuzberger@duke-enerqy.com.

Sincerely, Jr.A~~

J. Ed Burchfield, Jr Site Vice President Oconee Nuclear Station Attachments:

Figure 1. ONS Site Figure 2. ONS 6-mile Vicinity 2 Of 41 1' .I g ,'

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Figure 2. ONS 6-mlle Vicinity Legend D Airport Surface Water D Heliport US. Route

_,., State Highway CJoNSSite

  • - _ 1B-Mile Radius Place Local Road l=:t eounty

- Railroad 4 of 41 1' a FI?

J..'-> DUKE

~ ENERGY.

J. Ed Burchfield, Jr.

Vice President Oconee Nuclear Slalion Duke Energy ON01 VP I 7800 Rochester Hwy Seneca, SC 29672

~ 864.873.3478 I 864.873.5791 Ed.Buit:hfleld Oduks-&nergy.com BY U.S. MAIL RETURN RECEIPT REQUESTED November 11 , 2019 Paul McCormack, Director SC Department of Parks, Recreation and Tourism 1205 Pendleton St Columbia, SC 29201 RE: Duke Energy- Oconee Nuclear Station Units 1, 2, and 3 Subsequent License Renewal

Dear Mr. McCormack:

Duke Energy Carolinas, LLC (Duke Energy) is seeking to renew the operating license for Oconee Nuclear Station Units 1, 2, and 3 (ONS) for an additional 20 years (see Table 1). Duke Energy and ONS have safely and reliably provided electricity to our Carolinas customers for decades. ONS has generated clean and cost-effective power, provided thousands of well-paying jobs, and produced substantial economic benefits for the Carolinas. Renewing the licenses of ONS is important for our customers, communities and environment.

Table 1. ONS Licensing Dates Extended License Expiration ONS Unit License Expiration Date Date Unit 1 February 6, 2033 February 6, 2053 Unlt2 October 6, 2033 October 6, 2053 Unit3 July 19, 2034 July 19, 2054 Duke Energy's nuclear fleet plays an important role in the company's efforts to lower carbon emissions. In 2018, the Duke Energy nuclear fleet generated more than 72 billion kilowatt-hours of electricity and avoided the release of about 54 million tons of carbon dioxide - equivalent to keeping more than 1Omillion passenger cars off the road. The company has set aggressive carbon reduction goals of at least 50% by 2030 and net-zero by 2050 and keeping its nuclear fleet operating Is key to achieving these goals.

Renewing the nuclear licenses will provide significant value to Duke Energy customers, as well as continue to support Carolinas communities through jobs, tax revenues and partnerships.

Duke Energy employs about 5,000 workers in its nuclear group, with additional contract workers supporting refueling outages and project work. In 2018, the Duke Energy also paid more than

$300 million in property and payroll taxes associated with the nuclear stations, benefiting local governments and school districts. In addition, nuclear employees support the communities where they live and work by donating time and funds through sponsorships and volunteer activities.

The ONS site is situated on 510 acres in eastern Oconee County, South Carolina (SC),

approximately eight miles northeast of Seneca, SC, on the southern shore of Lake Keowee.

During the license renewal term, Duke Energy proposes to continue operating the units as currently operated. There are currently no ground-disturbing activities or refurbishment anticipated at the ONS site during the subsequent license renewal period. Figures depicting the station site and the vicinity within a 6-mile radius of the station are enclosed.

Should you or your team have any questions or comments about ONS or the license renewal process, please contact Mikayla Kreuzberger at (864) 873-4204 / Mikayla.Kreuzberger@duke-energy.com or Scott Fletcher at (980) 875-6014 / Scott.Fletcher@duke-enerqy.com.

Sincerely, J. Ed Burchfield, Jr Site Vice President Oconee Nuclear Station Attachments:

Figure 1. ONS Site Figure 2. ONS 6-mile Vicinity

Figure 1. ONS Site Legend CoNSSite


===:::::iMiles 0 O.:Z5 0.5 3 of 41 1' a ~ ,:

Figure 2. ONS 6-mlle Vicinity L

Central Legend a Airport

  • Surface water D Heliport __ ..,

c:JoNSS1te U.S. Route ,_ .I &-Mile Radius

- - State Highway Place Local Road t::J eounty

~ Railroad 4 of 4 I P a ~* t*

el_~ DUKE J. Ed Burchflelcl, Jr.

~ ENERGY. Vice President Oa,nee Nuclear S1111/Cln Dukt1 Energy ONOI VP f 7800 RochesterHwy Seneca, SC 29672

, 864.873.3478 t 864.873.5791 Ed.Burdllield @duke-energy.com BVU.S.MAIL RETURN RECEIPT REQUESTED November 11, 2019 Jeff Phillips, Director of Water Resources Greenville Water System 407 West Broad Street Greenville, SC 29601 RE: Duke Energy - Oconee Nuclear Station Units 1, 2, and 3 Subsequent License Renewal

Dear Mr. Phillips:

Duke Energy Carolinas, LLC (Duke Energy) is seeking to renew the operating license for Oconee Nuclear Station Units 1, 2, and 3 (ONS) for an additional 20 years (see Table 1). Duke Energy and ONS have safely and reliably provided electricity to our Carolinas customers for decades. ONS has generated clean and cost-effective power, provided thousands of well-paying jobs, and produced substantial economic benefits for the Carolinas. Renewing the licenses of ONS is important for our customers, communities and environment.

Table 1. ONS Ucenslng Oates Extended License Expiration ONS Unit Ucense Expiration Date Date Unit 1 February 6, 2033 February 6, 2053 Unit2 October 6, 2033 October 6, 2053 Unit3 July 19, 2034 July 19, 2054 Duke Energy's nuclear fleet plays an important role in the company's efforts to lower carbon emissions. In 2018, the Duke Energy nuclear fleet generated more than 72 billion kilowatt-hours of electricity and avoided the release of about 54 million tons of carbon dioxide - equivalent to keeping more than 1O million passenger cars off the road. The company has set aggressive carbon reduction goals of at least 50% by 2030 and net-zero by 2050 and keeping its nuclear fleet operating is key to achieving these goals.

Renewing the nuclear licenses will provide significant value to Duke Energy customers, as well as continue to support Carolinas communities through jobs, tax revenues and partnerships.

Duke Energy employs about 5,000 workers in its nuclear group, with additional contract workers supporting refueling outages and project work. In 2018, the Duke Energy also paid more than

$300 million in property and payroll taxes associated with the nuclear stations, benefiting local governments and school districts. In addition, nuclear employees support the communities where they live and work by donating time and funds through sponsorships and volunteer activities.

The ONS site is situated on 510 acres in eastern Oconee County, South Carolina (SC),

approximately eight miles northeast of Seneca, SC, on the southern shore of Lake Keowee.

During the license renewal term, Duke Energy proposes to continue operating the units as currently operated. There are currently no ground-disturbing activities or refurbishment anticipated at the ONS site during the subsequent license renewal period. Figures depicting the station site and the vicinity within a 6-mile radius of the station are enclosed.

Should you or your team have any questions or comments about ONS or the license renewal process, please contact Ed Bruce at (704) 382-5239 / Ed.Bruce@duke-energy.com.

Sincerely,

~ Cl 6l-l/JIJ J. Ed Burchfield, Jr Site Vice President Oconee Nuclear Station Attachments:

Figure 1. ONS Site Figure 2. ONS 6-mile Vicinity 2 Of 41 p ,l )' \'

Figure 1. ONS Site Legend c:JoNSS1te


c::==:::JM1ln 0 0.25 0.5 3 Of 41 p ,l ::! L'

Figure 2. ONS 6-mile Vicinity Pickens County Cffllral ClemlC>h Legend a Airport 6 Surface \Nater D Heliport U.S. Route ~--

c:]oNSSite

  • - _, 8-Mile Radius

-- State Highway Place Local Road C] eounty 0

---==:::::1Milc1 2

.... *- Railroad 4 of 41 Pai'. l'

(~ DUKE J. Ed Burchfield, Jr.

<{; ENERGY. VlcBPr9sldent Oconee Nuclear SlatiOn Duke Energy ONOt VP { 7800 Rochest91' Hwy Sen8ca, SC 29672

  • 864.873.3478
864.873.5791 Ed.Burchfield @duke-energy.com BY U.S. MAIL RETURN RECEIPT REQUESTED November 11, 2019 Bob Faires, Director Seneca Light and Water PO Box4773 Seneca, SC 29679 RE
Duke Energy - Oconee Nuclear Station Units 1, 2, and 3 Subsequent License Renewal

Dear Mr. Faires:

Duke Energy Carolinas, LLC (Duke Energy) is seeking to renew the operating license for Oconee Nuclear Station Units 1, 2, and 3 (ONS) for an additional 20 years (see Table 1). Duke Energy and ONS have safely and reliably provided electricity to our Carolinas customers for decades. ONS has generated clean and cost-effective power, provided thousands of well-paying jobs, and produced substantial economic benefits for the Carolinas. Renewing the licenses of ONS is important for our customers, communities and environment.

Table 1. ONS Licensing Dates Extended License Expiration ON$ Unit License Expiration Date Date Unit 1 February 6, 2033 February 6, 2053 Unit2 October 6, 2033 October 6, 2053 Unit3 July 19, 2034 July 19, 2054 Duke Energy's nuclear fleet plays an important role in the company's efforts to lower carbon emissions. In 2018, the Duke Energy nuclear fleet generated more than 72 billion kilowatt-hours of electricity and avoided the release of about 54 million tons of carbon dioxide - equivalent to keeping more than 1o million passenger cars off the road. The company has set aggressive carbon reduction goals of at least 50% by 2030 and net-zero by 2050 and keeping its nuclear fleet operating is key to achieving these goals.

Renewing the nuclear licenses will provide significant value to Duke Energy customers, as well as continue to support Carolinas communities through jobs, tax revenues and partnerships.

Duke Energy employs about 5,000 workers in its nuclear group, with additional contract workers supporting refueling outages and project work. In 2018, the Duke Energy also paid more than

$300 million in property and payroll taxes associated with the nuclear stations, benefiting local governments and school districts. In addition, nuclear employees support the communities where they live and work by donating time and funds through sponsorships and volunteer activities.

The ONS site is situated on 510 acres in eastern Oconee County, South Carolina (SC),

approximately eight miles northeast of Seneca, SC, on the southern shore of Lake Keowee.

During the license renewal term, Duke Energy proposes to continue operating the units as currently operated. There are currently no ground-disturbing activities or refurbishment anticipated at the ONS site during the subsequent license renewal period. Figures depicting the station site and the vicinity within a 6-mile radius of the station are enclosed.

Should you or your team have any questions or comments about ONS or the license renewal process, please contact Tony Garland at (864) 873-4216 / Tony.Garland@duke-energy.com.

Sincerely, jEAli ... 1/J/

J. Ed Burchfield, Jr Site Vice President Oconee Nuclear Station Attachments:

Figure 1. ONS Site Figure 2. ONS 6-mile Vicinity 2 of 4 I t' .1 r ,*

Figure 1. ONS Site Legend C]ONSSite

- - - -===:::::iMdn 0 0.211 0.5 3 Of 4 11l ,I ~ L'

Figure 2. ONS 6-mlle Vicinity Pickens County Central Clemso"il Legend a Airport Surface \Nater a

Heliport US. Route Stale Highway C]oNSSite

  • - Mile Radius PlaCfl Local Road l:::J eounty

- Railroad 4 Of 41 1' . l ~ ~*

"-.a, DUKE J. Ed Burc:fllleld, Jr.

,? ENERGY. V-,ce Pn,sident Oconee Nuclear Station Dulce E1111rgy ON01 VP / 7800 Rochester Hwy Seneca, SC 29672

,, 864.873.3478 i 864.873.5791 Ed.&nchfiekJ @duke-ene,gy.com BY U.S. MAIL RETURN RECEIPT REQUESTED November 11, 2019 Scott Parris City of Walhalla PO Box 1099 Walhalla, SC 29691 RE: Duke Energy - Oconee Nuclear Station Units 1, 2, and 3 Subsequent License Renewal

Dear Mr. Parris:

Duke Energy Carolinas, LLC (Duke Energy) is seeking to renew the operating license for Oconee Nuclear Station Units 1, 2, and 3 (ONS) for an additional 20 years (see Table 1). Duke Energy and ONS have safely and reliably provided electricity to our Carolinas customers for decades. ONS has generated clean and cost-effective power, provided thousands of well-paying jobs, and produced substantial economic benefits for the Carolinas. Renewing the licenses of ONS is important for our customers, communities and environment.

Table 1. ONS Ucenslng Dates Extended License Expiration ONS Unit License Expiration Date Date Unit 1 February 6, 2033 February 6, 2053 Unit2 October 6, 2033 October 6, 2053 Unit3 July 19, 2034 July 19, 2054 Duke Energy's nuclear fleet plays an important role in the company's efforts to lower carbon emissions. In 2018, the Duke Energy nuclear fleet generated more than 72 billion kilowatt-hours of electricity and avoided the release of about 54 million tons of carbon dioxide - equivalent to keeping more than 1O million passenger cars off the road. The company has set aggressive carbon reduction goals of at least 50% by 2030 and net-zero by 2050 and keeping its nuclear fleet operating is key to achieving these goals.

Renewing the nuclear licenses will provide significant value to Duke Energy customers, as well as continue to support Carolinas communities through jobs, tax revenues and partnerships.

Duke Energy employs about 5,000 workers in its nuclear group, with additional contract workers supporting refueling outages and project work. In 2018, the Duke Energy also paid more than

$300 million in property and payroll taxes associated with the nuclear stations, benefiting local governments and school districts. In addition, nuclear employees support the communities where they live and work by donating time and funds through sponsorships and volunteer activities.

The ONS site is situated on 510 acres in eastern Oconee County, South Carolina (SC),

approximately eight miles northeast of Seneca, SC, on the southern shore of Lake Keowee.

During the license renewal term, Duke Energy proposes to continue operating the units as currently operated. There are currently no ground-disturbing activities or refurbishment anticipated at the ONS site during the subsequent license renewal period. Figures depicting the station site and the vicinity within a 6-mile radius of the station are enclosed.

Should you or your team have any questions or comments about ONS or the license renewal process, please contact Ed Bruce at (704) 382-5239 / Ed.Bruce@duke-enerqy.com.

Sincerely, J. Ed Burchfield, Jr Site Vice President Oconee Nuclear Station Attachments:

Figure 1. ONS Site Figure 2. ONS 6-rnile Vicinity

Legend c::::JoNS Site

0---~==::::i 0.25 05 MilN 3 Of 41 f' ,l )' _.

Figure 2. ONS 6-mlle Vicinity Clemson Legend a Airport Surface Water a Heliport U.S. Route State Highway

~--

c:]oNSSrte

, __ , 8-Mile Radius Place Local Road c:J eounty 0

- - - = = = = i MDes 2

_,._ Railroad 4 of 41 P .i ~ ,.

( ~ DUKE

~ ENERGY.

J. Ed Burchfield, Jr.

\r,ce President Oconee Nuclear Station Dulce Energy ON01VP / 7800 Rochester Hwy Seneca, SC 29672

, 864,873.3478 I 864.873.5791 Ed.Burchfield Oduke-6nergy.com BYU.S.MAIL RETURN RECEIPT REQUESTED November 7, 2019 Mr. Chris Eleazer, Director Oconee Joint Regional Sewer Authority 623 Return Church Road Seneca, SC 29678 RE: Duke Energy- Oconee Nuclear Station Units 1, 2, and 3 Subsequent License Renewal

Dear Mr. Eleazer:

Duke Energy Carolinas, LLC (Duke Energy) is seeking to renew the operating license for Oconee Nuclear Station Units 1, 2, and 3 (ONS) for an additional 20 years (see Table 1). Duke Energy and ONS have safely and reliably provided electricity to our Carolinas customers for decades. ONS has generated clean and cost-effective power, provided thousands of well-paying jobs, and produced substantial economic benefits for the Carolinas. Renewing the licenses of ONS is important for our customers, communities and environment.

Table 1. ONS Licensing Dates ONS Unit License Expiration Date Extended License Expiration Date Unit l February 6, 2033 February 6, 2053 Unit 2 October 6, 2033 October 6, 2053 Unit 3 July 19, 2034 July 19, 2054 Duke Energy's nuclear fleet plays an important role in the company's efforts to lower carbon emissions. In 2018, the Duke Energy nuclear fleet generated more than 72 billion kilowatt-hours of electricity and avoided the release of about 54 million tons of carbon dioxide - equivalent to keeping more than 10 million passenger cars off the road. The company has set aggressive carbon reduction goals of at least 50% by 2030 and net-zero by 2050 and keeping its nuclear fleet operating is key to achieving these goals.

Renewing the nuclear licenses will provide significant value to Duke Energy customers, as well as continue to support Carolinas communities through jobs, tax revenues and partnerships. Duke Energy employs about 5,000 workers in its nuclear group, with additional contract workers supporting refueling outages and project work. In 2018, the Duke Energy also paid more than

$300 million in property and payroll taxes associated with the nuclear stations, benefiting local governments and school districts. In addition, nuclear employees support the communities where they live and work by donating time and funds through sponsorships and volunteer activities.

The ONS site is situated on 510 acres in eastern Oconee County, South Carolina (SC),

approximately eight miles northeast of Seneca, SC, on the southern shore of Lake Keowee.

During the license renewal term, Duke Energy proposes to continue operating the units as currently operated. There are currently no ground-disturbing activities or refurbishment anticipated at the ONS site during the subsequent license renewal period. Figures depicting the station site and the vicinity within a 6-mile radius of the station are enclosed.

Should you or your team have any questions or comments about ONS or the license renewal process, please contact Tony Garland at (864) 873-4216 / Ton *.Garland@duke-ener_g)'..com.

Sincerely, J. Ed Burchfield, Jr Site Vice President Oconee Nuclear Station Attachments:

Figure 1. ONS Site Figure 2. ONS 6-mile Vicinity

Figure 1. ONS Site Legend c::::J ONS Site

- - - ~:;===::;iMilea 0 0,2$ 0,5

Figure 2. ONS 6-mile Vicinity Legend a Airport 6 Surface 'Nater D Heliport U.S. Route

  • - - State Highway c:JONSSite

, __ , 8-Mile Radius Place Local Road D eounty

- - - Railroad 4 of 4 I I' a ~ ,:

ef_~ DUKE J. Ed Burchfield, Jr.

~ ENERGY. V-<<:e President Oconee Nuclear Station Duke Energy ON01VP / 7800 RochestfJf Hwy Seneca, SC 29672

,, 864.873.3478

' 864.873.5791 Ed.BurchflfJld Oduke-ene,gy.com BY U.S. MAIL RETURN RECEIPT REQUESTED November 11, 2019 Stan Simpson United States Army Corps of Engineers Savannah District 100 W. Oglethorpe Ave Savannah, GA 31401 RE: Duke Energy - Oconee Nuclear Station Units 1, 2, and 3 Subsequent License Renewal

Dear Mr. Simpson:

Duke Energy Carolinas, LLC (Duke Energy) is seeking to renew the operating license for Oconee Nuclear Station Units 1, 2, and 3 (ONS) for an additional 20 years (see Table 1). Duke Energy and ONS have safely and rellably provided electricity to our Carolinas customers for decades. ONS has generated clean and cost-effective power, provided thousands of well-paying jobs, and produced substantial economic benefits for the Carolinas. Renewing the licenses of ONS is important for our customers, communities and environment.

Table 1. ONS Licensing Dates Extended License Expfration ONS Unit License Expiration Date Date Unit 1 February 6, 2033 February 6, 2053 Unit2 October 6, 2033 October 6, 2053 Unit 3 July 19, 2034 July 19, 2054 Duke Energy's nuclear fleet plays an important role in the company's efforts to lower carbon emissions. In 2018, the Duke Energy nuclear fleet generated more than 72 billion kilowatt-hours of electricity and avoided the release of about 54 million tons of carbon dioxide - equivalent to keeping more than 10 million passenger cars off the road. The company has set aggressive carbon reduction goals of at least 50% by 2030 and net-zero by 2050 and keeping its nuclear fleet operating is key to achieving these goals.

Renewing the nuclear licenses will provide significant value to Duke Energy customers, as well as continue to support Carolinas communities through jobs, tax revenues and partnerships.

Duke Energy employs about 5,000 workers in its nuclear group, with additional contract workers supporting refueling outages and project work. In 2018, the Duke Energy also paid more than

$300 million in property and payroll taxes associated with the nuclear stations, benefiting local governments and school districts. In addition, nuclear employees support the communities where they live and work by donating time and funds through sponsorships and volunteer activities.

The ONS site is situated on 510 acres in eastern Oconee County, South Carolina (SC),

approximately eight miles northeast of Seneca, SC, on the southern shore of Lake Keowee.

During the license renewal term, Duke Energy proposes to continue operating the units as currently operated. There are currently no ground-disturbing activities or refurbishment anticipated at the ONS site during the subsequent license renewal period. Figures depicting the station site and the vicinity within a 6-mile radius of the station are enclosed.

Should you or your team have any questions or comments about ONS or the license renewal process, please contact Alan Stuart at (980) 373-2079 / Alan .Stuart@duke-enerqy.com.

Sincerely,

~-;:~

Site Vice President Oconee Nuclear Station Attachments:

Figure 1. ONS Site Figure 2. ONS 6-mile Vicinity 2 of 4 IP"~: c

Figure 1. ONS Site Legend CJoNSSite


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Figure 2. ONS 6-mlle Vicinity Central f

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ENCLOSURE 4 OCONEE NUCLEAR STATION NON-PROPRIETARY REFERENCE DOCUMENTS AND A REDACTED VERSION OF A PROPRIETARY REFERENCE DOCUMENT (PUBLIC VERSION)

ENCLOSURE 4 ATTACHMENT 1 OCONEE NUCLEAR STATION Framatome Topical Report ANP 3898NP, Revision 0, Framatome Reactor Vessel TLAA and Aging Management Review Input to the ONS SLRA, May 2021

framatome Framatome Reactor Vessel and RCP ANP-3898NP Revision 0 TLAA and Aging Management Review Input to the ONS SLRA May 2021 (c) 2021 Framatome Inc.

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

ANP-3898NP Revision 0 Copyright © 2021 Framatome Inc.

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

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page i Nature of Changes Section(s)

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

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page ii Contents Page

1.0 INTRODUCTION

............................................................................................... 1-1 1.1 References for Section 1.0 ..................................................................... 1-4 2.0 RPV NEUTRON FLUENCE PROJECTIONS .................................................... 2-1 2.1 Introduction ............................................................................................. 2-1 2.2 Regulatory Guidance for Subsequent License Renewal ......................... 2-1 2.3 Methodology ........................................................................................... 2-3 2.4 Summary and Conclusions ..................................................................... 2-8 2.5 References for Section 2.0 ..................................................................... 2-8 3.0 UPPER SHELF ENERGY.................................................................................. 3-1 3.1 Introduction ............................................................................................. 3-1 3.2 Regulatory Guidance for Subsequent License Renewal ......................... 3-1 3.3 Methodology ........................................................................................... 3-3 3.3.1 CvUSE at 72 EFPY Based on Regulatory Guide 1.99 Revision 2 .................................................................................... 3-4 3.3.2 Equivalent Margins Analyses ....................................................... 3-5 3.4 Traditional Beltline and Extended Beltline 72 EFPY Upper Shelf Energy ........................................................................................... 3-6 3.4.1 Material Properties Including Copper and Initial Upper-Shelf Energy...................................................................... 3-7 3.4.2 Comparison to Current Licensing Basis and NRC RVID2-Traditional Beltline .......................................................... 3-10 3.5 Confirmation that Oconee is Bounded by NRC Approved Generic Technical Reports ................................................................... 3-23 3.6 ONS Unit 1 Weld SA-1135 EMA Reconciliation.................................... 3-25 3.7 References for Section 3.0 ................................................................... 3-25 4.0 EMA OF ONS UNIT 3 RPV OUTLET NOZZLES AND TRANSITION FORGING .......................................................................................................... 4-1 4.1 Introduction ............................................................................................. 4-1 4.2 Regulatory Requirements ....................................................................... 4-2 4.2.1 10 CFR 50 Appendix G ................................................................ 4-2 4.2.2 COMPLIANCE WITH 10 CFR 50 APPENDIX G AND ACCEPTANCE CRITERIA ........................................................... 4-3

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page iii 4.3 Transition Forging Equivalent Margins Analysis (Levels A and B Service Loadings) ......................................................................... 4-4 4.3.1 Material Properties and Levels A and B Service Loadings ...................................................................................... 4-4 4.3.2 Fracture Mechanics Analysis ....................................................... 4-7 4.4 RPV Outlet Nozzle Forgings Equivalent Margins (Levels A and B Service Loadings) ....................................................................... 4-23 4.4.1 Material Properties and Levels A and B Service Loadings .................................................................................... 4-23 4.4.2 Fracture Mechanics Analysis ..................................................... 4-26 4.5 RPV Outlet Nozzle Forgings Equivalent Margins Analysis (Levels C and D Service Loadings) ...................................................... 4-39 4.5.1 Material Properties and Levels C and D Service Loadings .................................................................................... 4-39 4.5.2 Fracture Mechanics Evaluation .................................................. 4-41 4.6 References for Section 4.0 ................................................................... 4-51 5.0 PRESSURIZED THERMAL SHOCK ................................................................. 5-1 5.1 Introduction ............................................................................................. 5-1 5.2 Regulatory Guidance for Subsequent License Renewal ......................... 5-1 5.3 Methodology ........................................................................................... 5-3 5.4 Traditional Beltline and Extended Beltline 72 EFPY Pressurized Thermal Shock .................................................................... 5-4 5.4.1 Material Properties Including Copper, Nickel, and Initial RTNDT, ................................................................................. 5-6 5.4.2 Comparison to Current Licensing Basis ..................................... 5-13 5.5 References for Section 5.0 ................................................................... 5-23 6.0 P-T LIMITS ........................................................................................................ 6-1 6.1 Introduction ............................................................................................. 6-1 6.2 Regulatory Guidance for Subsequent License Renewal ......................... 6-1 6.3 P-T Limits at 72 EFPY ............................................................................ 6-3 6.4 Adjusted Reference Temperatures ......................................................... 6-5 6.5 Assessment of 72 EFPY Operating Window......................................... 6-12 6.6 References for Section 6.0 ................................................................... 6-20 7.0 FRACTURE MECHANICS EVALUATION OF UNDERCLAD CRACKS ........................................................................................................... 7-1 7.1 Introduction ............................................................................................. 7-1

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page iv 7.2 Regulatory Guidance for Subsequent License Renewal ......................... 7-1 7.3 Description of RV Shell Region with Postulated Underclad Cracks..................................................................................................... 7-2 7.3.1 Geometry ..................................................................................... 7-3 7.3.2 Adjusted Reference Temperatures .............................................. 7-3 7.3.3 Material Properties ....................................................................... 7-4 7.3.4 Service Loadings.......................................................................... 7-6 7.4 Fracture Mechanics Analysis .................................................................. 7-8 7.4.1 Methodology ................................................................................ 7-8 7.4.2 Acceptance Criteria .................................................................... 7-11 7.4.3 Fracture Toughness Curve ........................................................ 7-11 7.4.4 Fatigue Crack Growth Model...................................................... 7-12 7.4.5 Calculation of KI by Polynomial Stress Representation .............. 7-14 7.4.6 Calculation of KI by Weight Function .......................................... 7-15 7.4.7 Irwin Plastic Zone Correction ..................................................... 7-15 7.4.8 Evaluation Results ..................................................................... 7-19 7.5 References for Section 7.0 ................................................................... 7-20 8.0 REACTOR VESSEL ENVIRONMENTALLY-ASSISTED FATIGUE................... 8-1 8.1 Introduction ............................................................................................. 8-1 8.2 Regulatory Guidance for Subsequent License Renewal ......................... 8-2 8.3 Methodology ........................................................................................... 8-2 8.4 Assumptions ........................................................................................... 8-3 8.5 Summary of Results................................................................................ 8-4 8.6 References for Section 8.0 ..................................................................... 8-6 9.0 IRRADIATION EMBRITTLEMENT OF ONS RPV SUPPORTS FOR SLR ................................................................................................................... 9-1 9.1 Introduction ............................................................................................. 9-1 9.2 Regulatory Guidance for Subsequent License Renewal ......................... 9-1 9.3 Description of RPV Support and Evaluation Methodology ...................... 9-5 9.4 Irradiation Embrittlement Further Evaluation......................................... 9-13 9.4.1 Assessment of Current Condition............................................... 9-13 9.4.2 Design Stress Summary ............................................................ 9-14 9.4.3 Lowest Service Temperature ..................................................... 9-17 9.4.4 Establish EOL NDT Temperature............................................... 9-20 9.4.5 Comparison of EOL NDT to LST for Critical RPV Support Locations ...................................................................... 9-27 9.5 References for Section 9.0 ................................................................... 9-30

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page v 10.0 REACTOR COOLANT PUMPS THERMAL EMBRITTLEMENT (CASS) ............................................................................................................ 10-1 10.1 Introduction ........................................................................................... 10-1 10.2 Regulatory Guidance for SLR ............................................................... 10-4 10.3 Identification of RCP Items Subject to Aging Management Review .................................................................................................. 10-8 10.4 Calculation of Ferrite Numbers for CASS Parts .................................... 10-8 10.5 Temperatures of ONS Unit 2 and ONS Unit 3 Stuffing Boxes During Normal Operation .................................................................... 10-11 10.6 Calculation of Thermally Aged Jd ........................................................ 10-11 10.7 JIc for Limiting ONS Unit 2 RC-P1A1, Lower Half-Heat 17368-1............................................................................................... 10-14 10.8 Summary and Conclusions ................................................................. 10-15 10.9 References for Section 10.0 ............................................................... 10-15 11.0 ONS UNIT 2 AND ONS UNIT 3 RCP BOUNDING FLAW TOLERANCE EVALUATION ........................................................................... 11-1 11.1 Introduction ........................................................................................... 11-1 11.2 Methodology ......................................................................................... 11-2 11.3 Assumptions ......................................................................................... 11-5 11.4 Summary of Results.............................................................................. 11-9 11.5 References for Section 11.0 ............................................................... 11-11

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page vi List of Tables Table 2-1 72 EFPY ONS Unit 1 RPV Shell Locations with Fluence > 1.0E17 n/cm2 (E > 1.0 MeV) .............................................................................. 2-16 Table 2-2 72 EFPY ONS Unit 2 RPV Shell Locations with Fluence > 1.0E17 n/cm2 (E > 1.0 MeV) .............................................................................. 2-17 Table 2-3 72 EFPY ONS Unit 3 RPV Shell Locations with Fluence > 1.0E17 n/cm2 (E > 1.0 MeV) .............................................................................. 2-18 Table 3-1 Linde 80 Initial CvUSE Values ................................................................ 3-8 Table 3-2 Upper-Shelf Energy Data for Extended Beltline Forging Materials Used to Determine Generic Upper-Shelf Energy .................................... 3-9 Table 3-3 Traditional beltline Cu Content Differences between CLB and SLR ..... 3-13 Table 3-4 Traditional Beltline CvUSE Differences between CLB and SLR ........... 3-14 Table 3-5 Copper Content Data for Pre-1971 ASTM A508 Class 2 Reactor Vessel Forgings: Bethlehem Steel Ingots ............................................. 3-15 Table 3-6 Evaluation of 72 EFPY Charpy V-Notch Upper-Shelf Energy -

Oconee Unit 1 ....................................................................................... 3-16 Table 3-7 Evaluation of 72 EFPY Charpy V-Notch Upper-Shelf Energy -

Oconee Unit 2 ....................................................................................... 3-19 Table 3-8 Evaluation of 72 EFPY Charpy V-Notch Upper-Shelf Energy -

Oconee Unit 3 ....................................................................................... 3-21 Table 3-9 BAW-2178 and BAW-2192 Reactor Vessel Weld Locations--

Copper Content and 80-Year Fluence Projections ............................... 3-26 Table 4-1 Material Properties for SA-508 Class 2 ................................................... 4-6 Table 4-2 Material SA 508 Class 2 (3/4Ni-1/2Mo-1/3Cr-V) ................................... 4-24 Table 4-3 Service Levels A and B Outlet Nozzle Loads........................................ 4-25 Table 4-4 Service Levels C and D Outlet Nozzle Loads ....................................... 4-41 Table 4-5 ONS Unit 3 Outlet Nozzle EMA Results (Levels C and D) .................... 4-46 Table 5-1 Reference Temperature Data for Forging Materials Used to Determine Generic Initial Reference Temperature ................................. 5-9 Table 5-2 Generic Initial RTNDT for ASTM A508 Class 2 Forgings Supplied to Rotterdam Dockyard ............................................................................. 5-10 Table 5-3 Reference Temperature Data for Linde 80 Weld Metals Used to Determine Generic Initial Reference Temperature ............................... 5-11 Table 5-4 Reference Temperature Data for Weld Wire Heat No. 299L44 Used to Determine Generic Initial Reference Temperature ........................... 5-12 Table 5-5 Traditional Beltline Ni Content Differences between CLB and SLR ...... 5-18

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page vii Table 5-6 Traditional Beltline Initial RTNDT Differences between CLB and SLR .... 5-19 Table 5-7 Reference Temperature Data for Plate Materials Used to Determine Generic Initial Reference Temperature ................................................. 5-20 Table 5-8 Reference Temperature Data for Forging Materials Used to Determine Generic Initial Reference Temperature ............................... 5-21 Table 5-9 Heat Specific and Generic Initial RTTo with Associated Initial Margin ... 5-22 Table 5-10 Evaluation of Reactor Vessel Pressurized Thermal Shock Toughness Properties at 72 EFPY - Oconee Unit 1 ............................. 5-25 Table 5-11 Evaluation of Reactor Vessel Pressurized Thermal Shock Toughness Properties at 72 EFPY - Oconee Unit 2 ............................. 5-26 Table 5-12 Evaluation of Reactor Vessel Pressurized Thermal Shock Toughness Properties at 72 EFPY - Oconee Unit 3 ............................. 5-27 Table 6-1 Limiting Adjusted Reference Temperature (ART) Values for ONS Unit 1 Reactor Vessel ............................................................................. 6-6 Table 6-2 Limiting Adjusted Reference Temperature (ART) Values for ONS Unit 2 Reactor Vessel ............................................................................. 6-7 Table 6-3 Limiting Adjusted Reference Temperature (ART) Values for ONS Unit 3 Reactor Vessel ............................................................................. 6-8 Table 6-4 Adjusted Reference Temperature (ART) Evaluation for the ONS Unit 1 Reactor Vessel Traditional Beltline and Extended Beltline Materials at 72 EFPY .............................................................................. 6-9 Table 6-5 Adjusted Reference Temperature (ART) Evaluation for the ONS Unit 2 Reactor Vessel Traditional Beltline and Extended Beltline Materials at 72 EFPY ............................................................................ 6-10 Table 6-6 Adjusted Reference Temperature (ART) Evaluation for the ONS Unit 3 Reactor Vessel Traditional Beltline and Extended Beltline Materials at 72 EFPY ............................................................................ 6-11 Table 7-1 Reactor Vessel Shell Dimensions ........................................................... 7-3 Table 7-2 Summary of RTPTS Temperatures for 72 EFPY ...................................... 7-4 Table 7-3 Properties for Base Metal (3/4Ni 1/2Mo Cr-V) ......................................... 7-5 Table 7-4 Properties for Cladding (18Cr-8Ni) ......................................................... 7-5 Table 7-5 List of Applicable Design Transients ....................................................... 7-7 Table 7-6 Results Summary ................................................................................. 7-19 Table 8-1 CUFen Results Summary ......................................................................... 8-5 Table 9-1 RPV Support Load Comparison: LBB Loads / Original Design Loads .................................................................................................... 9-16

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page viii Table 9-2 ONS RPV Support Skirt Assembly Base Metal Materials ..................... 9-21 Table 9-3 ONS RPV Support Skirt Assembly Weld Materials ............................... 9-22 Table 9-4 Calculated NDTT Values for the ONS RPV Support Skirt Assembly Base Metal Materials ............................................................ 9-24 Table 9-5 Calculated NDTT Values for the ONS RPV Support Skirt Assembly Weld Materials...................................................................... 9-24 Table 9-6 Comparison of the NDTT dpa at ONS RPV Support Skirt Assembly Base Metal LST to Projected 72 EFPY dpa.......................... 9-25 Table 9-7 Comparison of the NDTT dpa at ONS RPV Support Skirt Assembly Weld Metal LST to Projected 72 EFPY dpa.......................... 9-26 Table 10-1 Thermal Embrittlement Susceptibility from NUREG-2191, Table XI.M12-1 ............................................................................................... 10-7 Table 10-2 ONS RCP Items Susceptible to Thermal Embrittlement Based on Ferrite % ............................................................................................. 10-10 Table 10-3 Thermally Aged Jd (0.1-inch) for ONS Unit 2 and ONS Unit 3 RCP Casings ............................................................................................... 10-13 Table 11-1 Summary of Results............................................................................ 11-10

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page ix List of Figures Figure 2-1 ONS Unit 1 Reactor Vessel Shell ......................................................... 2-10 Figure 2-2 ONS Unit 2 Reactor Vessel Shell ......................................................... 2-11 Figure 2-3 ONS Unit 3 Reactor Vessel Shell ......................................................... 2-12 Figure 2-4 Attenuated Neutron Fluence Comparison - Core Center ..................... 2-13 Figure 2-5 Attenuated Neutron Fluence Comparison - Nozzle Region.................. 2-14 Figure 2-6 ANP-10348P, Figure 1-1, Calculations and Measurements.................. 2-15 Figure 4-1 Stress Path Lines (Transition Forging).................................................. 4-14 Figure 4-2 J-Integral versus Circumferential Flaw Extension - BASE_1 (Transition Forging) .............................................................................. 4-15 Figure 4-3 J-Integral versus Circumferential Flaw Extension - BASE_2 (Transition Forging) .............................................................................. 4-16 Figure 4-4 J-Integral versus Circumferential Flaw Extension - BASE_3 (Transition Forging) .............................................................................. 4-17 Figure 4-5 J-Integral versus Circumferential Flaw Extension - BASE_4 (Transition Forging) .............................................................................. 4-18 Figure 4-6 J-Integral versus Axial Flaw Extension - BASE_1 (Transition Forging) ................................................................................................ 4-19 Figure 4-7 J-Integral versus Axial Flaw Extension - BASE_2 (Transition Forging) ................................................................................................ 4-20 Figure 4-8 J-Integral versus Axial Flaw Extension - BASE_3 (Transition Forging) ................................................................................................ 4-21 Figure 4-9 J-Integral versus Axial Flaw Extension - BASE_4 (Transition Forging) ................................................................................................ 4-22 Figure 4-10 Path Lines of Interest (Outlet Nozzle Levels A and B) .......................... 4-29 Figure 4-11 J-Integral versus Axial Flaw Extension - Nozzle Corner (Outlet Nozzle Levels A and B)......................................................................... 4-36 Figure 4-12 J-Integral versus Axial Flaw Extension - Taper Transition (Outlet Nozzle Levels A and B)......................................................................... 4-37 Figure 4-13 J-Integral versus Circumferential Flaw Extension - Taper Transition (Outlet Nozzle Levels A and B) ............................................................. 4-38 Figure 4-14 J-integral versus Flaw Extension for Nozzle Corner (Outlet Nozzle Levels C and D) .................................................................................... 4-47 Figure 4-15 J-integral versus Flaw Extension for Nozzle Boss (Outlet Nozzle Levels C and D) .................................................................................... 4-48

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page x Figure 4-16 J-integral versus Flaw Extension for Nozzle Taper (Outlet Nozzle Levels C and D) .................................................................................... 4-49 Figure 4-17 Locations of Interest (Outlet Nozzle Levels C and D) ........................... 4-50 Figure 6-1 ONS Unit 1 72 EFPY Uncorrected P-T Limits for Normal Heatup with RCP Start at 170°F ........................................................................ 6-14 Figure 6-2 ONS Unit 1 72 EFPY Uncorrected P-T Limits for Normal Cooldown with DHRS Initiation at 190°F and RCP Stop at 155°F (CD2) .............. 6-15 Figure 6-3 ONS Unit 2 and ONS Unit 3 72 EFPY Uncorrected P-T Limits for Normal Heatup with RCP Start at 170°F ............................................... 6-16 Figure 6-4 ONS Unit 2 and ONS Unit 3 72 EFPY Uncorrected P-T Limits for Normal Cooldown with DHRS Initiation at 190°F and RCP Stop at 155°F (CD2).......................................................................................... 6-17 Figure 6-5 ONS Unit 3 Low Range 54 EFPY Cooldown Curves ............................ 6-18 Figure 6-6 ONS Unit 3 Wide Range 54 EFPY Cooldown Curves .......................... 6-19 Figure 7-1 Regions with Clad Forging Material ...................................................... 7-16 Figure 7-2 Reactor Vessel Shell High Stress Locations ......................................... 7-17 Figure 7-3 Postulated Surface Flaw Dimensions ................................................... 7-18 Figure 9-1 Reactor Pressure Vessel Support Assembly .......................................... 9-9 Figure 9-2 Reactor Pressure Vessel Support Embedment Detail .......................... 9-10 Figure 9-3 NUREG-1509 Figure 4-2, Preliminary Evaluation ................................. 9-11 Figure 9-4 NUREG-1509 Figure 4-4, Transition Temperature Approach ............... 9-12 Figure 9-5 RPV Skirt Temperatures from Anchor Bolts Location Model ................ 9-18 Figure 9-6 RPV Skirt Temperatures from Shear Pin Location Model ..................... 9-19 Figure 10-1 Sulzer Bingham Reactor Coolant Pump ............................................... 10-2 Figure 10-2 Westinghouse Type 93 A Reactor Coolant Pump................................. 10-3

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page xi Nomenclature Acronym Definition BTP Branch Technical Position CASS Cast Austenitic Stainless Steel CLB Current Licensing Basis CMTR Certified Material Test Report CVN Charpy V-Notch CVNp Preirradiation Charpy V-Notch CvUSE Charpy V-Notch Upper Shelf Energy DORT Discrete Ordinate Transport DPA Displacement per Atom EAF Environmentally-Assisted Fatigue EFPY Effective Full Power Years EMA Equivalent Margins Analysis EOL End of Life FME Fracture Mechanics Evaluation IE Irradiation Embrittlement INF Inlet Nozzle Forging IS Intermediate Shell Jd J deformation LEFM Linear Elastic Fracture Mechanics LNB Lower Nozzle Belt LS Lower Shell LTOP Low Temperature Overpressure Protection MCNP Monte Carlo N-Particle MUR Measurement Uncertainty Recapture ONF Outlet Nozzle Forging ONS Oconee Nuclear Station ONS Unit 1 Oconee Nuclear Station Unit 1 ONS Unit 2 Oconee Nuclear Station Unit 2 ONS Unit 3 Oconee Nuclear Station Unit 3 PTS Pressurized Thermal Shock P-T Pressure-Temperature RCP Reactor Coolant Pump RG Regulatory Guide RPV Reactor Pressure Vessel

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page xii RVID2 Reactor Vessel Integrity Database, Version 2.0.1 RVWG Reactor Vessel Working Group TLAA Time Limited Aging Analyses TR Topical Report UCC Underclad Cracking US Upper Shell USE Upper Shelf Energy

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page xiii ABSTRACT The purpose of this document is to provide supplemental information relative to the information reported in the subsequent license renewal application for Oconee Nuclear Station Units 1, 2, and 3 reactor pressure vessel (RPV) and reactor coolant pump (RCP) time limited aging analyses (TLAA) and aging management review topics that are reported in NUREG-2192, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants. NUREG-2192, Section 4.2.2.1, Table 4.7-1, and Section 4.3.3.1.2 include the following TLAA topics that apply to the Oconee RPVs.

  • Section 4.2.2.1.1Neutron Fluence
  • Table 4.7-1Reactor Pressure Vessel Underclad Cracking
  • Section 4.3.3.1.2Components Evaluated for CUFen Aging Management Review TopicsRPV Supports and RCPs The NRC has developed guidance for aging management review (further evaluation) of RPV supports for irradiation embrittlement through draft interim staff guidance contained in ADAMS ML20049H359. In addition, the aging management program for evaluation of thermal embrittlement of cast austenitic steel, NUREG-2191, Volume 2, XI.M12, contains acceptable methods to manage thermal embrittlement of reactor coolant pump casings.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page xiv The supplemental information provided in this ANP report is intended to assist the NRC with review of the Oconee RPV and RCP TLAA aging management review topics listed above relative to the applicable Review Procedures reported in NUREG-2192. The topics addressed in this ANP report are as follows.

  • Upper Shelf Energy, Section 3.0
  • Equivalent Margins Analysis of Oconee Unit 3 RPV Outlet Nozzle and Transition Forgings, Section 4.0
  • Pressurized Thermal Shock, Section 5.0
  • P-T Limits, Section 6.0
  • Fracture Mechanics Evaluation of Underclad Cracks, Section 7.0
  • Reactor Vessel Environmentally-Assisted Fatigue, Section 8.0
  • Irradiation Embrittlement of RPV Supports, Section 9.0
  • Reactor Coolant Pumps Thermal Embrittlement (CASS), Section 10.0
  • ONS Unit 2 and ONS Unit 3 RCP Flaw Tolerance Evaluation, Section 11.0

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 1-1

1.0 INTRODUCTION

The purpose of this document is to provide supplemental technical information relative to the information reported in the subsequent license renewal application for Oconee Nuclear Station Units 1, 2, and 3 reactor pressure vessel (RPV) time limited aging analyses (TLAA) and aging management review topics that are reported in NUREG-2192 (Reference 1-1), Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants. NUREG-2192, Section 4.2.2.1, Table 4.7-1, and Section 4.3.3.1.2 include the following TLAA topics that apply to the Oconee RPVs.

  • Section 4.2.2.1.1Neutron Fluence
  • Table 4.7-1Reactor Pressure Vessel Underclad Cracking
  • Section 4.3.3.1.2Components Evaluated for CUFen Aging Management Review TopicsRPV Supports and RCPs The NRC has developed guidance for aging management review (further evaluation) of RPV supports for irradiation embrittlement through draft interim staff guidance contained in ADAMS ML20049H359. In addition, the aging management program for evaluation of thermal embrittlement of cast austenitic steel, NUREG-2191, Volume 2, XI.M12, contains acceptable methods to manage thermal embrittlement of reactor coolant pump casings.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 1-2 Oconee Current Licensing Basis (CLB)

The above topics were all identified as TLAA for the Oconee Unit 1, Unit 2, and Unit 3 RPVs for 60 years and were evaluated in the following Oconee CLB documents.

  • Letter from Duke Energy Corporation forwarding application for renewal of operating licenses for the Oconee Nuclear Station, Unit Nos. 1, 2, and 3, U. S.

Nuclear Regulatory Commission, ACN: 9807200136, Fiche: A4344:001-A4347:255, July 6, 1998 (Reference 1-2).

  • The ONS LRA references BAW-2251A (Reference 1-4) relative to evaluation of various RPV TLAA for 60-years. All TLAA evaluations were reported at 48 EFPY and include upper shelf energy (10 CFR 50 Appendix G), pressurized thermal shock (10 CFR 50.61), underclad cracking, metal fatigue, and environmentally-assisted fatigue of the RPV.
  • OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3, Issuance of Amendments Regarding Revised Pressure-Temperature Limits (TAC NOS. MF0763, MF0764, AND MF0765) ADAMS Accession Number ML14041A093 (Reference 1-5).

- NRC approval of ONS P-T limits to 54 EFPY and reconciliation of USE from 48 EFPY to 54 EFPY. Upper shelf energy was reconciled from 48 EFPY, as reported in BAW-2251A, to 54 EFPY by the NRC.

- Reduces the Applicability for the RCS Heatup and Cooldown limit curves from 54 EFPY to 44.6 Effective Full Power Years (EFPY) for Unit 1, to 45.3 EFPY for Unit 2, and to 43.8 EFPY for Unit 3 based on updated reactor vessel (RV) material evaluations discussed in Section IV.1 of the MUR submittal.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 1-3

- Maintains RTPTS and Underclad Cracking at 48 EFPY and USE at 54 EFPY for all 3 units.

In accordance with NUREG-2192, Standard Review Plan for Subsequent License Renewal, aging management review recommendations for reactor pressure vessel supports are addressed in Section 3.5.2.2.2.6, and Table 3.5-1, Item 097. Following review of the first subsequent license renewal applications for Turkey Point 3 and 4, Surry 1 and 2, and Peach Bottom, the NRC determined that additional guidance was needed to clarify aging management review expectations by the NRC for Reactor Pressure Vessel (RPV) supports for subsequent license renewal. As such, the NRC has issued draft interim staff guidance (Reference 1-7) to revise the SLR Standard Review Plan NUREG-2192 to add a new Section 3.5.2.2.2.7 that includes additional guidance for evaluation of RPV supports relative to irradiation embrittlement. This topic is addressed in Section 9.0 of this report.

EFPY Based on accrued EFPY through Cycles 31, 29, and 30 for Oconee Units 1 through 3 and assuming breaker-to-breaker operation and no outages per cycle (Capacity Factor

= 1) to 80 years of operation, the bounding projected EFPY for 80 years for each Oconee Unit is less than 72 EFPY. Therefore, the Oconee Nuclear Station RPV TLAA evaluations are completed to 72 EFPY for SLR.

Measurement Uncertainty Recapture All TLAA for subsequent license renewal reported in this ANP document consider the revised operating conditions (e.g., 1.64% increase in power) associated with MUR as reported in Reference 1-6. MUR is assumed at the beginning of Cycle 30 for ONS Unit 1, Cycle 29 for ONS Unit 2, and Cycle 29 for ONS Unit 3. At present, MUR has not been initiated and each unit is currently operating in Cycles 32, 30, and 31, respectively.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 1-4 The supplemental information provided in this ANP report is intended to assist the NRC with review of the Oconee RPV and RCP topics above relative to the applicable Review Procedures reported in NUREG-2192. The topics addressed in this ANP report are as follows.

  • Upper Shelf Energy, Section 3.0
  • Equivalent Margins Analysis of Oconee Unit 3 RPV Outlet Nozzle and Transition Forgings, Section 4.0
  • Pressurized Thermal Shock, Section 5.0
  • P-T Limits, Section 6.0
  • Fracture Mechanics Evaluation of Underclad Cracks, Section 7.0
  • Reactor Vessel Environmentally-Assisted Fatigue, Section 8.0
  • Irradiation Embrittlement of RPV Supports, Section 9.0
  • Reactor Coolant Pumps Thermal Embrittlement (CASS), Section 10.0
  • ONS Unit 2 and ONS Unit 3 RCP Flaw Tolerance Evaluation, Section 11.0 1.1 References for Section 1.0 1-1. NUREG-2192, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants 1-2. Letter from Duke Energy Corporation forwarding application for renewal of operating licenses for the Oconee Nuclear Station, Unit Nos. 1, 2, and 3, U. S.

Nuclear Regulatory Commission, ACN: 9807200136, Fiche: A4344:001-A4347:255, July 6, 1998 (ADAMS Accession No. ML15254A151 and ML15112A661) 1-3. NUREG-1723, Safety Evaluation Report Related to the License Renewal of Oconee Nuclear Station, Units 1, 2, and 3 (ADAMS Accession No. ML003695154)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 1-5 1-4. BAW-2251A, Demonstration of Management of Aging Effects for the Reactor Vessel, ADAMS Accession Numbers ML20212G894 and ML20212G911 1-5. OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3, Issuance of Amendments Regarding Revised Pressure-Temperature Limits (TAC NOS. MF0763, MF0764, and MF0765), ADAMS Accession Number ML14041A093 1-6. ONS License Amendment Request for Measurement Uncertainty Recapture Power Uprate, February 19, 2020, ADAMS Accession Number ML20050D379 and NRC SER, ADAMS Accession Number ML20335A001 1-7. RV Supports ISG SLR Document Changes: Add FE Section 3.5.2.2.2.7 and AMR Items for Irradiation Embrittlement of Reactor Vessel (RV) Steel Supports and Other Steel Structural Support Components near RV, ADAMS Accession Number ML20049H359

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 2-1 2.0 RPV NEUTRON FLUENCE PROJECTIONS 2.1 Introduction Neutron fluence is defined as the time integral of the neutron flux density, expressed as number of particles (neutrons) per cm2. Neutron fluence is used as an input to quantify the change in material properties of the reactor vessel (traditional beltline and extended beltline) regions, as required by 10 CFR 50, Appendix G, 10 CFR 50.61, over the life of the plant. Because these neutron embrittlement analyses are evaluated for the plant's service lifetime (10 CFR 50, Appendix G, IV., 1.) or end-of-life (10 CFR 50.61(a)(6)),

they are identified as time-limited aging analyses.

2.2 Regulatory Guidance for Subsequent License Renewal The regulatory guidance for NRC review of neutron fluence calculations performed in accordance with 10 CFR 54.21(c)(1)(ii) is reported in NUREG-2192, Section 4.2.3.1.1.2 (Reference 2-1).

The reviewer confirms that the applicant adequately reevaluated its RPV neutron fluence analysis for the subsequent period of extended operation. As part of its review, the review confirms that the applicant identifies (a) the neutron fluence for each beltline material at the end of the subsequent period of extended operation, (b) the NRC staff-approved methodology used to determine the neutron fluence or submits the methodology for NRC staff review, and (c) whether the methodology is consistent with the guidance in NRC RG 1.190.

Guidance for NRC review of acceptable methods and assumptions for determining reactor vessel neutron fluence for subsequent license renewal is provided in NUREG-2191 (Reference 2-2), X.M2, Neutron Fluence Monitoring as supplemented by the final NRC interim staff guidance for X.M2 (Reference 2-10).

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 2-2 Guidance on acceptable methods and assumptions for determining reactor vessel neutron fluence is described in the U.S. Nuclear Regulatory Commission (NRC)

Regulatory Guide (RG) 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. The methods developed and approved using the guidance contained in RG 1.190 are specifically intended for determining neutron fluence in the region of the RPV close to the active fuel region of the core and are not intended to apply to vessel regions significantly above and below the active fuel region of the core, nor to RVI components. Therefore, the use of RG 1.190-adherent methods to estimate neutron fluence for the RPV regions significantly above and below the active fuel region of the core and RVI components may require additional justification, even if those methods were approved by the NRC for RPV neutron fluence calculations. This program monitors in-vessel or ex-vessel dosimetry capsules and evaluates the dosimetry data, as needed. Additional in-vessel or ex-vessel dosimetry capsules may be needed when the reactor surveillance program has exhausted the available capsules for in-vessel exposure.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 2-3 2.3 Methodology The calculation based fluence analysis methodology contained in BAW-2241P-A, Revision 2 (Reference 2-3) is used to predict the 72 EFPY fluence at reactor vessel shell locations (both traditional beltline and extended beltline) for Oconee Units 1, 2, and

3. BAW-2241P-A, Revision 2 is fully compliant with Regulatory Guide 1.190 (Reference 2-4) for RPV traditional beltline locations. Traditional beltline locations for the Oconee RPVs are defined as those items reported in the NRC approved topical report BAW-2251A (Reference 2-5), Tables 4-5 through 4-7, and Tables A-2 through A-4, which is referenced in the Oconee 60-year license renewal application. The RPV beltline items identified in BAW-2251A are consistent with the beltline items reported in the NRCs reactor vessel integrity database (RVID2, Reference 2-6). For SLR, extended beltline items are defined as those RPV locations, in addition to traditional beltline locations, that will receive projected neutron fluence values greater than 1.0E+17 n/cm2 (E > 1 MeV) at the end of the subsequent period of extended operation (i.e., 72 EFPY for Oconee). A threshold of >1.0E+17 n/cm2 (E > 1 MeV) is established through RIS 2014-11 (Reference 2-7).

For the Oconee Nuclear Station, neutron transport calculations, using the methodology from BAW-2241P-A, Revision 2, were completed for Cycles 27-29 for ONS Unit 1, Cycles 25-28 for ONS Unit 2, and Cycles 26-28 for ONS Unit 3, and used to project fast neutron fluence at the reactor vessel shell (i.e., traditional beltline and extended beltline locations) to 72 effective full power years (EFPYs). A measurement uncertainty recapture (MUR) power uprate is conservatively factored in at 2% and is assumed at the beginning of Cycle 30 for ONS Unit 1, Cycle 29 for ONS Unit 2, and Cycle 29 for ONS Unit 3.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 2-4 Neutron fluence projections at 72 EFPY are provided for ONS Unit 1, ONS Unit 2, and ONS Unit 3 in Table 2-1, Table 2-2, and Table 2-3, respectively. Inside wetted surface fluence at 72 EFPY is reported for all RPV shell locations that exceed 1.0E+17 n/cm2 (E > 1.0 MeV). Fluence is attenuated through the thickness of the shell by taking the ratio of dpa at the depth in question to the dpa at the inner surface, as permitted by Regulatory Guide 1.99, Revision 2, Section 1.1. Based on these tables, traditional beltline and extended beltline items for SLR are as follows.

ONS Unit 1 Traditional Beltline and Extended Beltline RPV Items (Figure 2-1)

  • Forgings are ASTM A 508 Class 2

- RPV outlet nozzles (2) and RPV inlet nozzles (4)**

- Lower nozzle belt forging (NBF) (AHR 54: ZV 2861)

- Transition forging (122S347VA1)** The transition forging is also known by Framatome as the dutchman forging. This part is identified as transition forging in this ANP report.

  • Plates are ASTM A 302B, Modified

- Intermediate shell (C2197-2)

- Upper shell (C3265-1 and C3278-1)

- Lower shell (C2800-1 and C2800-2)

  • Linde 80 Welds

- RPV inlet and outlet nozzle to NBF welds**

- Lower NBF to intermediate shell circumferential weld (SA-1135)

- Intermediate shell axial welds (SA-1073)

- Intermediate shell to upper shell circumferential weld (SA-1229, 61% ID; WF25, 39% OD)

- Upper shell axial welds (SA-1493)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 2-5

- Upper shell to lower shell circumferential weld SA-1585

- Lower shell axial welds (SA-1430 and SA-1426)

- Lower shell forging to transition forging circumferential weld (WF-9)

    • Extended beltline items ONS Unit 2 Traditional Beltline and Extended Beltline RPV Items (Figure 2-2)
  • Forgings are all ASTM A 508 Class 2

- RPV outlet nozzles (2) and RPV inlet nozzles (4)**

- Lower nozzle belt forging (AMX 77: 123T382)

- Upper shell forging (AAW 163:3P2359)

- Lower shell forging (AWG 164: 4P1885)

- Transition forging (122T293VA1)**

  • Linde 80 Welds

- RPV inlet and outlet nozzle to NBF welds**

- Lower nozzle belt forging to upper shell forging circumferential weld (WF-154)

- Upper shell forging to lower shell forging circumferential weld (WF-25)

- Lower shell forging to transition forging circumferential weld (WF-112)

    • Extended beltline items ONS Unit 3 Traditional Beltline and Extended Beltline RPV Items (Figure 2-3)
  • Forgings are all ASTM A 508 Class 2

- RPV outlet nozzles (2) and RPV inlet nozzles (4)**

- Lower nozzle belt forging (4680)

- Upper shell forging (AWS 192: 522314)

- Lower shell forging (ANK 191: 522194)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 2-6

- Transition forging (417543-1)**

  • Linde 80 Welds

- RPV inlet and outlet nozzle to USF welds**

- Lower nozzle belt forging to upper shell forging circumferential weld (WF-200)

- Upper shell forging to lower shell forging circumferential weld (WF-67, 75%

ID; WF-70, 25% OD)

- Lower shell forging to transition forging circumferential weld (WF 169-1)

    • Extended beltline items As the lifetime of the reactor vessel for each Oconee unit is extended to 80-years, reactor vessel regions that may be susceptible to reduction of fracture toughness may extend beyond traditional beltline locations. As such, estimation of the fluence in reactor vessel regions adjacent to those that surround the effective height of the active core (traditional beltline) is required and the accepted fluence threshold for irradiation damage of reactor vessel materials, relative to the monitoring of limiting materials, is RPV inside wetted surface fluence greater than 1.0E+17 n/cm2 (E > 1.0 MeV)

(Reference 2-7). For locations of interest beyond the reactor vessel regions that surround the effective height of the active core, the model developed using the methodology from BAW-2241P-A, Revision 2 is extended to provide estimates of the fluence rates, and calculated fluence at 72 EFPY, at the RPV nozzle and transition forging.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 2-7 An adjustment of the method is employed to obtain best estimate cumulative (E > 1.0 MeV) fluence values for extended beltline locations. Specifically, a displacement per atom (dpa) adjustment is used wherein a ratio of the discrete ordinate transport (DORT) calculated dpa at the extended beltline location of interest (e.g., RPV nozzle) to the discrete ordinate transport (DORT) calculated dpa at the circumferential weld that connects the upper shell to the lower shell (SA-1585 for Unit 1, WF-25 for Unit 2, and WF-67 for Unit 3), is multiplied by the fluence at the circumferential weld to obtain the fluence at the extended beltline location of interest (e.g., RPV nozzle). The equation for dpa adjustment is as follows.

= x Where f is fluence n/cm2 (E > 1.0 MeV)

This formulation is also used to obtain the fluence at the 1/4T and 3/4T locations reported in Table 2-1 through Table 2-3 versus the Regulatory Guide 1.99, Revision 2 (Reference 2-8) attenuation formulation.

= ( 0.24 )

Figure 2-4 and Figure 2-5 illustrate the fluence rate (flux) attenuation with and without using the technique above. Fast flux (E > 1 MeV) is obtained from the DORT calculated fluence as a function of thickness through the RPV (traditional beltline and extended beltline regions). [

]

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 2-8 This dpa adjustment factor method is shown to be conservative (i.e., provides a positive bias) for extended beltline nozzle belt locations relative to recent 3-D Monte Carlo NParticle (MCNP) calculations reported in ANP-10348P, Revision 0, Figure 1-1, reproduced as Figure 2-6 below (Reference 2-9). That is, Framatome utilized a complex system of computer codes to analyze reactor regions beyond the traditional beltline that includes SOLIDWORKS3, VICTORIA 4 (a plugin for ANSYS 5), ADVANTG 6 and MCNP7; this new system is known as SVAM.

2.4 Summary and Conclusions The 72 EFPY fluence analyses have been projected to the end of the subsequent period of extended operation, and these projections are used as inputs to the RPV neutron embrittlement TLAA evaluations. The methodology used to generate Oconee 72 EFPY RPV fluence is in accordance with Regulatory Guide 1.190 compliant methodology BAW-2241P-A, Revision 2. These Oconee 72 EFPY calculations meet the RG 1.190 uncertainty requirements of 20% (1 ) or less for traditional beltline locations. For extended beltline locations, i.e., RPV outlet nozzle forgings, RPV inlet nozzle forgings, and transition forging, 72 EFPY fluence was estimated using the RG 1.190 compliant BAW-2241P-A, Revision 2 methodology that was modified by extending the model to include extended beltline locations and to implement dpa adjustment as described above. Therefore, the 72 EFPY fluence reported herein for extended beltline locations is justified for use in compliance with the Reactor Pressure Vessel Neutron Embrittlement Analyses TLAA guidance reported in NUREG-2192 for subsequent license renewal.

2.5 References for Section 2.0 2-1. NUREG-2192, Standard Review Plan Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants 2-2. NUREG-2191, Volume 2, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 2-9 2-3. Framatome Inc. Topical Report BAW-2241P-A, Revision 2, Fluence and Uncertainty Methodologies, ADAMS Accession No. ML073310655 (Proprietary),

ML073310660 (Non-Proprietary) 2-4. U.S. Nuclear Regulatory Commission, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, Regulatory Guide 1.190 (available @ https://www.nrc.gov/docs/ML0108/ML010890301.pdf) 2-5. Framatome Inc. Topical Report BAW-2251A, Demonstration of the Management of Aging Effects for the Reactor Vessel ADAMS, Accession Numbers ML20212G894 and ML20212G911 2-6. Reactor Vessel Integrity Database Version 2.0.1, https://www.nrc.gov/reactors/operating/ops-experience/reactor-vessel-integrity/database-overview.html 2-7. NRC Regulatory Issue Summary 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components 2-8. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials 2-9. Framatome Topical Report ANP-10348P and NP, Revision 0, Fluence Methodologies for SLR, ADAMS Accession No. ML20223A019 2-10. SLR-ISG-2021-02-MECHANICAL, Updated Aging Management Criteria for Mechanical Portions of Subsequent License Renewal Guidance, ADAMS Accession No. ML20181A434

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 2-10 Figure 2-1 ONS Unit 1 Reactor Vessel Shell

'I t

LE FORGING ELD SA 036 DSA 110 1 I G I WELD SA 1135 I I ER lE IATE S ELL C 1Q7-E DSA Q I S J ~ 6 1% ;

F 25 O UTS IDE 39%

SA 1493 I B.0S 1585 I

I I

I I

SA LOWERS ELL C 800- 1 AN C 800-

\

RA S I 10 22S347VA I

I DSA 263

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 2-11 Figure 2-2 ONS Unit 2 Reactor Vessel Shell


O UTL ET OZ2iLE FO RG[ G SA 1135 U PPE S HELJL FO RG I MG AAW 63; 3 Z3'5 Q W E UD W F 25 L01NER S E LJL FO GINIG A W G 1 fli4 ; 4 8 85

\

W EIL SA 147 1 A D SA 14184 LOWE EAD C 2 663- 1

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 2-12 Figure 2-3 ONS Unit 3 Reactor Vessel Shell UPP ER OZZLE B ELT FORGING 467Q OZZLE FORGING i

I WEUDWF25

\1 I

LOWE R OZZLE BBL FO GING 468[)

GING WELD WF 67 INSIDE 75%;

WF 7D Ol.11 S DE 25%

G, WBLD W F 169- 1 (1/2f..........- ~ - ~ - - - . ~~ TRA SI 10."-I FOF<<3 1 G, 4 17543- i

- . WE W F 70

~ l OWE EAD 13-0078-2

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 2-13 Figure 2-4 Attenuated Neutron Fluence Comparison - Core Center

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 2-14 Figure 2-5 Attenuated Neutron Fluence Comparison - Nozzle Region

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 2-15 Figure 2-6 ANP-10348P, Figure 1-1, Calculations and Measurements

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 2-16 Table 2-1 72 EFPY ONS Unit 1 RPV Shell Locations with Fluence > 1.0E17 n/cm2 (E > 1.0 MeV)

Base Inside Wetted Metal/Clad Material ID and/or Surface 1/4T 3/4T Reactor Vessel Material Interface 2 2 Heat Number Fluence (n/cm ) (n/cm )

2 Fluence (n/cm ) 2 (n/cm )

19 19 19 18 Upper Shell (US) Plates C3265-1, C3278-1 2.10x10 2.02x10 1.32x10 4.35x10 19 19 19 18 Lower Shell (LS) Plates C2800-1, C2800-2 2.10x10 2.02x10 1.31x10 4.32x10 19 19 19 18 Intermediate Shell (IS) Plates C2197-2 1.85x10 1.78x10 1.15x10 3.72x10 Lower Nozzle Belt (LNB) 18 18 18 17 AHR 54; ZV 2861 2.68x10 2.60x10 1.81x10 7.91x10 Forging Bottom of 12 Thickness of 18 18 17 17 AHR 54; ZV 2861 1.48x10 1.43x10 8.82x10 4.02x10 LNB Forging Top of 8.438 Thickness of 18 18 18 17 AHR 54; ZV 2861 2.19x10 2.06x10 1.41x10 7.91x10 LNB Forging Bottom of 8.438 Thickness of 17 17 17 17 C2800-1, C2800-2 9.93x10 9.59x10 7.14x10 3.80x10 LS 17 17 17 17 Transition Forging* 122S347VA1 2.70x10 2.52x10 2.17x10 1.96x10 Inlet Nozzle Forging (INF) 17 NA 1.11x10 NA NA NA Postulated Flaw Outlet Nozzle Forging (ONF) 17 NA 2.20x10 NA NA NA Postulated Flaw 8T1762; 299L44; 17 17 17 17 LNB to Bottom of ONF Welds 3.49x10 3.38x10 2.29x10 2.52x10 8T1554B 8T1762; 299L44; 17 17 17 17 LNB to Bottom of INF Welds 1.62x10 1.57x10 1.13x10 1.73x10 8T1554B 18 18 18 17 LNB to IS Circ. Weld SA-1135 2.91x10 2.82x10 1.97x10 8.80x10 19 19 18 18 IS Long. Welds (Both) SA-1073 1.38x10 1.33x10 8.75x10 2.97x10 19 19 19 18 IS to US Circ. Weld SA-1229; WF 25 1.86x10 1.79x10 1.16x10 3.79x10 19 19 19 18 LS to US Circ. Weld SA-1585 2.05x10 1.97x10 1.29x10 4.30x10 19 19 18 18 US Long. Welds (Both) SA-1493 1.36x10 1.31x10 8.58x10 2.94x10 19 19 19 18 LS Long. Weld (Both) SA-1426, SA-1430 1.68x10 1.62x10 1.07x10 3.59x10 17 17 17 17 LS to Transition Circ. Weld WF-9 2.70x10 2.52x10 2.17x10 1.96x10 Notes

  • Values are the same as the LS to Transition Forging Circumferential Weld

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 2-17 Table 2-2 72 EFPY ONS Unit 2 RPV Shell Locations with Fluence > 1.0E17 n/cm2 (E > 1.0 MeV)

Base Inside Wetted Metal/Clad Material ID and/or Surface 1/4T 3/4T Reactor Vessel Material Interface 2 2 Heat Number Fluence (n/cm ) (n/cm )

2 Fluence (n/cm ) 2 (n/cm )

19 19 19 18 Upper Shell (US) Forging AAW 163; 3P2359 1.98x10 1.90x10 1.24x10 4.14x10 019 19 19 18 Lower Shell (LS) Forging AWG 164; 4P1885 1.97x1 1.89x10 1.24x10 4.10x10 Lower Nozzle Belt (LNB) 19 19 19 18 AMX 77; 123T382 1.74x10 1.67x10 1.09x10 3.52x10 Forging Bottom of 12 Thickness of 18 18 18 17 AMX 77; 123T382 1.70x10 1.66x10 1.04x10 3.70x10 LNB Forging Top of 8.438 Thickness of 18 18 18 17 AMX 77; 123T382 2.42x10 2.32x10 1.62x10 7.45x10 LNB Forging Bottom of 8.438 Thickness of 17 17 17 17 AWG 164; 4P1885 9.19x10 8.87x10 6.63x10 3.58x10 LS 17 17 17 17 Transition Forging* 122T293VA1 2.50x10 2.34x10 2.02x10 1.84x10 Inlet Nozzle Forging (INF) 17 NA 1.03x10 NA NA NA Postulated Flaw Outlet Nozzle Forging (ONF) 17 NA 2.04x10 NA NA NA Postulated Flaw LNB to Bottom of ONF Welds [ ] 3.24x10 17 3.14x10 17 2.13x10 17 2.41x10 17 LNB to Bottom of INF Welds [ ] 1.50x10 17 1.46x10 17 1.06x10 17 1.67x10 17 19 19 19 18 LNB to US Circ. Weld WF 154 1.75x10 1.68x10 1.09x10 3.59x10 19 19 19 18 LS to US Circ. Weld WF 25 1.92x10 1.85x10 1.22x10 4.08x10 17 17 17 17 LS to Transition Circ. Weld WF 112 2.50x10 2.34x10 2.02x10 1.84x10 Notes

  • Values are the same as the LS to Transition Forging Circumferential Weld

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 2-18 Table 2-3 72 EFPY ONS Unit 3 RPV Shell Locations with Fluence > 1.0E17 n/cm2 (E > 1.0 MeV)

Base Inside Wetted Metal/Clad Material ID and/or Surface 1/4T 3/4T Reactor Vessel Material Interface 2 2 Heat Number Fluence (n/cm ) (n/cm )

2 Fluence (n/cm ) 2 (n/cm )

19 19 19 18 Upper Shell (US) Forging AWS 192; 522314 2.06x10 1.98x10 1.30x10 4.30x10 19 19 19 18 Lower Shell (LS) Forging ANK 191; 522194 2.05x10 1.97x10 1.29x10 4.26x10 Lower Nozzle Belt (LNB) 19 19 19 18 4680 1.81x10 1.74x10 1.13x10 3.66x10 Forging Bottom of 12 Thickness of 18 18 18 17 4680 1.81x10 1.77x10 1.11x10 3.87x10 LNB Forging Top of 8.438 Thickness of 18 18 18 17 4680 2.58x10 2.48x10 1.72x10 7.86x10 LNB Forging Bottom of 8.438 Thickness of 17 17 17 17 ANK 191; 522194 9.80x10 9.45x10 7.06x10 3.75x10 LS 17 17 17 17 Transition Forging* 417543-1 2.68x10 2.50x10 2.15x10 1.95x10 Inlet Nozzle Forging (INF) 17 NA 1.14x10 NA NA NA Postulated Flaw Outlet Nozzle Forging (ONF) 17 NA 2.21x10 NA NA NA Postulated Flaw LNB to Bottom of ONF Welds [ ] 3.50x10 17 3.39x10 17 2.30x10 17 2.51x10 17 LNB to Bottom of INF Welds

[ 1.62x10 17 1.58x10 17 1.14x10 17 1.73x10 17

]

19 19 19 18 LNB to US Circ. Weld WF 200 1.82x10 1.75x10 1.14x10 3.74x10 19 19 19 18 LS to US Circ. Weld WF 67; WF 70 2.01x10 1.93x10 1.27x10 4.24x10 17 17 17 17 LS to Transition Circ. Weld WF 169-1 2.68x10 2.50x10 2.15x10 1.95x10 Notes

  • Values are the same as the LS to Transition Forging Circumferential Weld

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-1 3.0 UPPER SHELF ENERGY 3.1 Introduction Appendix G of 10 CFR 50 (Reference 3-1) requires that reactor vessel beltline materials have Charpy upper-shelf energy ... of no less than 75 ft-lb (102J) initially and must maintain Charpy upper-shelf energy throughout the life of the vessel of no less than 50 ft-lb (68J) unless it is demonstrated in a manner approved by the Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, that lower values of Charpy USE will provide margins of safety against fracture equivalent to those required by Appendix G of Section XI of the ASME Code. Since USE reduction is a function of the end of license fluence, which is associated with the 60-year licensed operating period, the USE calculations meet the criteria of 10 CFR 54.3(a) and have been identified as Time Limited Aging Analyses (TLAA) requiring evaluation for 80 years.

Current Licensing Basis Compliance with the requirements of 10 CFR 50 Appendix G by Duke Energy for 60-years of operation is reported in Section 1.0 of this ANP report.

3.2 Regulatory Guidance for Subsequent License Renewal The regulatory guidance for NRC review of upper-shelf energy evaluations performed in accordance with 10 CFR 54.21(c)(1)(ii) is reported in NUREG-2192, Section 4.2.3.1.2.2 (Reference 3-2) and is repeated below.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-2 The documented results of the revised USE analysis (or revised EMA analysis, as applicable) based on the projected neutron fluence at the end of the subsequent period of extended operation are reviewed for compliance with 10 CFR Part 50, Appendix G.

The applicant may use NRC RG 1.99 Rev. 2 as the basis for using the 1/4T neutron fluence values for the reactor vessel beltline components (as projected to the end of the SLR period) to project the USE values for the reactor vessel beltline components at the end of the subsequent period of extended operation. The applicant also may use ASME Code Section XI Appendix K for the purpose of performing an equivalent margins analysis to demonstrate that adequate protection for ductile failure is maintained to the end of the subsequent period of extended operation.

The NRC staff reviews the applicants methodology for this evaluation. Branch Technical Position (BTP) 5-3, Fracture Toughness Requirements, in Standard Review Plan, Section 5.3.2, Pressure Temperature Limits, Upper-Shelf Energy, and Pressurized Thermal Shock, provides additional NRC positions on estimations of USE values for RPV beltline materials.

The NRC staff confirms that the applicant has provided sufficient information for all USE and/or equivalent margins analysis calculations for the subsequent period of extended operation as follows:

The applicant identifies the neutron fluence at the 1/4T location for each beltline material at the expiration of the subsequent period of extended operation.

To confirm that the USE analysis meets the requirements of Appendix G of 10 CFR Part 50 at the end of the subsequent period of extended operation, the NRC staff determines whether:

1. For each beltline material, the applicant provides the unirradiated USE and the projected USE at the end of the subsequent period of extended operation, and whether the drop in USE was determined using the limit lines in Figure 2 of NRC RG 1.99, Rev 2, based on the material copper content, or from surveillance data.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-3

2. If an equivalent margins analysis is used to demonstrate compliance with the USE requirements in Appendix G of 10 CFR Part 50, the applicant provides the analysis or identifies an NRC-approved topical report that contains the analysis which is applicable to the subsequent period of extended operation.

Information the NRC staff considers to assess the equivalent margins analysis includes the unirradiated USE (if available) for the material, its copper content, the neutron fluence (1/4T and at 1-inch depth), the projected SLR USE, the operating temperature in the downcomer at full power, the vessel radius, the vessel wall thickness, the J-applied analysis for Service Level C and D, the vessel accumulation pressure, and the vessel bounding heat-up/cool-down rate during normal operation.

3.3 Methodology Appendix G of 10 CFR 50 requires that reactor vessel beltline materials have Charpy upper-shelf energy ... of no less than 75 ft-lb (102J) initially and must maintain Charpy upper-shelf energy throughout the life of the vessel of no less than 50 ft-lb (68J) unless it is demonstrated in a manner approved by the Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, that lower values of Charpy USE will provide margins of safety against fracture equivalent to those required by Appendix G of Section XI of the ASME Code. Since USE reduction is a function of the end-of-life fluence, which is associated with the 60-year licensed operating period, the USE calculations meet the criteria of 10 CFR 54.3(a) and have been identified as Time Limited Aging Analyses (TLAA) requiring evaluation for 80 years.

The 72 EFPY Charpy V-notch Upper Shelf Energy (CvUSE) projections are calculated at the 1/4T thickness (1/4T) for Oconee Nuclear Station Units 1, 2, and 3 base and weld metals that have peak inside wetted surface fluence >1.0E+17 n/cm2 (E > 1 MeV) at 72 Effective Full Power Years (EFPY). Fluence at the 1/4T location for traditional and extended beltline items is reported in Table 2-1, Table 2-2, and Table 2-3 for ONS Units 1, 2, and 3, respectively.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-4 3.3.1 CvUSE at 72 EFPY Based on Regulatory Guide 1.99 Revision 2 Regulatory Guide 1.99, Revision 2 (Reference 3-3) provides two methods for determining CvUSE:

  • Position 1.2 for material that does not have surveillance data available, and
  • Position 2.2 for material that does have surveillance data.

For Position 1.2, the percent drop in CvUSE, for a stated copper content and neutron fluence, is determined by reference to Figure 2 of Regulatory Guide 1.99, Revision 2.

This percentage drop is applied to the initial CvUSE to obtain the adjusted CvUSE. For Position 2.2, the percent drop in CvUSE is determined by plotting the available data on Figure 2 and fitting the data with a line drawn parallel to the existing lines that bound all the plotted points. These calculations are reported in Section 3.4 below.

Material property input required for calculating 72 EFPY upper shelf energy using Positions 1.2 and 2.2, Figure 2, includes copper content (weight percent) of base metal and weld metal, and initial unirradiated upper shelf energy. Fluence is reported in Section 2.0 and material properties are addressed in Section 3.4.1 below.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-5 3.3.2 Equivalent Margins Analyses The ASME Code,Section XI, Appendix K acceptance criteria for Levels A through D Service Loadings for Oconee Units 1, 2, and 3 reactor vessel traditional beltline and extended beltline Linde 80 welds are satisfied and are reported in Framatome reports BAW-2192, Revision 0, Supplement 1P-A, Revision 0, Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessels of B&W Owners Reactor Vessel Working Group for Levels A and B Service Loadings (Reference 3-4), and BAW-2178, Revision 0, Supplement 1PA, Revision 0, Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessels of B&W Owners Reactor Vessel Working Group for Levels C and D Service Loadings (Reference 3-5). These reports contain equivalent margins analyses for all Oconee Unit 1, 2, and 3 Linde 80 welds identified as traditional beltline and extended beltline material for 80 years. Confirmation that the Oconee Nuclear Station reactor pressure vessels are bounded by these topical reports is provided in Section 3.5 below.

The ASME Code,Section XI, Appendix K acceptance criteria for Levels A through D Service Loadings for Oconee Unit 3 reactor vessel outlet nozzle forgings and the transition forging are satisfied and are reported in Section 4.0 of this ANP report.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-6 3.4 Traditional Beltline and Extended Beltline 72 EFPY Upper Shelf Energy The 72 EFPY Regulatory Guide 1.99, Revision 2 (Position 1.2) USE values of the vessel materials are predicted using the corresponding 1/4T fluence projection, copper content of the materials, and Figure 2 in Regulatory Guide 1.99, Revision 2. The predicted Position 2.2 USE values are determined for the reactor vessel materials that are contained in the surveillance program by using surveillance data along with the corresponding 1/4T fluence projection. The 1/4T fluence projections are 1/4T dpa adjusted values reported in Table 2-1 for ONS Unit 1, Table 2-2 for ONS Unit 2, and Table 2-3 for ONS Unit 3 unless the attenuation of fluence from the inside wetted surface using RG 1.99, Revision 2, Equation (3), exceed the 1/4T dpa adjusted values. The projected USE values were calculated to determine if the ONS Unit 1, Unit 2, and Unit 3 traditional beltline and extended beltline materials remain above the 50 ft-lb limit at 72 EFPY and are reported for Oconee Units 1, 2, and 3 in Table 3-6, Table 3-7, and Table 3-8, respectively.

All ONS Unit 1 reactor vessel traditional beltline and extended beltline plate and forging materials maintain a USE value greater than 50 ft-lbs at 72 EFPY. All ONS Unit 1 reactor vessel traditional beltline and extended beltline Linde 80 welds have USE values less than 50 ft-lbs at 72 EFPY and require an equivalent margins analysis to demonstrate that lower values of Charpy USE will provide margins of safety against fracture equivalent to those required by Appendix G of Section XI of the ASME Code.

All ONS Unit 2 reactor vessel traditional beltline and extended beltline forging materials maintain a USE value greater than 50 ft-lbs at 72 EFPY. All ONS Unit 2 reactor vessel traditional beltline and extended beltline Linde 80 welds have USE values less than 50 ft-lbs at 72 EFPY and require an equivalent margins analysis to demonstrate that lower values of Charpy USE will provide margins of safety against fracture equivalent to those required by Appendix G of Section XI of the ASME Code.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-7 All ONS Unit 3 reactor vessel traditional beltline and extended beltline forging materials maintain a USE value greater than 50 ft-lbs at 72 EFPY with the exception of the RPV outlet nozzle forgings and the transition forging. All reactor vessel traditional beltline and extended beltline Linde 80 welds have USE values less than 50 ft-lbs at 72 EFPY.

The RPV outlet nozzle forgings, transition forging, and traditional beltline and extended beltline Linde 80 welds require an equivalent margins analysis to demonstrate that lower values of Charpy USE will provide margins of safety against fracture equivalent to those required by Appendix G of Section XI of the ASME Code.

3.4.1 Material Properties Including Copper and Initial Upper-Shelf Energy Material properties required for the calculation of 72 EFPY USE are reported in Table 3-6, Table 3-7, and Table 3-8 and include copper content (weight %) and initial unirradiated upper-shelf energy (ft-lbs). For traditional beltline locations, the copper content and initial upper shelf energy are consistent with the current licensing basis (Section 1.0, Reference 1-2 and Reference 1-5) with some exceptions as reported in Section 3.4.2. For extended beltline locations (i.e., RPV inlet and outlet nozzles and associated welds and the transition forging), material properties have not been reported to the NRC and initial copper content and initial upper shelf energy for SLR are obtained as follows.

Initial Upper Shelf Energies A generic upper shelf energy for Linde 80 welds is reported herein at [ ]

and is obtained as follows. The upper-shelf energy values for Linde 80 material in the Reactor Vessel Working Group (RVWG) data base [ ] were compiled and are reported in Table 3-1. The mean was found to be [ ], the standard deviation is [ ] and the one-sided tolerance factor (k), 95% confidence for 95% coverage, is [ ]. The generic value is calculated to be [ ] (mean less k*sigma).

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-8 Table 3-1 Linde 80 Initial CvUSE Values Generic initial upper shelf energy is reported for the ONS Unit 1 and ONS Unit 2 RPV inlet and outlet nozzle forgings and transition forgings and the ONS Unit 3 RPV inlet nozzle forgings procured from Bethlehem Steel at [ ] and is obtained from the data reported in Table 3-2 below. The upper-shelf energy values in the weak direction for all such material in the RVWG data base [ ] are reported in Table 3-2 . The mean was found to be [ ], the standard deviation is

[ ] and the one-sided tolerance factor (k), 95% confidence for 95% coverage, is

[ ]. The generic value is calculated to be [ ] (mean less k*sigma).

The ONS Unit 3 RPV outlet nozzle forgings and transition forging were procured from Klockner-Werke and there are no initial upper shelf energy data available for these forgings from the Certified Material Test Reports (CMTRs).

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-9 Table 3-2 Upper-Shelf Energy Data for Extended Beltline Forging Materials Used to Determine Generic Upper-Shelf Energy Copper Weight PercentExtended Beltline Locations The ONS Unit 1 and ONS Unit 2 RPV inlet and outlet nozzle forgings and transition forgings and the ONS Unit 3 RPV inlet nozzle forgings were procured from Bethlehem Steel. The copper wt% for all such material in the RVWG data base [ ] are reported in Table 3-5. The mean was found to be [ ], the standard deviation is

[ ] and the one-sided tolerance factor (k), 95% confidence for 95% coverage, is [ ]. The generic value is calculated to be [ ] (mean plus k*sigma).

Data used to develop the generic mean are reported in (Table 3-5 below).

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-10 The copper weight percent of the RPV inlet and outlet nozzle forging to nozzle belt forging Linde 80 welds for all 3 Oconee Units are reported in Table 3-6 through Table 3-8 and in each case are the mean of measured values of the actual weld wire heats listed. Use of the mean of measured values for a plate or forging or for weld samples made with the weld wire heat number that matches the critical vessel weld is consistent with Regulatory Guide 1.99, Revision 2, Section 1.1 (Reference 3-3).

3.4.2 Comparison to Current Licensing Basis and NRC RVID2-Traditional Beltline For traditional beltline materials, differences between the current licensing basis (see Section 1.0) and the Cu wt% content and initial upper-shelf energy values utilized for subsequent license renewal are identified in Table 3-3 and Table 3-4, respectively. For items not listed, the CLB and SLR Cu wt% and CvUSE values are identical. In addition, Cu wt% and CvUSE values reported in the NRC Reactor Vessel Integrity Database (RVID2-Reference 2-6) database are included for information. The data and information provided by licensees in their responses to the staff's Generic Letter (GL) 92-01, Revision 1 close-out letters, and in response to GL 92-01, Revision 1, Supplement 1, are included in RVID2. This database includes updates from June 1999

- July 2000. Duke Energys responses to GL92-01, GL 92-01, Revision 1, and GL 92-01, Revision 1, Supplement 1 are based on the following generic topical reports in chronological order.

  • BAW-2222, Response to Closure Letters to GL 92-01, June 1994, ADAMS Forwards rept BAW-2222, "Reactor Vessel Working Group Response to Closure Ltrs to NRC GL 92-01, Rev 1, ADAMS Legacy Accession 9407060101 (microfiche).

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-11

  • BAW-2257, Revision 1, October 1995, B&WOG Reactor Vessel Working Group Response to GL 92-01, Rev 1, Suppl 1, ADAMS Accession Number ML20094D978

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-12 Based on a comparison of the NRC RVID2 data and updated SLR data for traditional beltline locations, additional explanatory information is provided for the following items due to the potential to increase margins to embrittlement limits.

  • ONS Unit 2 LNB Forging AMX-77. Cu wt% decrease from 0.13% to [ ]

A generic Cu wt % of [ ] is selected for forging AMX-77, which is derived in Section 3.4.1 above for Bethlehem forgings. Based on a review of the ONS Unit 2 RPV Quality Assurance Data Package, ONS Unit 2 LNB forging AMX-77 was forged by the Ladish Co. As reported in BAW-2222, Section 4.1, for Ladish forgings for which copper content was not determined, a mean was found to be 0.05 wt %, the standard deviation is 0.03 and the one-sided factor (k) is 2.453. The generic value is calculated to be 0.13 wt % (mean plus k).

[

]

  • ONS Unit 3 LNB 4680. Initial USE increase from 109 ft-lbs to 113 ft-lbs.

With recent (2014) measured values for heat 4680 (100% shear) from a nozzle dropout at 111, 108, 114, and 119 ft-lbs, the mean is 113 ft-lbs. The CLB initial USE for heat 4680, originally at 109 ft-lbs, was based on BAW-2222. Recent testing of heat 4680 permitted the increase from 109 ft-lbs to 113 ft-lbs.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-13 Table 3-3 Traditional beltline Cu Content Differences between CLB and SLR

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-14 Table 3-4 Traditional Beltline CvUSE Differences between CLB and SLR

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-15 Table 3-5 Copper Content Data for Pre-1971 ASTM A508 Class 2 Reactor Vessel Forgings: Bethlehem Steel Ingots

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-16 Table 3-6 Evaluation of 72 EFPY Charpy V-Notch Upper-Shelf Energy - Oconee Unit 1 Material Description 72 EFPY, Fluence 72 EFPY Initial 1/4T CVN Predicted CvUSE Location Decrease CvUSE 2

RV Material Mat. ID Heat # Type Cu, wt% (ft-lb) (n/cm ) (%) (ft-lb)

Regulatory Guide 1.99, Revision 2, Position 1.2 Lower Nozzle A-AHR Belt (LNB) 54 ZV-2861 508, [ ] [ ] 1.81E+18 [ ] 90.3 Forging Cl. 2 SA-Intermediate C2197- 302 C2197-2 0.15 80 1.15E+19 24.8 60.2 Shell (IS) Plate 2 Gr. B Mod.

SA-Upper Shell (US) Plate C3265-1 C3265-1 302 Gr. B [ ] [ ] 1.32E+19 [ ] 86.1 Mod.

SA-C3278- 302 US Plate C3278-1 0.12 80 1.32E+19 22.4 62.1 1 Gr. B Mod.

SA-Lower Shell C2800- 302 C2800-1 0.11 80 1.31E+19 21.3 62.9 (LS) Plate 1 Gr. B Mod.

SA-C2800- 302 LS Plate C2800-2 0.11 119 1.31E+19 21.3 93.6 2 Gr. B Mod.

Outlet Nozzle A-Forging (ONF) NA 122S316VA2 508, [ ] [ ] 2.29E+17 [ ] 96.0 1 Cl. 2 A-ONF 2 NA 122S316VA1 508, [ ] [ ] 2.29E+17 [ ] 96.0 Cl. 2 A-Inlet Nozzle Forging (INF) 1 NA 123S346VA1 508, [ ] [ ] 1.13E+17 [ ] 96.0 Cl. 2 A-INF 2 NA 123S346VA2 508, [ ] [ ] 1.13E+17 [ ] 96.0 Cl. 2 A-INF 3 NA 124S502VA1 508, [ ] [ ] 1.13E+17 [ ] 96.0 Cl. 2 A-INF 4 NA 124S502VA2 508, [ ] [ ] 1.13E+17 [ ] 96.0 Cl. 2 A-Transition Forging NA 122S347VA1 508, [ ] [ ] 2.17E+17 [ ] 96.0 Cl. 2

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-17 Material Description 72 EFPY, Fluence 72 EFPY Initial 1/4T CVN Predicted CvUSE Location Decrease CvUSE 2

RV Material Mat. ID Heat # Type Cu, wt% (ft-lb) (n/cm ) (%) (ft-lb)

NA 8T1762 Linde 80 [ ] [ ] 2.29E+17 [ ] 45.0 LNB to ONF Linde Welds NA 299L44 80 [ ] [ ] 2.29E+17 [ ] 40.1 NA 8T1554B Linde 80 [ ] [ ] 2.29E+17 [ ] 45.9 NA 8T1762 Linde 80 [ ] [ ] 1.13E+17 [ ] 45.0 LNB to INF Linde Welds NA 299L44 80 [ ] [ ] 1.13E+17 [ ] 40.1 Linde NA 8T1554B 80 [ ] [ ] 1.13E+17 [ ] 45.9 LNB to IS Circ.

Weld (100%)

SA-1135 61782 Linde 80 [ ] [ ] 1.97E+18 [ ] 41.6 IS Long. Welds SA- Linde (Both 100%) 1073 1P0962 80 [ ] [ ] 8.75E+18 [ ] 36.7 IS to US Circ. SA- Linde Weld (ID 61%) 1229 71249 80 [ ] [ ] 1.16E+19 [ ] 34.3 IS to US Circ.

Weld (OD WF-25 299L44 Linde 80 [ ] [ ] 3.95E+18 [ ] 35.1 39%)

US Long.

SA- Linde Welds (Both 1493 8T1762 80 [ ] [ ] 8.58E+18 [ ] 37.9 100%)

US to LS Circ.

Weld (100%)

SA-1585 72445 Linde 80 [ ] [ ] 1.29E+19 [ ] 34.3 LS Long. Weld SA- Linde (100%) 1426 8T1762 80 [ ] [ ] 1.07E+19 [ ] 37.0 LS Long. Weld (100%)

SA-1430 8T1762 Linde 80 [ ] [ ] 1.07E+19 [ ] 37.0 LS to Transition Linde Forging Circ.

WF-9 72445 80 [ ] [ ] 2.17E+17 [ ] 44.0 Weld (100%)

Regulatory Guide 1.99, Revision 2, Position 2.2 SA-Upper Shell (US) Plate C3265-1 C3265-1 302 Gr. B [ ] [ ] 1.32E+19 [ ] 86.1 Mod.

LNB to ONF Linde Weld NA 299L44 80 [ ] [ ] 2.29E+17 [ ] 39.4 LNB to INF Linde Welds NA 299L44 80 [ ] [ ] 1.13E+17 [ ] 39.4 LNB to IS Circ.

Weld (100%)

SA-1135 61782 Linde 80 [ ] [ ] 1.97E+18 [ ] 41.9 IS to US Circ. SA- Linde Weld (ID 61%) 1229 71249 80 [ ] [ ] 1.16E+19 [ ] 35.8

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-18 Material Description 72 EFPY, Fluence 72 EFPY Initial 1/4T CVN Predicted CvUSE Location Decrease CvUSE 2

RV Material Mat. ID Heat # Type Cu, wt% (ft-lb) (n/cm ) (%) (ft-lb)

IS to US Circ.

Linde Weld (OD WF-25 299L44 80 [ ] [ ] 3.95E+18 [ ] 33.1 39%)

US to LS Circ. SA- Linde Weld (100%) 1585 72445 80 [ ] [ ] 1.29E+19 [ ] 33.0 LS to Transition Linde Forging Circ.

WF-9 72445 80 [ ] [ ] 2.17E+17 [ ] 43.3 Weld (100%)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-19 Table 3-7 Evaluation of 72 EFPY Charpy V-Notch Upper-Shelf Energy - Oconee Unit 2 Material Description Fluence 72 EFPY Initial 1/4T CVN Predicted Mat. CvUSE Location Decrease CvUSE 2

RV Material ID Heat # Type Cu, wt% (ft-lb) (n/cm ) (%) (ft-lb)

Regulatory Guide 1.99, Revision 2, Position 1.2 LNB AMX A-508, Forging 77 123T382 Cl. 2 [ ] [ ] 1.09E+19 [ ] 86.8 AAW A-508, US Forging 3P2359 0.04 133 1.24E+19 20.0 106.4 163 Cl. 2 AWG A-508, LS Forging 4P1885 0.02 138 1.24E+19 20.0 110.4 164 Cl. 2 A-508, ONF 1 NA [ ] Cl. 2 [ ] [ ] 2.13E+17 [ ] 96.0 A-508, ONF 2 NA [ ] Cl. 2 [ ] [ ] 2.13E+17 [ ] 96.0 INF 1 NA [ ] A-508, Cl. 2 [ ] [ ] 1.06E+17 [ ] 96.0 A-508, INF 2 NA [ ] Cl. 2 [ ] [ ] 1.06E+17 [ ] 96.0 A-508, INF 3 NA [ ] Cl. 2 [ ] [ ] 1.06E+17 [ ] 96.0 INF 4 NA [ ] A-508, Cl. 2 [ ] [ ] 1.06E+17 [ ] 96.0 Transition Forging NA [ ] A-508, Cl. 2 [ ] [ ] 2.02E+17 [ ] 96.0 Linde LNB to ONF NA 8T1762 80 [ ] [ ] 2.13E+17 [ ] 45.0 Welds NA 72445 Linde 80 [ ] [ ] 2.13E+17 [ ] 44.0 LNB to INF NA 8T1762 Linde 80 [ ] [ ] 1.06E+17 [ ] 45.0 Welds Linde NA 72445 80 [ ] [ ] 1.06E+17 [ ] 44.0 LNB to US Circ. Weld WF-154 406L44 Linde 80 [ ] [ ] 1.09E+19 [ ] 32.3 (100%)

US to LS WF- Linde Circ. Weld 25 299L44 80 [ ] [ ] 1.22E+19 [ ] 31.3 (100%)

LS to Transition WF- Linde Forging 112 406L44 80 [ ] [ ] 2.02E+17 [ ] 42.4 Circ. Weld (100%)

Regulatory Guide 1.99, Revision 2, Position 2.2 AAW A-508, US Forging 3P2359 0.04 133 1.24E+19 23.8 101.4 163 Cl. 2 LNB to ONF Weld NA 72445 Linde 80 [ ] [ ] 2.13E+17 [ ] 43.3

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-20 Material Description Fluence 72 EFPY Initial 1/4T CVN Predicted Mat. CvUSE Location Decrease CvUSE 2

RV Material ID Heat # Type Cu, wt% (ft-lb) (n/cm ) (%) (ft-lb)

LNB to INF Weld NA 72445 Linde 80 [ ] [ ] 1.06E+17 [ ] 43.3 LNB to US WF- Linde Circ. Weld 154 406L44 80 [ ] [ ] 1.09E+19 [ ] 32.6 (100%)

US to LS Circ. Weld WF-25 299L44 Linde 80 [ ] [ ] 1.22E+19 [ ] 26.3 (100%)

LS to Transition WF- Linde Forging 112 406L44 80 [ ] [ ] 2.02E+17 [ ] 42.5 Circ. Weld (100%)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-21 Table 3-8 Evaluation of 72 EFPY Charpy V-Notch Upper-Shelf Energy - Oconee Unit 3 Material Description Fluence 72 EFPY Initial 1/4T CVN Predicted Mat. CvUSE Location Decrease CvUSE 2

RV Material ID Heat # Type Cu, wt% (ft-lb) (n/cm ) (%) (ft-lb)

Regulatory Guide 1.99, Revision 2, Position 1.2 A-508, LNB Forging 4680 4680 0.15 113 1.13E+19 24.7 85.1 Cl. 2 AWS A-508, US Forging 522314 0.01 112 1.30E+19 20.2 89.4 192 Cl. 2 ANK A-508, LS Forging 522194 0.02 144 1.29E+19 20.2 114.9 191 Cl. 2 A-508, ONF 1 NA [ ] Cl. 2 [ ] [ ] 2.30E+17 [ ] ---

A-508, ONF 2 NA [ ] Cl. 2 [ ] [ ] 2.30E+17 [ ] ---

INF 1 NA [ ] A-508, Cl. 2 [ ] [ ] 1.14E+17 [ ] 96.0 A-508, INF 2 NA [ ] Cl. 2 [ ] [ ] 1.14E+17 [ ] 96.0 A-508, INF 3 NA [ ] Cl. 2 [ ] [ ] 1.14E+17 [ ] 96.0 INF 4 NA [ ] A-508, Cl. 2 [ ] [ ] 1.14E+17 [ ] 96.0 Transition Forging NA [ ] A-508, Cl. 2 [ ] [ ] 2.15E+17 [ ] 48.8 Linde LNB to ONF NA 72105 80 [ ] [ ] 2.30E+17 [ ] 40.8 Welds NA 406L44 Linde 80 [ ] [ ] 2.30E+17 [ ] 42.4 NA 72105 Linde 80 [ ] [ ] 1.14E+17 [ ] 40.8 LNB to INF Linde Welds NA 72102 80 [ ] [ ] 1.14E+17 [ ] 44.3 NA 82102 Linde 80 [ ] [ ] 1.14E+17 [ ] 39.8 LNB to US WF- Linde Circ. Weld 200 821T44 80 [ ] [ ] 1.14E+19 [ ] 33.8 (100%)

US to LS Circ. Weld WF-67 72442 Linde 80 [ ] [ ] 1.27E+19 [ ] 32.1 (ID 75%)

US to LS WF- Linde Circ. Weld 70 72105 80 [ ] [ ] 4.27E+18 [ ] 34.9 (OD 25%)

LS to Transition WF-Forging 169- 8T1554 Linde 80 [ ] [ ] 2.15E+17 [ ] 45.9 Circ. Weld 1 (100%)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-22 Material Description Fluence 72 EFPY Initial 1/4T CVN Predicted Mat. CvUSE Location Decrease CvUSE 2

RV Material ID Heat # Type Cu, wt% (ft-lb) (n/cm ) (%) (ft-lb)

Regulatory Guide 1.99, Revision 2, Position 2.2 AWS A-508, US Forging 522314 0.01 112 1.30E+19 20.7 88.8 192 Cl. 2 Linde LNB to ONF NA [ ] 80 [ ] [ ] 2.30E+17 [ ] 42.1 Welds NA [ ] Linde 80 [ ] [ ] 2.30E+17 [ ] 42.5 LNB to INF Linde Weld NA [ ] 80 [ ] [ ] 1.14E+17 [ ] 42.1 ANK A-508, LS Forging 522194 0.02 144 1.29E+19 26.8 105.5 191 Cl. 2 LNB to US Circ. Weld WF-200 821T44 Linde 80 [ ] [ ] 1.14E+19 [ ] 33.4 (100%)

US to LS WF- Linde Circ. Weld 67 72442 80 [ ] [ ] 1.27E+19 [ ] 37.1 (ID 75%)

US to LS Circ. Weld WF-70 72105 Linde 80 [ ] [ ] 4.27E+18 [ ] 36.6 (OD 25%)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-23 3.5 Confirmation that Oconee is Bounded by NRC Approved Generic Technical Reports The ASME Code,Section XI, acceptance criteria for Levels A through D Service Loadings for Oconee Units 1, 2, and 3 reactor vessel traditional beltline and extended beltline Linde 80 welds are satisfied and are reported in Framatome reports BAW-2192, Revision 0, Supplement 1P-A, Revision 0, Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessels of B&W Owners Reactor Vessel Working Group for Levels A and B Service Loadings (Reference 3-4), and BAW-2178, Revision 0, Supplement 1P-A, Revision 0, Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessels of B&W Owners Reactor Vessel Working Group for Levels C and D Service Loadings (Reference 3-5). These reports contain equivalent margins analyses for all Oconee Unit 1, 2, and 3 Linde 80 welds identified as traditional beltline and extended beltline material for 80 years.

As illustrated in Table 3-9, the 80-year inside wetted surface fluence values reported in Table 3-1 of BAW-2178, Revision 0, Supplement 1P-A, Revision 0 and Table 3-1 of BAW-2192, Revision 0, Supplement 1P-A, Revision 0 bound the 72 EFPY fluence values reported in Table 2-1, Table 2-2, and Table 2-3 of this ANP report with the exception of ONS Unit 1 weld SA-1135, circumferential weld that connects the lower nozzle belt forging to the intermediate shell. Note that the EMAs reported in the topical reports (TRs) conservatively utilized 80-year fluence values of at least an order of magnitude higher than the 72 EFPY nozzle fluence reported herein.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-24 In addition, the weld chemistry (Copper wt %) data reported in Table 3-1 of BAW-2178, Revision 0, Supplement 1P-A, Revision 0 and Table 3-1 of BAW-2192, Revision 0, Supplement 1PA, Revision 0 is consistent with weld chemistry copper content reported in Table 3-6, Table 3-7, and Table 3-8. The Levels C and D limiting design transients reported in Section 4.3.1 of BAW-2178, Revision 0, Supplement 1P-A, Revision 0 are applicable to Oconee Units and are based on a review of the Oconee Reactor Vessel Design Specification transients and the UFSAR Chapter 15 events relative to transients that would result in the highest thermal stresses coupled with pressure stresses relative to the EMA analysis; this satisfies Regulatory Guide 1.161 with respect to Levels C and D transient selection. The materials of construction, RPV geometry, and range of explanatory variables for the J-R model (Section A.5 of BAW-2192, Revision 0, Supplement 1P-A, Revision 0) reported in the topical reports are confirmed to be applicable to Linde 80 traditional beltline and extended beltline welds at Oconee Units 1 through 3.

As such, Framatome has confirmed that Oconee Units 1, 2, and 3 are bounded by the topical report submittals BAW-2178, Revision 0, Supplement 1P-A, Revision 0, and BAW-2192, Revision 0, Supplement 1P-A, Revision 0 relative to fluence (with the exception of ONS Unit 1 weld SA-1135), weld chemistry, geometry, materials of construction, design transients, and the J-R model applicability. The results of the Oconee Units 1, 2, and 3 EMA are reported in BAW-2178, Revision 0, Supplement 1PA, Revision 0 and BAW-2192, Revision 0, Supplement 1P-A, Revision 0 and are not repeated herein.

The equivalent margins analyses for the ONS Unit 3 RPV outlet nozzle forgings and transition forging are not included in the above NRC approved topical reports but are reported in Section 4.0 of this ANP report.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-25 3.6 ONS Unit 1 Weld SA-1135 EMA Reconciliation Oconee Unit 1 weld SA-1135 was evaluated in BAW-2192, Revision 0, Supplement 1PA, Revision 0, at a fluence of 1.90E+18 n/cm2 yet the fluence for weld SA-1135 is reported in Table 2-1 as 2.91E+18 n/cm2. Therefore, an EMA reconciliation was performed for this weld at a fluence of 2.91E+18 n/cm2. The ratio J0.1/J1 for weld SA1135 decreased from [ ] to [ ], is greater than the required value of 1.0, and flaw extensions remain ductile and stable. The limiting weld remains Oconee Unit 1 axial weld SA-1073 as reported in BAW-2192, Revision 0, Supplement 1P-A, Revision 0.

3.7 References for Section 3.0 3-1. 10 CFR Part 50.60, Appendix G, Code of Federal Regulations, Title 10, Part 50 -

Domestic Licensing of Production and Utilization Facilities, Appendix G -

Fracture Toughness Requirements, Federal Register Vol. 84, Page 65644, November 29, 2019.

3-2. NUREG-2192, Standard Review Plan Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants 3-3. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials 3-4. BAW-2192, Revision 0, Supplement 1P-A, Revision 0, Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessels of B&W Owners Reactor Vessel Working Group for Levels A & B Service Loads. ADAMS Accession Number ML19101A355.

3-5. BAW-2178, Revision 0, Supplement 1P-A, Revision 0, Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessels of B&W Owners Reactor Vessel Working Group for Levels C & D Service Loads. ADAMS Accession Number ML19101A355.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-26 Table 3-9 BAW-2178 and BAW-2192 Reactor Vessel Weld Locations--Copper Content and 80-Year Fluence Projections (IS) Inside Wetted (IS) Inside Wetted Material ID and/or Surface Fluence or Reactor Vessel Material Cu, wt% Ni, wt% Surface Fluence or Cu, wt% 2 Heat Number 2 n/cm E> 1.0 MeV) n/cm E> 1.0 MeV)

Tables 2-1 to 2-3 Oconee Unit 1, 80-YearFluence (E > 1.0 MeV) 17 Wire Ht. 8T1762 0.19 0.57 1.50E+18 0.19 3.49x10 Lower Nozzle Belt (LNB) to Outlet Nozzle 17 Wire Ht. 299L44 0.34 0.68 1.50E+18 0.34 3.49x10 Forging (ONF) Welds 17 Wire Ht. 8T1554B 0.16 0.57 1.50E+18 0.16 3.49x10 17 Wire Ht. 8T1762 0.19 0.57 1.50E+18 0.19 1.62x10 17 LNB to Inlet Nozzle Forging (INF) Welds Wire Ht. 299L44 0.34 0.68 1.50E+18 0.34 1.62x10 17 Wire Ht. 8T1554B 0.16 0.57 1.50E+18 0.16 1.62x10 SA-1135 18 LNB to Intermediate Shell (IS) Circ. Weld 0.23 0.52 1.90E+18 0.23 2.91x10 (Wire Ht. 61782)

SA-1073 19 IS Long. Welds (Both) 0.21 0.64 1.58E+19 0.21 1.38x10 (Wire Ht. 1P0962)

SA-1229 19 IS to Upper Shell (US) Circ. Weld (ID 61%) 0.23 0.59 2.02E+19 0.23 1.86x10 (Wire Ht. 71249)

SA-1493 19 US Long. Welds (Both) 0.19 0.57 2.05E+19 0.19 1.36x10 (Wire Ht. 8T1762)

SA-1585 19 US to Lower Shell (LS) Circ. Weld 0.22 0.54 2.14E+19 0.22 2.05x10 (Wire Ht. 72445)

SA-1426 19 LS Long. Weld (1) 0.19 0.57 1.82E+19 0.19 1.68x10 (Wire Ht. 8T1762)

SA-1430 19 LS Long. Weld (2) 0.19 0.57 1.82E+19 0.19 1.68x10 (Wire Ht. 8T1762)

WF-9 17 LS to Transition Forging Circ. Weld 0.22 0.54 4.88E+17 0.22 2.70x10 (Wire Ht. 72445)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 3-27 (IS) Inside Wetted (IS) Inside Wetted Material ID and/or Surface Fluence or Reactor Vessel Material Cu, wt% Ni, wt% Surface Fluence or Cu, wt% 2 Heat Number 2 n/cm E> 1.0 MeV) n/cm E> 1.0 MeV)

Tables 2-1 to 2-3 Oconee Unit 2, 80-Year Fluence (E > 1.0 MeV) 17 Lower Nozzle Belt (LNB) to Outlet Nozzle Wire Ht. 8T1762 0.19 0.57 1.50E+18 0.19 3.24x10 17 Forging (ONF) Welds Wire Ht. 72445 0.22 0.54 1.50E+18 0.22 3.24x10 17 LNB to Inlet Nozzle Forging (INF) Welds Wire Ht. 8T1762 0.19 0.57 1.50E+18 0.19 1.50x10 17 Wire Ht. 72445 0.22 0.54 1.50E+18 0.22 1.50x10 WF-154 19 LNB to Upper Shell (US) Circ. Weld 0.27 0.59 1.99E+19 0.27 1.75x10 (Wire Ht. 406L44)

WF-25 19 US to Lower Shell (LS) Circ. Weld 0.34 0.68 2.18E+19 0.34 1.92x10 (Wire Ht. 299L44)

WF-112 17 LS to Transition Forging Circ. Weld 0.27 0.59 5.20E+17 0.27 2.50x10 (Wire Ht. 406L44)

Oconee Unit 3, 80-Year Fluence (E > 1.0 MeV) 17 Lower Nozzle Belt (LNB) to Outlet Nozzle Wire Ht. 72105 0.32 0.58 1.50E+18 0.32 3.50x10 17 Forging (ONF) Welds Wire Ht. 406L44 0.27 0.59 1.50E+18 0.27 3.50x10 17 Wire Ht. 72105 0.32 0.58 1.50E+18 0.32 1.62x10 17 LNB to Inlet Nozzle Forging (INF) Welds Wire Ht. 72102 0.21 0.58 1.50E+18 0.21 1.62x10 17 Wire Ht. 82102 0.35 1.00 1.50E+18 0.35 1.62x10 WF-200 19 LNB to Upper Shell (US) Circ. Weld 0.24 0.63 1.92E+19 0.24 1.82x10 (Wire Ht. 821T44)

WF-67 19 US to Lower Shell (LS) Circ. Weld (ID 75%) 0.26 0.60 2.04E+19 0.26 2.01x10 (Wire Ht. 72442)

WF-169-1 17 LS to Transition Forging Circ. Weld 0.16 0.57 4.78E+17 0.16 2.68x10 (Wire Ht. 8T1554)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-1 4.0 EMA OF ONS UNIT 3 RPV OUTLET NOZZLES AND TRANSITION FORGING 4.1 Introduction Appendix G to 10 CFR Part 50, which addresses low upper-shelf toughness of the reactor pressure vessel (RPV) forgings, requires that an Equivalent Margins Analysis (EMA) be completed when the upper-shelf Charpy energy level falls below 50 ft-lb in order to demonstrate adequate material toughness. The 72 EFPY Charpy V-notch Upper Shelf Energy (CvUSE) calculations at the quarter thickness wall locations for Oconee (ONS) Units 1, 2, and 3, documented in Table 3-6, Table 3-7, and Table 3-8 concludes that the predicted CvUSE is above 50 ft-lbs at all three Oconee Units, with the following exceptions:

1. ONS Unit 3 Transition Forging: The ONS Unit 3 Transition forging (Heat 417543-1) predicted 72 EFPY CvUSE value is 48.8 ft-lb, which is below the 50 ft-lb threshold.
2. ONS Unit 3 Outlet Nozzle Forgings: The ONS Unit 3 outlet nozzle forgings have no recorded initial CvUSE. Consequently, predicted CvUSE values are not explicitly calculated.
3. All traditional and extended beltline Linde 80 welds, which are evaluated at 72 EFPY in Sections 3.5 and 3.6, and References 4-1 and 4-2.

The ONS Unit 3 RPV outlet nozzle forgings (2) and the transition forging (1) were procured from Klockner-Werke, are extended beltline materials for SLR, and Charpy VNotch (CVN) testing on the upper shelf was not required by the construction code for these forgings. Therefore, 72 EFPY EMAs are required for the ONS Unit 3 RPV outlet nozzle forgings and transition forging to demonstrate compliance with 10 CFR Part 50, Appendix G. Consistent with BAW-2192 and BAW-2178, Revision 0, Supplement 1P-A, Revision 0, the Levels A and B Service Loadings for the RPV outlet nozzle forgings and transition forging are computed first, and only the limiting of these locations is evaluated for Levels C and D Service Loadings.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-2 In addition, the EMAs for the ONS RPV outlet nozzle forgings and transition forging reported herein conservatively utilize a 72 EFPY peak inside surface fluence of 1.0x1018 n/cm2 (E > 1.0 MeV) compared to the projected fluence of 2.68x1017 n/cm2 for the transition forging and 2.21x1017 n/cm2 for the RPV outlet nozzles (Table 2-3).

4.2 Regulatory Requirements 4.2.1 10 CFR 50 Appendix G In accordance with 10 CFR 50, Appendix G (Reference 4-3), IV, A, 1, Reactor Vessel Upper Shelf Energy Requirements are as follows:

  • Reactor vessel beltline materials must have Charpy upper-shelf energy in the transverse direction for base material and along the weld for weld material according to the ASME Code, of no less than 75 ft-lb (102 J) initially and must maintain Charpy upper-shelf energy throughout the life of the vessel of no less than 50 ft-lb (68 J), unless it is demonstrated in a manner approved by the Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, that lower values of Charpy upper-shelf energy will provide margins of safety against fracture equivalent to those required by Appendix G of Section XI of the ASME Code. This analysis must use the latest edition and addenda of the ASME Code incorporated by reference into 10 CFR 50.55a(b)(2) at the time the analysis is submitted.
  • Additional evidence of the fracture toughness of the beltline materials after exposure to neutron irradiation may be obtained from results of supplemental fracture toughness tests for use in the analysis specified in section IV.A.1.a.
  • The analysis for satisfying the requirements of section IV.A.1 of this appendix must be submitted, as specified in 10 CFR 50.4, for review and approval on an individual case basis at least three years prior to the date when the predicted Charpy upper-shelf energy will no longer satisfy the requirements of section IV.A.1 of this appendix, or on a schedule approved by the Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-3 In accordance with NRC Regulatory Guide 1.161, the NRC has determined that the analytical methods described in the ASME Code,Section XI, Appendix K, provide acceptable guidance for evaluating reactor pressure vessels when the Charpy upper-shelf energy falls below the 50 ft-lb limit of Appendix G to 10 CFR Part 50. However, the staff noted that Appendix K does not provide information on the selection of transients and gives very little detail on the selection of material properties. Selection of design transients and the methodology for calculation of applied J-integrals for the RPV outlet nozzles and transition forging are consistent with the methodology used to calculate applied Jintegrals for Linde 80 welds (References 4-1 and 4-2).

4.2.2 COMPLIANCE WITH 10 CFR 50 APPENDIX G AND ACCEPTANCE CRITERIA The current edition of the ASME Code,Section XI, listed in 10 CFR 50.55a is the 2017 Edition. Therefore, the analyses reported herein are performed in accordance with the 2017 Edition (Reference 4-4) of Section XI of the ASME Code, Appendix K. The material properties used in this analysis are based on the ASME Code,Section II, Part D, 2017 Edition. The ASME Code,Section XI, Subarticle K-2200, provides acceptance criteria for Levels A and B Service Loadings, and Subarticle K-2300 for Level C Service Loadings, and K-2400 for Level D Service Loadings. Both the RPV outlet forgings and transition forging are assessed against Subarticle K-2200. The limiting of these locations is evaluated against the ASME Code,Section XI, Subarticle K-2400, which provides acceptance criteria for Level D Service Loadings. However, for the Subarticle K-2400 evaluation reported herein the Jintegral resistance versus flaw extension curve is a conservative representation for the vessel material under analytical evaluation and not a best estimate curve.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-4 4.3 Transition Forging Equivalent Margins Analysis (Levels A and B Service Loadings)

The analytical procedure used for the 72 EFPY equivalent margins analysis of the ONS Unit 3 transition forging is in accordance with the ASME Code,Section XI, Appendix K, 2017 Edition (Reference 4-4), with selection of design transients based on the guidance in Regulatory Guide 1.161 (Reference 4-5).

4.3.1 Material Properties and Levels A and B Service Loadings 4.3.1.1 J-Integral Resistance Model The J-integral resistance model is from NUREG/CR-5729 (Reference 4-6), i.e., RPV Jd base metal models reported in Table 11, Charpy Model and CVNP Model. The J-integral resistance model (Charpy Model or CVNP) that provides the most conservative Jintegral resistance at a crack extension of 0.10 inches is selected for each evaluation.

In order to utilize the NUREG/CR-5729 (Reference 4-6) base metal Jd Charpy and CVNP models, initial unirradiated upper shelf energy is established for these extended beltline forgings procured from Klockner-Werke.

The ONS Unit 3 transition forging unirradiated CvUSE was established through NRC review of the ONS 54 EFPY P-T limits (NRC SER-ADAMS Accession Number ML14041A093, Page 8). The SER references an e-mail (ADAMS Accession Number ML13350A098) wherein the unirradiated CvUSE of 59.2 ft-lbs in the weak direction is reported for the ONS Unit 3 transition forging. This value is not a measured upper shelf energy but was obtained as follows.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-5 Data taken from the actual RPV certified material test report where limited CVN test data were performed at a single test temperature (+10 °F) to confirm that at least 30 ftlbs was obtained for the material. CVN test data does not represent the impact energy upper-shelf (i.e., >95% shear), but is a mean value taken at a reduced percent shear fracture and is a conservative estimate of the CvUSE in that the impact energy will increase as the percent shear fracture increases. This test report data was conservatively considered to be in the strong direction and the mean value of the reported CVN data at +10 °F was multiplied by 65% to obtain 59.2 ft-lbs in the weak direction.

4.3.1.2 Mechanical Properties of Transition Forgings The reactor vessel lower shell and transition forgings are fabricated from A-508 Class 2 that is identified in the current ASME Code as SA-508 Grade 2 Class 1 material (Table 4-1). For the stainless steel cladding, 18Cr-8Ni Type 304 material properties are used.

The predicted Charpy impact energy value at 72 EFPY of 48.8 ft-lbs (Table 3-8) is used for the NUREG/CR-5729 Charpy Model and the associated initial Charpy impact energy value of 59.2 ft-lbs is used for the CVNp model. The weak direction upper shelf energies, 72 EFPY and unirradiated, are conservatively used for both axial and circumferential flaws.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-6 Table 4-1 Material Properties for SA-508 Class 2 Temp. E C k y

-6 6

(°F) (10 1/°F) (10 psi) (typ.) (BTU/lb-°F) (Btu/hr-in-°F) (lb/in3) (ksi) 70 6.4 27.8 0.3 0.105 1.975 0.2841 50.00 100 6.5 27.6 0.3 0.107 1.967 0.2839 50.00 200 6.7 27.1 0.3 0.113 1.958 0.2831 47.15 300 6.9 26.7 0.3 0.120 1.950 0.2823 45.25 400 7.1 26.2 0.3 0.125 1.925 0.2817 44.50 500 7.3 25.7 0.3 0.131 1.892 0.2809 43.20 600 7.4 25.1 0.3 0.136 1.850 0.2802 42.00 650 7.5 24.9 0.3 0.139 1.825 0.2797 41.40 700 7.6 24.6 0.3 0.142 1.800 0.2794 40.60 4.3.1.3 Levels A and B Service Loadings The Levels A and B Service Loadings required by Appendix K are an accumulation pressure (internal pressure load) and a cooldown transient (thermal load). In accordance with Article K-1300 of Appendix K (Reference 4-4), the accumulation pressure used for flaw evaluations should not exceed 1.1 times the design pressure.

With the design pressure of 2500 psig, the accumulation pressure is then 2750 psig (=

1.1x2500 psig). A conservative cooldown rate of 100°F/hour (from 560°F to 70°F in 4.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />) is used to represent the Levels A and B Service Loadings. This is the maximum attainable cooldown rate at ONS Unit 3 and the maximum cooldown rate required by Appendix K. This is consistent with BAW-2192, Revision 0, Supplement 1P-A, Revision 0, Page 22.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-7 4.3.1.4 Reactor Vessel Design Data Pertinent design data for the thickness of the transition regions of the ONS Unit 3 reactor vessel are listed below.

4.3.2 Fracture Mechanics Analysis 4.3.2.1 Methodology and Acceptance Criteria In accordance with the ASME Code,Section XI, Appendix K (Reference 4-4), Subarticle K-1200, the following analytical procedure was used for Levels A and B Service Loadings.

a. The flaws in the transition forging were postulated in accordance with Subarticle K-2200. When analytically evaluating adequacy of the upper shelf toughness for the base material, both interior axial and circumferential flaws with depths one quarter of the wall thickness and lengths six times the depth shall be postulated, and toughness properties for the corresponding orientation shall be used.

Smaller flaw sizes may be used when justified. The path lines of interest wherein flaws are postulated are illustrated in Figure 4-1.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-8

b. Loading conditions at the locations of the postulated flaws were determined for Levels A and B Service Loadings. For Levels A and B Service Loadings the equations to calculate the stress intensity factor (SIF) due to pressure and thermal gradients for a given pressure and cooldown rate are given in Article K4210. Consistent with BAW-2192, Revision 0, Supplement 1P-A, Revision 0 (Reference 4-1), the accumulation pressure is taken as ten percent above the design pressure and the maximum cooldown rate is 100°F/hr.
c. Material properties, including E, , y, and the Jintegral resistance curve (JR curve), were determined at the locations of the postulated flaws. Youngs modulus, mean coefficient of thermal expansion and yield strength are addressed in Section 4.3.1.2. The JR curve is discussed in Section 4.3.1.1.
d. The postulated flaws were evaluated in accordance with the acceptance criteria of Subarticle K-2200. Requirements for evaluating the applied Jintegral are provided in Subarticle K-3200, and for determining flaw stability in Subarticle K3400. Subarticle K-3500(a) invokes the procedure provided in Subarticle K4200 for evaluating the applied J-integral for a specified amount of ductile flaw extension. Three permissible evaluation methods to address flaw stability are described in Subarticle K-3500(b). The evaluation method selected herein is the J-R curve crack driving force diagram procedure described in Subarticle K-4310.

4.3.2.2 Procedure for Evaluating Levels A and B Service Loadings The EMA of the transition forging uses the fracture mechanics analytical procedures of the ASME Code,Section XI, Appendix K (Reference 4-4), augmented as necessary to incorporate location specific stresses determined by a 2 dimensional (2-D) finite element analysis and stress intensity factors calculated from an influence coefficient based solution for axial or circumferential surface flaws.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-9 Stress intensity factors (SIFs) for Levels A and B Service Loadings are calculated based on the Raju and Newman solution (Reference 4-7) for internal semi-elliptical axial surface flaws in a cylinder, and the [ ] solution (Reference 4-8) for internal semi-elliptical circumferential surface flaws in cylindrical shells. An effective flaw depth is used to account for small scale yielding at the crack tip in computing the applied Jintegral per Appendix K, Article K-4210. The adequacy of the upper-shelf toughness margin and demonstration of flaw stability is then evaluated according to Articles K4220 and K-4310, respectively. The overall evaluation procedure for both axial and circumferential flaws is outlined below.

a. Axial Flaw

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-10

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-11

b. Circumferential Flaw

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-12

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-13 Temperature Range for Upper-Shelf Fracture Toughness Evaluations: Upper-shelf fracture toughness is determined through use of Charpy V-notch impact energy versus temperature plots by noting the temperature above which the Charpy energy remains on a plateau, maintaining a relatively high constant energy level. Similarly, fracture toughness can be addressed in three different regions on the temperature scale, i.e., a lower-shelf toughness region, a transition region, and an upper-shelf toughness region.

Fracture toughness of reactor vessel steel and associated weld metals are conservatively predicted by the ASME Code initiation toughness curve, KIc, in lower-shelf and transition regions. In the upper-shelf region, the upper-shelf toughness curve, KJc, is derived from the upper-shelf J-integral resistance model described. When upper-shelf toughness is plotted versus temperature, a plateau-like curve develops that decreases slightly with increasing temperature. Since the present analysis addresses the low upper-shelf fracture toughness issue, only the upper-shelf temperature range, which begins at the intersection of the KIc and upper-shelf toughness curves, is considered.

4.3.2.3 Evaluation for Levels A and B Service Loadings The ASME Code,Section XI, Appendix K acceptance criteria have been satisfied for Levels A and B Service Loadings for ONS Unit 3 transition forging. For circumferential flaws, the minimum ratio J0.1/J1 is [ ] by the Charpy model, and [ ] by the CVNp model. For axial flaws, the minimum ratio is [ ] by the Charpy model and

[ ] by the CVNp model. The overall minimum J0.1/J1 ratio is [ ], which is greater than the minimum acceptable value of 1.0 required by Appendix K of the Code.

In addition, using structural factors of 1.25 on pressure and 1.0 on thermal loading, flaw extension is demonstrated to be ductile and stable for the transition forging since the slope of the applied J-integral curve is less than the slope of the lower bound J-R curve at the point where the two curves intersect.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-14 The J-R curve data and applied J-integrals are plotted in Figure 4-2 through Figure 4-5 for circumferential flaws, and in Figure 4-6 through Figure 4-9 for axial flaws at each of the path lines shown in Figure 4-1. The limiting path line location is Base_1.

Figure 4-1 Stress Path Lines (Transition Forging)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-15 Figure 4-2 J-Integral versus Circumferential Flaw Extension - BASE_1 (Transition Forging)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-16 Figure 4-3 J-Integral versus Circumferential Flaw Extension - BASE_2 (Transition Forging)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-17 Figure 4-4 J-Integral versus Circumferential Flaw Extension - BASE_3 (Transition Forging)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-18 Figure 4-5 J-Integral versus Circumferential Flaw Extension - BASE_4 (Transition Forging)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-19 Figure 4-6 J-Integral versus Axial Flaw Extension - BASE_1 (Transition Forging)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-20 Figure 4-7 J-Integral versus Axial Flaw Extension - BASE_2 (Transition Forging)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-21 Figure 4-8 J-Integral versus Axial Flaw Extension - BASE_3 (Transition Forging)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-22 Figure 4-9 J-Integral versus Axial Flaw Extension - BASE_4 (Transition Forging)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-23 4.4 RPV Outlet Nozzle Forgings Equivalent Margins (Levels A and B Service Loadings)

The analytical procedure used for the 72 EFPY equivalent margins analysis, Levels A and B Service Loadings, of the ONS Unit 3 RPV outlet nozzle forging is in accordance with the ASME Code,Section XI, Appendix K, 2017 Edition (Reference 4-4), with selection of design transients based on the guidance in Regulatory Guide 1.161 (Reference 4-5).

4.4.1 Material Properties and Levels A and B Service Loadings 4.4.1.1 J-Integral Resistance Model The J-integral resistance model is from NUREG/CR-5729 (Reference 4-6), i.e., RPV Jd base metal models reported in Table 11, i.e., Charpy Model and CVNP Model. The J-integral resistance model (Charpy Model or CVNP) that provides the most conservative J-integral resistance at a crack extension of 0.10 inches is selected for each evaluation.

In order to utilize the NUREG/CR-5729 (Reference 4-6) base metal Jd Charpy and CVNP models, initial unirradiated upper shelf energy is established for these extended beltline forgings procured from Klockner-Werke.

Unirradiated upper shelf energy for the ONS RPV outlet nozzle forgings procured from Klockner-Werke is not available from the certified material test report and was conservatively estimated at 78 ft-lbs in the strong direction and 52 ft-lbs in the weak direction based on the data reported in PWROG-17090-NP-A, Revision 0, Generic Rotterdam Forging and Weld Initial Upper-Shelf Energy Determination (Reference 4-9),

Table 6. The mean minus 2 standard deviations CvUSE, in the strong and weak directions, for Fried-Krupp Huttenwerke AG Forgings reported in Table 6 were conservatively applied to the Klockner-Werke forgings. This assumption is justified based on review of PWROG-17090-NP-A, Revision 0, Table 7, wherein use of actual Klockner-Werke measured data and statistical mean in the strong direction minus 2 standard deviations would result in an unirradiated CvUSE greater than 78 ft-lbs in the strong direction.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-24 4.4.1.2 Mechanical Properties The reactor vessel outlet nozzle forgings are fabricated from A-508 Class 2 that is identified in the current ASME Code as A508-64 Grade 2 Class 1 material (Table 4-2).

For the stainless steel cladding, 18Cr-8Ni Type 304 material properties are used.

Table 4-2 Material SA 508 Class 2 (3/4Ni-1/2Mo-1/3Cr-V)

Temperature Modulus of Elasticity Poissons Ratio Yield Strength

(°F) (psi) (-) (ksi) 70 2.78E+07 0.30 50.00 100 2.76E+07 0.30 50.00 200 2.71E+07 0.30 47.00 (1) 270 2.68E+07 0.30 45.95 300 2.67E+07 0.30 45.50 (1) 317 2.66E+07 0.30 45.28 400 2.62E+07 0.30 44.20 500 2.57E+07 0.30 43.20 600 2.51E+07 0.30 42.10 (1) 650 2.49E+07 0.30 41.40 700 2.46E+07 0.30 40.70 Note(s):

1. Interpolated value The NUREG/CR-5729 RPV Jd base metal Charpy Model requires upper shelf energy at 72 EFPY as input. As such, a conservative fluence value of 1.0E+18 n/cm2 at the inside surface was assumed at the RPV nozzles, with no attenuation, and the 72 EFPY CvUSE values at the RPV nozzles were calculated using RG 1.99, Position 1.2. The NUREG/CR-5729 RPV Jd CVNp model requires unirradiated upper shelf energy, discussed above, and fluence (assumed at 1.0E+18 n/cm2) as inputs. The estimate of 72 EFPY fluence of 1.0E+18 n/cm2 is conservative relative to that reported in Table 2-3 for the ONS Unit 3 RPV outlet nozzles.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-25 4.4.1.3 Levels A and B Service Loadings The Levels A and B Service Loadings required by Appendix K are an accumulation pressure (internal pressure load) and a cooldown transient (thermal load). Per article K1300 of Appendix K (Reference 4-4), the accumulation pressure used for flaw evaluation should not exceed 1.1 times the design pressure. With the design pressure of 2500 psig, the accumulation pressure is then 2750 psig. The ONS units design cooldown rate is 100 °F/hr, which is the maximum required by Appendix K of the ASME Code,Section XI and is consistent with BAW-2192, Revision 0, Supplement 1P-A, Revision 0. Pipe loads for Levels A and B Service Loadings are reported in Table 4-3.

Table 4-3 Service Levels A and B Outlet Nozzle Loads 4.4.1.4 Reactor Vessel Design Data Pertinent design data including the design pressure and geometry of the RPV shell and outlet nozzle forgings of the ONS Unit 3 reactor vessel are listed below.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-26 4.4.2 Fracture Mechanics Analysis 4.4.2.1 Methodology and Acceptance Criteria The analytical procedure was used for the evaluation of Levels A and B Service Loadings is in accordance with the ASME Code,Section XI, Appendix K (Reference 4-4), Subarticle K-1200.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-27

a. Flaws in the RPV outlet nozzle forgings were postulated in accordance with the Subarticle K-2200. When analytically evaluating adequacy of the upper shelf toughness for the base material, both interior axial and circumferential flaws with depths onequarter of the wall thickness and lengths six times the depth shall be postulated, and toughness properties for the corresponding orientation shall be used. Smaller flaw sizes may be used when justified. Path lines of interest in the RPV outlet nozzle are illustrated in Figure 4-10. Circumferential and axial flaws with depths onequarter of the wall thickness and lengths six times the depth are postulated in the taper region (Path ntt). The defect in the corner region (Path nct) is an axial flaw in accordance with the ASME Code,Section XI, 2017 Edition, G2223 and Figure G-2223-1, as modified below.

- For the postulated nozzle corner defect, Appendix K requires that a 1/4T flaw size be postulated when evaluating for Levels A and B Service Loadings.

However, Appendix K, Subarticle K-2200 permits the analysis to be performed considering a smaller flaw size, if it can be justified. Herein, a 3inch flaw is postulated at the nozzle corner, which is consistent with the maximum allowable postulated defect for the shell section that is 12-inches thick or greater. This postulated nozzle corner flaw is substantially greater than the allowable planar flaw at this location per the inservice inspection standards reported in Table IWB-3512-1 for the inside corner region.

b. Loading conditions at the locations of the postulated flaws were determined for Levels A and B Service Loadings. For Levels A and B Service Loadings the equations to calculate the stress intensity factor (SIF) due to pressure and thermal gradients for a given pressure and cooldown rate are given in Article K4210. Consistent with BAW-2192P, Revision 0, Supplement 1P-A, Revision 0 (Reference 4-1), the accumulation pressure is taken as ten percent above the design pressure and the maximum cooldown rate is 100°F/hr.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-28

c. Material properties, including E, y, and the Jintegral resistance curve (JR curve), were determined at the locations of the postulated flaws. Youngs modulus and yield strength are addressed in Section 4.4.1.2. The JR curve is discussed in Section 4.4.1.1.
d. The postulated flaws were evaluated in accordance with the acceptance criteria of Subarticle K-2200. Requirements for evaluating the applied Jintegral are provided in Subarticle K-3200, and for determining flaw stability in Subarticle K3400. Subarticle K-3500(a) invokes the procedure provided in Subarticle K4200 (K-4220) for evaluating the applied J-integral for a specified amount of ductile flaw extension. Three permissible evaluation methods to address flaw stability are described in Subarticle K-3500(b). The evaluation method selected herein is the J-R curve crack driving force diagram procedure described in Subarticle K-4310.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-29 Figure 4-10 Path Lines of Interest (Outlet Nozzle Levels A and B)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-30 4.4.2.2 Procedure for Evaluation

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Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-32

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3. Upper Shelf Toughness at an extension of 0.1 inch: Evaluation of upper-shelf toughness at a flaw extension of 0.1 inch is performed for the flaw depths, a = 3 in. + 0.10 in. (for path line nct) a = 0.25t +0.10 in. (for path line ntt)

Using:

SF = 1.15 p = Pa

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-34 Where Pa is the accumulation pressure for Levels A and B Service Loadings, such that:

J1 < J0.1 Where:

J1 = the applied J-integral for a structural factor of 1.15 on pressure, and a structural factor of 1.0 on thermal loading J0.1 = the J-integral resistance at a ductile flaw extension of 0.10 in.

4. Evaluation of Flaw Stability: Evaluation of flaw stability is performed through use of a crack driving force diagram procedure by comparing the slopes of the applied Jintegral curve and the material J-integral resistance curve, or J-R curve. The applied J-integral is calculated for a series of flaw depths corresponding to increasing amounts of ductile flaw extension. The applied pressure is the accumulation pressure for Levels A and B Service Loadings, Pa, and the structural factor (SF) on pressure is 1.25. Flaw stability at a given applied load is verified when the slope of the applied J-integral curve is less than the slope of the J-R curve at the point on the J-R curve where the two curves intersect.
5. Postulated Flaw: For nozzle corner path line (nct) crack depth is selected as 3 inches (that corresponds to 1/4 of the nozzle belt shell wall thickness) and for taper-transition region path line (ntt) crack depth is selected as 1/4 of the wall thickness at the taper section.

4.4.2.3 Evaluation for Levels A and B Service Loadings The ONS Unit 3 outlet nozzle EMA, in accordance with Appendix K, K-4200 applied stress intensity factor formulations at 72 EFPY demonstrate that the ASME Code,Section XI, Appendix K (Reference 4-4) acceptance criteria have been satisfied for Levels A and B Service Loadings considering both the CVNp and the Charpy methods.

The results of the analysis are as follows:

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-35

1. Figure 4-11 for the nozzle corner axial flaw, Figure 4-12 for the tapered transition axial flaw, and Figure 4-13 for the tapered transition circumferential flaw, show that with a factor of safety of 1.15 on pressure and 1.0 on thermal loading, the applied Jintegral (Japplied) is less than the lower bound J-integral of the material resistance curve at a ductile flaw extension of 0.10 inch (J0.1). The ratio of J0.1/Japplied for the CVNp method is [ ] for the nozzle corner axial flaw, tapered transition axial flaw, and tapered transition circumferential flaw, respectively. Also, the ratio of J0.1/Japplied for the Charpy method is [ ] for the nozzle corner axial flaw, tapered transition axial flaw, and tapered transition circumferential flaw, respectively, which all are greater than the required value of 1.0.
2. Figure 4-11 for the nozzle corner axial flaw, Figure 4-12 for the tapered transition axial flaw, and Figure 4-13 for the tapered transition circumferential flaw, show that with a factor of safety of 1.25 on pressure and 1.0 on thermal loading, flaw extensions are ductile and stable because the slope of the applied J-integral curve is less than the slope of the lower J-R curve at the point where the two curves intersect.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-36 Figure 4-11 J-Integral versus Axial Flaw Extension - Nozzle Corner (Outlet Nozzle Levels A and B)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-37 Figure 4-12 J-Integral versus Axial Flaw Extension - Taper Transition (Outlet Nozzle Levels A and B)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-38 Figure 4-13 J-Integral versus Circumferential Flaw Extension - Taper Transition (Outlet Nozzle Levels A and B)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-39 4.5 RPV Outlet Nozzle Forgings Equivalent Margins Analysis (Levels C and D Service Loadings)

The analytical procedure used for the 72 EFPY equivalent margins analysis, Levels C and D Service Loadings, of the ONS Unit 3 RPV outlet nozzle forging is in accordance with the ASME Code,Section XI, Appendix K, 2017 Edition (Reference 4-4), with selection of design transients based on the guidance in Regulatory Guide 1.161 (Reference 4-5) and consistent with Levels C and D design transients reported in BAW-2178, Revision 0, Supplement 1P-A, Revision 0 (Reference 4-2). Consistent with BAW-2178, Revision 0, Supplement 1P-A, Revision 0, only the most limiting item (i.e., RPV outlet nozzle) is evaluated for Levels C and D Service Loadings based upon the Levels A and B EMAs reported in Sections 4.3 and 4.4 above.

4.5.1 Material Properties and Levels C and D Service Loadings 4.5.1.1 J-Integral Resistance Model The J-integral resistance model is from NUREG/CR-5729 (Reference 4-6), i.e., RPV Jd base metal models reported in Table 11, i.e., Charpy Model and CVNP Model as described in Section 4.4.1.1.

4.5.1.2 Mechanical Properties The mechanical properties for the Levels C and D EMA are consistent with Section 4.4.1.2. The calculation of the plastic zone correction uses the yield stress limit, Sy = [ ]

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-40 The NUREG/CR-5729 RPV Jd base metal Charpy Model requires upper shelf energy at 72 EFPY as input. As such, a conservative fluence value of 1.0E+18 n/cm2 at the inside surface was assumed at the RPV nozzles with no attenuation, and the 72 EFPY CvUSE values at the RPV nozzles were calculated using RG 1.99, Position 1.2. The NUREG/CR-5729 RPV Jd CVNp model requires unirradiated upper shelf energy, discussed above, and fluence (assumed at 1.0E+18 n/cm2) as inputs. The estimate of 72 EFPY fluence of 1.0E+18 n/cm2 is conservative relative to that reported in Table 2-3 for the ONS RPV outlet nozzles.

4.5.1.3 Levels C and D Service Loadings Thermal Transients Levels C and D transients for the outlet nozzle forging are consistent with Levels C and D transients reported in BAW-2178, Revision 0, Supplement 1P-A, Revision 0 (Reference 4-2). There is one Level C transient and the other four (4) are Level D transients.

Pipe Loads In order to account for the influence of the nozzle loads, loads are developed for faulted service conditions and are shown in Table 4-4. These loads are also used for emergency (Level C) service level transient.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-41 Table 4-4 Service Levels C and D Outlet Nozzle Loads 4.5.1.4 Reactor Vessel Design Data Pertinent design data including the design pressure and geometry of the RPV shell and outlet nozzle forgings of the ONS Unit 3 reactor vessel are reported in Section 4.4.1.4.

4.5.2 Fracture Mechanics Evaluation 4.5.2.1 Methodology and Acceptance Criteria In accordance with the ASME Code,Section XI, Appendix K, Subarticle K-1200, the following analytical procedure was used for Levels C and D Service Loadings.

a. Flaws in the RPV outlet nozzle forging were postulated in accordance with the Subarticles K-2300 and K-2400. Three locations of postulated defects are illustrated in Figure 4-17: nozzle corner, nozzle boss, and taper region.
b. Loading conditions at the locations of the postulated flaws were determined for Levels C and D Service Loadings.
c. Material properties, including E, y, and the Jintegral resistance curve (JR curve), were determined at the locations of the postulated flaws. Youngs modulus and yield strength are addressed in Section 4.5.1.2. The JR curve is discussed in Section 4.5.1.1.
d. The postulated flaws were evaluated in accordance with the acceptance criteria of Subarticle K-2300 (both Levels C and D Service Loadings) by calculating the applied J-integral according to the procedure provided by Subarticle K-5210.

The applied J-integral was then evaluated to satisfy the criteria for flaw extension in Subarticle K-5220 and flaw stability in Subarticle K-5300.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-42 4.5.2.2 Procedure for Evaluation

[

]

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1. Procedure for Evaluating Levels C and D Service Loadings: The methodology used herein to evaluate Levels C and D Service Loadings is outlined as follows:

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-44

2. Temperature Range for Upper-Shelf Fracture Toughness Evaluations: Upper-shelf fracture toughness is determined through use of Charpy V-notch impact energy versus temperature plots by noting the temperature above which the Charpy energy remains on a plateau, maintaining a relatively high constant energy level. Similarly, fracture toughness can be addressed in three different regions on the temperature scale, i.e., a lower-shelf toughness region, a transition region, and an upper-shelf toughness region. Fracture toughness of the outlet nozzle forgings material is conservatively predicted by the ASME Code initiation toughness curve, KIc, in lower-shelf and transition regions. In the upper-shelf region, the upper-shelf toughness curve, KJc, is derived from the upper-shelf J-integral resistance model. When upper-shelf toughness is plotted versus temperature, a plateau-like curve develops that decreases slightly with increasing temperature. Since the present analysis addresses the low upper-shelf fracture toughness issue, only the upper-shelf temperature range, which begins at the intersection of KIc and the upper-shelf toughness curves, is considered.
3. Effect of Cladding Material on Stress Intensity Factor: The Finite Element model utilized to develop inputs for this analysis features cladding; therefore, the effect of cladding is included.

Significant stress discontinuities can be observed at plots of the through wall thermal stress at the cladding-to-base metal transition region. These stress discontinuities generate (in most cases) compressive stresses on the wetted surface of the shell and consequently relieve the overall tensile stresses which would open the postulated flaw. For this reason, the compressive stresses are neglected in the calculation of the final safety margin.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-45 4.5.2.3 Evaluation for Levels C and D Service Loadings The results of the low upper shelf toughness calculations for Levels C and D Service Loadings with projected fluence at 72 EFPY (80 calendar years) for Oconee Unit 3 Outlet Nozzle forging demonstrate that the acceptance criteria of the ASME Code,Section XI, Appendix K for Levels C and D Service Loadings have been met.

As shown in Table 4-5, the minimum equivalent margin for the outlet nozzle forging is

[ ] (minimum required is 1.0). The limiting location on the outlet nozzle forging is the nozzle corner and the limiting transient is the [

] transient.

The J-R curve data and applied J-integrals are plotted in Figure 4-14, Figure 4-15 and Figure 4-16 for the path lines shown in Figure 4-17, and demonstrate that:

1. With a structural factor of 1.0 on all loading, the applied J-integral (J1) is less than the J-integral of the material at a ductile flaw extension of 0.10 inch (J0.1).
2. With a structural factor of 1.0 on all loading, the flaw extensions are ductile and stable since the slope of the applied J-integral curve is less than the slope of the lower bound J-R curve at the point of intersection for both (CVNp and Charpy) material models.
3. The remaining ligament (limiting location at the safe end) is sufficient to preclude tensile instability by sufficient margin.

Note that the margins presented in Table 4-5 are calculated using lower bound J-R curve for both the Levels C and D Service Loadings, and that applied J-integral (J1) neglects the effect of cladding and attached piping which in these specific configurations decrease the value of applied J-integral.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-46 Table 4-5 ONS Unit 3 Outlet Nozzle EMA Results (Levels C and D)

Temperature Japplied Jresistance Location Margin T (°F) (lb/inch)

Nozzle Corner 590 [ ] [ ] [ ]

Nozzle Boss 467 [ ] [ ] [ ]

Nozzle Taper 492 [ ] [ ] [ ]

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-47 Figure 4-14 J-integral versus Flaw Extension for Nozzle Corner (Outlet Nozzle Levels C and D)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-48 Figure 4-15 J-integral versus Flaw Extension for Nozzle Boss (Outlet Nozzle Levels C and D)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-49 Figure 4-16 J-integral versus Flaw Extension for Nozzle Taper (Outlet Nozzle Levels C and D)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-50 Figure 4-17 Locations of Interest (Outlet Nozzle Levels C and D)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-51 4.6 References for Section 4.0 4-1. BAW-2192, Revision 0, Supplement 1P-A, Revision 0, Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessels of B&W Owners Reactor Vessel Working Group for Levels A & B Service Loads. ADAMS Accession Number ML19101A355 4-2. BAW-2178, Revision 0, Supplement 1P-A, Revision 0, Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessels of B&W Owners Reactor Vessel Working Group for Levels C & D Service Loads. ADAMS Accession Number ML19101A355 4-3. 10 CFR 50 Appendix G, Code of Federal Regulations, Title 10, Part 50 -

Domestic Licensing of Production and Utilization Facilities, Appendix G -

Fracture Toughness Requirements, Federal Register Vol. 60. No. 243, December 19, 1995 4-4. ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 2017 Edition 4-5. Regulatory Guide 1.161, Evaluation of Reactor Pressure Vessels With Charpy Upper Shelf Energy Less Than 50 ft-lb 4-6. NUREG/CR-5729, Multivariable Modeling of Pressure Vessel and Piping J-R Data 4-7. I.S. Raju and J.C. Newman, Jr., Stress-Intensity Factors for Internal and External Surface Cracks in Cylindrical Vessels, Journal of Pressure Vessel Technology, Vol. 104, pp. 293-298, November 1982 4-8. [

]

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 4-52 4-9. PWROG-17090-NP-A, Revision 0, Generic Rotterdam Forging and Weld Initial Upper-Shelf Energy Determination, ADAMS Accession Number ML20024E238

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-1 5.0 PRESSURIZED THERMAL SHOCK 5.1 Introduction Title 10 of the Code of Federal Regulation (CFR) 50.61 provides requirements for protection against pressurized thermal shock events for pressurized water reactors.

Licensees are required to perform an assessment of the projected values of pressurized thermal shock reference temperature (RTPTS) whenever a significant change occurs in projected values of RTPTS, or upon request for a change in the expiration date for the operation of the facility. Since RTPTS is a function of the end of license fluence, which is associated with the 60-year licensed operating period, the RTPTS calculations meet the criteria of 10 CFR 54.3(a) and have been identified as Time Limited Aging Analyses (TLAA) requiring evaluation for 80 years.

5.2 Regulatory Guidance for Subsequent License Renewal The regulatory guidance for NRC review of pressurized thermal shock evaluations performed in accordance with 10 CFR 54.21(c)(1)(ii) is reported in NUREG-2192, Section 4.2.3.1.3.2 (Reference 5-1) and is repeated below.

The documented results of the revised PTS analysis based on the projected neutron fluence at the end of the subsequent period of extended operation are reviewed for compliance with 10 CFR 50.61 or 10 CFR 50.61a.

The NRC staff confirms that the applicant has provided sufficient information for PTS for the subsequent period of extended operation as follows:

The applicant identified the neutron fluence at the clad-to-base metal interface for each beltline material at the expiration of the subsequent period of extended operation.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-2 There are two methodologies from 10 CFR 50.61 that can be used in the PTS analysis, based on the projected neutron fluence at the end of the subsequent period of extended operation. RTNDT is the reference temperature (NDT means nil-ductility temperature) used as an indexing parameter to determine the fracture toughness and the amount of embrittlement of a material. RTPTS is the reference temperature used in the PTS analysis and is related to RTNDT at the end of the facilitys operating license.

The first methodology does not rely on plant-specific surveillance data to calculate delta RTNDT (RTNDT, the mean value of the adjustment or shift in reference temperature caused by irradiation). The RTNDT is determined by multiplying a chemistry factor from the tables in 10 CFR 50.61 by a neutron fluence factor calculated from the neutron flux using Equation 3 in 10 CFR 50.61.

For the surveillance data to be defined as credible, the difference in the predicted values and the measured values for RTNDT must be less than 15.6 °C (28 °F) for weld metal components or less than 9.4 °C (17 °F) for base metal components. When a credible surveillance data set exists, the chemistry factor can be determined from these data in lieu of a value from the table in 10 CFR 50.61. The standard deviation for the RTNDT used in the margin term assessment (e.g., ) of the RTPTS calculations may be reduced from 15.6 °C (28 °F) to 7.8 °C (14 °F) for welds or from 9.4 °C (17 °F) to 4.7 °C (8.5 °F) for base metal materials. However, need not exceed one-half of the RTNDT value used in the RTPTS calculations.

To confirm that the PTS analysis results in RTPTS values below the screening criteria in 10 CFR 50.61 at the end of the subsequent period of extended operation, the applicant provides the following:

1. For each beltline material, provide the unirradiated RTNDT, the method of calculating the unirradiated RTNDT (either generic or plant-specific), the margin, chemistry factor, the method of calculating the chemistry factor, the mean value for the shift in transition temperature, and the RTPTS value.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-3

2. If there are two or more data for a surveillance material that is from the same heat of material as the beltline material, provide analyses to determine whether the data are credible in accordance with NRC RG 1.99, Revision 2, and whether the margin value used in the analysis is appropriate.
3. If a surveillance program does not include the vessel beltline controlling material but two or more data sets are available from other beltline materials, then provide an analysis of the data in accordance with NRC RG 1.99, Revision 2, Regulatory Position C.2.1, to show that the results either bound or are comparable to the values that would be calculated for the same materials using Regulatory Position C.1.1.

If the PTS screening criteria in 10 CFR 50.61 are projected to be exceeded during the subsequent period of extended operation, an analysis based on NRC RG 1.154 or 10 CFR 50.61a may be submitted for review. For applicants with PTS analysis based upon an NRC-approved submittal using 10 CFR 50.61a, the analysis is revised to reflect the subsequent period of extended operation.

5.3 Methodology 10 CFR 50.61(c) provides two methods for determining RTPTS. These methods are also described in Section 1.1 (i.e., Position 1.1) and Section 2.1 (i.e., Position 2.1) of Regulatory Guide 1.99, Revision 2 (Reference 5-2). Position 1.1 applies for material without credible surveillance data available and Position 2.1 is used for material with two or more credible surveillance data sets available. The RTPTS values are calculated for both Positions 1.1 (all 3 Units) and 2.1 (ONS Unit 3 only) by following the guidance in Regulatory Guide 1.99, Revision 2, using the copper and nickel content of the Units 1, 2, and 3 RPV materials, and subsequent period of extended operation 72 EFPY fluence projections at the clad/base metal interface.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-4 5.4 Traditional Beltline and Extended Beltline 72 EFPY Pressurized Thermal Shock The 10 CFR 50.61(c) methods were used with the 72 EFPY base metal/clad interface fluence values reported in Table 2-1, Table 2-2, and Table 2-3 to calculate RTPTS values for Oconee Units 1, 2, and 3 at 72 EFPY. The subsequent period of extended operation 72 EFPY RTPTS calculations for Oconee Units 1, 2, and 3 are summarized in Table 5-10, Table 5-11, and Table 5-12.

10 CFR 50.61(b)(2) establishes screening criteria for RTPTS as 270°F for plates, forgings, and longitudinal welds and 300°F for circumferential welds. In accordance with SECY 82-465, Enclosure A (Reference 5-3), which provides the bases for the PTS screening criteria, the NRC staff reported that further justification was needed from the B&W owners for longer term use of the PTS screening criteria for B&W plants. The B&WOG provided justification for long term use of the PTS screening criteria for B&W plants through BAW-1791 (Reference 5-4), and Duke Energy provided a separate justification for applicability of the PTS screening criteria to B&W-designed plants through Reference 5-5.

As reported in BAW-1791, the methods of probabilistic risk assessment were applied to a comprehensive set of initiating events to determine the frequencies of possible PTS sequences for a generic configuration of the B&W-designed 177-fuel assembly plant.

Best-estimate thermal-hydraulics analyses were used to bound the temperature and pressure histories of PTS sequences in nine categories. The thermal responses and frequencies were input to a probabilistic fracture mechanics correlation to relate the probabilities of through-wall vessel cracking to mean surface transition temperatures.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-5 The aggregate results were compared to the NRC evaluation set forth in SECY 82-465, Enclosure A. It was concluded that the aggregate frequencies for crack initiation without arrest in B&W designed plants are more than one order of magnitude less than those established by the NRC for other pressurized water reactor designs. BAW-1791 supplements and extends earlier pressurized thermal shock evaluations for B&W-designed plants, and the information presented confirms the applicability, without modification, of the proposed PTS screening criteria of SECY 82-465 to all B&W-designed plants. Therefore, the PTS screening criteria of 270°F for plates, forgings, and longitudinal welds and 300°F for circumferential welds remain applicable to the Oconee units for the subsequent period of extended operation.

All of the ONS Units 1, 2, and 3 RPV materials are below the RTPTS screening criteria values of 270°F for base metal and longitudinal welds, and 300°F for circumferentially oriented welds through the subsequent period of extended operation. For Unit 1, the limiting materials are axial welds SA-1426 and SA-1430 with a projected value of RTPTS at 72 EFPY of 207.2°F (screening criterion of 270°F). For Unit 2, the limiting material is circumferential weld WF-25 with a projected value of RTPTS at 72 EFPY of 245.1°F (screening criterion of 300°F). For Unit 3, the limiting material is circumferential weld WF-67 with a projected value of RTPTS at 72 EFPY of 240.2°F (screening criterion of 300°F).

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-6 5.4.1 Material Properties Including Copper, Nickel, and Initial RTNDT, Material properties required for the calculation of 72 EFPY RTPTS are reported in Table 5-10, Table 5-11, and Table 5-12, and include copper content (weight %), nickel content (weight %), and initial RTNDT. Copper content (weight %) and a comparison to the current licensing basis is addressed in Sections 3.4.1 and 3.4.2, above. For traditional beltline locations, the nickel content, and initial RTNDT are consistent with the current licensing basis (Section 1.0, Reference 1-2) with some exceptions as reported in Section 5.4.2. For extended beltline locations (i.e., RPV inlet and outlet nozzles and associated welds and the transition forging), material properties have not been reported to the NRC and nickel content, initial RTNDT, and I for SLR for these items are reported below.

Nickel Content Nickel content for ONS Unit 1, ONS Unit 2, and ONS Unit 3 RPV outlet nozzle forgings and inlet nozzle forgings are plant-specific and are directly from the product analysis reported in the certified material test reports for the specific heat of material. Nickel content of the transition forgings are from the product analysis of specific heats of material measurements as reported in BAW-1820 (Reference 5-6). The nickel content of extended beltline Linde 80 welds are mean values from weld wire heat specific measurements, as permitted by Regulatory Guide 1.99, Revision 2. The nickel content in weight % for extended beltline locations (base metal and welds) are reported in Table 5-10, Table 5-11, and Table 5-12 for ONS Unit 1, ONS Unit 2, and ONS Unit 3, respectively.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-7 Initial RTNDT and I The initial RTNDT for the ONS Unit 1 and ONS Unit 2 RPV inlet and outlet nozzle forgings, ONS Unit 1 and ONS Unit 2 transition forgings, and the ONS Unit 3 RPV inlet nozzle forgings is a generic mean from the data presented in Table 5-1 below. For ASME SA-508 (ASTM A 508) Class 2 forgings where no initial (unirradiated) reference temperatures are available, a statistical evaluation was performed on the available data to obtain a generic mean value. The mean RTNDT was found to be [ ], and the standard deviation is [ ].

The data in Table 5-1 was obtained from forgings manufactured by US suppliers. A review of forging data from non-US suppliers indicates that the data for the non-US forging materials may not be representative of US-supplied materials. Therefore, the data illustrated in Table 5-1 is used for forgings manufactured by US suppliers, and not for initial RTNDT values for forgings from a non-US supplier, such as Klockner Werke.

ONS Unit 3 Klockner-Werke RPV Outlet Nozzle Forgings and Transition Forging The initial RTNDT and I of the ONS Unit 3 Klockner Werker forgings are estimated from the Rotterdam Dockyard data reported in Table 5-2 below. The generic mean is

[ ] and standard deviation of [ ]. The actual suppliers of the forgings to Rotterdam Dockyards for the plants listed in Table 5-2 are not known but likely included Klockner-Werke, Fried-Krupp Huttenwerke AG, and Rheinstahl Huttenwerke AG based on review of PWROG-17090, Table 2 (Reference 5-7). As such, the above generic mean of [ ] and standard deviation of [ ] are acceptable due to consideration of the class of European forgings that are representative of the ONS Unit 3 RPV outlet nozzle forgings and transition forging.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-8 Extended Beltline Linde 80 Welds For weld metals fabricated with Linde 80 flux where no initial (unirradiated) reference temperatures are available (e.g., RPV nozzle welds), a statistical evaluation was performed on the available data to obtain a generic mean value. The generic mean was found to be [ ], and the standard deviation is [ ], based on test data reported in Table 5-3 below.

For weld wire heat number 299L44, the initial RTNDT is [ ] and its standard deviation is [ ], respectively, based on the group data for this weld wire based on test data reported in Table 5-4 below. This data may be applied outside the traditional RPV beltline.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-9 Table 5-1 Reference Temperature Data for Forging Materials Used to Determine Generic Initial Reference Temperature

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-10 Table 5-2 Generic Initial RTNDT for ASTM A508 Class 2 Forgings Supplied to Rotterdam Dockyard

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-11 Table 5-3 Reference Temperature Data for Linde 80 Weld Metals Used to Determine Generic Initial Reference Temperature

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-12 Table 5-4 Reference Temperature Data for Weld Wire Heat No. 299L44 Used to Determine Generic Initial Reference Temperature

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-13 5.4.2 Comparison to Current Licensing Basis For traditional beltline materials, differences between the current licensing basis (see Section 1.0) and the Ni wt% content and initial RTNDT values utilized for subsequent license renewal are identified in Table 5-5 and Table 5-6, respectively. For items not listed, the CLB and SLR Ni wt% and initial RTNDT values are consistent. In addition, Ni wt% and Initial RTNDT values reported in the NRC Reactor Vessel Integrity Database (RVID2, Reference 5-8) database are included for information. The data and information provided by licensees in their responses to the staff's Generic Letter (GL) 92-01, Revision 1 close-out letters, and in response to GL 92-01, Revision 1, Supplement 1, are included in RVID2. This database includes updates from June 1999

- July 2000. Duke Energys responses to GL 92-01, GL 92-01, Revision 1, and GL 92-01, Revision 1, Supplement 1 are based on generic topical reports. The generic Topical Reports that provided Oconees response to GL 92-01 are reported in Section 3.4.2.

The Ni wt% values and initial RTNDT reported herein for SLR were calculated with additional information and subsequent to those found in the CLB; therefore, they are an improved representative of the materials.

Nickel differences-CLB versus SLR-Traditional Beltline Materials Differences for Ni content values are identified in Table 5-5. The value of 0.57% for heat 8T1762 is the mean value of 6 weld qualification samples and is identical to the nickel wt% reported in the NRC RVID2.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-14 Initial RTNDT-CLB versus SLR-Traditional Beltline Materials Differences for the mean and standard deviation for initial RTNDT traditional beltline items are identified in Table 5-6. The NRC RVID2 database has not been updated to reflect the changes to Linde 80 weld initial RTNDT based on BAW-2308, Revisions 1-A and 2-A (Reference 5-9). The mean and standard deviation for initial RTNDT values reported in the CLB for forgings and plates were calculated based a dataset comprised of Bellefonte Unit 1 and Unit 2 (BLN-1 and BLN-2) reactor vessel forgings and reactor vessel & steam generator plates, respectively. BLN-1 and BLN-2 are later generation B&W units and, therefore, may have more favorable initial RTNDT values than would be representative of the earlier vintage B&W units.

Therefore, material data that were more representative of the earlier vintage B&W units are utilized and initial RTNDT and associated I of traditional beltline plates, forgings, and welds were recalculated for subsequent license renewal with removal of the BLN-1 and BLN-2 data. These more representative results for the mean and standard deviation for initial RTNDT values are reported herein and apply to forgings, plates, and welds fabricated with Linde 80 flux and are listed in Table 5-6. Details of the revised initial RTNDT values relative to NRC RVID2 are provided below if the revision for SLR provides greater margins to embrittlement limits (e.g., if either initial RTNDT is reduced or I is reduced). Traditional beltline forging and plate material items for which material properties have been revised such that margins may be reduced to embrittlement limits for SLR are as follows. Traditional beltline Linde 80 welds have all been revised based on BAW-2308.

  • Oconee Unit 1LNB forging AHR 54, IS plates C2197-2, US plates C3265-1 and C3278-1, LS plate C2800-1 and C2800-2
  • Oconee Unit 2LNB forging AMX 77
  • Oconee Unit 3LNB forging 4680

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-15 For those instances where test data is not available for the determination of unirradiated initial reference temperature (RTNDT), for base metal plate, forgings, and weld metals fabricated with Linde 80 flux, generic mean values have been established as follows.

Plates For ASME SA-533 (ASTM A 533) Grade B Class 1 plates and ASME SA-302 (ASTM A 302) Grade B Modified plates where no initial (unirradiated) reference temperatures are available, a statistical evaluation was performed on the available data to obtain a generic mean value. The mean was found to be [ ], and the standard deviation is [ ]. The data is reported in Table 5-7.

Forgings For ASME SA-508 (ASTM A 508) Class 2 forgings where no initial (unirradiated) reference temperatures are available, a statistical evaluation was performed on the available data to obtain a generic mean value. The mean was found to be [ ],

and the standard deviation is [ ]. The data is reported in Table 5-8. This data was obtained from forging manufactured from US suppliers and is not applicable to the following ONS Unit 3 forgings supplied by Klockner Werke and Rotterdam Dockyard:

ONS Unit 3 RPV outlet nozzle and transition forgings (see Section 5.4.1) by Klockner Werke, and ONS Unit 3 lower nozzle belt forging by Rotterdam Dockyard.

LNB Forging (4680) for ONS Unit 3 was procured from the Rotterdam Dockyard, for which measured values were not available from the time of manufacturing; therefore, the values reported in the CLB and NRC RVID2 are generic initial RTNDT values determined by B&W from U.S. forging companies, e.g., Ladish Co, Bethlehem Steel Corporation, or Midvale-Hepppenstall. Framatome located archive ONS Unit 3 LNB forging material and testing was performed in accordance with ASTM E208-06 to determine the nil-ductility transition temperature. Testing resulted in an initial RTNDT of

+10 °F with I=0.0 °F, which is reported in Table 5-6.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-16 For direct test measurements, ASTM E208, Section 17.1, addresses precision and bias and contains the following: no information is presented about either the precision or bias of Test Method E208 since the test result is non-quantitative. As such, Framatomes position is that when actual drop weight test data is available for a specific plate, forging, or weld to establish TNDT in accordance with ASTM E208, determination of uncertainty in the test method is not applicable since the test result is non-quantitative and I=0.0. This position is consistent with that reported in BAW-2222 (Reference 5-10), Reactor Vessel Working Group Response to Closure Letters to NRC Generic Letter 9201, Revision 1 for Oconee Units 1, 2, and 3, for all plant-specific entries. When establishing a generic mean, the margin (I) is calculated from the statistical evaluation of the test data using the standard deviation.

As such, I values have been reduced to 0.0 °F for ONS Unit 1 plates C3265-1 and C2800-2 based on test data reported in Table 5-7. In each case, initial RTNDT values have increased and are based on the measured values reported in Table 5-7.

Weld Metal Fabricated with Linde 80 Flux For weld metals fabricated with Linde 80 flux where no initial (unirradiated) reference temperatures are available, a statistical evaluation was performed on the available data to obtain a generic mean value. The mean value initial RTNDT is [ ], and the standard deviation is [ ] as reported in Section 5.4.1 above.

For weld wire heat number 299L44, the mean value initial RTNDT is [ ], and standard deviation is [ ], based on the group data for this weld wire as reported in Section 5.4.1 above.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-17 BAW-2308 (Reference 5-9) was prepared for the B&W Owners Group (B&WOG)

Reactor Vessel Working Group (RVWG) to justify alternative initial reference temperatures (IRTNDT) for the Linde 80 beltline welds in B&W fabricated reactor vessels.

The alternative IRTNDT was determined based on brittle-to-ductile transition range fracture toughness test data of these weld metals obtained in accordance with ASTM Standard E1921 and using ASME Boiler and Pressure Vessel Code Case N-629.

BAW-2308 was submitted to the NRC for review and acceptance as a B&WOG topical for application to the pressurized thermal shock (PTS) rule (10 CFR 50.61) and 10 CFR Part 50, Appendix G, pressure temperature limits. The heat-specific and generic RTTo and initial margin terms are tabulated in Table 5-9. The data in Table 5-9 is an excerpt from Table 9 of BAW-2308 Revision 2-A.

Duke Energy requested an exemption (Reference 5-11) from the requirements of 10 CFR 50.61 and 10 CFR Part 50, Appendix G to revise plant-specific initial RTNDT values using BAW-2308, Rev 1-A, and its supplement BAW-2308, Rev 2A. This exemption request was approved by the NRC staff on April 26, 2012 (Reference 5-12). There is no specific time dependency associated with this exemption request and it may appropriately be applied to traditional RPV beltline locations for subsequent license renewal. The exemption may not be applied to extended beltline locations, as neither the Duke Energy exemption request nor the NRC SER addressed applicability of the exemption request to extended beltline locations.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-18 Table 5-5 Traditional Beltline Ni Content Differences between CLB and SLR

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-19 Table 5-6 Traditional Beltline Initial RTNDT Differences between CLB and SLR

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-20 Table 5-7 Reference Temperature Data for Plate Materials Used to Determine Generic Initial Reference Temperature

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-21 Table 5-8 Reference Temperature Data for Forging Materials Used to Determine Generic Initial Reference Temperature

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-22 Table 5-9 Heat Specific and Generic Initial RTTo with Associated Initial Margin With Proposed ASTM E1921 Loading Rate Adjustment Linde 80 Heat Initial Margin IRTT0 (°F) aI (°F) 406L44 -98.0 11.6 71249 -53.5 12.8 72105 -31.1 13.7 821T44 -84.2 9.6 299L44 -74.3a 12.8a 72442 -33.2 12.2 72445 -72.5 12.0 61782b -58.5 15.4 Other heats -48.6a 18.0a Notes:

a. This value includes new data from that reported within BAW-2308-1A.
b. This heat was not reported within BAW-2308-1A.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-23 5.5 References for Section 5.0 5-1. NUREG-2192, Standard Review Plan Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants 5-2. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials 5-3. SECY 82-465, Pressurized Thermal Shock, ADAMS Accession Number ML16232A574 5-4. BAW-1791, B&W Owners Group Probabilistic Evaluation of Pressurized Thermal Shock, Phase 1 Report, ADAMS Accession Number ML20024A860, 8307010051 (legacy) 5-5. Duke Power Company Comments on Pressurized Thermal Shock Screening Criteria, ADAMS Accession Number ML20028E598, 8301280040 (legacy) 5-6. BAW-1820, Babcock & Wilcox Owner's Group 177-Fuel Assembly Reactor Vessel and Surveillance Program Materials Information, ADAMS Accession Number ML20108D545 5-7. PWROG-17090-NP-A, Revision 0, Generic Rotterdam Forging and Weld Initial Upper-Shelf Energy Determination, ADAMS Accession Number ML20024E238 5-8. Reactor Vessel Integrity Database Version 2.0.1, https://www.nrc.gov/reactors/operating/ops-experience/reactor-vessel-integrity/database-overview.html 5-9. BAW-2308, Revision 1-A and 2-A, Initial RTNDT of Linde 80 Weld Metals ADAMS, Accession Number ML081270388 for 2-A. Final Safety Evaluation for Revision 1 is ML052070408 5-10. BAW-2222, Reactor Vessel Working Group Response to Closure Letters to NRC Generic Letter 92-01, Revision 1, ADAMS Legacy No. 9407060101

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-24 5-11. Duke Energy Letter Request for Exemption from Certain Requirements Contained in 10 CFR 50.61 and 10 CFR 50, Appendix G, ADAMS Accession No. ML11223A010 5-12. Oconee Nuclear Station, Units 1, 2, and 3, Exemption From the Requirements of 10 CFR PART 50.61 and 10 CFR Part 50, Appendix G (TAC NOS. ME7000, ME7001, ME7002, ME7003, ME7004, AND ME7005), ADAMS Accession No. ML120580196

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-25 Table 5-10 Evaluation of Reactor Vessel Pressurized Thermal Shock Toughness Properties at 72 EFPY - Oconee Unit 1 Applicable 72 EFPY RTPTS at 72 RTPTS at 72 Reactor Vessel Beltline/Beltline Extended Initial RTNDT Chemistry Margin at 72 Screening Pass Matl. Ident. Heat Number Type Cu wt% Ni wt% Fluence EFPY EFPY Region Location (°F) Factor 2 EFPY (°F) Criteria Criteria (n/cm ) (°F) (°F)

(°F)

Lower Nozzle Belt (LNB) Forging AHR 54 ZV-2861 A-508, Cl. 2 [ ] [ ] [ ] 115.3 2.60E+18 73.1 [ ] 149.9 270 Y SA-302 Gr. B Intermediate Shell (IS) Plates C2197-2 C2197-2 0.15 0.50 6.7 104.5 1.78E+19 121.0 55.6 183.4 270 Y Mod.

SA-302 Gr. B Upper Shell (US) Plate C3265-1 C3265-1 0.10 0.50 21 65.0 2.02E+19 77.5 34.0 132.5 270 Y Mod.

SA-302 Gr. B US Plate C3278-1 C3278-1 0.12 0.60 6.7 83.0 2.02E+19 98.9 55.6 161.2 270 Y Mod.

SA-302 Gr. B Lower Shell (LS) Plate C2800-1 C2800-1 0.11 0.63 6.7 74.5 2.02E+19 88.8 55.6 151.1 270 Y Mod.

SA-302 Gr. B LS Plate C2800-2 C2800-2 0.11 0.63 20 74.5 2.02E+19 88.8 34.0 142.8 270 Y Mod.

Outlet Nozzle Forging (ONF) 1 NA 122S316VA2 A-508, Cl. 2 [ ] [ ] [ ] [ ] 3.38E+17 [ ] [ ] 86.1 270 Y ONF 2 NA 122S316VA1 A-508, Cl. 2 [ ] [ ] [ ] [ ] 3.38E+17 [ ] [ ] 85.9 270 Y Inlet Nozzle Forging (INF) 1 NA 123S346VA1 A-508, Cl. 2 [ ] [ ] [ ] [ ] 1.57E+17 [ ] [ ] 78.7 270 Y INF 2 NA 123S346VA2 A-508, Cl. 2 [ ] [ ] [ ] [ ] 1.57E+17 [ ] [ ] 78.8 270 Y INF 3 NA 124S502VA1 A-508, Cl. 2 [ ] [ ] [ ] [ ] 1.57E+17 [ ] [ ] 78.9 270 Y INF 4 NA 124S502VA2 A-508, Cl. 2 [ ] [ ] [ ] [ ] 1.57E+17 [ ] [ ] 78.9 270 Y Transition Forging NA 122S347VA1 A-508, Cl. 2 [ ] [ ] [ ] [ ] 2.52E+17 [ ] [ ] 82.8 270 Y NA 8T1762 Linde 80 [ ] [ ] [ ] [ ] 3.38E+17 [ ] [ ] 84.2 270 Y LNB to ONF Welds NA 299L44 Linde 80 [ ] [ ] [ ] [ ] 3.38E+17 [ ] [ ] 111.2 270 Y NA 8T1554B Linde 80 [ ] [ ] [ ] [ ] 3.38E+17 [ ] [ ] 80.9 270 Y NA 8T1762 Linde 80 [ ] [ ] [ ] [ ] 1.57E+17 [ ] [ ] 63.0 270 Y LNB to INF Welds NA 299L44 Linde 80 [ ] [ ] [ ] [ ] 1.57E+17 [ ] [ ] 78.1 270 Y NA 8T1554B Linde 80 [ ] [ ] [ ] [ ] 1.57E+17 [ ] [ ] 61.1 270 Y LNB to IS Circ. Weld (100%) SA-1135 61782 Linde 80 0.23 0.52 -58.5 167.0 2.82E+18 109.3 63.9 114.7 300 Y IS Long. Welds (Both 100%) SA-1073 1P0962 Linde 80 0.21 0.64 -48.6 170.6 1.33E+19 184.1 66.6 202.1 270 Y IS to US Circ. Weld (ID 61%) SA-1229 71249 Linde 80 0.23 0.59 -53.5 167.6 1.79E+19 194.4 61.6 202.5 300 Y US Long. Welds (Both 100%) SA-1493 8T1762 Linde 80 0.19 0.57 -48.6 167.0 1.31E+19 179.5 66.6 197.5 270 Y US to LS Circ. Weld (100%) SA-1585 72445 Linde 80 0.22 0.54 -72.5 167.0 1.97E+19 197.9 60.9 186.3 300 Y LS Long. Weld (100%) SA-1426 8T1762 Linde 80 0.19 0.57 -48.6 167.0 1.62E+19 189.2 66.6 207.2 270 Y LS Long. Weld (100%) SA-1430 8T1762 Linde 80 0.19 0.57 -48.6 167.0 1.62E+19 189.2 66.6 207.2 270 Y LS to Transition Forging Circ. Weld (100%) WF-9 72445 Linde 80 [ ] [ ] [ ] 158.0 2.52E+17 31.3 [ ] 76.7 300 Y

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-26 Table 5-11 Evaluation of Reactor Vessel Pressurized Thermal Shock Toughness Properties at 72 EFPY - Oconee Unit 2 72 EFPY RTPTS at 72 RTPTS at 72 Applicable Reactor Vessel Beltline/Beltline Initial RTNDT Chemistry Margin at 72 Pass Matl. Ident. Heat Number Type Cu wt% Ni wt% Fluence EFPY EFPY Screening Extended Region Location (°F) Factor 2 EFPY (°F) Criteria (n/cm ) (°F) (°F) Criteria (°F)

LNB Forging AMX 77 123T382 A-508, Cl. 2 [ ] [ ] [ ] [ ] 1.67E+19 [ ] [ ] 155.1 270 Y US Forging AAW 163 3P2359 A-508, Cl. 2 0.04 0.75 20 26.0 1.90E+19 30.6 30.6 81.1 270 Y LS Forging AWG 164 4P1885 A-508, Cl. 2 0.02 0.80 20 20.0 1.89E+19 23.5 23.5 67.0 270 Y ONF 1 NA [ ] A-508, Cl. 2 [ ] [ ] [ ] [ ] 3.14E+17 [ ] [ ] 85.2 270 Y ONF 2 NA [ ] A-508, Cl. 2 [ ] [ ] [ ] [ ] 3.14E+17 [ ] [ ] 85.4 270 Y INF 1 NA [ ] A-508, Cl. 2 [ ] [ ] [ ] [ ] 1.46E+17 [ ] [ ] 78.2 270 Y INF 2 NA [ ] A-508, Cl. 2 [ ] [ ] [ ] [ ] 1.46E+17 [ ] [ ] 78.4 270 Y INF 3 NA [ ] A-508, Cl. 2 [ ] [ ] [ ] [ ] 1.46E+17 [ ] [ ] 78.4 270 Y INF 4 NA [ ] A-508, Cl. 2 [ ] [ ] [ ] [ ] 1.46E+17 [ ] [ ] 78.3 270 Y Transition Forging NA [ ] A-508, Cl. 2 [ ] [ ] [ ] [ ] 2.34E+17 [ ] [ ] 82.2 270 Y NA 8T1762 Linde 80 [ ] [ ] [ ] [ ] 3.14E+17 [ ] [ ] 81.8 270 Y LNB to ONF Welds NA 72445 Linde 80 [ ] [ ] [ ] [ ] 3.14E+17 [ ] [ ] 83.9 270 Y NA 8T1762 Linde 80 [ ] [ ] [ ] [ ] 1.46E+17 [ ] [ ] 61.5 270 Y LNB to INF Welds NA 72445 Linde 80 [ ] [ ] [ ] [ ] 1.46E+17 [ ] [ ] 62.6 270 Y LNB to US Circ. Weld (100%) WF-154 406L44 Linde 80 0.27 0.59 -98.0 182.6 1.68E+19 208.7 60.6 171.3 300 Y US to LS Circ. Weld (100%) WF-25 299L44 Linde 80 0.34 0.68 -74.3 220.6 1.85E+19 257.8 61.6 245.1 300 Y LS to Transition Forging Circ. Weld (100%) WF-112 406L44 Linde 80 [ ] [ ] [ ] 182.6 2.34E+17 [ ] [ ] 82.1 300 Y

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 5-27 Table 5-12 Evaluation of Reactor Vessel Pressurized Thermal Shock Toughness Properties at 72 EFPY - Oconee Unit 3 72 EFPY RTPTS at 72 RTPTS at 72 Applicable Reactor Vessel Beltline/Beltline Initial RTNDT Chemistry Margin at 72 Pass Matl. Ident. Heat Number Type Cu wt% Ni wt% Fluence EFPY EFPY Screening Extended Region Location (°F) Factor 2 EFPY (°F) Criteria (n/cm ) (°F) (°F) Criteria (°F) 10 CFR 50.61 (CF Tables)

LNB Forging 4680 4680 A-508, Cl. 2 0.15 0.91 10 116.1 1.74E+19 133.8 34.0 177.8 270 Y US Forging AWS 192 522314 A-508, Cl. 2 0.01 0.73 40 20.0 1.98E+19 23.7 23.7 87.5 270 Y LS Forging ANK 191 522194 A-508, Cl. 2 0.02 0.76 40 20.0 1.97E+19 23.7 23.7 87.4 270 Y ONF 1 NA [ ] A-508, Cl. 2 [ ] [ ] [ ] [ ] 3.39E+17 [ ] [ ] 128.6 270 Y ONF 2 NA [ ] A-508, Cl. 2 [ ] [ ] [ ] [ ] 3.39E+17 [ ] [ ] 128.6 270 Y INF 1 NA [ ] A-508, Cl. 2 [ ] [ ] [ ] [ ] 1.58E+17 [ ] [ ] 79.0 270 Y INF 2 NA [ ] A-508, Cl. 2 [ ] [ ] [ ] [ ] 1.58E+17 [ ] [ ] 79.0 270 Y INF 3 NA [ ] A-508, Cl. 2 [ ] [ ] [ ] [ ] 1.58E+17 [ ] [ ] 78.8 270 Y INF 4 NA [ ] A-508, Cl. 2 [ ] [ ] [ ] [ ] 1.58E+17 [ ] [ ] 78.8 270 Y Transition Forging NA [ ] A-508, Cl. 2 [ ] [ ] [ ] [ ] 2.50E+17 [ ] [ ] 121.9 270 Y NA 72105 Linde 80 [ ] [ ] [ ] [ ] 3.39E+17 [ ] [ ] 103.4 270 Y LNB to ONF Welds NA 406L44 Linde 80 [ ] [ ] [ ] [ ] 3.39E+17 [ ] [ ] 96.5 270 Y NA 72105 Linde 80 [ ] [ ] [ ] [ ] 1.58E+17 [ ] [ ] 73.9 270 Y LNB to INF Welds NA 72102 Linde 80 [ ] [ ] [ ] [ ] 1.58E+17 [ ] [ ] 65.0 270 Y NA 82102 Linde 80 [ ] [ ] [ ] [ ] 1.58E+17 [ ] [ ] 91.8 270 Y LNB to US Circ. Weld (100%) WF-200 821T44 Linde 80 0.24 0.63 -84.2 178.0 1.75E+19 205.4 59.2 180.4 300 Y US to LS Circ. Weld (ID 75%) WF-67 72442 Linde 80 0.26 0.60 -33.2 180.0 1.93E+19 212.4 61.1 240.2 300 Y LS to Transition Forging Circ. Weld (100%) WF-169-1 8T1554 Linde 80 [ ] [ ] [ ] 143.9 2.50E+17 28.4 [ ] 72.0 300 Y 10 CFR 50.61 (Surveillance Data)- Position 2.1 US Forging AWS 192 522314 A-508, Cl. 2 0.01 0.73 40 36.0 1.98E+19 42.7 34.0 116.7 270 Y LS Forging ANK 191 522194 A-508, Cl. 2 0.02 0.76 40 17.4 1.97E+19 20.6 20.6 81.2 270 Y

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 6-1 6.0 P-T LIMITS 6.1 Introduction 10 CFR Part 50, Appendix G requires that the reactor pressure vessel (RPV) be maintained within established pressure-temperature (P-T) limits, including heatup and cooldown operations. These limits specify the maximum allowable pressure as a function of reactor coolant temperature. As the RPV is exposed to increased neutron irradiation, its fracture toughness is reduced. The P-T limits must account for the anticipated RPV fluence. The current ONS 54 EFPY P-T limits are based upon fluence projections for 60 years of plant operation (Reference 6-1). Because they were based upon a fluence assumption of 60 years of operation, the P-T limits analyses meet the definition of 10 CFR 54.3(a) and have been identified as time-limited aging analyses (TLAA).

6.2 Regulatory Guidance for Subsequent License Renewal Since the P-T limits will be updated through the 10 CFR 50.90 process at a later, more appropriate date, the effects of aging on the intended function(s) of the RPVs will be adequately managed for the subsequent period of extended operation. The ONS Reactor Vessel Material Surveillance Program (XI.M31), ONS Neutron Fluence Monitoring program (X.M2), and plant Technical Specifications will ensure that updated P-T limits based upon updated ART values will be submitted to the NRC for approval prior to exceeding the period of applicability for Units 1, 2, and 3. The regulatory guidance for NRC review of P-T limit evaluations performed in accordance with 10 CFR 54.21(c)(1)(iii) is reported in NUREG-2192, Section 4.2.3.1.4.3 (Reference 6-2), and is reported below.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 6-2 Updated P-T limits for the subsequent period of extended operation must be established and implemented prior to entry into the subsequent period of extended operation. The 10 CFR 50.90 process for P-T limits located in the TS LCOs or the TS Administrative Controls Process for P-T limits that are administratively amended through a PTLR process can be considered adequate AMPs within the scope of 10 CFR 54.21(c)(1)(iii), such that P-T limits will be maintained through the subsequent period of extended operation.

For plants whose P-T limits are controlled by an applicable Administrative Controls TS Section and an NRC-approved PTLR process, the methodologies referenced in the applicable TS Section are reviewed to verify that they will comply with the requirements in 10 CFR Part 50, Appendix G and conform to the recommended position for minimum methodology contents in GL 96-03. Otherwise, the methodology bases for generating updates of the P-T limits during the subsequent period of extended operation are reviewed to determine whether a 10 CFR 54.22-implemented license amendment and TS change of the methodology requirements is necessary for the SLRA.

At present, the Oconee reactor coolant system P-T limits reside within plant Technical Specifications (Reference 6-3), Section 3.4.3, and are applicable to 54 EFPY (60 years). The ONS License Amendment Request for Measurement Uncertainty Recapture Power Uprate (MUR), February 19, 2020, ADAMS Accession Number ML20050D379 (Reference 6-4) reduces the applicability of these HU/CD curves as follows.

  • Reduces the Applicability for the RCS Heatup and Cooldown limit curves from 54 EFPY to 44.6 Effective Full Power Years (EFPY) for Unit 1, to 45.3 EFPY for Unit 2, and to 43.8 EFPY for Unit 3 based on updated reactor vessel (RV) material evaluations discussed in Section IV.1 of the MUR submittal.

Therefore, the 10 CFR 50.90 process will be used by Duke Energy to update the P-T limits prior to entering the subsequent period of extended operation.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 6-3 6.3 P-T Limits at 72 EFPY Consistent with the Oconee 60-year license renewal application, NUREG-1723 (Reference 6-5), Section 4.2.4.3.1, Duke Energy prepared 72 EFPY (equivalent to 80 years of operation) P-T limits to demonstrate that the predicted operating window is sufficient to conduct heatups and cooldowns. The 72 EFPY P-T limits were developed with consideration of traditional and extended beltline materials (i.e., locations where RPV neutron fluence > 1.0E+17 n/cm2, E > 1 MeV) using the analytical methods and flaw acceptance criteria of topical report BAW-10046A, Revision 2 (Reference 6-6) and ASME Code,Section XI, Appendix G (2013 Edition, which permits use of KIc).

Geometric discontinuities at the lower shell to lower transition forging and the nozzle belt forging to intermediate/upper shell (i.e., taper transition regions) were not explicitly modeled for the development of the ONS 72 EFPY P-T limits. However, to support 60-year P-T limits for a B&W-designed plant, detailed 2-D ANSYS finite element models were prepared to evaluate the impact of taper transition on P-T limits. The taper transition region P-T limits were compared to the existing uncorrected 60-year P-T limits and the existing limits were found to be more restrictive. This conclusion is not expected to change for operation to 72 EFPY.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 6-4 Limiting materials relative to ART at 54 EFPY (MUR submittal maintained the 54 EFPY P-T limit curves and reduced their EFPY applicability) and 72 EFPY are reported in Table 6-1 for Oconee Unit 1, Table 6-2 for Oconee Unit 2, and Table 6-3 for Oconee Unit 3. The 72 EFPY ARTs are higher than the 54 EFPY ARTs, with the exception of the lower shell to transition forging, at all locations for Oconee Unit 1. The 72 EFPY ARTs are higher than the 54 EFPY ARTs at all locations for Oconee Unit 2 with the exception of the lower shell to transition forging weld and nozzle belt forging AMX-77; the reduction in ART for AMX-77 is attributed to a reduction in the chemistry factor owing to an updated and revised copper content of forging AMX-77. The 72 EFPY ARTs are higher than the 54 EFPY ARTs at all locations for Oconee Unit 3 with the exception of the lower shell to transition forging weld and nozzle belt forging 4680.

Reduction in ART at 72 EFPY for nozzle belt forging 4680 is attributed to testing that was completed on a lower nozzle belt forging nozzle cutout thus significantly reducing the initial RTNDT and margin term relative to what was reported for 54 EFPY. The 54 EFPY ART reported for the lower shell to transition forging weld of <115°F conservatively includes consideration of the lower shell to transition forging weld, transition forging, and lower head.

Two sets of uncorrected heatup and cooldown 72 EFPY P-T limits are developed to assess the 72 EFPY operating window: one set (Figure 6-1 and Figure 6-2) for ONS Unit 1, and a second set (Figure 6-3 and Figure 6-4) for combined ONS Unit 2 and ONS Unit 3, where the limiting adjusted reference temperatures considering ONS Unit 2 and ONS Unit 3 were utilized. Note that typical P-T limit development includes various RCP start and decay heat removal system (DHRS) initiation combinations. The heatup and cooldown curves reported herein are for illustrative purposes only to demonstrate that plant heatups and cooldowns may be conducted at 72 EFPY.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 6-5 6.4 Adjusted Reference Temperatures Limiting ARTs for each RPV region are reported in Table 6-1, Table 6-2, and Table 6-3.

Adjusted reference temperatures for all traditional beltline and extended beltline locations used to develop the 72 EFPY P-T limits are reported in Table 6-4 for ONS Unit 1, Table 6-5 for ONS Unit 2, and Table 6-6 for ONS Unit 3. Material properties of extended beltline locations and changes to material properties tor traditional beltline locations for SLR are reported in Sections 3.4 and 5.4. The limiting ARTs for ONS Unit 2 and ONS Unit 3 were used to develop combined P-T curves for ONS Unit 2 and ONS Unit 3.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 6-6 Table 6-1 Limiting Adjusted Reference Temperature (ART) Values for ONS Unit 1 Reactor Vessel Wall Limiting RTNDT Limiting RTNDT Vessel Item Location (°F) at 54 EFPY (°F) at 72 EFPY 171.0 188.1 Beltline Base 1/4T (SA-1493) (SA-1426/1430)

Metal/Axial Weld 132.9 139.8 3/4T (C2197-2) (C2197-2) 164.2 182.6 1/4T Beltline Circ. (SA-1229) (SA-1229)

Weld 132.1 151.1 3/4T (WF-25) (WF-25) 111.9 122.1 Nozzle Belt (at 1/4T (AHR-54) (AHR-54) 12-inch Wall Thickness) 83.5 104.9 3/4T (AHR-54) (AHR-54)

Outlet Nozzle (Postulated (1) (2) 81.6 Corner Flaw) 72.4 Lower Shell to 1/4T < 115 (3)

(WF-9)

Transition Forging Weld 69.6 3/4T < 115 (3)

(WF-9)

1. Inside wetted surface
2. No shift assumed; initial RTNDT at 60°F per BAW-10046A, Revision 2. Fluence at 54 EFPY <

2 1.0E+17 n/cm (E > 1.0 MeV)

3. ANP-3127P, Q1-- Response to Second Bullet of RAI-1. Includes consideration of lower shell to transition forging weld, transition forging, and lower head.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 6-7 Table 6-2 Limiting Adjusted Reference Temperature (ART) Values for ONS Unit 2 Reactor Vessel Wall Limiting RTNDT Limiting RTNDT Vessel Item Location (°F) at 54 EFPY (°F) at 72 EFPY 161.8 147.1 Beltline Base 1/4T (AMX-77) (AMX-77)

Metal/Axial Weld 135.7 126.6 3/4T (AMX-77) (AMX-77) 193.1 220.1 1/4T Beltline Circ. (WF-25) (WF-25)

Weld 132.5 153.0 3/4T (WF-25) (WF-25) 102.4 112.5 Nozzle Belt (at 1/4T (AMX-77) (AMX-77) 12-inch Wall Thickness) 79.4 97.5 3/4T (AMX-77) (AMX-77)

Outlet Nozzle (Postulated (1) (2) 81.0 Corner Flaw) 77.2 Lower Shell to 1/4T < 115 (3)

(WF-112)

Transition Forging Weld 74.3 3/4T < 115 (3)

(WF-112)

1. Inside wetted surface
2. No shift assumed; initial RTNDT at 60°F per BAW-10046A, Revision 2, Table 3-1. Fluence at 2

54 EFPY < 1.0E+17 n/cm (E > 1.0 MeV)

3. ANP-3127P, Q1-- Response to Second Bullet of RAI-1. Includes consideration of lower shell to transition forging weld, transition forging, and lower head.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 6-8 Table 6-3 Limiting Adjusted Reference Temperature (ART) Values for ONS Unit 3 Reactor Vessel Wall Limiting RTNDT Limiting RTNDT Vessel Item Location (°F) at 54 EFPY (°F) at 72 EFPY 190.8 164.1 Beltline Base 1/4T (4680) (4680)

Metal/Axial Weld 160.0 129.4 3/4T (4680) (4680) 195.6 219.9 1/4T Beltline Circ. (WF-67) (WF-67)

Weld 162.1 183.4 3/4T (WF-70) (WF-70) 106.3 106.0 Nozzle Belt (at 1/4T (4680) (4680) 12-inch Wall Thickness) 88.8 87.0 3/4T (4680) (4680)

Outlet Nozzle (Postulated (1) (2) 118.4 Corner Flaw) 68.1 Lower Shell to 1/4T < 115 (3)

(WF-169-1)

Transition Forging Weld 65.8 3/4T < 115 (3)

(WF-169-1)

1. Inside wetted surface
2. No shift assumed; initial RTNDT at 60°F per BAW-10046A, Revision 2, Table 3-1. Fluence at 2

54 EFPY < 1.0E+17 n/cm (E > 1.0 MeV)

3. ANP-3127P, Q1-- Response to Second Bullet of RAI-1. Includes consideration of lower shell to transition forging weld, transition forging, and lower head

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 6-9 Table 6-4 Adjusted Reference Temperature (ART) Evaluation for the ONS Unit 1 Reactor Vessel Traditional Beltline and Extended Beltline Materials at 72 EFPY

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 6-10 Table 6-5 Adjusted Reference Temperature (ART) Evaluation for the ONS Unit 2 Reactor Vessel Traditional Beltline and Extended Beltline Materials at 72 EFPY

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 6-11 Table 6-6 Adjusted Reference Temperature (ART) Evaluation for the ONS Unit 3 Reactor Vessel Traditional Beltline and Extended Beltline Materials at 72 EFPY

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 6-12 6.5 Assessment of 72 EFPY Operating Window

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 6-13

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For subsequent license renewal, the temperature range where ONS Unit 3 RPV outlet nozzles are predicted to restrict the operation window (i.e., 60 F to 170 F) will have a reduced margin to the NPSH limit. This margin may be regained by removing conservatisms in the 72 EFPY analysis (e.g., revision of heat transfer coefficients to be consistent with RCS flow during 1 and 2 pump operation at low RCS temperatures, reducing the HU/CD rates at low RCS temperatures, and utilizing isothermal conditions to establish LTOP P-T limits) or utilizing provisions permitted by Generic Letter (GL) 8811.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 6-14 Figure 6-1 ONS Unit 1 72 EFPY Uncorrected P-T Limits for Normal Heatup with RCP Start at 170°F

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 6-15 Figure 6-2 ONS Unit 1 72 EFPY Uncorrected P-T Limits for Normal Cooldown with DHRS Initiation at 190°F and RCP Stop at 155°F (CD2)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 6-16 Figure 6-3 ONS Unit 2 and ONS Unit 3 72 EFPY Uncorrected P-T Limits for Normal Heatup with RCP Start at 170°F

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 6-17 Figure 6-4 ONS Unit 2 and ONS Unit 3 72 EFPY Uncorrected P-T Limits for Normal Cooldown with DHRS Initiation at 190°F and RCP Stop at 155°F (CD2)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 6-18 Figure 6-5 ONS Unit 3 Low Range 54 EFPY Cooldown Curves

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 6-19 Figure 6-6 ONS Unit 3 Wide Range 54 EFPY Cooldown Curves

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 6-20 6.6 References for Section 6.0 6-1. Oconee Nuclear Station, Units 1, 2, and 3, Issuance of Amendments Regarding Revised Pressure-Temperature Limits (TAC NOS. MF0763, MF0764, and MF0765), ADAMS Accession Number ML14041A093 6-2. NUREG-2192, Standard Review Plan Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants 6-3. Oconee Nuclear Station Units 1, 2, and 3, Plant Technical Specifications, November 11, 2020 6-4. ONS License Amendment Request for Measurement Uncertainty Recapture Power Uprate, February 19, 2020, ADAMS Accession Number ML20050D379 and NRC SER, ADAMS Accession Number ML20335A001 6-5. NUREG-1723, Safety Evaluation Report Related to the License Renewal of Oconee Nuclear Station, Units 1, 2, and 3 (ADAMS Accession No. ML003695154) 6-6. BAW-10046A, Revision 2, Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR 50, Appendix G, ADAMS Accession Number ML20207G642

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 7-1 7.0 FRACTURE MECHANICS EVALUATION OF UNDERCLAD CRACKS 7.1 Introduction Intergranular separations in the heat-affected zone (HAZ) of RPV low-alloy steel under austenitic stainless steel (SS) cladding (i.e., underclad cracking) is a potential TLAA, as indicated in Table 4.7-1 of the subsequent license renewal standard review plan, NUREG-2192 (Reference 7-1). Duke Energy identified and evaluated this TLAA for 60-years in the ONS License Renewal Application, Section 5.4.2.3, wherein reference is made to BAW-2274, which is contained in BAW-2251A (Reference 7-2) as Appendix C.

NRC approval of the application of BAW-2251A, Appendix C, to the ONS 60-year LRA is reported in NUREG-1723 (Reference 7-3), Section 4.2.4.4. Underclad cracking was recently evaluated for MUR conditions as reported in the ONS License Amendment Request (Reference 7-4)Section IV.1.C.vii. The MUR evaluation concluded that the generic 48 EFPY evaluation reported in BAW-2251A, Appendix C remained bounding for ONS when considering MUR conditions at 54 EFPY.

7.2 Regulatory Guidance for Subsequent License Renewal The regulatory guidance for NRC review of plant-specific TLAA evaluations performed in accordance with 10 CFR 54.21(c)(1)(ii) is reported in NUREG-2192, Section 4.7.3.1.2 (Reference 7-1), and is reported below.

The documented results of the revised analyses are reviewed to verify that their period of evaluation is extended such that they are valid for the subsequent period of extended operation. The applicable analysis technique can be the one that is in effect in the plants CLB at the time that the subsequent license renewal application (SLRA) is filed.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 7-2 The applicant may recalculate the TLAA using an 80-year period to show that the acceptance criteria continue to be satisfied for the subsequent period of extended operation. The applicant also may revise the TLAA by recognizing and reevaluating any overly conservative conditions and assumptions. Examples include relaxing overly conservative assumptions in the original analysis, using new or refined analytical techniques, and performing the analysis using an 80-year period. The applicant should provide a sufficient description of the analysis and document the results of the reanalysis to show that it is satisfactory for the subsequent period of extended operation.

As applicable, the plants code of record is used for the reevaluation, or the applicant may update to a later code edition pursuant to 10 CFR 50.55a. In the latter case, the reviewer verifies that the requirements in 10 CFR 50.55a are met.

In some cases, the applicant may identify activities to be performed to verify the assumption basis for the calculation (e.g., cycle counting). An evaluation of that activity is provided by the applicant. The reviewer assures that the applicants verification activities are sufficient to confirm the validity of the calculation assumptions for the subsequent period of extended operation.

7.3 Description of RV Shell Region with Postulated Underclad Cracks The reactor vessel shell is an approximately 14-foot inner diameter, 37-foot high vertical cylindrical shell with a concave lower head (Reference 7-2), Figure 7-1. The closure heads at Oconee Units 1, 2, and 3 have been replaced and are not susceptible to the postulated underclad cracks owing to compliance with Regulatory Guide 1.43. The reactor vessel shell consists of three sub-assemblies:

1. Upper shell assembly (upper shell flange and upper shell forgings, also called the nozzle belt region)
2. Shell assembly (intermediate and lower shell areas)
3. Lower vessel head assembly

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 7-3 For the evaluation of underclad cracks, there are five locations of interest (Figure 7-2).

  • Nozzle belt
  • Shell taper
  • Shell
  • Transition forging 7.3.1 Geometry There are three locations without discontinuities: bottom sphere dome, shell, and nozzle belt shell. Thicknesses and radii of the RPV shell at these locations are listed in Table 7-1.

Table 7-1 Reactor Vessel Shell Dimensions 7.3.2 Adjusted Reference Temperatures The adjusted reference temperatures for nil-ductility transition, RTNDT, at locations applicable to axially and circumferentially oriented flaws are listed in Section 5.0 (RTPTS Values at 72 EFPY) for the three Oconee units. The values are listed for thickness t = 0.0 inch, that is for cladding to base metal interface, and projected fluence corresponding to 72 EFPY (80 calendar years). The overall summary of RTPTS temperatures according to the unit and applicable flaw orientation is reported in Table 7-2. The table lists only the bounding values for forgings, plates, and welds.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 7-4 Table 7-2 Summary of RTPTS Temperatures for 72 EFPY The highest value for axially oriented flaw occurs at Oconee Unit 1: RTPTS = 207.2°F; the highest value for circumferentially oriented flaw occurs at Oconee Unit 2:

RTPTS = 245.1°F. [

]

7.3.3 Material Properties There are two materials of construction in the reactor vessel shell: (1) low-alloy base metal, and (2) stainless steel cladding. Table 7-3 and Table 7-4 list physical properties for both of these materials. (Note that density , thermal conductivity TC, and specific heat list only one value. The thermal run is made using one constant value for the entire temperature range. Thermal stresses are calculated using temperature-dependent modulus of elasticity and temperature-dependent coefficient of thermal expansion).

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 7-5 Table 7-3 Properties for Base Metal (3/4Ni 1/2Mo Cr-V)

Table 7-4 Properties for Cladding (18Cr-8Ni)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 7-6 7.3.4 Service Loadings 7.3.4.1 Design Basis Transients The inside surface of the reactor vessel is subjected to transient loads in the form of primary coolant cold leg temperatures and pressures as defined in the RPV ASME Code,Section III, Design Specification. The normal, upset, emergency and faulted condition transient events that are considered in the underclad cracking analysis are listed in Table 7-5 along with number of occurrences over complete projected life span of the plant. Note that Transients [ ] are listed for completeness, but do not contribute to crack growth of the postulated underclad cracks since these transients are describing events occurring either in the pressurizer or in the steam generator, and their representation in the reactor vessel downcomer region is as a steady state condition. These transients do not contribute to the growth of postulated underclad cracks in the reactor vessel and are therefore not considered.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 7-7 Table 7-5 List of Applicable Design Transients

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 7-8 7.3.4.2 Piping Loads Stresses at the five locations of interest are included due to deadweight, operating basis earthquake (OBE), thermal expansion at steady state conditions, safe shutdown earthquake (SSE), and loss of coolant accident (LOCA) from attached piping to the reactor vessel, such as inlet and outlet nozzles. These stresses include localized effects from the RPV inlet/outlet nozzles as well.

7.3.4.3 Effects of Discontinuities Discontinuity (interaction) stresses are stresses due to taper-transition regions as they occur in flange top, shell taper, and transition forging location. These stresses were added to the cylindrical shell stresses at locations near structural discontinuities to account for the effects of closure head bolting or impact of taper-transition between two shells of different thicknesses.

7.4 Fracture Mechanics Analysis 7.4.1 Methodology The methodology used to evaluate intergranular separations for the ONS reactor vessels at 80-years is consistent with the methodology reported in the update of BAW10013 included as Appendix C of BAW-2251A (Reference 7-2). The Oconee-specific analysis was performed for 80-years (72 EFPY and MUR conditions) using current fracture toughness information, applied stress intensity factor solutions, fatigue crack growth correlations for SA-508 Class 2 materials, and is evaluated in accordance with the criteria prescribed in the ASME Code,Section XI, 2013 Edition, IWB-3612 (Reference 7-5).

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 7-9 The underclad cracks are conservatively postulated as surface flaws with an initial flaw depth of 0.353 inches (Figure 7-3, maximum known flaw depth plus the thickness of clad) length of the flaw at 2.12 inches, and aspect ratio at 6:1. Furthermore, the flaws are conservatively postulated at the highly stressed regions of the reactor vessel which includes structural discontinuities. The effect of additional stresses and the applied stress intensity factor due to the higher thermal expansion coefficient of the clad is also considered in the analysis. All the significant normal and upset condition transients with the associated number of cycles given in Table 7-5 are considered in the fatigue crack growth analysis. The final flaw size at the end of 80-years is thereby predicted. The stress intensity factor due to this flaw size is then computed for both the Levels A and B (normal/upset) service level conditions as well as the Levels C and D (emergency/faulted) service level conditions. The applied stress intensity factors are then compared against the ASME Code,Section XI, IWB-3612 standards to ensure that the required safety margins for each of the loading conditions are met.

7.4.1.1 Procedure Transient through-wall temperatures are calculated using a constant heat transfer (film) coefficient. The value of a constant heat transfer coefficient used is the maximum value occurring during the applicable thermal transients for any given location.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 7-10

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 7-11 7.4.2 Acceptance Criteria In accordance with the ASME Code,Section XI, IWB-3612 criteria, a flaw is acceptable if the applied stress intensity factor at the final flaw size is less than the available fracture toughness (at the crack tip temperature) with appropriate safety factors, satisfies the following criteria:

1. For Levels A and Level B (Normal and Upset) service conditions:

_f K I (a f ) < K Ic / 10

2. For Levels C and Level D (Emergency and Faulted) service conditions:

_f K I (a f ) < K Ic / 2 Where KI(af) is the maximum applied stress intensity factor at the final flaw size af, and KIc is the fracture toughness (based on fracture initiation) of the material at the corresponding crack tip temperature and irradiation level obtained from Figure A-4200-1 (Reference 7-5).

7.4.3 Fracture Toughness Curve From Article A-4200 (Section XI, Reference 7-5), the lower bound KIc fracture toughness for critical crack initiation is expressed as:

KIc = 33.2 + 20.734 exp [0.02*(T - RTNDT)].

where T is the crack tip temperature, RTNDT is the nil-ductility reference temperature of the material, KIc is in units of ksiin, and T and RTNDT are in units of °F.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 7-12 In this present flaw evaluation, KIc is limited to a maximum value of 200 ksiin (upper-shelf fracture toughness).

7.4.4 Fatigue Crack Growth Model Flaw growth due to cyclic loading is calculated using the fatigue crack growth rate model from Reference 7-5, Appendix A, Section A-4300:

da

= Co ( K I )n dN ,

where KI is the stress intensity factor range in ksiin and da/dN is in inches/cycle. For a surface flaw in a water environment, KI = KImax - KImin Rpwr = KImin / KImax (Rpwr = 0 for KImin 0) 0 Rpwr 0.25: KI < 17.74: n = 5.95 Co = 1.02 x 10-12 x S S = 1.0 KI 17.74: n = 1.95 Co = 1.01 x 10-7 x S S = 1.0 0.25 < Rpwr < 0.65: KI < 17.74 [(3.75Rpwr + 0.06) / (26.9Rpwr - 5.725)]0.25:

n = 5.95 Co = 1.02 x 10-12 x S S = 26.9Rpwr - 5.725

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 7-13 KI 17.74 [(3.75Rpwr + 0.06) / (26.9Rpwr - 5.725)]0.25:

n = 1.95 Co = 1.01 x 10-7 x S S = 3.75Rpwr + 0.06 0.65 Rpwr 1.0: KI < 12.04: n = 5.95 Co = 1.02 x 10-12 x S S = 11.76 KI 12.04: n = 1.95 Co = 1.01 x 10-7 x S S = 2.5 The same section also specifies that the following in-air rates must be used if it is greater than the in-water rates specified above.

Rair = KImin / KImax n = 3.07 Co = 1.99 x 10-10 x S 0 Rair 1: S = 25.72 x (2.88 - Rair)-3.07 KI = KImax - KImin

-2 Rair < 0 and KImax - KImin (0.8)x1.12fa: S=1 KI = KImax

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 7-14 Rair < -2 and KImax - KImin (0.8)x1.12fa: S=1 KI = (1 - Rair) x KImax / 3 Rair < 0 and KImax - KImin > (0.8)x1.12fa: S=1 KI = KImax - KImin Where f is the flow stress defined by f = 1/2(ys + ult), where ys is the yield strength and ult is the ultimate tensile strength calculated at the maximum crack tip temperature between the transient time points where the maximum and minimum stress intensity factor occur. The (0.8) reduction factor is established by NRC in 10 CFR 50.55a, Item (xxviii),Section XI condition: Analysis of Flaws, (Reference 7-7).

The fatigue crack growth calculations contain an explicit check to ensure that the maximum crack growth rate is used on the present flaw evaluations.

7.4.5 Calculation of KI by Polynomial Stress Representation The Polynomial Stress Representation method is used to calculate KI due to pressure, pipe loads, and (shell) interaction stresses.

Reference 7-8, the ASME Code, Appendix A, paragraph A-3211 solution for an internal semi-elliptical axial and circumferential surface flaw in a cylindrical vessel is used as a general form of stress intensity factor equation for calculating crack tip stress intensity factor for arbitrary through-wall stress.

2 3 4

= 0 + 1 + 2 + 3 + 4

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 7-15 Where a is the total crack depth of a surface flaw, x is the distance through the wall, where the origin is at the center of the flaw, A0, A1, A2, A3, and A4 are fitting constants in units of stress derived from fitting the actual stress distribution at the flaw plane.

Coefficients for inside circumferential semi-elliptical flaws at the deepest point (point 1) are found in Reference 7-8, Tables A-3630-1, 3, 5, and 7, for Ri/t = 1, 5, 10, and 20 respectively. The coefficients for inside axial semi-elliptical flaws at the deepest point (point 1) are given in Reference 7-8, Tables A-3650-1, 3, 5, and 7, respectively for Ri/t = 1, 5, 10, and 20.

7.4.6 Calculation of KI by Weight Function A weight function solution is used to calculate KI from transient thermal stresses with cladding effect. This solution allows direct calculation of KI for both the axial and circumferential orientations of the flaw. Weight function method solution is selected over cladding stress extrapolation solutions due to its superior ability to represent non-linear stress profiles.

7.4.7 Irwin Plastic Zone Correction The final crack size endpoint of the fatigue crack growth is adjusted by the plastic zone 2

1 ()

correction calculated as: = + 6

, Reference 7-9, where y is material yield strength. Conservatively use the value of y @ 700°F rounded down (see Table 7-3),

i.e., y = 40 ksi, and the final stress intensity factor is calculated as: ( ) = () .

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 7-16 Figure 7-1 Regions with Clad Forging Material CLOSURE HEAD REMOVED FROM SCOPE Inlet Nozzle (4) Single Wire Clad w/SMAW Clad or ASA Single Arc Outlet Nozzle (2)

Cladw/SMAW or ASA Sing le Arc Single Wire Clad Behind Nozzle Welds Plate Note : All inside surfaces six-wire clad except as noted All material ASTM A508 , Class 2 except plate, as noted.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 7-17 Figure 7-2 Reactor Vessel Shell High Stress Locations CLOSURE HEAD OUT OF SCOPE FLANGE NOZZLE BELT SHELL TAPER SHELL Transition Forging

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 7-18 Figure 7-3 Postulated Surface Flaw Dimensions 6:1 sem i-elliptical surface flaw length Not to scale 0.165" separation reg ion 0.353" flaw depth 2.118" flaw length 0.188" cladding thickness

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 7-19 7.4.8 Evaluation Results Table 7-6 reports the final crack sizes at each of the five locations along with the calculated safety margin that accounts for plastic zone correction. It was demonstrated that at all the investigated locations, postulated flaws in both the axial and circumferential orientations meet the acceptance criteria of Article IWB-3612 (Reference 7-5) for Levels A and Level B (normal/upset), and Levels C and Level D (emergency/faulted) loading conditions after accounting for plastic zone correction.

The lowest fracture toughness margin for Levels A and B (normal/upset) loading conditions is 3.94 and occurs at the flange top axially oriented flaw, which is higher than the minimum required margin of 3.16 (10 = 3.16). The lowest fracture toughness margin for Levels C and D (emergency/faulted) loading condition is 1.62 and occurs at the shell taper due to a circumferentially oriented flaw, which is greater than the minimum required margin of 1.41 (2 = 1.41). Therefore, all the postulated flaws for the underclad cracking in the reactor vessel for all the three Oconee Units remain acceptable at the end of the 80year service evaluation period.

Table 7-6 Results Summary

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 7-20 7.5 References for Section 7.0 7-1. NUREG-2192, Standard Review Plan Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants 7-2. BAW-2251A, Demonstration of the Management of Aging Effects for the Reactor Vessel, ADAMS Accession Numbers ML20212G894 and ML20212G911 7-3. NUREG-1723, Safety Evaluation Report Related to the License Renewal of Oconee Nuclear Station, Units 1, 2, and 3 (ADAMS Accession No. ML003695154) 7-4. ONS License Amendment Request for Measurement Uncertainty Recapture Power Uprate, February 19, 2020, ADAMS Accession Number ML20050D379 and NRC SER, ADAMS Accession Number ML20335A001 7-5. ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 2013 Edition 7-6. [

]

7-7. Code of Federal Regulations, Title 10, Part 50.55a, Domestic Licensing of Production and Utilization Facilities, Codes and Standards, 85 FR 26576, Jun. 3, 2020; 85 FR 34088, Jun. 3, 2020 7-8. ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 2019 Edition 7-9. T.L. Anderson, Fracture Mechanics - Fundamentals and Applications, Taylor and Francis, 3rd Edition

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 8-1 8.0 REACTOR VESSEL ENVIRONMENTALLY-ASSISTED FATIGUE 8.1 Introduction As outlined in Section X.M1 of NUREG-2191 (Reference 8-1) and Section 4.3 of NUREG-2192 (Reference 8-2), the effects of the reactor water environment on fatigue cumulative usage factors (CUFs) must be examined for a set of sample critical components for the plant. This sample set includes the locations identified in NUREG/CR-6260, Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components (Reference 8-3) and additional plant-specific locations in the reactor coolant pressure boundary if they may be more limiting than those considered in NUREG/CR-6260, Table 6-1. Any additional limiting locations are identified through an environmentally-assisted fatigue (EAF) screening evaluation. The EAF screening process evaluates existing CLB fatigue usage values for the ASME Code,Section III components, including the NUREG/CR-6260 locations, to determine the lead indicator (also referred to as sentinel) locations for EAF.

The current ONS analysis of record for reactor vessel EAF NUREG/CR-6260 locations is reported in BAW-2251A (Reference 8-4) and NUREG-1723 (Reference 8-5, Page 4-16). Review of non-NUREG/CR-6260 locations was not required for 60-years by the NRC for the ONS in accordance with NUREG-1723.

The purpose of this section is to summarize the environmentally-assisted fatigue usage factors for the Oconee RPVs for 80 years. The EAF evaluation is based on the Fen methodology provided in NUREG/CR-6909, Revision 1 (Reference 8-6), for the following Reactor Pressure Vessel (RPV) NUREG/CR-6260 locations at Oconee Nuclear Station (ONS) Units 1, 2, and 3:

1. Reactor vessel shell and lower head
2. Reactor vessel inlet and outlet nozzles
3. ICI nozzle to lower head weld
4. Reactor vessel core flood nozzle

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 8-2 8.2 Regulatory Guidance for Subsequent License Renewal The regulatory guidance for NRC review of components evaluated for CUFen TLAA evaluations performed in accordance with 10 CFR 54.21(c)(1)(ii) is reported in NUREG-2192, Section 4.3.3.1.2.2 (Reference 8-2), and is reported below.

The operating cyclic load experience and a list of the assumed transients used in the existing fatigue parameter calculations is reviewed for the current operating term to ensure that the number of assumed occurrences for each transient are projected to the end of the subsequent period of extended operation. The reviewer verifies that a comparison of the operating cyclic load severity to the severity assumed in the existing fatigue parameter calculations for each transient has been made to demonstrate that the cyclic load severities used in the fatigue parameter calculations remain bounding. In addition, the reviewer verifies that a comparison of the water chemistry conditions to those assumed in the Fen calculations has been made to demonstrate that the water chemistry conditions used in the Fen calculations are appropriate. For consistency purposes, the review also includes an assessment of the TLAA information against relevant design basis information and CLB information. The review includes verification that the applicant has updated the CUFen calculations for the applicable NUREG/CR-6260 or more limiting component locations using the methods of analysis in either RG 1.207, Revision 1, NUREG/CR-6909, Revision 0 (with average temperature used consistent with the clarification that was added to NUREG/CR-6909, Revision 1); or other subsequent NRC-endorsed alternatives.

The Code of Record should be used for the reevaluation, or the applicant may update to a later Code edition pursuant to 10 CFR 50.55a using an appropriate Code reconciliation. In the latter case, the reviewer verifies that the requirements in 10 CFR 50.55a are met.

8.3 Methodology The following methodology was used for the determination of the EAF usage factors (CUFen).

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 8-3

1. Retrieve the bounding in-air fatigue usage factor (CUFin-air) for each location of interest from the current license basis (CLB) ASME Code,Section III design reports for Oconee Units 1, 2, and 3.
2. Calculate the maximum environmentally-assisted fatigue correction factor (Fen) for each location of interest using NUREG/CR-6909, Rev. 1 (Reference 8-6).
3. For locations constructed of low alloy steel and featuring relatively low CUFin-air magnitudes, CUFen (= CUFin-air x Fen).
4. For locations featuring high fatigue usage factor due to high level of conservatism, recalculate the in-air fatigue usage factor (CUFin-air) using appropriate stress ranges and number of operating transient cycles using NUREG/CR-6909, Rev. 1, fatigue curves. Locations featuring high fatigue usage factor due to high level of conservatism are: inlet nozzle, inlet nozzle safe end (inlet nozzles are buttered with carbon steel at their terminal ends), outlet nozzle, outlet nozzle safe end (outlet nozzles are buttered with carbon steel at their terminal ends), ICI J-weld, core flood nozzle, core flood nozzle safe end, and venturi.
5. Calculate EAF usage factor based on recalculated CUFin-air per step 4, CUFen

(= CUFin-air x Fen).

8.4 Assumptions

1. In NUREG/CR-6909 Rev. 1, Reference 8-6, when calculating T* in Appendix A, the upper bound limit of equation for T is 325°C (617°F). For the purposes of calculating a Fen, this temperature is considered bounding for locations exposed to LWR coolant.
2. Since the strain rate is unknown for stress intensity ranges, the slowest strain rate used in Reference 8-6 is considered for conservatism.
3. Sulfur content of 0.015 wt.% is assumed. This value is taken from Reference 8-6 as bounding value.
4. Low alloy steel material with cladding is conservatively assumed to be exposed to the LWR coolant environment.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 8-4

5. The in-air cumulative fatigue usage factors extracted from the ASME Code,Section III design report summary documents and used in the calculation may be of locations not exposed to the RCS fluid as detailed location of fatigue usages reported are not readily available. However, since it is the maximum usage factor for that component, for conservatism, it is used herein.
6. The number of NSSS design transients defined in the RPV ASME Code,Section III, Design Specifications, which are applicable to 60-years, are not revised for 80-years.

8.5 Summary of Results Table 8-1 includes a summary of the results of the environmentally-assisted fatigue usage factors for the RPV NUREG/CR-6260 items. The final CUFen is based on the original design number of operating transient cycles. Note that the operating transient cycles are not adjusted or prorated for the 80 years plant operation from the current license basis.

The maximum CUFen is 10.01 for the core flood venturi; however, the core flood venturi is not a pressure retaining item and was not evaluated in NUREG/CR-6260. The CUFen values for the remaining locations are less than 1.0 and therefore acceptable.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 8-5 Table 8-1 CUFen Results Summary Component Material Bounding Fen CUFin-air CUFen Lower Head SA533-Gr2, Code Case 1339 [ ] [ ] 0.756 Nozzle Belt A-508-64 Cl.2 [ ] [ ] 0.107 Shell A-508-64 Cl.2 [ ] [ ] 0.624 Inlet Nozzle A-508-64 Cl.2 [ ] [ ] 0.832 Inlet Nozzle Safe End SA-106 Gr.C [ ] [ ] 0.832 Outlet Nozzle A-508-64 Cl.2 [ ] [ ] 0.832 Outlet Nozzle Safe End SA-106 Gr.C [ ] [ ] 0.832 ICI-J-Weld SB-166 [ ] [ ] 0.744 Core Flood Nozzle ID A-508-64 Cl.2 [ ] [ ] 0.882 Core Flood Safe End1 A-336-F8m [ ] [ ] 0.525 Core Flood Venturi SA-376 TP304 [ ] [ ] 10.010 Notes:

1. NUREG/CR-6260, Page 5-41, reported that the CUF for the safe end was essentially zero and was not evaluated in NUREG/CR-6260.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 8-6 8.6 References for Section 8.0 8-1. NUREG-2191, Volume 2, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report 8-2. NUREG-2192, Standard Review Plan Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants 8-3. NUREG/CR-6260, Application of NUREG/CR-5999 Interim Fatigue Curve to Selected Nuclear Power Plant Components, March 1995 8-4. BAW-2251A, Demonstration of the Management of Aging Effects for the Reactor Vessel, ADAMS Accession Numbers ML20212G894 and ML20212G911 8-5. NUREG-1723, Safety Evaluation Report Related to the License Renewal of Oconee Nuclear Station, Units 1, 2, and 3, ADAMS Accession Number ML003695154 8-6. NUREG/CR-6909, Effect of LWR Water Environments on the Fatigue Life of Reactor Materials, Rev. 1, May 2018 Final Report

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-1 9.0 IRRADIATION EMBRITTLEMENT OF ONS RPV SUPPORTS FOR SLR 9.1 Introduction The purpose of this section is to evaluate the ONS RPV supports relative to the draft NRC interim staff further evaluation draft guidance reported in Reference 9-1 with respect to determination of the applicability of reduction of fracture toughness for subsequent license renewal. The evaluation reported herein is consistent with the guidance reported in NUREG-1509 (Reference 9-2).

9.2 Regulatory Guidance for Subsequent License Renewal In accordance with NUREG-2192, Standard Review Plan for Subsequent License Renewal (Reference 9-3), aging management review recommendations for reactor vessel (RV) supports are addressed in Section 3.5.2.2.2.6, and Table 3.5-1, Item 097.

Following review of the first subsequent license renewal applications for Turkey Point 3 and 4, Surry 1 and 2, and Peach Bottom, the NRC determined that additional guidance was needed to clarify aging management review expectations by the NRC for Reactor Pressure Vessel (RPV) supports for subsequent license renewal. As such, the NRC has issued draft interim staff guidance (Reference 9-1) to revise the SLR Standard Review Plan NUREG-2192 to add a new Section 3.5.3.2.2.7 that includes the following review guidance.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-2 Further evaluation is recommended of a plant-specific program (or plant-specific enhancements to selected GALL-SLR AMPs) to manage reduction of fracture toughness due to irradiation embrittlement from accumulated neutron fluence and gamma dose, which could occur in BWR and PWR steel structural support components (including associated weldments and bolted connections) located in the vicinity of the Reactor Vessel (RV) during the subsequent period of extended operation. These components include the RV steel supports, neutron shield tank, steel structural support components of reactor shield wall and sacrificial shield wall, or other steel structural support components located in the vicinity of the RV. Further, loss of function due to radiation exposure (neutron and/or gamma) for non-steel, non-concrete materials (e.g., Lubrite or other lubricant/coating in support sliding feet) that are used in these structural support components should also be evaluated and dispositioned, with supporting technical information, on a plant-specific basis for the subsequent period of extended operation. If a plant-specific program or program enhancements are determined to be necessary, the reviewer confirms that the acceptance criteria for AMP program elements described in BTP RLSB-1 {Appendix A.1 of NUREG-2192 (SRP-SLR)}.

Otherwise, the reviewer confirms the adequacy of the justification provided for the aging effects not requiring management such that intended function(s) are maintained consistent with the CLB for the subsequent period of extended operation.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-3 The applicant may address irradiation aging effects by analysis, testing, aging management inspections (e.g., one-time, periodic), or a combination of these methods; the reviewer confirms the evaluation is supported by an adequate technical basis that also accounts for uncertainties and limitations of available technical data and knowledge. NUREG-1509, Radiation Effects on Reactor Pressure Vessel Supports, (Ref. 211) provides general guidance and an acceptable approach (with exception of structural consequence analysis approach) to evaluate loss of fracture toughness via a screening evaluation, a fracture mechanics analysis or a transition temperature analysis (relative to the lowest operating temperature). A plant-specific AMP or plant-specific enhancements to selected GALL-SLR AMPs may not be necessary for the steel structural support components if (i) the screening criteria or criteria for reevaluation in Chapter 4, RPV Support Reevaluation Criteria, of NUREG-1509 (with exception of the structural consequence analysis approach in Section 4.5) are satisfied on a plant-specific basis using the technical evaluation procedures described therein; (note that adequacy of technical justification provided for other applicant-proposed demonstrated technical evaluation approach used will be reviewed on a case-by-case basis) and (ii) there is no plant-specific operating experience of irradiation embrittlement observed to date. The methodology in Nonmandatory Appendix A of the ASME Code,Section XI may be used for analytical evaluation of postulated flaws in a fracture mechanics analysis. For related non-steel, non-concrete structural support components, the applicant should provide supporting technical information and data to justify its determination regarding the need for a plant-specific program to manage irradiation aging effects; the reviewer evaluates this on a case-by-case basis.

1 Reference 21 is per NUREG-2192

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-4 It should be noted that the conclusions in NUREG-1509 were based on analyses and limited to neutron fluences relevant to 40 years of operation. Since the neutron fluences for the subsequent period of extended operation to 80 years often exceed the neutron fluence values used in the analyses of NUREG-1509, the analytical methodologies (except the structural consequence analysis approach in Section 4.5) in NUREG-1509 remain applicable but the conclusions are no longer supported.

The reviewer confirms that the applicants technical evaluation estimated the neutron and gamma fluence incident on the RV steel supports and other applicable steel structural support components for the subsequent period of extended operation and evaluated the susceptibility of components to loss of fracture toughness. The evaluation should demonstrate that the RV steel supports and other steel structural support components noted above will remain capable of performing their intended function consistent with the current licensing basis through the subsequent period of extended operation. The reviewer confirms that the evaluation included sufficient description that demonstrate that appropriate methodologies and conservative assumptions were implemented to estimate the levels of neutron fluence and gamma dose for the subsequent period of extended operation. The damage parameter displacements per atom (dpa) should include neutrons of all energies (high and low) rather than only those with E > 1 MeV (i.e., embrittlement predictions should include damage from the entire neutron energy spectrum based on E > 0.1 MeV). Alloying elements, such as copper, can increase the rate of radiation embrittlement. Therefore, the evaluation should take into consideration the material properties and chemical composition of the steel (e.g., initial nil ductility temperature (NDT), type of steel, copper content, weld material) and the lowest service (operating) temperature to which the components are exposed.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-5 A structural integrity evaluation of the RV steel supports and other steel structural support components should include all design basis loading combinations referenced in the current licensing basis (CLB). It is essential that the evaluation account for: (1) plant specific operating experience of RV steel supports and other structural components degradation due irradiation embrittlement and other susceptible aging effects; (2) the current as-found physical condition of the supports and structural components; and (3) account for the effects of observed signs of degradation (including but not limited to corrosion, cracks, or permanent deformation) and potential future degradations reasonably projected to the end of the subsequent period of extended operation (i.e., 80 years). This should typically be based on a detailed physical examination (to the extent possible) of the RV supports that is documented in an inspection report that serves as the basis for the evaluation and decisions regarding further actions. Where a detailed physical examination is not feasible, adequately justified assumptions of potential degradations through the end of the subsequent period of extended operation (i.e., 80 years) should be made, or appropriate monitoring actions proposed to manage the susceptible aging effects consistent with the assumptions made in the evaluation.

9.3 Description of RPV Support and Evaluation Methodology The RPV support assembly and embedment detail are illustrated in Figure 9-1 and Figure 9-2, and consist of a support skirt, a support flange, anchor bolts and associated washers and hex nuts, sole plate, vertical bearing plate and associated nelson studs, grout, and reinforced concrete pedestal that contains the embedded anchor bolts. For the discussion that follows the RPV support assembly is defined as the RPV support skirt and the RPV support flange, which were attached to the reactor vessel by Babcock

& Wilcox during fabrication of the reactor vessel. The RPV embedment includes the anchor bolts and associated washers and hex nuts, sole plate, vertical bearing plate and associated nelson studs, grout, and reinforced concrete pedestal that contain the embedded anchor bolts, and these items were supplied by the Architect Engineer.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-6 The support skirt is fabricated from two semi-circular carbon steel (SA-516 Grade 70) rolled plates welded together longitudinally to form a 2-inch thick, 60-inch high, 175.5-inch ID cylinder. The top of the support skirt cylinder is welded to the bottom of the reactor vessel transition forging. Twelve 9.25-inch diameter holes and twelve 2-inch diameter holes are included in the support skirt for ventilation of the reactor vessel cavity, emergency flooding of the cavity, and steam venting in the event of flooding.

The support skirt cylinder is welded at the bottom to the support flange, which is constructed from four 90-degree carbon steel (SA-515 Grade 70) segments welded together to form a 3.5-inch high cylinder (ID of 161.5 inches and OD of 193.5 inches).

The carbon steel support flange contains forty-eight 3-inch diameter holes, equally spaced at 7 degrees 30 minutes, on both the inside (on a diameter of 167.5 inches) and outside (on a diameter of 187.5 inches) of the support flange to accept 96 high strength alloy steel (A490) anchor bolts. Each anchor bolt is 2-inches in diameter and 6 ft 10-inches in length with the top of each anchor bolt threaded and extending approximately 7-inches above the sole plate. Each anchor bolt is secured to the embedment by one heavy hex nut, one jamb nut, one standard size hardened washer, and a 2 1/8-inch ID by 4 1/8-inch OD by 1-inch thick plate washer. In addition, the support flange contains forty-eight 1.5 inch diameter holes (on a diameter of 187.5 inches) for shear pins (A490). The shear pin holes are equally spaced and located azimuthally midway between adjacent anchor bolts; forty-two 7-inch long shear pins are inserted into the 1.5-inch diameter holes in the support flange and extend into the underlying sole plate (since the sole plate is formed by six sub-sections, only 42 shear pins are installed).

The anchor bolts are specified to be prestressed (approximately 80 ksi each) such that lift off of the support flange will not occur during a postulated operating basis earthquake.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-7 Evaluation Methodology The evaluation process for the RPV support structural steel is documented in NUREG-1509 (Reference 9-2), and is illustrated through flow charts (i.e., Figures 4-2 and 4-4 of NUREG-1509), reproduced as Figure 9-3 and Figure 9-4 below. The initial step for the evaluation involves assessment of the existing condition of the RPV support at the time of re-evaluation, comparison with the initial construction condition, and the degree of degradation predicted by the end of plant life. In addition, a review of the original design and safety margin is performed. This review includes the original design methodology, load combinations for which the supports were designed, allowable stresses and their margins with respect to the actual stresses in the members, and codes governing the original design. To confirm that there is adequate RPV support fracture resistance, the assessment is based on a transition temperature analysis (Figure 9-4) wherein it is sufficient to demonstrate that there is an adequate margin between the minimum operating temperature (i.e., lowest service temperature) and the nil-ductility temperature (NDT) at 72 EFPY.

Based on the configuration of the RPV support assembly and embedment detail described above, the following items directly support the RPV support intended function, i.e., to provide structural support for the reactor vessel, and are evaluated for susceptibility to reduction of fracture toughness by irradiation embrittlement.

  • RPV support skirt (SA-516 Grade 70)
  • Anchor bolts (A490)
  • Anchor bolt jamb nuts, hex nuts, and washers (assumed to be equivalent to A490)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-8 The embedment vertical bearing plate and associated nelson studs do not support the RPV support assembly intended function and are not evaluated for susceptibility to reduction of fracture toughness by irradiation embrittlement herein. NUREG-2192, Section 3.5.2.2.2.6, Reduction of Strength and Mechanical Properties of Concrete Due to Irradiation, includes the following guidance relative to grout and concrete. Data related to the effects and significance of neutron and gamma radiation on concrete mechanical and physical properties is limited, especially for conditions (dose, temperature, etc.) representative of light-water reactor (LWR) plants. However, based on literature review of existing research, radiation fluence limits of 1x1019 neutrons/cm2 neutron radiation (fluence cutoff energy E > 0.1 MeV) and 1x108 Gy (1x1010 rad) gamma dose are considered conservative radiation exposure levels beyond which concrete material properties may begin to degrade markedly. 72 EFPY fluence at the ONS RPV support embedment is estimated at 1.63E+18 n/cm2 (E > 0.1 MeV) and gamma dose at 1.75E+09 rad. As such, the embedment concrete and grout are not susceptible to irradiation embrittlement.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-9 Figure 9-1 Reactor Pressure Vessel Support Assembly I ra si:io

/ .eac: or V ss I Lov,i r 11 a ppor- Fla e

/

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-10 Figure 9-2 Reactor Pressure Vessel Support Embedment Detail Reactor Vessel Support Skirt Anchor Bolt Vertical Bearing Plate Nelson Studs Support Flange 0

. ** 0 0 * \? .o Sole Plate ~ -o Grout

=:*:= .

0 0 0 0 0

\? 0 0 0

\? 0 Oo 0 0 Oo 0 0 0

\?

Pedestal Concrete

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-11 Figure 9-3 NUREG-1509 Figure 4-2, Preliminary Evaluation Reexamine Reexamine Reexamine Original Original Original Identify Assess Document Unusual Methodology Criteria Allowables Residual Fabrication Features and Stresses Procedure i i i I

Stresses I

Evaluate Existing

~

OK -

Physical Conditions ReviewRPV Assess i "

)\

Support Design Magnitude and Type of Stresses Damaged or Degraded

-- Repair or Replace

- i (3)

Was Fracture YES- Comply with YES~ RPV Supports Prevention - Criteria? Satisfactory Considered ?

NOi NO Evaluate Supports -

~

FIRST OPTION Fracture Mechanics _ EITHER Approach, Figure 4 -3 -

Note: The numbers next to the blocks above SECOND OPTION refer to the corresponding paragraphs of "Explanatory Notes" at the end of Transition OR this section.

Temperature Approach, Figure 4-4

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-12 Figure 9-4 NUREG-1509 Figure 4-4, Transition Temperature Approach R.eviewRPV !fuifial TT Dete:rmhle EOL Support Da~ Anlhlbfe Filu.eTu:e I dpa, Desijn YESI Revil!W' Bs * 'blish OJ!era1iing Material.

Condltlmis Condi&

S'e urdi ll>ata. stablish Dase or do DOLTT Til'Ti!st, Dete:rmm:e ATI Detennlne - Fro- - -- {5)

,I

~ - - - . . . i i' HOLTT I F' ,eJ-1 I-_ _- _ _- _ -

(1) (2) i,y,:ihmte: . I YES 1---- = - --~ ~-----.....,JJ!ol 'ffSO!i..+ Marglnli------ - -~ I RPV Su polilS S lisfiickr-.ey s.,LST

~:ilffl C)f Rt.Cu:rt>ls.h Note: 1n w to dte lllodts abQve

refer lo the ,co - p;tragn1p _
  • o l "Ellpanal:Ol'J' the end of diis~ .

Fi;,mrc 4-'1 'fransiti.un 'f/cmpcn.turc Approach 7.7 NURBG-1500

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-13 9.4 Irradiation Embrittlement Further Evaluation 9.4.1 Assessment of Current Condition Assessment of the existing condition of the ONS Units 1, 2, and 3 RPV supports includes a mandatory, visual physical condition inspection of the vital parts of the supports. Rust, cracks, or permanent deformation of any part of the RPV supports should be noted as evidence that some distress has been sustained.

Visual inspections of the ONS RPV supports were performed in 2012 (ONS Unit 1),

2013 (ONS Unit 2), and 2014 (ONS Unit 3). For each ONS unit, a VT-3 visual examination was performed on accessible surfaces of the RPV support with a calculated coverage of approximately 66.5% of the support surface areas within the examination boundary. The VT-3 visual examinations were performed using personnel, equipment, and procedures qualified in accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, 1998 Edition with the 2000 Addenda, and there were no unacceptable conditions or indications detected during these examinations. The ONS Unit 1 examination record indicates that the inspection was rejected; this was due to limited coverage that is discussed below.

These examinations were considered limited because the ASME Code,Section XI inspection required a VT-3 visual examination of 100% of the RV support. Based on this limited inspection, Duke Energy submitted Relief Request 15-ON-004 to the NRC requesting relief from the ASME Code,Section XI examination requirements for the RPV supports at ONS. After review of Relief Request 15-ON-004, the NRC staff, through a Safety Evaluation, granted Duke Energy relief for the limited examination of the ONS RPV supports, and further concluded that the examinations performed to the extent practical provide reasonable assurance of structural integrity of the RPV support.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-14 9.4.2 Design Stress Summary The design stresses for the ONS RPV supports (ONS Unit 1, ONS Unit 2, and ONS Unit

3) were developed in accordance with the ASME Code, 1965 Edition with Addenda through Summer 1967. Stress intensities were calculated for the following load cases and locations.

Normal/Upset Condition Loads

1. Primary Stresses - Dead Weight (DW) + Operational Basis Earthquake (OBE) +

Pressure:

  • Support Skirt: Maximum Calculated Stress Intensity = [

] Location - towards bottom of skirt at 9-1/4 hole section

  • Support Flange Bending Stress = [ ]
  • Support Flange Shear Stress = [ ]
  • Shear Pins Stress = [ ] (includes thermal)
2. Primary + Secondary Stresses - DW + OBE + Pressure + Thermal transients:
  • Support Skirt: Stress Intensity Range = [ ]
  • Location - bottom of skirt just above flange Faulted Condition Loads:
1. Primary Stresses - DW + SSE + LOCA + Pressure:
  • Support Skirt: Maximum Calculated Stress Intensity = [

]

  • Location - towards bottom of skirt at 9.25 hole section
  • Bolts: Maximum Calculated Shear Stress [ ]

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-15

  • Shear Pins: Maximum Calculated Shear Stress [

]

Notes:

(1) The faulted stress analyses looked at the bolts as the weak link such that if the bolts were acceptable, the flange was acceptable.

(2) The faulted stresses above consider Large Break LOCAs. Considering Leak Before Break (LBB) along with OTSG replacement, the loads (thus stresses) are less but the component stresses with regards to ranking are expected to be the same.

(3) Definitions:

- DW: Deadweight

- OBE: Operating Basis Earthquake

- SSE Safe Shutdown Earthquake

- LOCA: Loss of Coolant Accident

[

]

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-16 9.4.2.1 Load Reduction Considering LBB When the steam generators were replaced at the Oconee Units, the RCS loop model was reanalyzed considering LBB of the primary piping (hot and cold leg piping). With regard to the RPV support, load comparisons were performed. The load ratios for the faulted condition loading of DW + SSE + LOCA are reported in Table 9-1.

Table 9-1 RPV Support Load Comparison: LBB Loads / Original Design Loads

[

]

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-17 9.4.3 Lowest Service Temperature The lowest service temperature (LST), as defined in the NUREG-1509, Page 40, is estimated for the ONS Unit 1, ONS Unit 2, and ONS Unit 3 RPV support assembly and embedment. The lowest service temperature (LST) is defined as the minimum temperature of the most vulnerable part of the fracture-critical member when design-basis accident loads occur. RPV support temperatures can be established either from measurements or theoretical calculations.

For the ONS, design basis accidents (DBAs) are addressed in Chapter 15 of the ONS updated final safety analysis report (UFSAR). Based on a review of the ONS Chapter 15 DBAs, lowest service temperature may be defined by calculating the temperature distribution in the RPV support skirt at 100% steady state conditions. The B&W-designed reactors operate at a constant average temperature between approximately 15% power and 100% power; thus 100% power represents a condition with the lowest cold leg and RV downcomer temperatures prior to the postulated DBA.

A 2-D model of the lower RPV, RPV skirt and support flange, and anchoring and supporting structure was prepared. This model is based on an existing 3-D solids model of the Oconee RPV and internals. The resulting 2-D model is supplied to the ANSYS Mechanical application. A solution mesh is generated, thermal boundary conditions (loads) are applied, and a steady-state thermal solution is calculated. Two different temperature contours of the ONS Unit 1, ONS Unit 2, and ONS Unit 3 RPV support skirt assembly were developed using models based on two different locations (i.e., 2-D cross sections); the anchor bolt location, Figure 9-5, and the shear pin location, Figure 9-6. The RPV support assembly temperatures, which are estimated using the anchor bolt location model, are used for conservatism. Based on the anchor bolt location model temperature contours, the minimum LST for the RPV support assembly and embedment is 139.05°F (59.5°C) at the location of the anchor bolts, and is conservatively applied for all the RPV support assembly materials in this assessment.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-18 Figure 9-5 RPV Skirt Temperatures from Anchor Bolts Location Model

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-19 Figure 9-6 RPV Skirt Temperatures from Shear Pin Location Model

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-20 9.4.4 Establish EOL NDT Temperature 9.4.4.1 DPA of RPV Support Skirt, Flange, Anchor Bolts, and Shear Pins The projected 72 EFPY dpa of the ONS Unit 1, ONS Unit 2, and ONS Unit 3 RPV support skirt weld and embedment anchor bolts is conservatively estimated at 5.53E-04 dpa. This projected 72 EFPY dpa is obtained by calculating dpa at the bottom of a RG 1.190 compliant RPV DORT model (BAW-2241P-A, Revision 2, Reference 9-5), which is approximately 17.49 inches above the transition forging to RPV skirt weld, and then by generating RPV skirt weld flux and dpa rate profiles on the outer RPV surface in the air cavity region and extrapolating dpa from the bottom of the DORT model to the RPV skirt to transition forging weld. The dpa of 5.53E-04 is conservatively assumed to be applicable to all the RPV support assembly and embedment materials (i.e., skirt, flange, anchor bolts, nuts, washers, and shear pins).

Neutron fluence and gamma dose at 80-years (72 EFPY) are calculated using source terms that bound all Oconee units to provide bounding estimates for RPV support and the biological shield wall. [

]

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-21 9.4.4.2 RPV Support Assembly Materials The ONS Unit 1, ONS Unit 2, and ONS Unit 3 RPV support skirt base metal materials (SA-516 Grade 70) are classified as carbon-manganese steels and the RPV support flange materials (SA-515 Grade 70) are classified as plain carbon steels in accordance with NUREG-1509, Table 4-2, Classification of Wrought Grades into Groups. The anchor bolts were furnished in the field as ASTM A490 steel; the shear pins are also manufactured per ASTM A490. Materials of construction for the jamb nut and hex nut are assumed to be the same as the anchor bolts with respect to material properties.

The ONS Unit 1, ONS Unit 2, and ONS Unit 3 RPV support assembly welds are produced using either the manual metal arc process or semi-automatic gas shielded metal arc process. Table 9-2 lists the material descriptions for the RPV skirt/flange base metals and Table 9-3 lists the RPV skirt/flange associated welds for ONS Unit 1, ONS Unit 2, and ONS Unit 3.

Table 9-2 ONS RPV Support Skirt Assembly Base Metal Materials

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-22 Table 9-3 ONS RPV Support Skirt Assembly Weld Materials 9.4.4.3 Radiation Induced NDT Shift In accordance with NUREG-1509, Section 4.3.4.2, when using the transition temperature approach to evaluate the RPV support integrity, the NDT temperature at end-of-life should include the irradiation-induced shift (NDTT) and account for uncertainties related to the NDT determination. Therefore, the NDT temperature at end-of-life (ART) is expressed by the following equation:

= + +

To determine the dpa at which the RPV support skirt assembly materials will equal the estimated LST, the above equation can be rearranged as follows:

= ( + )

where ART = LST (see Section 9.4.3 above).

The NDTT values for the ONS Unit 1, ONS Unit 2, and ONS Unit 3 RPV support assembly and embedment materials are reported in Table 9-4 and Table 9-5.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-23 Based on the calculated NDTT for the RPV support assembly materials, a dpa is estimated from Figure 3-1, upper bound curve, of NUREG-1509 for each ONS Unit 1, ONS Unit 2, and ONS Unit 3 RPV support assembly material. These estimated dpa values along with the projected 72 EFPY dpa are illustrated in Table 9-6 and Table 9-7 for comparison. Based on this comparison, the projected 72 EFPY dpa for the ONS Unit 1, ONS Unit 2, and ONS Unit 3 RPV support flange and the ONS Unit 1 and ONS Unit 2 RPV support flange welds are greater than the estimated dpa for the minimum LST of the RPV support skirt assembly, indicating the expected ART values for these materials will be greater than the LST at 72 EFPY of operation. These items require further evaluation to demonstrate that the intended function(s) of the RPV support assembly will be maintained consistent with the CLB during the Subsequent Period of Extended Operation (SPEO) when considering irradiation embrittlement.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-24 Table 9-4 Calculated NDTT Values for the ONS RPV Support Skirt Assembly Base Metal Materials Table 9-5 Calculated NDTT Values for the ONS RPV Support Skirt Assembly Weld Materials

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-25 Table 9-6 Comparison of the NDTT dpa at ONS RPV Support Skirt Assembly Base Metal LST to Projected 72 EFPY dpa

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-26 Table 9-7 Comparison of the NDTT dpa at ONS RPV Support Skirt Assembly Weld Metal LST to Projected 72 EFPY dpa

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-27 9.4.5 Comparison of EOL NDT to LST for Critical RPV Support Locations In accordance with Table 9-6 and Table 9-7, the ONS Unit 1, ONS Unit 2, and ONS Unit 3 RPV support flange and the ONS Unit 1 and ONS Unit 2 RPV support flange welds connecting the 90 degree segments together to form a circular support plate are potentially susceptible to reduction of fracture toughness by irradiation embrittlement at 72 EFPY of operation. All remaining RPV support assembly and embedment items that support the RPV support intended function are not susceptible to irradiation embrittlement.

The design stresses for the ONS RPV support flange and its associated welds indicate the normal/upset loads show low bending and shear stresses in the support flange compared to their respective allowables, and that the RPV support skirt is more highly stressed. The faulted stresses show that the RPV support skirt is the highly stressed item with little margin. The RPV support flange was not explicitly addressed for faulted loads as the bolts were determined to be the weak link and the bolts were shown acceptable with margin. The faulted loads considered were large break primary piping breaks. With consideration of LBB of the primary piping, the loads for the faulted condition have been reduced considerably in the RPV support skirt, flange, and bolts.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-28 The support flange is bolted to the concrete with 48 bolts spaced equally around the outside of the flange, and 48 bolts spaced equally around the inside part of the flange;

[

]

Based on consideration of the potential for irradiation embrittlement of the RPV support assembly and embedment items, only the ONS Unit 1, ONS Unit 2, and ONS Unit 3 RPV support flange and the ONS Unit 1 and ONS Unit 2 RPV support flange welds are potentially susceptible to irradiation embrittlement. [

]

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-29 As discussed in Section 9.4.2.1, load reduction considering LBB is approximately one-quarter of the original design loads with the exception of Fy. Consideration of LBB would significantly reduce the faulted stress intensities in the RPV support skirt and support flange and would likely significantly increase the margin to the allowable stress intensities.

Since the RPV support skirt is the critical location of the RPV support assembly and is not susceptible to irradiation embrittlement based on the NDT evaluation reported above, the RPV support intended function will be maintained consistent with the CLB during the SPEO when considering damage due to irradiation.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 9-30 9.5 References for Section 9.0 9-1. RV Supports ISG SLR Document Changes: Add FE Section 3.5.2.2.2.7 and AMR Items for Irradiation Embrittlement of Reactor Vessel (RV) Steel Supports and Other Steel Structural Support Components near RV, ADAMS Accession Number ML20049H359 9-2. U.S. Nuclear Regulatory Commission, NUREG-1509, Radiation Effects on Reactor Pressure Vessel Supports, May 1996 9-3. NUREG-2192, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants 9-4. Framatome Inc. Topical Report BAW-1621, Effects of Asymmetric LOCA Loads, Phase II Analysis, ADAMS Accession Number ML19320B058 9-5. Framatome Inc. Topical Report BAW-2241P-A, Revision 2, Fluence and Uncertainty Methodologies, ADAMS Accession No. ML073310655 (Proprietary),

ML073310660 (Non-Proprietary)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 10-1 10.0 REACTOR COOLANT PUMPS THERMAL EMBRITTLEMENT (CASS) 10.1 Introduction The Oconee Nuclear Station (ONS) reactor coolant pumps (RCPs) were fabricated by Sulzer Bingham, 28 by 28 by 41 Type RQV (ONS Units 2 and 3) and Westinghouse, Type 93A (ONS Unit 1). Pressure retaining RCP items that are subject to aging management review, fabricated from cast austenitic stainless steel (CASS), and potentially subject to reduction of fracture of fracture toughness by thermal embrittlement include the following: Unit 1 - main flange and casing, Units 2 and 3 -

cover/stuffing box and casing (Figure 10-1 and Figure 10-2). RCP pressure boundary items that are fabricated from cast austenitic steel may be susceptible to thermal embrittlement if the ferrite content of the castings exceeds the screening criteria established by the NRC in NUREG-2191,Section XI.M12, Evaluation and Technical Basis (Reference 10-1) as supplemented by the final interim staff guidance for mechanical, SLR-ISG-2021-02-MECHANICAL (Reference 10-2).

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 10-2 Figure 10-1 Sulzer Bingham Reactor Coolant Pump

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 10-3 Figure 10-2 Westinghouse Type 93 A Reactor Coolant Pump g

~ Q-G' ~

"' ...J

1 Casing, CF8

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 10-4 10.2 Regulatory Guidance for SLR Regulatory guidance for NRC review of thermal embrittlement of reactor coolant system components fabricated from cast austenitic stainless steel is contained in NUREG-2191, Volume 2 (Reference 10-1), aging management program XI.M12, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS) as supplemented by the final interim staff guidance for mechanical, SLR-ISG-2021-02-MECHANICAL (Reference 10-2). The XI.M12 program description is as follows.

The reactor coolant system components are inspected in accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI. This inspection is augmented to detect the effects of loss of fracture toughness due to thermal aging embrittlement of cast austenitic stainless steel (CASS) piping components except for valve bodies. This aging management program (AMP) includes determination of the potential significance of thermal aging embrittlement of CASS components based on casting method, molybdenum content, and percent ferrite.

For components for which thermal aging embrittlement is potentially significant as defined below, aging management is accomplished through either (a) qualified visual inspections, such as enhanced visual examination (EVT-1); (b) a qualified ultrasonic testing (UT) methodology; or (c) a component-specific flaw tolerance evaluation in accordance with the ASME Code,Section XI. Additional inspection or evaluations to demonstrate that the material has adequate fracture toughness are not required for components for which thermal aging embrittlement is not significant. The scope of the program includes ASME Code Class 1 piping components constructed from CASS with service conditions above 250 °C (482 °F).

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 10-5 For pump casings, as an alternative to the screening and other actions described above, no further actions are needed if applicants demonstrate that the original flaw tolerance evaluation performed as part of Code Case N-481 implementation remains bounding and applicable for the SLR period or the evaluation is revised to be applicable for 80 years. For valve bodies, based on the results of the assessment documented in the letter dated May 19, 2000, from Christopher Grimes, U.S. Nuclear Regulatory Commission (NRC), to Douglas Walters, Nuclear Energy Institute (May 19, 2000 NRC letter), screening for significance of thermal aging embrittlement is not required. The existing ASME Code,Section XI inspection requirements are adequate for valve bodies.

RCP items that are fabricated from cast austenitic steel may be susceptible to thermal embrittlement if the ferrite content of the castings exceeds the screening criteria established by the NRC in NUREG-2191,Section XI.M12, Evaluation and Technical Basis, as repeated below.

Based on the criteria set forth in the May 19, 2000, NRC letter, the potential significance of thermal aging embrittlement of CASS materials is determined in terms of casting method, molybdenum content, and ferrite content. For low-molybdenum content steels (SA-351 Grades CF3, CF3A, CF8, CF8A or other steels with 0.5 weight percent (wt.%) Mo), only static-cast steels with >20 percent ferrite are potentially susceptible to thermal embrittlement. Static-cast low-molybdenum steels with 20 percent ferrite and all centrifugal-cast low-molybdenum steels are not susceptible. For high-molybdenum content steels (SA-351 Grades CF3M, CF3MA, and CF8M or other steels with 2.0 to 3.0 wt.% Mo), static-cast steels with >14 percent ferrite and centrifugal-cast steels with >20 percent ferrite thermal embrittlement can be potentially significant (i.e., screens in). For static-cast high-molybdenum steels with 14 percent ferrite and centrifugal-cast high-molybdenum steels with 20 percent ferrite, thermal aging embrittlement is not significant (i.e., screens out). The thermal embrittlement screening criteria of CASS with different molybdenum and ferrite contents are summarized in NUREG-2191, Table XI.M12-1, reproduced as Table 10-1 below.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 10-6 In the significance screening method, ferrite content is calculated by using the Hulls equivalent factors (described in NUREG/CR-4513, Revision 1) or a staff-approved method for calculating delta ferrite in CASS materials. A fracture toughness value of 255 kilo-joules per square meter (kJ/m2) (1,450 inch-pounds per square inch) at a crack extension of 2.5 millimeters (0.1 inch) is used to differentiate between CASS materials for which thermal aging embrittlement is not significant and those for which thermal aging embrittlement is potentially significant. Extensive research data indicate that for CASS materials without the potential for significant thermal aging embrittlement, the saturated lower-bound fracture toughness is greater than 255 kJ/m2 (NUREG/CR-4513, Revision 1).

The final interim staff guidance (Reference 10-2) provides clarification regarding acceptance criteria for a flaw tolerance approach. NUREG-2191, XI.M12, specifies that flaws detected in CASS components are evaluated in accordance with the applicable procedures of the ASME Code,Section XI. The most recent version of the ASME Code,Section XI incorporated by reference in 10 CFR 50.55a (2013 Edition), does not contain evaluation procedures applicable to CASS with ferrite content 20 percent.

The final ISG permits use of the 2019 Edition of ASME Code,Section XI, Appendix C, wherein the flaw evaluation procedures in the 2019 Edition of the Code were developed by considering the ferrite content, fracture toughness, tensile data of CASS materials, and the relevant elastic-plastic correction factors (Z-factors) as a function of ferrite content.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 10-7 Table 10-1 Thermal Embrittlement Susceptibility from NUREG-2191, Table XI.M12-1 Table XI.M12-1. Thermal Embrittlement Susceptibility Potentially Not Molybdenum (Mo) Fe Casting Susceptible Susceptible Content Content Method (Screens In) (Screens Out)

>20%

Low or 0.5 wt.% Static X ferrite 20%

Low or 0.5 wt.% Static X ferrite Low or 0.5 wt.% Any Centrifugal X

>14%

High or 2.0-3.0 wt.% Static X ferrite

>20%

High or 2.0-3.0 wt.% Centrifugal X ferrite 14%

High or 2.0-3.0 wt.% Static X ferrite 20%

High or 2.0-3.0 wt.% Centrifugal X ferrite

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 10-8 10.3 Identification of RCP Items Subject to Aging Management Review RCP items subject to aging management review are reported in the ONS 60-year license renewal application (Reference 10-3), Section 3.4.8, and include the following:

Casing, Cover, and Pressure-Retaining Bolting. Reference 10-3, Table 3.4-1, identifies that the casing and cover are fabricated from CASS with an applicable aging effect of reduction of fracture toughness, and aging management in accordance with the ASME Code,Section XI, In-service Inspection, as supplemented by the CASS Flaw Evaluation Procedure reported in Section 4.18.2.1.

Oconee reactor coolant pump quality assurance data packages, which contain material certification reports for pressure retaining items, were used to confirm the materials of construction for RCP CASS items to obtain material chemistry data required to calculate ferrite content using Hulls Equivalent Factors. In some instances, non-pressure retaining items (e.g., volute and suction piece adapter for ONS Unit 2 and ONS Unit 3) were included for conservatism.

The following Oconee RCP pressure retaining items are subject to aging management review and made from CASS.

  • ONS Unit 1main flange and pump casing (CF8), Figure 10-2.
  • ONS Units 2 and 3cover/stuffing box and pump casing (CF8M), Figure 10-1.

Non-pressure retaining parts volute and suction piece adapter are made from CF8 and are shown for more information only.

10.4 Calculation of Ferrite Numbers for CASS Parts Ferrite numbers are calculated for the RCP pressure boundary items. Ferrite % is calculated using Hulls Equivalent factors and heat-specific chemistry as follows (Reference 10-4, Equations (3), (4), and (5)). There is no indication that niobium was used in any of the casting heats.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 10-9 Creq = Cr + 1.21(Mo) + 0.48(Si) - 4.99 Nieq = (Ni) + 0.11(Mn) - 0.0086(Mn)2+ 18.4(N) + 24.5(C) + 2.77, where the concentrations of the various alloying and interstitial elements are given in wt.%. The concentration of N is often not available in a CMTR; if not known, it is assumed to be 0.04 wt % (Reference 10-4). The maximum molybdenum content for CF8 material is assumed to be 0.5% in accordance with ASTM-A351. The ferrite content c is given by c = 100.3(Creq/ Nieq)2 - 170.72(Creq/ Nieq) + 74.22 Results of RCP items (CF8, CF8M) that exceed the ferrite screening criteria listed in Section 10.2 are summarized in Table 10-2.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 10-10 Table 10-2 ONS RCP Items Susceptible to Thermal Embrittlement Based on Ferrite %

ONS Unit 1 Not susceptible to thermal embrittlement-all casings and main flanges are made from CF8 and ferrite % of each item is below 20%.

ONS Unit 2 Items potentially susceptible to thermal embrittlement include selected CF8M casings (upper and lower) and selected stuffing boxes (CF8M) based on ferrite percentages. Volute and suction piece adapters are made from CF8, ferrite %

are all below 20%, and are not susceptible to thermal embrittlement. In all instances, static castings are assumed.

RCP Item Heat Number Ferrite % >14%

RCP Casings (lower and upper)-CF8M 2RC-P1A1-Serial Number 1 Lower Half-Heat 17368-1 14.46 2RC-P1A2-Serial Number 5 Upper Half-Heat 19427-1 17.55 2RC-P1A2-Serial Number 5 Lower Half-Heat 20683 18.59 2RC-P1B2-Serial Number 3 Upper Half-Heat 20723 16.46 Stuffing Boxes-CF8M Heat 20011-1 14.72 Heat-19880-2 16.98 Heat 16858-3 16.45 Heat 19969-1 19.57 Heat 19880-3 17.33 ONS Unit 3 Items potentially susceptible to thermal embrittlement include CF8M casings (upper and lower) and selected stuffing boxes based on ferrite percentages. Volute and suction piece adapters are made from CF8, ferrite % are all below 20%, and are not susceptible to thermal embrittlement. In all instances, static castings are assumed.

RCP Item Heat Number Ferrite %>14%

RCP Casings (lower and upper)-CF8M 3RC-P3A1-Serial Number 8 Lower Half-Heat 21795 14.20 3RC-P3A1-Serial Number 8 Upper Half-Heat 24245-1 15.96 3RC-P3A2-Serial Number 10 Upper Half-Heat 13011 19.94 3RC-P3B1-Serial Number 6 Upper Half-Heat 22270-1 15.65 3RC-P3B1-Serial Number 6 Lower Half-Heat 20910-1 15.38 3RC-P3B2-Serial Number 7 Lower Half-Heat 24685 20.06 Stuffing Boxes-CF8M Heat 15384-2 17.25 Heat 16496-4 14.86 Heat 17434 16.46 Heat 19969-3 20.32 Heat 17278-2 16.93

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 10-11 ONS Unit 1 CF8 castings are not susceptible to thermal embrittlement since all casings and main flanges are made from CF8 and ferrite % of each item is below 20%. For ONS Unit 2 and ONS Unit 3, items potentially susceptible to thermal embrittlement include CF8M casings (upper and lower) and selected stuffing boxes based on ferrite percentages. Volute and suction piece adapters are made from CF8, ferrite % are all below 20%, and are not susceptible to thermal embrittlement. In all instances, static castings are assumed.

10.5 Temperatures of ONS Unit 2 and ONS Unit 3 Stuffing Boxes During Normal Operation In accordance with Reference 10-1, Volume 2, Program XI.M12, the program includes screening criteria to determine which CASS components have the potential for significant reduction of fracture toughness due to thermal aging embrittlement and require augmented inspection. The screening criteria are applicable to all primary pressure boundary components constructed from CASS with service conditions above 250 °C (482 °F). CASS components that see normal service at 482 °F and below are not susceptible to reduction of fracture toughness by thermal embrittlement. The stuffing boxes for ONS Unit 2 and ONS Unit 3 are located above the thermal barrier and the temperature of the CASS stuffing box is below 482 °F during normal full power operation, assuming a conservative cold leg temperature of 580 °F. As such, the stuffing boxes are below the temperature threshold for thermal embrittlement and are not susceptible to reduction of fracture toughness.

10.6 Calculation of Thermally Aged Jd The methodology for calculating thermally aged Jd for each affected material heat (ONS Unit 2 and ONS Unit 3) is in accordance with NUREG/CR-4513, Revision 2 (Reference 10-4), Section 3.1, Estimation of Thermal Embrittlement of CASS Materials of Known Composition and Service Condition - Service Time Values. Equations are as follows.

  • Material heat chemistry is identified and ferrite numbers are calculated based on Reference 10-4, Equation (5)

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 10-12

  • For each pump casing half, calculate Room Temperature CV sat using Reference 10-4, CF8M equations (21) through (23) where Ni < 10%, and equations (24) through (26) where Ni > 10%
  • Calculate decrease in room temperature CV using Reference 10-4 equations (11), (13) to (17) and assume Cv init is 200 J/cm2 per page 25 since that is not available from the QADP. Aging parameter P based on 72 EFPY at 555 °F.
  • Develop J-R curve Jd=C(delta a)n by using Reference 10-4 Equations (31) and (32) for C and at 290 °C to 320 °C, and assume static castings. For n, use Reference 10-4, Equation (44) for 290°C to 320 °C.

This results in a fracture toughness estimate for each pump casing half for the temperature range 290 °C (554 °F) - 320 °C (608 °F). Pump casings with thermally aged Jd > 1450 in-lb/in2 (see Section 10.2) at a crack extension of 0.1 inches are not susceptible to reduction of fracture toughness by thermal embrittlement.

There were only four heats of material whose Jd (0.1) inches are less than 1450 inlb/in2, and are reported in Table 10-3. The pump heat with minimum Jd (0.1) is 2RC-P1A1-Serial Number 1, Lower Half-Heat 17368-1.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 10-13 Table 10-3 Thermally Aged Jd (0.1-inch) for ONS Unit 2 and ONS Unit 3 RCP Casings ONS Unit 2 2

RCP Heat Ferrite Jd (0.1-inch) Thermally Aged (in-lb/in ) Comment 2

less than 1450 in-lb/in . Susceptible 2RC-P1A1-Serial Number 1 Lower Half-Heat 17368-1 14.46 837.81 to TE and candidate for flaw tolerance 2RC-P1A2-Serial Number 5 Upper Half-Heat 19427-1 17.55 1182.42 ""

2RC-P1A2-Serial Number 5 Lower Half-Heat 20683 18.59 1063.91 ""

2RC-P1B2-Serial Number 3 Upper Half-Heat 20723 16.46 1835.41 Not susceptible since Jd > 1450 ONS Unit 3 3RC-P3A1-Serial Number 8 Lower Half-Heat 21795 14.2 2105.39 Not susceptible since Jd > 1450 3RC-P3A1-Serial Number 8 Upper Half-Heat 24245-1 15.96 2084.61 Not susceptible since Jd > 1450 2

less than 1450 in-lb/in . Susceptible 3RC-P3A2-Serial Number 10 Upper Half-Heat 13011 19.94 1214.81 to TE and candidate for flaw tolerance 3RC-P3B1-Serial Number 6 Upper Half-Heat 22270-1 15.65 2133.33 Not susceptible since Jd > 1450 3RC-P3B1-Serial Number 6 Lower Half-Heat 20910-1 15.38 2070.08 Not susceptible since Jd > 1450 3RC-P3B2-Serial Number 7 Lower Half-Heat 24685 20.06 2143.71 Not susceptible since Jd > 1450 Note:

Items potentially susceptible include CF8M casings (upper and lower). Stuffing boxes made from CF8M not susceptible due to temperature < 482 °F at full power.

Ferrite > 14% potentially susceptible to TE for CF8M

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 10-14 10.7 JIc for Limiting ONS Unit 2 RC-P1A1, Lower Half-Heat 17368-1 Should a linear elastic flaw tolerance evaluation be performed, a value of JIc must be obtained from the formulation of Jd reported in Section 10.6, so that a value of KIc may be calculated. A bounding JIc may be obtained from RCP Heat 17368-1. The procedure used to calculate JIc from the Jd power law formulation is consistent with ASTM E813-89, Section 9.2, Figure 3. Steps are as follows for Heat 17368-1.

  • Develop a blunting line in accordance with the following equation, J = 2*y*a

- y = yield strength from the QADP, psi; [ ]

- a=crack extension, inches

  • Develop and plot an offset line parallel to the blunting line at an offset value of 0.2 mm (0.008 in.)
  • The estimate of JIc shall be at the intersection of the power law J-R curve and the offset line
  • Calculate KIc from the following equation

- KIc = (E'JIc)1/2, where the normalized elastic modulus is given by E' = E/(1 2), E is the elastic modulus, and is the Poisson ratio. Modulus of elasticity and Poissons ratio are from the ASME Code,Section II, 2017 Edition, at 555 °F.

E=25.9E06-(25.9E06-25.3E06)*55/100=25.6E06 psi from Table TM-1 (Group G Material)

Poissons ratio = 0.30 from Table PRD Jd = [ ].

Jd = [

].

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 10-15 JIc = [ ] and KIc= [ ].

10.8 Summary and Conclusions In accordance with Table 10-2, four RCP casing heats of material for ONS Unit 2 are potentially susceptible to thermal embrittlement, and six RCP pump casing heats of material for ONS Unit 3 are potentially susceptible to thermal embrittlement based on ferrite screening. When considering thermally aged Jd (0.1-inch) for the affected heats and comparing those values to a screening criterion of 1,450 in-lb/in2, RCP pump Heats 17368-1, 19427-1, 20683 for ONS Unit 2, and Heat 13011 for ONS Unit 3 require further evaluation for subsequent license renewal. As permitted by NUREG-2191, Volume 2, XI.M12, further evaluation may be performed for these RCP casing pump heats by completing a bounding flaw tolerance evaluation to demonstrate that the thermally aged material toughness adequately protect against a loss of structural integrity in CASS components. This evaluation is reported in Section 11.0 herein.

10.9 References for Section 10.0 10-1. NUREG-2191, Volume 2, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report 10-2. SLR-ISG-2021-02-MECHANICAL, Updated Aging Management Criteria for Mechanical Portions of Subsequent License Renewal Guidance Interim Staff Guidance (ADAMS Accession No. ML20181A434) 10-3. Letter from Duke Energy Corporation forwarding application for renewal of operating licenses for the Oconee Nuclear Station, Unit Nos. 1, 2, and 3, U.S.

Nuclear Regulatory Commission, ACN: 9807200136, Fiche: A4344:001-A4347:255, July 6, 1998 (ADAMS Accession No. ML15254A151 and ML15112A661) 10-4. NUREG/CR-4513, Revision 2, Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems, ADAMS Accession Number ML16145A082

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 11-1 11.0 ONS UNIT 2 AND ONS UNIT 3 RCP BOUNDING FLAW TOLERANCE EVALUATION 11.1 Introduction As permitted by NUREG-2192, Volume XI,M12 (see Section 10.2), a bounding flaw tolerance evaluation of the ONS Units 2 and 3 RCP casings that are potentially susceptible to reduction of fracture toughness by thermal embrittlement (Section 10.8) was completed by performing a fatigue crack growth analysis and a Linear Elastic Fracture Mechanics (LEFM) based fracture toughness margin evaluation using the existing 2-D linearized finite element stresses from the analyses of the suction side of the casing. These LEFM evaluations for subsequent license renewal are based, in part, on the Code Case N-481 (Reference 11-1) evaluations for ONS Unit 2, Pump Casing Heats 17368-1 and 16729-1, reported in Reference 11-2, and for ONS Unit 3, Pump Casing Heats 21795-1 and 24245-1, reported in Reference 11-3.

In accordance with NRC Regulatory Guide 1.147 (Reference 11-4), Code Case N-481 was annulled on 3/28/2004 and ASME Code,Section XI, Examination Category B-L-1, was removed from the ASME Code,Section XI, in the 2007 Edition, 2008 Addenda (Reference 11-5). At present, Duke is utilizing the 2007 Edition through the 2008 Addenda of the ASME Code,Section XI and portions of the 2017 Edition at Oconee (Reference 11-6) for the fifth ISI interval at Oconee. As such, Code Case N-481 is no longer utilized at ONS Unit 2 and ONS Unit 3 for inservice inspection of RCP casings.

However, as reported in NUREG-2191, Volume 2, XI.M12, for RC pump casings, as an alternative to the screening and other actions described above, no further actions are needed if applicants demonstrate that the original flaw tolerance evaluation performed as part of Code Case N-481 implementation remains bounding and applicable for the SLR period or the evaluation is revised to be applicable for 80 years.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 11-2 As such, the ONS Unit 2 and ONS Unit 3 Code Case N-481 evaluations are revisited with modifications discussed below for SLR to demonstrate that no further actions are needed for the CASS for Unit 2 and Unit 3 RCP heats that are susceptible to thermal embrittlement. In accordance with Regulatory Guide 1.147, an annulled Code Case cannot be used in a subsequent ISI interval unless it is implemented as an approved alternative under 10 CFR 50.55a(z). However, NUREG-2191, XI.M12, places no restrictions on the Code Case N-481 evaluation applicability duration and NRC approval of the updated flaw tolerance analysis, based in part on the current Code Case N-481 evaluation, will be provided through NRC review of the Oconee subsequent license renewal application.

11.2 Methodology The methodology for the RCP LEFM evaluation for SLR is as follows. The limiting ONS Unit 2 RCP 2RC-P1A1 is selected for the bounding evaluation.

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Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 11-9 11.4 Summary of Results The circumferential and axial fatigue flaw growth for an initial 0.3-inch semielliptical flaw with a 6:1 aspect ratio for the most limiting locations are summarized in Table 11-1.

Since the final a/t ratio is less than 0.25 (quarter thickness), the flaw tolerance objective is met. In addition, since the maximum stress intensity factor, Kmax of [ ] is less than the limiting KIc value of [ ] (limiting KIc value occurs in 2RC-P1A1 Heat 17368-1), the RCP pump casings demonstrate adequate fracture toughness margin for the 80 year SLR period. Note that the limiting KIC reported for Heat 17368-1 in the Code Case N-481 analysis in Reference 11-2 is 140.6 ksi-in, which is a saturated lower bound value. The reduction at 80 years to [ ] is due to the use of NUREG/CR-4513, Revision 2, and consideration of thermally aged properties of Heat 17368-1 (i.e., Jd and calculation of JIC) at 72 EFPY.

Aging management of the potentially susceptible RCP pump heats 17368-1, 19427-1, 20683 for ONS Unit 2, and heat 13011 for ONS Unit 3 is accomplished through a bounding flaw tolerance evaluation using guidance from the evaluation procedures and acceptance criteria contained in ASME Code, Code Case N-481. Since Code Case N-481 does not provide guidance on safety factors to be used in the evaluation; safety factors consistent with Appendix G of the ASME Code,Section XI are used. The flaw tolerance evaluation is done by performing a fatigue crack growth analysis (using ASME B&PV Code Case N-809) and Linear Elastic Fracture Mechanics (LEFM) based fracture toughness evaluation to demonstrate that the postulated flaw does not grow to the reference quarter thickness (1/4T) flaw size and the maximum stress intensity factor (Kmax) at the final flaw size does not exceed the critical stress intensity factor, KIc determined for the susceptible heats. The results from the bounding evaluation for the potentially susceptible four RCP casing heats at ONS Units 2 and 3 demonstrate that there is adequate fracture toughness margin for the 80 year Subsequent License Renewal (SLR) period and therefore the thermally aged material toughness adequately protects against a loss of structural integrity in CASS components.

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 11-10 Table 11-1 Summary of Results Flaw Type Location Actual Allowable Kmax,AppG KIc, Margin (1-Actual/Allowable)*100 Fracture Toughness ksiin ksiin Circumferential Path 4 Axial Path 2 Flaw Size Circumferential Path 4 Axial Path 1

- I I

Framatome Inc. ANP-3898NP Revision 0 Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 11-11 11.5 References for Section 11.0 11-1. ASME Boiler and Pressure Vessel Code, Code Case N-481, Alternate Examination Requirements for Cast Austenitic Pump Casings,Section XI, Division 1, March 5, 1990 11-2. ISI Report, Unit 2 Oconee 1998 Refueling Outage 16, ADAMS Accession Number ML15239A572 (SIA Report SIR-98-039-Duke Report OSC-7225-provides CC-N481 report for ONS-2, RCP heat 17368-1, but not available on ADAMS) 11-3. ISI Report for Oconee Unit 3 Refueling Outage 17, ADAMS Accession Number ML15239A177 (SIR-98-077 included in ML15239A177) 11-4. Regulatory Guide 1.147, Revision 19, Inservice Inspection Code Case Acceptability, ASME Code,Section XI, Division 1 11-5. ASME Boiler and Pressure Vessel Code,Section XI, 2007 Edition through 2008 Addenda 11-6. Brunswick 1 & 2; Catawba 1 & 2; Robinson 2; McGuire 1 & 2; Oconee 1, 2, & 3; Harris 1 - Request to Use a Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI (EPID L-2020-LLR-0126), ADAMS Accession Number ML21029A335 11-7. ASME Boiler and Pressure Vessel Code,Section XI, 2019 Edition.

11-8. ASME Boiler and Pressure Vessel Code, Code Case N-809, Reference Fatigue Crack Growth Rate Curves for Austenitic Stainless Steels in Pressurized Water Environment,Section XI, Division 1, June 23, 2015.

11-9. EPRI Report NP-1406-SR, Nondestructive Examination Acceptance Standards, Technical Basis and Development of Boilers and Pressure Vessel Code, ASME Code,Section XI, Division 1, May 1980

ENCLOSURE 4 ATTACHMENT 2 OCONEE NUCLEAR STATION Framatome Topical Report ANP 3899NP, Revision 0, Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA, May 2021

framatome Framatome Reactor Vessel Internals ANP-3899NP Revision 0 TLAA Input to the ONS SLRA May 2021 (c) 2021 Framatome Inc.

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

ANP-3899NP Revision 0 Copyright © 2021 Framatome Inc.

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

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page i Nature of Changes Section(s)

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

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page ii Contents Page

1.0 INTRODUCTION

............................................................................................... 1-1 1.1 References ............................................................................................. 1-5 2.0 REDUCTION OF FRACTURE TOUGHNESS DUE TO NEUTRON EMBRITTLEMENT (BAW-10008) (SLRA SECTION 4.7.1.1) ............................ 2-1 2.1 Introduction ............................................................................................. 2-1 2.2 Methodology ........................................................................................... 2-4 2.3 Evaluation ............................................................................................... 2-6 2.3.1 Phase I ......................................................................................... 2-6 2.3.2 Phase II ...................................................................................... 2-27 2.4 Summary and Conclusions ................................................................... 2-31 2.5 References ........................................................................................... 2-31 3.0 REACTOR VESSEL INTERNALS FLOW-INDUCED VIBRATION ENDURANCE LIMITS, BAW-10051 (SLRA SECTION 4.7.1.2)......................... 3-1 3.1 Introduction ............................................................................................. 3-1 3.2 Methodology ........................................................................................... 3-2 3.3 Evaluation (Update of FIV Endurance Limit TLAA for 80 years) ...................................................................................................... 3-2 3.3.1 Environmentally-Assisted Fatigue Affects .................................... 3-2 3.3.2 ASME Fatigue Curves for Design Analysis (80 Years) ................ 3-4 3.3.3 Fatigue Usage Factor for RV Internals ......................................... 3-5 3.4 Summary and Conclusions ..................................................................... 3-9 3.5 References ............................................................................................. 3-9 4.0 REACTOR VESSEL INTERNALS REPLACEMENT BOLTING METAL FATIGUE (SLRA SECTION 4.3.2.2)..................................................... 4-1 4.1 Introduction ............................................................................................. 4-1 4.2 Methodology ........................................................................................... 4-2 4.3 Evaluation (Update of Metal fatigue CUF for 80 years) .......................... 4-2 4.3.1 Environmentally-Assisted Fatigue Effects .................................... 4-2 4.3.2 ASME Fatigue Curves for Design Analysis .................................. 4-2 4.3.3 Fatigue Usage Factor for the Replacement Bolts......................... 4-3 4.4 Summary and Conclusions ..................................................................... 4-6 4.5 References ............................................................................................. 4-6

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page iii 5.0 REACTOR VESSEL INTERNALS FLUENCE/DPA AT 72 EFPY (SLRA SECTION 4.7.1.3 AND XI.M16A)........................................................... 5-1 5.1 Introduction ............................................................................................. 5-1 5.2 Methodology ........................................................................................... 5-2 5.3 Oconee Specific 72 EFPY Reactor Internals Fluence/dpa...................... 5-2 5.3.1 DORT 2-D Calculation of RVI Neutron Exposure to 72 EFPY............................................................................................ 5-3 5.3.2 MCNP Calculation of RVI Neutron Exposure to 72 EFPY............................................................................................ 5-6 5.3.3 Estimate of Uncertainty .............................................................. 5-11 5.4 Oconee TLAA (BAW-10008, Part 1, Revision 1) Fluence Margin ................................................................................................... 5-12 5.5 Oconee Gap Analysis (XI.M16A) Fluence Margin ................................ 5-12 5.6 Summary and Conclusions ................................................................... 5-13 5.7 References ........................................................................................... 5-18

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page iv List of Tables Table 2-1 ASME B&PV Code Yield Strength Values at 600°F .................................. 2-7 Table 2-2 Faulted Stress Summary From BAW-10008, Part 1, Revision 1 vs ASME Code Strengths at 600°F ................................................................ 2-8 Table 2-3 Projected Fast Fluence (E > 1.0 MeV) of Select Reactor Vessel Internals Component Items from Table 1 of BAW-10008, Part 1, Revision 1 ................................................................................................ 2-15 Table 2-4 Core Support Shield Stress Analysis Results for Faulted Conditions Reproduced from BAW-1621 Supplement 1 & ASME Code Yield Strength at 600ºF .................................................................................... 2-18 Table 2-5 Lower Grid Assembly Stress Analysis Results for Faulted Conditions Reproduced from BAW-1621 Supplement 1 & ASME Code Yield Strength at 600ºF .................................................................................... 2-19 Table 2-6 Flow Distributor Assembly Stress Analysis Results for Faulted Conditions Reproduced from BAW-1621 Supplement 1 & ASME Code Yield Strength at 600ºF .................................................................. 2-20 Table 2-7 Core Barrel Assembly Stress Analysis Results for Faulted Conditions Reproduced from BAW-1621 Supplement 1 & ASME Code Yield Strength at 600ºF .................................................................................... 2-21 Table 2-8 Plenum Assembly Stress Analysis Results for Faulted Conditions Reproduced from BAW-1621 Supplement 1 & ASME Code Yield Strength at 600ºF .................................................................................... 2-22 Table 2-9 Projected Fast Fluence (E > 1.0 MeV) of Select Reactor Vessel Internals Component Items from BAW-1621 Supplement 1 .................... 2-26 Table 2-10 Select Component/Weld Recalculated Stress Analysis Results for Faulted Conditions and ASME Code Yield Strength at 600°F ................. 2-29 Table 2-11 High Temperature Alloy A-286 Yield Strength Data ................................ 2-30 Table 3-1 ASME Fatigue Curve for Austenitic Steels (RV Internals) ......................... 3-6 Table 3-2 ASME Fatigue Curve for High Strength Steel Bolts (RV Internal Bolts)..... 3-7 Table 3-3 Summary of FIV Stress Results for the SLR for 80 Years ......................... 3-8 Table 4-1 Fatigue Curve for RV Internal Replacement Bolts (Inconel X-750 Material) .................................................................................................... 4-5 Table 5-1 Projected Fast Fluence (E > 1.0 MeV) of Reactor Vessel Internals Component Items in Support of TLAA in Section 2.0 .............................. 5-14 Table 5-2 Projected 72 EFPY, 80 EFPY Fluence and dpa, MRP-189 Revision 3, used to Support the Gap Analysis ....................................................... 5-15

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page v Table 5-3 Margin between Projected Fast Fluence (E > 1.0 MeV) of Reactor Vessel Internals Component Items in Support of TLAA in Section 2.0 and the ONS-Specific Best-Estimate 72 EFPY Fluence Values .............. 5-16 Table 5-4 Margin between Projected Fast Fluence (E > 1.0 MeV) of Reactor Vessel Internals Component Items in Support of the Gap Analysis and the ONS-Specific Best-Estimate 72 EFPY Fluence Values .............. 5-17

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page vi List of Figures Figure 2-1 Change in Yield Strength as a Function of Neutron Dose for Solution-Annealed Type 304, 304L, and 347 Stainless Steels at Elevated Temperature, 270-380°C (Figure 13(a) of NUREG/CR-7027) ................. 2-11 Figure 5-1 Sketch of the Oconee RVI DORT model ................................................... 5-5 Figure 5-2 View of Lower Grid Assembly and Core Barrel ......................................... 5-8 Figure 5-3 View of Upper Vessel Base Metal ............................................................. 5-9 Figure 5-4 View of Incore Guide Support Plate......................................................... 5-10

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page vii Nomenclature Acronym Definition 10 CFR Title 10, Code of Federal Regulations AMP Aging Management Program ASME American Society of Mechanical Engineers B&PV Boiler and Pressure Vessel CLB Current Licensing Basis dpa Displacements per atom EAF Environmentally-Assisted Fatigue EFPY Effective Full Power Year FIV Flow-Induced Vibration LBB Leak-Before-Break LCB Lower Core Barrel LOCA Loss of Coolant Accident LRA License Renewal Application MRP Materials Reliability Program NRC U.S. Nuclear Regulatory Commission ONS Oconee Nuclear Station ONS-1 Oconee Nuclear Station Unit 1 ONS-2 Oconee Nuclear Station Unit 2 ONS-3 Oconee Nuclear Station Unit 3 RAI Request for Additional Information RPV Reactor Pressure Vessel RV Reactor Vessel RVI Reactor Vessel Internals SER Safety Evaluation Report SSHT Surveillance Specimen Holder Tube Sy Yield Strength TLAA Time Limited Aging Analysis UCB Upper Core Barrel

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page viii ABSTRACT The purpose of this document is to provide supplemental information relative to the information reported in the subsequent license renewal application for Oconee Nuclear Station Units 1, 2, and 3 reactor vessel internals (RVI) time limited aging analyses (TLAA) topics that are reported in NUREG-2192, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants. NUREG-2192, Table 4.1-2, Section 4.7, and Section 4.3.3.1.1 include the following TLAA topics that apply to the RVI.

  • Table 4.1 Ductility Reduction Evaluation for Reactor Internals (B&W designed PWRs only)
  • Section 4.3.3.1.1 - Components Evaluated for Fatigue Parameters Other Than CUFen

The supplemental information provided in this ANP report is intended to assist the NRC with review of the Oconee RVI TLAA topics listed above relative to the applicable Review Procedures reported in NUREG-2192. The topics addressed in this ANP report are as follows.

  • Reduction of Ductility, BAW-10008, Part 1, Revision 1, addressed in Section 2.0
  • Flow-Induced Vibration, BAW-10051, Revision 1 & BAW-10051A, Supplement 1, addressed in Section 3.0
  • Metal Fatigue of RVI Replacement Bolting, addressed in Section 4.0
  • Reactor Internals Fluence/dpa, as input to the BAW-10008, Part 1, Revision 1 update for SLR and the XI.M16A gap analysis, addressed in Section 5.0

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 1-1

1.0 INTRODUCTION

The purpose of this document is to provide supplemental technical information relative to the information reported in the subsequent license renewal application for Oconee Nuclear Station Units 1, 2, and 3 reactor vessel internals (RVI) time limited aging analyses (TLAA) topics that are reported in NUREG-2192 (Reference 1-1), Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants. NUREG-2192, Table 4.1-2, Section 4.7, and Section 4.3.3.1.1 include the following TLAA topics that apply to the RVI.

  • Table 4.1 Ductility Reduction Evaluation for Reactor Internals (B&W designed PWRs only)
  • Section 4.3.3.1.1 - Components Evaluated for Fatigue Parameters Other Than CUFen

Current Licensing BasisRVI TLAA The Oconee 60 year license renewal application (Reference 1-2) and associated NRC SER (Reference 1-3), Section 4.2.5.3, contain the following assessment of TLAA applicable to the Oconee RVI.

  • Reduction in ductility (BAW-10008, Part 1, Revision 1, Reference 1-4). The staff determined that the ONS RVI aging management program will adequately manage the irradiation aging effect in accordance with 10 CFR 54.21(c)(1)(iii).

This TLAA was addressed by Duke Energy in the ONS RVI inspection program and the NRC acceptance of this TLAA evaluation is contained in Reference 1-5.

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  • Flow-induced vibration endurance limit assumptions (BAW-10051, Revision 1 and BAW-10051A, Supplement 1, Reference 1-6, Reference 1-7). This TLAA was evaluated in BAW-2248A (Reference 1-8) and the NRC determined that component stress values were found to be less than the endurance limit, rendering the evaluation acceptable, according to the requirements of 10 CFR 54.21(c)(1)(ii). This TLAA must be updated for 80-years.
  • Flaw growth acceptance in accordance with the ASME B&PV Code, Section Xl ISI requirements. Duke Energy confirmed that there were no analytical evaluations of flaws applicable to the ONS reactor vessel internals for 60 years.

This topic is addressed as part of Metal Fatigue, Section 4.3, evaluation in NUREG-2192 and is not addressed herein.

EFPY Based on accrued EFPY through Cycles 31, 29, and 30 for Oconee Units 1 through 3 and assuming breaker-to-breaker operation and no outages per cycle (Capacity Factor

= 1), the bounding projected EFPY for 80 years for each Oconee Unit is less than 72 EFPY. Therefore, for the Oconee Nuclear Station, TLAA evaluations for SLRA are completed to 72 EFPY. A measurement uncertainty recapture (MUR) power uprate is conservatively factored in at 2% for all TLAA evaluations that utilize neutron fluence as an input.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 1-3 Measurement Uncertainty Recapture

- The MUR power uprate conditions were reviewed for impact on the existing design basis analyses for the reactor vessel internals. No changes to the RCS pressure were made as part of the power uprate. The existing analyses are based on the design conditions in the RCS functional specification. The MUR power uprate conditions are bounded by the design conditions. Since the operating transients will not change as a result of the power uprate and no additional transients have been proposed, the existing loads, stresses and fatigue values remain valid.

- The structural adequacy of RV internals and incore instrument nozzles of ONS Units 1, 2 and 3 was also reviewed with respect to flow-induced vibration (FIV) relative to the MUR power uprate. The components currently analyzed for FIV include the incore instrumentation nozzles, the flow distributor assembly, the thermal shield, and the inlet baffle. From the comparative analysis, the new operational condition of ONS Units 1, 2 and 3 after the MUR power uprate are bounded by the current analysis (BAW-10051, Revision 1 and BAW-10051A, Supplement 1). The RV internals and incore instrument nozzles are structurally adequate with regard to flow-induced vibration including the effects of the MUR power uprate.

- The MUR (Reference 1-9) was silent regarding the BAW-10008, Part 1, Revision 1. As such, this topic is addressed herein for subsequent license renewal.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 1-4 All RVI TLAA for subsequent license renewal reported in this ANP document consider the revised operating conditions (e.g., 1.64% increase in power) associated with MUR as reported in Reference 1-9. MUR is assumed at the beginning of Cycle 30 for ONS Unit 1, Cycle 29 for ONS Unit 2, and Cycle 29 for ONS Unit 2. At present, MUR has not been initiated and each unit is currently operating in Cycles 32, 30, and 31, respectively.

The supplemental information provided in this ANP report is intended to assist the NRC with review of the Oconee RVI topics above relative to the applicable Review Procedures reported in NUREG-2192. The topics addressed in this ANP report are as follows.

  • Reduction of Ductility, BAW-10008, Part 1, Revision 1, is addressed in Section 2.0
  • Flow-Induced Vibration, BAW-10051, Revision 1, & BAW-10051A, Supplement 1 is addressed in Section 3.0
  • Metal Fatigue of RVI Replacement Bolting is addressed in Section 4.0
  • Reactor Internals Fluence/dpa, as input to the BAW-10008, Part 1, Revision 1 update for SLR and the XI.M16A gap analysis is addressed in Section 5.0.

- The fluence/dpa input to BAW-10008, Part 1, Revision 1 is a TLAA.

Framatome does not consider fluence/dpa input to the XI.M16A gap analysis to be a TLAA since these parameters, while time dependent, are used for screening purposes only. Specifically, time dependent damage mechanism screening is used as input to the failure modes and effects criticality analysis (FMECA), which is a qualitative evaluation performed by a panel of subject matter experts.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 1-5

- The FMECA results are used to perform categorization of RVI items into one of 3 categories (A, B, or C). This categorization may be modified for Category B and C items by further evaluation as reported in MRP-229, and MRP-231, and MRP-355. The FMECA followed by further evaluation for Category B and C items are used to place all RVI items into the following inspection classifications reported in MRP-227: primary, expansion, existing programs, and no additional measures. The final inspection classifications are a result of quantitative input (e.g., fluence and CUFs) and qualitative assessment (FMECA), and as such are not considered by Framatome to meet the definition of a TLAA since fluence and CUFs are not used exclusively to determine the final inspections classifications reported in MRP-227.

1.1 References 1-1. NUREG-2192, Standard Review Plan Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants 1-2. Letter from Duke Energy Corporation forwarding application for renewal of operating licenses for the Oconee Nuclear Station, Unit Nos. 1, 2, and 3, U. S.

Nuclear Regulatory Commission, ACN: 9807200136, Fiche: A4344:001-A4347:255, July 6, 1998, ADAMS Accession Number. ML15254A151 and ML15112A661 1-3. NUREG-1723, Safety Evaluation Report Related to the License Renewal of Oconee Nuclear Station, Units 1, 2, and 3, Docket Numbers 50-269, 50-270, and 50-287, ADAMS Accession Number ML003695154 1-4. BAW-10008, Part 1, Revision 1, "Reactor Internals Stress and Deflection Due to Loss-of-Coolant Accident and Maximum Hypothetical Earthquake, Duke Energy Letter Incorporating BAW-10008, Part 1, Revision 1 into the Oconee Current Licensing Basis (CLB), ADAMS Accession Number ML19312B713. AEC letter to Duke regarding approved SER for BAW-10008, Part 1, Revision 1, ML19319B162

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 1-6 1-5. Oconee Nuclear Station, Units 1, 2, and 3 - Approval of Time-Limited Aging Analysis for Reactor Vessel Internals, ADAMS Accession Number ML13045A489 1-6. BAW-10051, Revision 1, Design of Reactor Internals and Incore Instrumentation Nozzles for Flow-Induced Vibration, September 1972, revised in November 1972, Acceptability of BAW-10051, Revision 1, ADAMS Accession Number ML19316A566.

1-7. BAW-10051A, Supplement 1, Structural Analysis of 177-Fuel Assembly Redesigned Surveillance Specimen Holder Tube," ADAMS Accession Number ML19248D133, 7908020516 (legacy) 1-8. BAW-2248A, Demonstration of the Management of Aging Effects for the Reactor Vessel Internals, March 2000, ADAMS Accession Number ML003708443 1-9. ONS License Amendment Request for Measurement Uncertainty Recapture Power Uprate, February 19, 2020, ADAMS Accession Number ML20050D379 and Safety Evaluation, January 26, 2021, ADAMS Accession Number ML20335A001

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-1 2.0 REDUCTION OF FRACTURE TOUGHNESS DUE TO NEUTRON EMBRITTLEMENT (BAW-10008) (SLRA SECTION 4.7.1.1) 2.1 Introduction Framatome Topical Report BAW-10008, Part 1, Revision 1, dated June 1970 (Reference 2-1) documents the acceptability of the reactor vessel internals under LOCA and a combination of LOCA and seismic loadings. The effect of irradiation on the material properties and deformation limits for the internals is presented in Appendix E where it is concluded that at the end of 40 years, the internals will have adequate ductility to absorb local strain at the regions of maximum stress intensity, and that irradiation will not adversely affect deformation limits. Because the conclusion is based on a fluence determination of 40 years of operation, this meets the definition of 10 CFR 54.3(a) and was identified as time-limited aging analysis (TLAA).

Framatome Topical Report BAW-2248A (Reference 2-2) states that this TLAA will be resolved on a plant-specific basis per 10 CFR 54.21(c)(1)(iii). Plant-specific analysis is required to demonstrate that, under LOCA and seismic loading and with irradiation accumulated at the expiration of the period of extended operation, the internals have adequate ductility to absorb local strain at the regions of maximum stress intensity and will meet the deformation limits. Subsequently, in a letter to the NRC dated December 17, 1999 (Reference 2-3), Duke committed to perform the plant-specific analysis and develop data to demonstrate that the internals will meet the deformation limits at the expiration of the renewal license. In response, the NRC in its SER related to the license renewal of three Oconee Units, determined that this program will adequately manage the irradiation aging effect in accordance with 10 CFR 54.21(c)(1)(iii) (Reference 2-4) for the 60 year period.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-2 In a letter to the NRC dated February 20, 2012 (Reference 2-5), Duke submitted the plant-specific analysis constituting the updated TLAA for reduction of fracture toughness of the reactor vessel internals for NRC staff review. The NRC reviewed the licensee's basis for the update of the reduction of ductility TLAA for the three Oconee Units (Reference 2-6) and determined that:

  • The licensee has projected the neutron fluence for the RVI for the 60 year period using an acceptable methodology consistent with RG 1.190.
  • The licensee compiled test data from materials irradiated in operating light-water reactors, plus the Halden test reactor. These materials were irradiated under conditions more similar to the conditions of the ONS, 1, 2, and 3 RVI and therefore should more accurately represent the behavior of the Type 304SA (solution annealed) material in the ONS 1, 2 and 3 RVI. The newer test data, when plotted on the original graph, confirms the conservatism of the original figure.
  • The licensee's evaluation of the deformation limits of BAW-10008, Part 1, Rev. 1, considering the change in tensile properties of the Type 304SA material due to irradiation, is correct.
  • The licensee appropriately revised Appendix E of BAW-10008, Part 1, Rev.1 to conclude the RVI would have adequate ductility at 60 years (54 EFPY) to withstand the postulated LOCA plus seismic event.
  • The disposition of the TLAA for loss of fracture toughness was not changed by this analysis. Since the NRC staff-approved disposition of this TLAA is that aging will be adequately managed in accordance with 10 CFR 54.21(c)(1)(iii), the licensee must reevaluate this TLAA if new relevant data on loss of ductility of irradiated stainless steel is generated.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-3 Based on the above, the NRC staff concluded that the licensee's evaluation of the TLAA for reduction of ductility of the RVI is acceptable, and the license renewal commitment documented in Section 4.2.5.3 of NUREG-1723 to perform a plant-specific analysis and develop data to demonstrate that the internals will meet the deformation limits at the expiration of the 60 year renewed license, is fulfilled. Therefore, this TLAA must be re-evaluated for subsequent license renewal.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-4 2.2 Methodology Section 3.2 of BAW-10008, Part 1, Revision 1 notes that there are two primary safety considerations that govern the deformation limits of the internals: deformation shall not prevent insertion of control rods, nor shall it prevent adequate post-accident cooling of the core. The resultant faulted condition stresses (Case IV) are provided in Table 1 of BAW-10008, Part 1, Revision 1. Each reactor vessel internals component item listed in Table 1 of BAW-10008, Part 1, Revision 1 is assessed in accordance with one of three process steps (Categories 1 Phase I) for Faulted Conditions to determine if each reactor vessel internals component item should be considered potentially susceptible to an unacceptable amount of reduction of ductility at 72 EFPY. This assessment is similar to that submitted and approved by the NRC for the Davis-Besse Nuclear Power Station, Unit 1 for the period of 60-year license renewal (References 2-7, 2-8). For completeness, more recent stress calculations (i.e., developed after the data reported in Table 1 of BAW-10008, Part 1, Revision 1) applicable to the ONS units were reviewed to ensure that the original design basis calculations reported in Table 1 of BAW-10008, Part 1, Revision 1 capture the limiting reactor vessel internals component items for reduction of ductility. This review included asymmetric loading stress calculations for Faulted Conditions reported in BAW-1621, related RAIs, and the ONS final safety evaluation report (SER) for BAW-1621 (References 2-9, 2-10, 2-11), and original and replacement high-strength bolting stress calculations. BAW-1621 is not identified as TLAA but is included for completeness since it contains faulted load stress intensities that post-date the stress intensities reported in BAW-10008, Part 1, Revision 1. The three process steps (Categories 1-3) to determine the impact of reduction of ductility at 72 EFPY on each component item (in order of evaluation) are as follows:

1. Determine if the faulted stress intensity for the reactor vessel internals component item is less than the unirradiated ASME B&PV Code yield strength at operating temperature (600°F) and therefore plasticity will not occur at 72 EFPY.

Since the material remains elastic (and neutron embrittlement would increase the yield strength), reduction of ductility is acceptable.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-5

2. Determine if the reactor vessel internals component item is already highly irradiated (i.e., only reactor vessel internals component items directly adjacent to the fuel assemblies - e.g., baffle plates), such that the faulted stress intensity remains below the irradiated yield strength (increases as fluence increases) and plasticity will not occur at 72 EFPY. Since the material remains elastic, with large margin to the irradiated yield strength, reduction of ductility is acceptable.
3. Determine if the expected fluence exposure is low enough such that neutron embrittlement is considered negligible, reduction of ductility is minimal or will not occur at 80 years/72 EFPY and unirradiated ductility properties are still applicable.

For each process step, further evaluation is required for those component items that do not satisfy the screening criterion defined within each process step. For example, component items that do not fall below the yield strength criterion in Category 1 are carried forward to Category 2. Category 2 items, developed from the first process step, that do meet the irradiated yield strength criterion are carried forward to Category 3.

Component items that exceed the Category 3 screening criterion require further evaluation to demonstrate that those reactor vessel internals component items will have sufficient ductility at 72 EFPY to meet the deformation limits at the expiration of the subsequent period of extended operation. Further evaluation of those component items are addressed in Phase II.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-6 2.3 Evaluation 2.3.1 Phase I 2.3.1.1 Categorization of Component Items from Table 1 of BAW-10008, Part 1, Revision 1 Each reactor vessel internals component item listed in Table 1 of BAW-10008, Part 1, Revision 1 is assessed in accordance with one of three categories to determine if each reactor vessel internals component item should be considered potentially susceptible to an unacceptable amount of reduction of ductility at 72 EFPY. The three categories to determine the impact of reduction of ductility at 72 EFPY on each component item are as follows:

1. The Case IV faulted stress intensity listed in Table 1 of BAW-10008, Part 1, Revision 1 for the reactor vessel internals component item is less than the unirradiated yield strength at operating temperature and therefore plasticity will not occur at 72 EFPY. Since the material remains elastic (and neutron embrittlement would increase the yield strength), reduction of ductility is acceptable.
2. The reactor vessel internals component item is already highly irradiated (i.e., only reactor vessel internals component items directly adjacent to the fuel assembles

- e.g., baffle plates), such that the faulted stress intensity remains below the irradiated yield strength (increases as fluence increases) and plasticity will not occur at 72 EFPY. Since the material remains elastic, with large margin to the irradiated yield strength, the reduction of ductility is acceptable.

3. The neutron embrittlement is negligible, reduction of ductility is minimal or will not occur at 72 EFPY and unirradiated ductility properties are still applicable.

Component items that are not shown acceptable for the three categories above are evaluated in Phase II.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-7 2.3.1.1.1 Assessment per Criteria Defined by Category #1 For the Category 1 assessment, the faulted stress intensity for each the reactor vessel internals component item is compared to the unirradiated ASME B&PV Code yield strength for the applicable material type at operating temperature (i.e., about 550°F to 600°F). For those reactor vessel internals component items that have a reported stress intensity value less than the unirradiated ASME B&PV Code yield strength at 600°F, the material should remain elastic (and neutron embrittlement would increase the yield strength), reduction of ductility is acceptable; therefore, no further analysis is required.

The unirradiated ASME B&PV Code yield strength (Sy) values at 600°F for the reactor vessel internals component items materials of interest are presented in Table 2-1.

Reactor vessel internals component items reported in BAW-10008, Part 1, Revision 1, Table 1, that satisfy Category 1 screening criterion are reported in Table 2-2. Seven component items were carried forward to Process Step 2: lower grid plate, plenum cover, plenum cylinder reinforcing plate, core support shield top flange, core support shield lower flange, baffle plates, and upper core barrel bolts for ONS-3 only.

Table 2-1 ASME B&PV Code Yield Strength Values at 600°F Yield Strength (psi) of Yield Strength (psi)

Yield Strength (psi) Yield Strength (psi)

Temperature Alloy A-286 of Alloy X-750 of Type 304 of Type 304L

(ºF) (e.g., SA-453 Grade (SB-637, Grade 688, (e.g., SA-240) (e.g., SA-240) 660) Type 3) 600 17,300 14,400 81,000 92,500

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-8 Table 2-2 Faulted Stress Summary From BAW-10008, Part 1, Revision 1 vs ASME Code Strengths at 600°F Case IV Allowable ASME Code Comparison of Case IV Stress Faulted Stress Stress (psi), Yield Faulted Stress Intensity vs Component Item Material Type Intensity not used in Strength (Sy) ASME Yield Strength (Sy) at (psi) Comparison at 600ºF (psi) 600ºF Lower grid plate Further evaluation is PL+Pb 17,300 Outlet pipe rupture required Type 304 Further evaluation is Inlet pipe rupture PL+Pb 17,300 required Further evaluation is Plenum cover Type 304 PL+Pb 17,300 required Further evaluation is Plenum cylinder reinforcing plate Type 304 PL+Pb 17,300 required Pm 14,400 No further analysis is required Upper guide tubes Type 304L PL+Pb 14,400 No further analysis is required Upper guide tube sectors Type 304L PL+Pb 14,400 No further analysis is required Further evaluation is Pm 17,300 Core support shield, top flange required Subcooled portion of LOCA Further evaluation is PL+Pb 17,300 Type 304 required Pm 17,300 No further analysis is required Saturated portion Further evaluation is PL+Pb 17,300 required Pm 17,300 No further analysis is required Core support shield, lower flange Subcooled portion of LOCA Further evaluation is PL+Pb 17,300 required Type 304 Pm 17,300 No further analysis is required Saturated portion of LOCA Further evaluation is PL+Pb 17,300 required Core barrel, top flange Pm 17,300 No further analysis is required Subcooled portion of LOCA PL+Pb 17,300 No further analysis is required Type 304 Pm 17,300 No further analysis is required Saturated portion of LOCA PL+Pb 17,300 No further analysis is required Further evaluation is Baffle plates Type 304 PL+Pb 17,300 required Further evaluation is Internals bolts Pm 81,000 Alloy required (ONS-3 only)

Core barrel-core support A286 Further evaluation is shield joint PL+Pb 81,000 required (ONS-3 only)

Core barrel-lower grid cylinder Alloy Pm 81,000 No further analysis is required joint A286 PL+Pb 81,000 No further analysis is required (1) Stress analysis of the current upper core barrel (UCB) bolting configurations was performed in 2008 for missing or deficient UCB bolts. The maximum calculated faulted stresses for the bolting are:

ONS-1 ( [ ]); ONS-2 ([ ]); and ONS-3 ([ ]).

(2) Stress analysis of current lower core barrel (LCB) bolting configuration was performed in 2008 for deficient LCB bolts. The maximum calculated faulted stresses for the bolting are: ONS-1 ( [

]); ONS-2 ([ ]); and ONS-3

( [ ]).

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-9 2.3.1.1.2 Assessment per Criteria Defined by Category #2 For the Category #2 assessment, the expected reactor vessel fluence (E > 1.0 MeV) exposure of the applicable ONS reactor vessel internals component items listed in Table 1 of BAW-10008, Part 1, Revision 1 are reviewed to determine if the component item is highly irradiated to the level where saturation of the materials yield strength has occurred as shown in Figure 2-1 (Figure 13(a) of NUREG/CR-7027). Only reactor vessel internals component items directly adjacent to the fuel assembles are expected to be highly irradiated. For the case where the component item is highly irradiated at 72 EFPY and the faulted stress intensity is below the irradiated yield strength (which increases as fluence increases), plasticity will not occur at 72 EFPY; therefore, the material remains elastic with large margin to the irradiated yield strength, and the reduction of ductility is acceptable.

Based on the 72 EFPY fluence estimates for the stress types of the seven component items identified in Section 2.3.1.1.1 where the faulted stress intensities are greater than the unirradiated yield strengths at 600°F, only the baffle plates are expected to be highly irradiated ([ ] - Table 5-1). Applying the baffle plate estimated fluence to the curve in Figure 2-1, the irradiated yield strength is expected to be about [

]. Note that, per the correlation in Figure 2-1, above about 5 dpa the irradiated yield strength is greater than a factor of three higher than the baffle plate stress intensity in Table 2-2. This correlation is consistent with Section 3.5.3 of MRP-135, Rev 2 (Reference 2-12), which describes factors for temperature and irradiation dose that can be used to calculate yield strength for Type 304 solution-annealed stainless steel based on updated unirradiated and irradiated tensile test data. The baffle plates faulted stress intensity in Table 1 of BAW-10008, Part 1, Revision 1

([ ]) is significantly less than this irradiated saturated yield strength, therefore the material remains elastic with large margin to the irradiated yield strength, and the reduction of ductility is acceptable and no further evaluation is required.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-10 For the remaining six reactor vessel internals component items identified in Table 2-2 where the faulted stress intensities are greater than the unirradiated yield strengths at 600°F, further assessment in accordance with criteria defined by Category #3 is performed to determine if the reactor vessel internals component item should be considered potentially susceptible to an unacceptable amount of reduction of ductility at 72 EFPY. The six component items carried forward to process step 3 include the lower grid plate, plenum cover, plenum cylinder reinforcing plate, core support shield top flange, core support shield lower flange, and upper core barrel bolting for ONS-3 only.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-11 Figure 2-1 Change in Yield Strength as a Function of Neutron Dose for Solution-Annealed Type 304, 304L, and 347 Stainless Steels at Elevated Temperature, 270-380°C (Figure 13(a) of NUREG/CR-7027)

Type 304 , 304L, & 347 SS (a) 1000 MRP Curve 0 304L-1 (A) SA, Bor-60 at 32o*c, 330*c

6. 304L-1 (A) SA, Samara a 3oo*c, 33o*c I> 304L-1 (A) SA, Samara at 3oo*c, 380"C o+ 0 V 304L-1 (A) SA, Bor-60 +Samara. 330*c 0

<l 304L-1 (CB) SA. Bor-60 at 320*c, 330*c ro a..

+

304-3 (FD) SA. Bo r-6 at 32o*c. 330*c 304 (EH or J7) SA, Bor-60 at 32o* c . 330*c

~ X 304-J3 (EH) 1 %CW, Bor-60 at 32o*c. 33o*c

,C i5>

C:

Q) 600 .*

347-2 (EC) SA. Bor-60 at 32o*c, 33o*c 304 Decom. PWR at 2ss-315*c. 320*c 304 Oskarshamn 112 at 2so*c, 270-290'C

<ii

-0 Q) 304L Oskarshamn 112 at 2so*c. 2sa-295*c 304L Barseback 1 at 2ss*c, 2sa*c 347 Point Beach 2. 32o*c

,(lo 304 BWR at 2ss* c , 288' C 304 Chooz A at 2sa*c. 32o*c B 2 till 304 BWR at 2ss*c. 300'C

!!'.:I 304L BWR at 2ss*c. 300*c 20 0 20 40 60 80 100 Neutron Dose (dpa)

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-12 2.3.1.1.3 Assessment Per Criteria Defined by Category #3 For the remaining six ONS reactor vessel internals component items not screened out per the Category #1 and #2 assessments, a review is performed to determine if the expected fluence (E > 1.0 MeV) exposure of the reactor vessel internals component item is low enough such that neutron embrittlement is considered negligible, thus reduction of ductility is minimal or will not occur at 72 EFPY for the component item.

The projected 72 EFPY dose and fluence values for the remaining six reactor vessel internals component items are listed in Table 2-3. The methodology used to develop these dose and fluence values is described in Section 5.2. If the projected fluences listed in Table 2-3 are applied to Figure E-3 of BAW-10008, Part 1, Revision 1, the decrease in uniform elongation for the plenum cover, plenum cylinder reinforcing plate, and core support shield top flange (fabricated from Type 304 stainless steel) at both 572°F and 752°F (temperatures between which these component items would be expected to experience) is such that the 20 percent uniform elongation of irradiated material credited for 40 years in Appendix E of BAW-10008, Part 1, Revision 1 and the 8.6 percent allowable strain specified in Appendix A of BAW-10008, Part 1, Revision 1 is met for these component items. Note that Figure 3-12 in Reference 2-13 provides irradiated Type 304 stainless steel solution annealed test data to compare to the curves in Figure E-3 of BAW-10008, Part 1, Revision 1. The test data validates the conservatism of the curves in Figure E-3 of BAW-10008, Part 1, Revision 1. This slight decrease in uniform elongation at the projected fluence levels for the plenum cover, plenum cylinder reinforcing plate, and core support shield top flange is confirmed in Figure 13(c) of NUREG/CR-7027 (Reference 2-14).

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-13 Section 3.5.3 of MRP-135, Rev 2 describes factors for temperature and irradiation dose that can be used to calculate uniform elongation for Type 304 solution-annealed stainless steel based on updated unirradiated and irradiated tensile test data. This data requires a dose value in dpa, so these equations can be used to calculate the uniform elongation for the plenum cover only, as the plenum cylinder reinforcing plate and core support shield top flange do not have calculated dose values listed in dpa in Table 2-3.

Using the equations in Section 3.5.3 of MRP-135, Rev 2, the percent uniform elongation calculated for a dose of [ ] (Table 2-3) and a temperature of 600°F (315°C) for solution-annealed Type 304 stainless steel is about [ ]. It should be noted that the dose value for the plenum cover is actually listed in Table 2-3 as less than (<)

[ ], so the percent uniform elongation expected is most likely greater than

[ ].

In addition, as shown in Figure 5 of Volume 5, Issue 3 of the Journal of Engineering Materials and Technology (Reference 2-15), the uniform elongation of unirradiated Type 304SA stainless steel at 600°F only decreases slightly with increasing strain rate.

Further observation of the data shows that even at the highest tested strain rates (101/sec and 102/sec) at 600°F, the uniform elongation is above the 20 percent uniform elongation of irradiated austenitic stainless steel material credited for 40 years in Appendix E and the 8.6 percent allowable strain specified in Appendix A of BAW-10008, Part 1, Revision 1. It is also observed that yield strength increases with increasing strain rate at 600°F as shown in Figure 3 of Volume 5, Issue 3 of the Journal of Engineering Materials and Technology. In addition to having sufficient ductility at 72 EFPY relative to the allowable stresses of Table 1 of BAW-10008, Part 1, Revision 1, the plenum cover, plenum cylinder reinforcing plate, and core support shield top flange will have greater resistance to plastic deformation at increased strain rates.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-14 For the three remaining reactor vessel internals component items (lower grid plate, core support shield lower flange, and upper core barrel bolts at ONS-3), further assessment will be performed in Phase II to determine if these reactor vessel internals component items should be considered potentially susceptible to an unacceptable amount of reduction of ductility at 72 EFPY.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-15 Table 2-3 Projected Fast Fluence (E > 1.0 MeV) of Select Reactor Vessel Internals Component Items from Table 1 of BAW-10008, Part 1, Revision 1 Case IV Faulted 72 EFPY 72 EFPY Neutron Stress (1) (1)

Component Item Stress Intensity Dose Fluence , E > 1.0 MeV Comment Type 2 (psi) (dpa) (n/cm )

Lower grid plate PL+Pb

- - Fluence sufficient to cause neutron embrittlement of Outlet pipe rupture component item. Further evaluation is required.

Fluence sufficient to cause neutron embrittlement of Inlet pipe rupture PL+Pb component item. Further evaluation is required.

Fluence sufficiently low that the effect of irradiation Plenum cover PL+Pb embrittlement on the ductility of component item is minimal; further analysis not required.

Fluence sufficiently low that the effect of irradiation Plenum cylinder reinforcing plate PL+Pb embrittlement on the ductility of component item is minimal; further analysis not required.

Fluence sufficiently low that the effect of irradiation Pm embrittlement on the ductility of component item is minimal; Core support shield, top flange further analysis not required.

Subcooled portion of LOCA Fluence sufficiently low that the effect of irradiation PL+Pb embrittlement on the ductility of component item is minimal; further analysis not required.

Fluence sufficiently low that the effect of irradiation Saturated portion PL+Pb embrittlement on the ductility of component item is minimal; further analysis not required.

Fluence sufficient to cause Core support shield, lower flange neutron embrittlement of PL+Pb Subcooled portion of LOCA component item. Further evaluation is required.

Fluence sufficient to cause neutron embrittlement of Saturated portion of LOCA PL+Pb component item. Further evaluation is required.

Fluence sufficient to cause neutron embrittlement of Pm Internals bolts component item. Further evaluation is required.

Core barrel-core support shield Fluence sufficient to cause joint (ONS-3 only) neutron embrittlement of PL+Pb component item. Further

- I I - evaluation is required.

(1) See Section 5.0 (Table 5-1 and Table 5-2)

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-16 2.3.1.2 Categorization of Component Items from BAW-1621 More recent asymmetric LOCA loading stress calculations (i.e., developed after the data reported in Table 1 of BAW-10008, Part 1, Revision 1) are available and reported in Framatome Topical Report BAW-1621, which is supplemented by NRC RAI responses.

The applicable ONS asymmetric loading stress intensity calculations (i.e., skirt-supported) are reviewed to ensure that the original design basis calculations reported in Table 1 of BAW-10008, Part 1, Revision 1 capture the limiting reactor vessel internals component items for reduction of ductility. Based on the stress limits criteria for faulted conditions (consistent with BAW-10008, Part 1, Revision 1), the assessment in the following sections is limited to only reactor vessel internals component items associated with the stress category Pm (primary membrane stress intensity) or Pm+Pb (primary membrane plus primary bending stress intensity).

Similar to the actions performed for the reactor vessel internals component items in Section 2.3.1.1, each reactor vessel internals component item listed in Tables 3.24-1, 3.24-2, 3.24-3, 3.24-4, and 3.24-5 of BAW-1621, Supplement 1 is assessed in accordance with one of three categories to determine if each component item should be considered potentially susceptible to an unacceptable amount of reduction of ductility at 72 EFPY. The three categories are the same as those listed in Section 2.3.1.1.

Component items that are not shown acceptable for the three categories are evaluated in Phase II.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-17 2.3.1.2.1 Assessment per Criteria Defined by Category #1 For the Category #1 assessment, the faulted condition maximum calculated Pm or Pm+Pb stress intensity values listed in Tables 3.24-1, 3.24-2, 3.24-3, 3.24-4, and 3.24-5 of BAW-1621, Supplement 1 for each reactor vessel internals component item is compared to the unirradiated yield strength at operating temperature (i.e., 600°F). For those reactor vessel internals component items that have reported faulted condition maximum calculated Pm or Pm+Pb stress intensity values less than the unirradiated yield strength at operating temperature, the material should remain elastic (and neutron embrittlement would increase the yield strength), reduction of ductility is acceptable; therefore, no further analysis is required.

The unirradiated yield strength (Sy) values at 600°F for the reactor vessel internals component items materials of interest are presented in Table 2-1.

The comparison of the faulted condition maximum calculated Pm or Pm+Pb stress intensity values listed in Tables 3.24-1, 3.24-2, 3.24-3, 3.24-4, and 3.24-5 of BAW-1621 Supplement 1 to the applicable unirradiated yield strengths at 600°F are shown in Table 2-4, Table 2-5, Table 2-6, Table 2-7, and Table 2-8. Eight component items were carried forward to Process Step 2: ONS-3 UCB bolts, core support shield lower flange, lower grid rib section, lower grid rib section/lower grid shell forging bolt, support post/support forging weld, plenum cylinder, upper grid rib section, and upper grid pad joint bolt.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-18 Table 2-4 Core Support Shield Stress Analysis Results for Faulted Conditions Reproduced from BAW-1621 Supplement 1 & ASME Code Yield Strength at 600ºF Max Maximum ASME Code Comparison of Case IV Stress Calculated Allowable Stress Yield Strength Faulted Stress Intensity Component Item Material Type Stress (psi), not used in (Sy) at 600ºF vs ASME Yield Strength (psi) Comparison (psi) (Sy) at 600ºF Core Support Shield Type 304 Pm - - 17,300 No further analysis is required Buckling N/A N/A Core Support Shield/Core Alloy Further evaluation is Pm 81,000 Barrel Bolted Joint A286 required (ONS-3 only)

No further analysis is Pm 17,300 required Core Barrel Upper Flange Type 304 Bearing N/A N/A Shear N/A N/A No further analysis is Pm 17,300 required Core Support Shield Lower Type 304 Further evaluation is Flange Pm+Pb 17,300 required Bearing N/A N/A Core Support Shield Upper Load Type 304 N/A N/A Flange Limit, kips


I -

(1) Stress analysis of the current UCB bolting configurations was performed in 2008 for missing or deficient bolts. The maximum calculated faulted stress for the bolting are: ONS-1 ( [

]), ONS-2 ([ ]), and ONS-3 ([

]).

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-19 Table 2-5 Lower Grid Assembly Stress Analysis Results for Faulted Conditions Reproduced from BAW-1621 Supplement 1 & ASME Code Yield Strength at 600ºF Maximum ASME Code Comparison of Case IV Max Calculated Stress Allowable Stress Yield Strength Faulted Stress Intensity Component Item Material Stress Type (psi), not used in (Sy) at 600ºF vs ASME Yield Strength (psi)

Comparison (psi) (Sy) at 600ºF

~

- No further analysis is Pm 81,000 Core Barrel/Lower Grid Alloy required Assembly Bolts A286 Stripping N/A N/A Bearing N/A N/A No further analysis is Core Barrel Lower Flange Type 304 Pm 17,300 required Grid Pad Type 304 Bearing N/A N/A Grid Pad/Rib Section Joint - No further analysis is Grade B8 Pm 17,300 Bolt required Grid Pad/Rib Section Joint - Alloy Shear N/A N/A Dowel X750 Further evaluation is Pm 17,300 required Rib Section Type 304 Further evaluation is Pm+Pb 17,300 required Support Post/Rib Section Joint - No, further analysis not Grade B8 Pm 17,300 Bolt required Rib Section/Lower Grid Shell Further evaluation is Grade B8 Pm 17,300 Forging Joint Bolt required No further analysis is Support Posts Type 304 Pm 17,300 required No further analysis is Flow Distributor Plate Type 304 Pm 17,300 required Flow Distributor Plate/Lower Type No further analysis is Pm 14,400 Grid Shell Forging Joint - Weld 308/308L required Support Post/Flow Distributor No further analysis is Type 308L Pm 14,400 Plate Welded Joint required No further analysis is Pm 17,300 required Support Forging Type 304 No further analysis is Pm + Pb 17,300 required Support Post/Support Forging Type Further evaluation is Pm 14,400 Joint - Weld 308/308L required No further analysis is Lower Grid Shell Forging Type 304 Pm


I

- 17,300 required

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-20 Table 2-6 Flow Distributor Assembly Stress Analysis Results for Faulted Conditions Reproduced from BAW-1621 Supplement 1 & ASME Code Yield Strength at 600ºF Maximum ASME Code Comparison of Case IV Max Calculated Stress Allowable Stress Yield Strength Faulted Stress Intensity Component Item Material Stress Type (psi), not used in (Sy) at 600ºF vs ASME Yield Strength (psi)

Comparison (psi) (Sy) at 600ºF Pm

- - 81,000 No further analysis is required Flow Distributor Head/Lower Alloy Grid Assembly Bolts A286 Tear-out N/A N/A Bearing N/A N/A No further analysis is Pm 17,300 required Flow Distributor Shell Type 304 No further analysis is Pm+Pb 17,300 required No further analysis is Pm 17,300 required Flow Distributor Flange Type 304 No further analysis is Pm+Pb 17,300

- I

- required

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-21 Table 2-7 Core Barrel Assembly Stress Analysis Results for Faulted Conditions Reproduced from BAW-1621 Supplement 1 & ASME Code Yield Strength at 600ºF Maximum Allowable ASME Code Comparison of Case IV Max Calculated Stress Stress (psi), Yield Strength Faulted Stress Intensity Component Item Material Stress Type not used in (Sy) at 600ºF vs ASME Yield Strength (psi)

Core Barrel Type 304 Pm Buckling

- Comparison

- (psi) 17,300 N/A (Sy) at 600ºF No further analysis is required N/A No further analysis is Baffle-Former A-A Bolts Grade B8 Pm 17,300 required No further analysis is Barrel-Former A-A Bolts Grade B8 Pm 17,300 required(1)

No further analysis is Thermal Shield Lower End Type 304 Pm 17,300 required No further analysis is Thermal Shield Upper End Type 304 Pm 17,300 required Thermal Shield/Lower No further analysis is Alloy X-750 Pm 92,500 Grid Shell Forging Bolts required Thermal Shield Upper Bearing N/A N/A Type 304 Restraint Shear N/A N/A No further analysis is Bolts Alloy A-286 Pm 81,000 required Dowels Alloy A-286 Shear (1) Based on B&W fabrication records, the lowest measured room temperature yield strength is N/A N/A

[ ] psi. Per empirically based correlations between yield strength and temperature for unirradiated stainless steels, it is estimated the barrel-former bolts measured yield strength will be reduced to [ ] psi at 600°F. Since the calculated stress of [ ] psi is less than

[ ] psi, reduction of ductility is acceptable.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-22 Table 2-8 Plenum Assembly Stress Analysis Results for Faulted Conditions Reproduced from BAW-1621 Supplement 1 & ASME Code Yield Strength at 600ºF Maximum ASME Code Comparison of Case IV Max Calculated Stress Allowable Stress Yield Strength Faulted Stress Intensity Component Item Material Stress Type (psi), not used in (Sy) at 600ºF vs ASME Yield Strength (psi)

Comparison (psi) (Sy) at 600ºF Plenum Cover Type 304 Bearing N/A N/A Further evaluation is Pm 17,300 Plenum Cylinder Type 304 required Buckling N/A N/A No further analysis is Pm 17,300 CRGT Slotted Region Type 304 required Buckling N/A N/A No further analysis is Pm 17,300 CRGT Perforated Region Type 304 required Buckling N/A N/A No further analysis is CRGT Lower Joint - Bolt Grade B8 Pm 17,300 required CRGT Lower Joint - Dowel Type 304 Shear N/A N/A No further analysis is CRGT/Plenum Cover Joint Type 308L Pm 14,400 required No further analysis is Pm 17,300 required Upper Grid Rib Section Type 304 Further evaluation is Pm + Pb 17,300 required Upper Grid Pad Joint - Dowel Alloy X-750 Dowel N/A N/A Further evaluation is Upper Grid Pad Joint - Bolt Grade B8 Bolt 17,300

- required

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-23 2.3.1.2.2 Assessment per Criteria Defined by Category #2 For the Category #2 assessment, the expected reactor vessel fluence (E > 1.0 MeV) exposure of the applicable ONS reactor vessel internals components items are reviewed to determine if the component item is highly irradiated to the level where saturation of the materials yield strength has occurred as shown in Figure 2-1 (Figure 13(a) of NUREG/CR-7027). Only reactor vessel internals component items directly adjacent to the fuel assembles are expected to be highly irradiated. For the case where the component is highly irradiated at 72 EFPY and the faulted condition maximum calculated Pm or Pm+Pb value is below the irradiated yield strength (which increases as fluence increases), plasticity will not occur at 72 EFPY; therefore, the material remains elastic with large margin to the irradiated yield strength, and the reduction of ductility is acceptable.

Based on the fluence estimates for the eight component item Pm or Pm+Pb stress category types identified in Table 2-4, Table 2-5, and Table 2-8 where the faulted condition maximum calculated Pm or Pm+Pb stress intensity values are greater than the unirradiated yield strengths at 600°F, none are expected to be highly irradiated.

For those eight reactor vessel internals component items identified in Table 2-4, Table 2-5 and Table 2-8 where the maximum calculated stress are greater than the unirradiated yield strengths at 600°F, further assessment in accordance with criteria defined by Category #3 is performed to determine if these reactor vessel internals component items should be considered potentially susceptible to an unacceptable amount of reduction of ductility at 72 EFPY.

2.3.1.2.3 Assessment per Criteria Defined by Category #3 For the eight remaining ONS reactor vessel internals component items not screened out per the Category #1 and #2 assessments, a review is performed to determine if the expected fluence (E > 1.0 MeV) exposure of the reactor vessel internals component item is low enough such that neutron embrittlement is considered negligible, thus reduction of ductility is minimal or will not occur at 72 EFPY for the component item.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-24 The projected 72 EFPY dose and fluence values for the remaining eight reactor vessel internals component items are listed in Table 2-9. The methodology used to develop these dose and fluence values is described in Section 5.2. If the projected fluences listed in Table 2-9 are applied to Figure E-3 of BAW-10008, Part 1, Revision 1, the decrease in uniform elongation for the plenum cylinder (fabricated from Type 304 stainless steel) at both 572°F and 752°F (temperatures between which these component items would be expected to experience) is such that the 20 percent uniform elongation of irradiated material credited for 40 years in Appendix E of BAW-10008, Part 1, Revision 1 and the 8.6 percent allowable strain specified in Appendix A of BAW-10008, Part 1, Revision 1 is met for this component item. Note that Figure 3-12 in Reference 2-13 provides irradiated Type 304 stainless steel solution annealed (SA) test data to compare to the curves in Figure E-3 of BAW-10008, Part 1, Revision 1. The test data validates the conservatism of the curves in Figure E-3 of BAW-10008, Part 1, Revision 1. This slight decrease in uniform elongation at the projected fluence levels for the plenum cylinder is confirmed in Figure 13(c) of NUREG/CR-7027.

Section 3.5.3 of MRP-135, Rev 2 describes factors for temperature and irradiation dose that can be used to calculate uniform elongation for Type 304 solution-annealed stainless steel based on updated unirradiated and irradiated tensile test data. This data requires a dose value in dpa, so these equations cannot be used to calculate the uniform elongation for the plenum cylinder as the plenum cylinder does not have calculated dose values listed in dpa in Table 2-9.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-25 In addition, as shown in Figure 5 of Volume 5, Issue 3 of the Journal of Engineering Materials and Technology, the uniform elongation of unirradiated Type 304SA stainless steel at 600°F only decreases slightly with increasing strain rate. Further observation of the data shows that even at the highest tested strain rates (101/sec and 102/sec) at 600°F, the uniform elongation is above the 20 percent uniform elongation of irradiated austenitic stainless steel material credited for 40 years in Appendix E and the 8.6 percent allowable strain specified in Appendix A of BAW-10008, Part 1, Revision 1. It is also observed that yield strength increases with increasing strain rate at 600°F as shown in Figure 3 of Volume 5, Issue 3 of the Journal of Engineering Materials and Technology. In addition to having sufficient ductility at 72 EFPY relative to the allowable stresses of Table 1 of BAW-10008, Part 1, Revision 1, the plenum cylinder will have greater resistance to plastic deformation at increased strain rates.

For the seven remaining reactor vessel internals component items (ONS-3 UCB bolts, core support shield lower flange, lower grid rib section, lower grid rib section/lower grid shell forging bolt, support post/support forging weld, upper grid rib section, and upper grid pad joint bolt), further assessment will be performed in Phase II to determine if these reactor vessel internals component item should be considered potentially susceptible to an unacceptable amount of reduction of ductility at 72 EFPY.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-26 Table 2-9 Projected Fast Fluence (E > 1.0 MeV) of Select Reactor Vessel Internals Component Items from BAW-1621 Supplement 1 Stress Max Calculated 72 EFPY Dose 72 EFPY Neutron Fluence, Component Item 2 Comment Type Stress (psi) (dpa) E > 1.0 MeV (n/cm )

Fluence sufficient to cause Core Support Shield/Core Barrel Bolted Joint (ONS3 Pm [ ] [ ] [ ] neutron embrittlement of component item. Further only) (Note 1) (Note 1) evaluation is required.

Fluence sufficient to cause Core Support Shield Lower Flange Pm+Pb [ ] [ ] [ ] neutron embrittlement of component item. Further (Note 1) (Note 1) evaluation is required.

Fluence sufficient to cause Pm [ ] [ ] [ ] neutron embrittlement of component item. Further (Note 1) (Note 1) evaluation is required.

(Lower Grid) Rib Section Fluence sufficient to cause Pm+Pb [ ] [ ] [ ] neutron embrittlement of component item. Further (Note 1) (Note 1) evaluation is required.

Fluence sufficient to cause (Lower Grid) Rib Section/Lower Grid Shell Forging Joint Bolt Pm [ ] [ ] [ ] neutron embrittlement of component item. Further evaluation is required.

Fluence sufficient to cause Support Post/Support Forging Joint - Weld Pm [ ] [ ] [ ] neutron embrittlement of component item. Further evaluation is required.

Fluence sufficiently low that Plenum Cylinder Pm [ ] [ ] [ ] the effect of irradiation embrittlement on the ductility (Note 1) (Note 1) of component item is minimal; further analysis not required.

Fluence sufficient to cause Upper Grid Rib Section Pm+Pb [ ] [ ] [ ] neutron embrittlement of component item. Further (Note 1) (Note 1) evaluation is required.

Fluence sufficient to cause Upper Grid Pad Joint - Bolt Bolt [ ] [ ] [ ] neutron embrittlement of component item. Further evaluation is required.

(1) See Section 5.0 (Table 5-1 and Table 5-2)

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-27 2.3.2 Phase II Based on the results of the Phase I categorical assessments, the following reactor vessel internals component items were determined to be potentially susceptible to an unacceptable amount of reduction of ductility at 72 EFPY. Duplicate component item results between the two evaluations are underlined.

From Table 1 of BAW-10008, Part 1, Revision 1 (Section 2.3.1.1):

1. Lower grid plate (i.e., lower grid rib section)
2. Core support shield lower flange
3. Upper core barrel bolts (ONS-3 only based on stress analysis of the current ONS-3 upper core barrel (UCB) bolting configuration)

From RAI responses to BAW-1621, Supplement 1 (Section 2.3.1.2):

1. Upper core barrel bolts (ONS-3 only based on stress analysis of the current ONS-3 upper core barrel (UCB) bolting configuration)
2. Core support shield lower flange
3. Lower grid rib section
4. Lower grid rib section/lower grid shell forging joint bolt
5. Support post/support forging joint - weld
6. Upper grid rib section
7. Upper grid pad joint - bolt

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-28 To reconcile these component items/welds, a recalculated set of loads were developed to determine the faulted condition stresses for the six component items/welds excluding the ONS-3 upper core barrel bolts. These recalculated loads were based on smaller break loadings due to Leak-Before-Break (LBB) of the primary piping. The recalculated stress intensities are compared to the applicable unirradiated ASME Code yield strength values at 600°F in Table 2-10. The recalculated faulted condition stress intensity values are less than the unirradiated ASME Code yield strength at 600°F for the six components items/welds. Thus reduction of ductility is acceptable for these six component items and no further analysis is required.

For the ONS-3 UCB bolts the calculated Pm+Pb stress is [ ] psi. Per Table 2-1, the yield strength for Alloy A-286 at 600°F that was used for the Category 1 assessments is 81,000 psi. A limited records search for measured high temperature yield strength data for Alloy A-286 was performed. The data has been gathered and summarized in Table 2-11. This data shows that at these elevated temperatures, the measured yield strength values of these tested materials are greater than [ ]

psi, with the lowest reported being 88.6 ksi at 800.6ºF. Therefore Category 1, as described in Sections 2.3.1.1.1 and 2.3.1.2.1 can be used for assessment and the reduction of ductility is expected to be acceptable for the ONS-3 UCB bolts.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-29 Table 2-10 Select Component/Weld Recalculated Stress Analysis Results for Faulted Conditions and ASME Code Yield Strength at 600°F ASME Code Comparison of Case IV Recalculated Faulted Stress Yield Strength Faulted Stress Intensity vs Component Item Material Stress Intensity Type (Sy) at 600ºF ASME Yield Strength (Sy) at (psi)

(psi) 600ºF Pm 17,300 No further analysis is required Lower grid plate (lower grid rib section) Type 304 Pm+Pb 17,300 No further analysis is required Core support shield lower flange Type 304 Pm+Pb 17,300 No further analysis is required Upper core barrel bolts (ONS-3) Alloy A-286 N/A 81,000 Further evaluation is required Lower grid rib section/lower grid shell Grade B8 Pm 17,300 No further analysis is required forging joint bolt Support post/support forging joint - weld Type 308/308L Pm 14,400 No further analysis is required Pm 17,300 No further analysis is required Upper grid rib section Type 304 Pm+Pb 17,300 No further analysis is required Upper grid pad joint - bolt Grade B8 Bolt

- - 17,300 No further analysis is required

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-30 Table 2-11 High Temperature Alloy A-286 Yield Strength Data Material Product Measured Heat Treatment Details Strain Rate Temperature Reference Heat Form Yield Strength

-5 -1 3 x 10 sec 88.6 ksi

-5 -1 3 x 10 sec 89.3 ksi K-58139-2 Strip 3 x 10-3 sec-1 100.0 ksi Annealed at 982ºC (1800ºF) for 30 minutes -2 -1 6 x 10 sec 427ºC 105.6 ksi Precipitation heat-treated at 718ºC (1325ºF) for 16 -1 (2-16) 1.0 sec (800.6ºF) 103.4 ksi hours -5 -1 3 x 10 sec 102.4 ksi 8790 Rod 6 x 10-2 sec-1 102.6 ksi

-1 1.0 sec 101.2 ksi 90456 Solution heat treated at 1800ºF for 30 minutes, air Sheet Unknown 94.5 ksi 1140ºR cooled (2-17) 82312 Sheet Unknown (680.33ºF) 94.3 ksi Age-hardened at 1325ºF for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, air cooled Solution annealed at 982ºC (1800ºF) for 30 minutes, 427ºC 730 MPa K-58139-2 water quenched Strip Unknown (2-18)

(800.6ºF) (105.877 ksi)

Aged at 718ºC (1325ºF) for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> Solution annealed at 1800ºF for 30 minutes, water K-58139-2 quenched Strip 3 x 10-4 sec-1 96.5 ksi Aged at 1325ºF for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, air cooled 600ºF (2-19)

Solution annealed at 1800ºF for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, water

-5 -1 8790 quenched Bar 3 x 10 sec 113.5 ksi Aged at 1325ºF for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, air cooled Solution annealed at 1088ºF (982ºC) for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, oil Unknown quenched Bar Unknown 600ºF ~92 ksi (2-20)

Aged at 1325ºF for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, air cooled

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-31 2.4 Summary and Conclusions The effect of irradiation on the material properties and deformation limits of the reactor vessel internals at all three units at Oconee is acceptable for an 72 EFPY lifetime such that the internals will have adequate ductility to absorb local strain at the regions of maximum stress intensity, and that irradiation will not adversely affect deformation limits under faulted condition loadings.

2.5 References 2-1. BAW-10008, Part 1, Revision 1, Reactor Internals Stress and Deflection Due to Loss-of-Coolant Accident and Maximum Hypothetical Earthquake, June 1970.

Duke Energy Letter Incorporating BAW-10008, Part 1, Revision 1 into the Oconee Current Licensing Basis (CLB), ADAMS Accession Number ML19312B713. AEC letter to Duke regarding approved SER for BAW-10008, Part 1, Revision 1, ML19319B162.

2-2. BAW-2248A, Demonstration of the Management of Aging Effects for the Reactor Vessel Internals, March 2000, ADAMS Accession Number ML003708443 2-3. Letter from M. S. Tuckman (Duke Energy), to U. S. Nuclear Regulatory Commission Document Control Desk License Renewal, Response to NRC Letter dated November 18, 1999, Oconee Nuclear Station, Docket Numbers 50-269, 50-270, and 50-287, ADAMS Accession Number ML993620451, December 17, 1999.

2-4. NUREG-1723, Safety Evaluation Report Related to the License Renewal of Oconee Nuclear Station, Units 1, 2, and 3, Docket Numbers 50-269, 50-270, and 50-287, ADAMS Accession Number ML003695154, March 2000.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-32 2-5. Letter from T. Preston Gillespie (Duke Energy) to U. S. Nuclear Regulatory Commission Document Control Desk, Duke Energy Carolinas, LLC, Oconee Nuclear Station (ONS), Units 1, 2, and 3, Docket Numbers 50-269, 50-270, and 50-287, License Renewal Commitment to Submit a Time Limited Aging Analysis for the Reactor Vessel Internals to the NRC for Review, ADAMS Accession Number ML12053A332, February 20, 2012.

2-6. Letter from John Boska (U. S. NRC) to Preston Gillespie (Duke Energy),

Oconee Nuclear Station, Units 1, 2, and 3 - Approval of Time-Limited Aging Analysis for Reactor Vessel Internals (TAC NOS. ME8436, ME8437, ME8438),

ADAMS Accession Number ML13045A489, February 19, 2013.

2-7. ANP-3542P Revision 1 (103-3542P-001), Time-Limited Aging Analysis (TLAA)

Regarding Reactor Vessel Internals Loss of Ductility for Davis-Besse Nuclear Power Station, Unit No. 1 at 60 Years, March 2017. ADAMS Accession Number ML17026A007, Davis-Besse Nuclear Power Station Unit 1 Regarding MRP-227-A Applicant Action Item 8: Time-Limited Aging Analysis (TLAA) Regarding Reactor Vessel Internals Loss of Ductility at 60 Years 2-8. Letter from Blake Purnell (U. S. NRC) to Mark B. Bezilla (FirstEnergy), Davis-Besse Nuclear Power Station, Unit No. 1 - Time-Limited Aging Analysis for the Reactor Vessel Internals Loss of Ductility at 60 Years (CAC NO. MF9126, EPID L-2017-LRO-0002), ADAMS Accession Number ML17256B190, October 12, 2017.

2-9. BAW-1621, B&W 177-FA Owners Group - Effects of Asymmetric LOCA Loadings - Phase II Analysis, July 1980, ADAMS Accession Number ML19320B058.

2-10. BAW-1621, Supplement 1, B&W 177-FA Owners Group - Effects of Asymmetric LOCA Loadings - Phase II Analysis - Supplement 1 Responses to NRC Questions, June 1981, ADAMS Accession Number ML20009B628.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-33 2-11. Letter from John F. Stolz, NRC to Mr. H. B. Tucker, Duke Power Company, Safety Evaluation Report on Asymmetric LOCA Loads, Oconee 1, 2, and 3, NRC Legacy Accession Number 8305310083, May 20, 1983, ADAMS Accession Numbers ML15238A802 and ML15238A804 2-12. Materials Reliability Program: Development of a Material Constitutive Model for Irradiated Austenitic Stainless Steels (MRP-135, Revision 2). EPRI, Palo Alto, CA: 2019. 300213216.

2-13. Information in Support of the EPRI Materials Reliability Program (MRP):

Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-Rev. 0) Review, ADAMS Accession Number ML103090248, October 29, 2010.

2-14. NUREG/CR-7027 (ANL-10/11), Degradation of LWR Core Internal Materials due to Neutron Irradiation, ADAMS Accession Number ML102790482, December 2010.

2-15. High Strain Rate Tensile Properties of AISI Type 304 Stainless Steel, J. M.

Steichen, Journal of Engineering Materials and Technology, Volume 95, Issue 3, pp 182-185, July 1973.

2-16. J. M. Steichen, Hanford Engineering Development Laboratory Report HEDL-TME 75-81, UC-70 h, p, The Effect of Strain Rate, Thermal Aging, and Irradiation on the Tensile Properties of A-286, March 1976.

2-17. Westinghouse Astronuclear Laboratory Report WANL-TME-2791, Neutron Irradiation Effects on the Tensile Properties of Inconel 718, Waspaloy and A-286, June 1971.

2-18. W. J. Mills and L. A. James, Fracture Mechanics Behavior of Neutron Irradiated Alloy A-286, 1983.

2-19. L. A. James, Hanford Engineering Development Laboratory Report HEDL-TME 75-82, UC-79 b, h, The Effect of Temperature on the Fatigue-Crack Propagation Behavior of A-286 Steel, January 1976.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 2-34 2-20. R. F. Muraca and J. S. Whittick, Western Applied Research & Development, Prepared for National Aeronautics and Space Administration, Materials Data Handbook Stainless Steel Alloy A-286, June 1972.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 3-1 3.0 REACTOR VESSEL INTERNALS FLOW-INDUCED VIBRATION ENDURANCE LIMITS, BAW-10051 (SLRA SECTION 4.7.1.2) 3.1 Introduction BAW-10051, Revision 1 and BAW-10051A, Supplement 1 (References 3-1, 3-2) calculated stress values for the ONS reactor vessel internals for 40 years and compared them to endurance limit (stress) values. This methodology was extended from 40 years to 60 years in Section 4.5.1 of BAW-2248A (Reference 3-3) by conservatively increasing the number of cycles and determining the endurance limit using the latest ASME fatigue curves. In the SER related to the License Renewal of the Oconee stations for 60 years (NUREG-1723, pages 4-23 and 4-24) (Reference 3-4), the staff notes that the TLAAs applicable to the ONS reactor vessel internals are addressed in BAW-2248A which include:

  • Flow-induced vibration endurance limit assumptions
  • Transient cycle count assumptions for the replacement bolting
  • Reduction in fracture toughness The finding in NUREG-1723 relative to the flow-induced vibration endurance limit is as follows.

The flow-induced vibration fatigue limit assumptions were based on 1012 cycles for 40 years. The analysis was extended into the period of extended operation for license renewal by conservatively increasing the number of cycles to 1013, then determining the endurance limit using the latest ASME fatigue curves. The component stress values were found to be less than the endurance limit, rendering the evaluation acceptable, according to the requirements of 10 CFR 54.21(c)(1)(ii).

Based on the above, the NRC staff concluded that the licensees evaluation of the TLAA for flow-induced vibration endurance limit assumptions is acceptable. Therefore, this 60-year TLAA must be re-evaluated for subsequent license renewal.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 3-2 3.2 Methodology The source references that were used to justify the B&W Reactor Pressure Vessel internals flow-induced vibration endurance limit assumptions for 60 years are reviewed and updated for applicability to 80 years. The rationale and methods that are applied to SLR for 80 years are identical to those previously considered for 60 years but augmented with additional justification to conservatively address environmentally-assisted fatigue (EAF) affects utilizing the EAF criteria developed in NUREG/CR-6909, Revision 1 (Reference 3-5). In addition, the more limiting fatigue curves published in the 2013 edition of the ASME Section III code (Reference 3-6) (compared to the 1986 edition used with the ONS LRA for 60 years) (Reference 3-7) are considered with the fatigue evaluation of the ONS RV internals.

3.3 Evaluation (Update of FIV Endurance Limit TLAA for 80 years) 3.3.1 Environmentally-Assisted Fatigue Affects 3.3.1.1 RV Internals (Austenitic Steels)

The fatigue curves published in the 2013 ASME Code for austenitic steels (Figure I-9.2) do not consider environmental effects. Appendix A of NUREG/CR-6909 provides an acceptable method to determine the environmental correction factor for austenitic steels. Per Section A.2 (Equation A.13) of NUREG/CR-6909, the environmental correction factor (Fen) is equal to 1.0 if the strain amplitude () is less than 0.10% (or 28.3 ksi stress amplitude). The summary of stress resulting from FIV shows the maximum amplitude of stress is equal to [ ] ksi (Table 3-3). Therefore, [

]

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 3-3 3.3.1.2 RV Internal Bolts (High Strength Bolt Steel)

The RV internal bolts have an axial preload. Provided the structural members jointed by the bolts do not separate during the cyclic loads, the bolt preload (in conjunction with the compressive clamping force imparted to the structural members by the bolt) greatly reduces the cyclic loading imparted to the bolts that result from high cycle FIV loadings created by the response of the RV internals to random turbulence. This source of flow excitation is relatively weak and thus, separation of the mating surfaces is not possible.

As such, the cyclic strain amplitude () resulting from FIV loadings in the bolts is

[ ] than 0.10% (or less than 28.3 ksi stress amplitude), as shown in Table 3-3. Per Section A.2 (Equation A.13) of NUREG/CR-6909, the environmental correction factor (Fen) is equal to 1.0 if the strain amplitude () is less than 0.10% (or stress amplitude less than 28.3 ksi). Thus, the environmental correction factor (Fen) for the bolts is equal to [ ]. Therefore, [

]

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 3-4 3.3.2 ASME Fatigue Curves for Design Analysis (80 Years) 3.3.2.1 RV Internals The ASME fatigue curves for austenitic steel were extended from 106 cycles to 1011 cycles in the 1983 edition of the ASME Section III code (Reference 3-8). The extended fatigue curves A, B and C are depicted in Figure I-9.2.2 of the aforementioned ASME code year. The 2013 edition of the ASME Section III code eliminated fatigue curves A and B of Figure I-9.2.2 (or Figure I-9.2 in 2013 edition).

Therefore, only one fatigue curve (curve C) is provided in the 2013 edition of the ASME code. The values from curve C are illustrated in Table 3-1 of this report in the column labeled SLR for 80 years. Table 3-1 shows that the allowable alternating stress for austenitic steels is 13.6 ksi at 1011 cycles. This allowable cyclic stress is conservatively extrapolated to [ ] ksi at 1013 cycles based upon the decay rate of [ ]% decay per decade of cycles used in the previous decade (1010 to 1011 cycles).

3.3.2.2 RV Internal Bolts As shown in Table 3-2, the number of cycles and allowable stress associated with the fatigue curves for the high strength steel bolting (Figure I-9.4) published in the ASME Code has not changed over the years of interest. The largest number of cycles published in the 2013 ASME code for this material is 106. The allowable cyclic stress (Sa) for cycles greater than 106 were extrapolated to 1013 cycles based upon the decay rate of [ ] consistent with the methodology applied in BAW-10051, Revision 1 and BAW-10051A, Supplement 1 and for the ONS LRA for 60 years.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 3-5 3.3.3 Fatigue Usage Factor for RV Internals In BAW-10051, Revision 1 and BAW-10051A, Supplement 1, 1012 cycles was postulated for the 40 year plant life. For an 80 year plant life, the number of cycles to be postulated is 2.0 x1012 cycles. However, 1013 cycles is conservatively evaluated herein.

A multiplication factor of [ ] is considered for the thermal adjustment of the fatigue curve (e.g., the modulus of elasticity at temperature is approximately [ ] smaller than the one at room temperature). The environmental fatigue correction factor of [

] is also considered (Section 3.3.1). The allowable cyclic stress for 80 years of operation is determined below.

For the RV Internals:

Allowable cyclic stress = [ ] ksi at 1013 (per Table 3-1)

For the RV Internal Bolts:

Allowable cyclic stress = [ ] ksi at 1013 (per Table 3-2)

Table 3-3 provides a comparison of the alternating stress value (Sa) for the ONS RV internals and bolting components. The component with the least margin reported in this table has a stress of [ ] psi for the upper support bolts of the thermal shield and is smaller than the allowable cyclic stress of [ ] psi. Therefore, at least [ ]

margin (or a safety factor of [ ]) exists between the maximum FIV stress and the allowable cyclic stress from the fatigue curve. Since this stress is below the endurance limit ([ ] psi at 1013 cycles), the maximum fatigue usage factor associated with FIV for the RV internals is effectively equal to [ ] for an 80 year life.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 3-6 Table 3-1 ASME Fatigue Curve for Austenitic Steels (RV Internals)

Allowable Cyclic Stress: Sa(allowable)

Number of Units: psi (0-peak)

Cycles SLR for 80 years  % Decay 1

10 870.0 N/A 2

10 287.0 67.0%

3 10 108.0 62.4%

4 10 53.4 50.6%

5 10 28.4 46.8%

6 10 18.3 35.6%

7 10 14.4 21.3%

8 10 14.1 2.1%

9 10 13.9 1.4%

10 10 13.7 1.4%

11 10 13.6 0.73%

1012 [ ] [ ]

1013 [ ] [ ]

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 3-7 Table 3-2 ASME Fatigue Curve for High Strength Steel Bolts (RV Internal Bolts)

Allowable Cyclic Stress: Sa(allowable)

Number of Units: psi (0-peak)

Cycles SLR for 80 Years  % Decay 1

10 1150.0 2

10 320.0 3

10 100.0 4

N/A 10 34.0 105 19.0 6

10 13.5 10 7 -

108 109 1010 1011 1012 1013

- I

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 3-8 Table 3-3 Summary of FIV Stress Results for the SLR for 80 Years Sa RV Internal Part Sa(allowable) Sa(allowable) / Sa (BAW-10051)

Incore Instrumentation Nozzle [ ] [ ] [ ]

Incore Instrumentation Guide Tubes Cantilevered Portion below flow distributor [ ] [ ] [ ]

Between flow distributor and support plate [ ] [ ] [ ]

Between support plate and spider casting [ ] [ ] [ ]

Gusset welds [ ] [ ] [ ]

J-groove welds [ ] [ ] [ ]

Flow Distributor Ligament stresses [ ] [ ] [ ]

Support plate ledge area [ ] [ ] [ ]

Flow distributor assembly to lower grid assembly [ ] [ ] [ ]

Flow Distributor Assembly Support Plate Ligament stresses [ ] [ ] [ ]

Thermal Shield Upper support bolts [ ] [ ] [ ]

Upper support blocks [ ] [ ] [ ]

Lower support bolts [ ] [ ] [ ]

Surveillance Holder Tube [ ] [ ] [ ]

Inlet Baffle [ ] [ ] [ ]

Redesigned Surveillance Specimen Holder Tubes Bracket [ ] [ ] [ ]

Tube [ ] [ ] [ -

]

Capsule [ ] [ ] [ ]

Notes:

(1) All stresses are in units of psi (0-peak)

(2) Portions of the SSHT were removed from the RV internals of ONS-1, ONS-2, and ONS-3 in 1976 due to structural damage that resulted from FIV (Section 3.3.2 of ANP-3186P, Rev 2) (Reference 3-9). The upper portion of the SSHT assembly attached to the core support shield was left intact.

The majority of the lower portion of the SSHT assembly attached to the thermal shield was removed, leaving only brackets, pins, etc. (e.g., the tube was removed). Therefore, the FIV results for the tube part of the SSHT are not determined for the SLR for 80 years.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 3-9 3.4 Summary and Conclusions Table 3-3 provides a summary of the FIV stress associated with the Oconee RV internals and bolting items. The item with the least margin reported in this table has a stress for the thermal shield upper support bolts that is smaller than the allowable cyclic stress. Therefore, a safety factor exists between the maximum FIV stress and the allowable cyclic stress from the fatigue curve for this limiting location. Since this stress is below the endurance limit, the maximum fatigue usage factor associated with FIV for the RV internals is below the endurance limit for an 80 year life.

Therefore, the FIV integrity of the Oconee RV internals and bolting is deemed acceptable for SLR for 80 years in accordance with 10 CFR 54.21(c)(1)(ii). The high cycle fatigue affects associated with FIV during the subsequent period of extended operation (60 to 80 years) and the concerns associated with the environmental fatigue affects have no impact upon the BAW-10051, Revision 1 and BAW-10051A, Supplement 1 analyses and thus have been demonstrated not to be detrimental to the FIV performance of the Oconee RV internals and bolting.

3.5 References 3-1. BAW-10051, Revision 1, Design of Reactor Internals and Incore Instrumentation Nozzles for Flow-Induced Vibration, September 1972, revised in November 1972, Acceptability of BAW-10051, Revision 1 ADAMS Accession Number 19316A566.

3-2. BAW-10051A, Supplement 1, Structural Analysis of 177-FA Redesigned Surveillance Specimen Holder Tube, ADAMS Accession Number ML19248D133, 7908020516 (legacy) 3-3. BAW-2248A, Demonstration of the Management of Aging Effects for the Reactor Vessel Internals, March 2000, ADAMS Accession Number ML003708443

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 3-10 3-4. NUREG-1723, Safety Evaluation Report Related to the License Renewal of Oconee Nuclear Station, Units 1, 2, and 3, Docket Numbers 50-269, 50-270, and 50-287, ADAMS Accession Number ML003695154, March 2000.

3-5. NUREG/CR-6909 (Revision 1), Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials, May 2018.

3-6. 2013 ASME Boiler and Pressure Vessel Code,Section III, Appendices.

3-7. 1986 ASME Boiler and Pressure Vessel Code,Section III, Appendices.

3-8. 1983 ASME Boiler and Pressure Vessel Code,Section III, Appendices.

3-9. ANP-3186P, Revision 2, ONS Licensing Renewal Scope and MRP-189, Rev 1 Comparison, August 2013, ADAMS Accession Number ML13275A297

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 4-1 4.0 REACTOR VESSEL INTERNALS REPLACEMENT BOLTING METAL FATIGUE (SLRA SECTION 4.3.2.2) 4.1 Introduction As described in BAW-2248A (Reference 4-1), Sections 2.0 and 4.5.1, the reactor vessel internals were designed and constructed prior to the development of ASME Code requirements for core support structures. Because of the lack of specific ASME design rules for core support structures at the time of design and construction of the ONS reactor vessel internals,Section III of the ASME code was used as a guideline for the design criteria for the reactor vessel internals. Qualification of the internals was accomplished by both analytical and test methods. The only specific fatigue analyses performed in the original design were those that addressed high cycle fatigue reported in BAW-10051, Revision 1 and BAW-10051A, Supplement 1 (References 4-2, 4-3),

which is a TLAA and is addressed in Section 3.0.

Inservice inspection at the three Oconee nuclear plants during 1981 and 1982 revealed failed bolts in the joint fastening the lower end of the reactor vessel thermal shield to the lower grid assembly. These bolts were made of A-286 material. The failed bolts were replaced with X-750 bolting and fatigue analyses were performed for the replacement bolts as reported in BAW-1843PA (Reference 4-4). These topical reports summarize fatigue analyses performed to the ASME code (Section III, Subsection NG) including both high cycle fatigue (FIV) and low cycle fatigue (design transients). The fatigue evaluations for internal replacement bolting reported in BAW-1843PA are TLAAs that were evaluated in BAW-2248A for 60-years and must be addressed for subsequent license renewal.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 4-2 4.2 Methodology The total usage factor was recalculated for the replacement bolting items by considering both the existing design transient usage factors, which are acceptable for 80-years, and by the recalculation of fatigue usage factors due to flow-induced vibration. The current CUFs of replacement bolting items are acceptable since 40-year design cycles (CLB cycles) are postulated to bound 80 years of plant operation. The methods that are applied to SLR for 80 years are identical to those previously considered for 60 years considering the fatigue analysis presented in BAW-1843PA for the replacement bolting items.

4.3 Evaluation (Update of Metal fatigue CUF for 80 years) 4.3.1 Environmentally-Assisted Fatigue Effects As discussed in Section 3.3.1.2, the environmental correction factor (Fen) is equal to 1.0 if the strain amplitude () is less than 0.10% (or stress amplitude less than 28.3 ksi).

The cyclic stress in the replacement bolts resulting from FIV is conservatively calculated to be 1.07 ksi (Section 4.3.3), which is less than 28.3 ksi. Thus, the environmental correction factor (Fen) for the replacement bolts is equal to [ ] Therefore, [

]

4.3.2 ASME Fatigue Curves for Design Analysis The fatigue curve for the RV internal replacement bolts shown in Table 4-1 has not been modified since its original determination. Therefore, reconciliation of differences associated with the number of cycles and alternating stress is not necessary for this fatigue curve as has been performed for the fatigue curves of the original bolts in Section 3.3.2.2.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 4-3 4.3.3 Fatigue Usage Factor for the Replacement Bolts The fatigue analysis calculation below for 80 years adheres to the analytical methodology applied for 40 years and 60 years and follows Crandalls method (Reference 4-5) for a random stress process.

The usage factor for low cycle fatigue is [ ]. This value is the same for 40 years since the RCS transients are unchanged from a 40 year life.

The allowable FIV usage factor is therefore 1.0 - UFlow cycle = [ ].

The number of FIV cycles at [ ] for 80 years is [ ].

Where [ ] is based on a modal analysis of the thermal shield.

The design fatigue curve for Inconel X-750 bolt material is reported in Table 4-1 and can be mathematically defined by the following expression:

(Ncycles)(Sa)b = c where:

b = [ ] (Constant for X-750 fatigue curve for cycles between 109 and 1011) c = [ ] (Constant for X-750 fatigue curve for cycles between 109 and 1011)

Sa = Allowable alternating / cyclic stress for a given number of cycles.

The usage factor for a random stress process when determined using Crandalls method can be expressed as:

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 4-4 UF = N:de, (2

  • sJ(r((l + ~))

log(UF) = log(N""" )- Iog(c)+ b

  • log(2}+ b
  • log(S.)+ log(r(1 + ~))

Solving for the log(SJ tenn yields; Substituting the variables into the above equation and considering r= [ ]

yields:

Sa = [ ] rms for 80 years The stress due to FIV (SFIV) for the replacement bolts = [ ] rms.

The margin is thus Sa / SFIV = [ ].

UFFIV = [ ] for 80 years UFcumulative = UFFIV+UFlow_cycle= [ ]

Since the FIV stress for the Oconee replacement lower thermal shield to lower grid assembly bolts is less than the 80 year allowable FIV stress, and the 80-year CUF that includes consideration of FIV is less than 1.0, these bolts will be acceptable for the period of SLR.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 4-5 Table 4-1 Fatigue Curve for RV Internal Replacement Bolts (Inconel X-750 Material)

Number of Allowable Cyclic Stress: Sa(allowable)

Cycles Units: ksi (0-peak) 4 x 103 104 2 x 104 4 x 104 105 2 x 105 4 x 105 106 2 x 106 4 x 106 107 108 109 1010 1011

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 4-6 4.4 Summary and Conclusions The stress for the replacement bolts (lower thermal shield to lower grid assembly) is smaller than the allowable cyclic stress. As such, a safety factor exists between the maximum FIV stress and the allowable cyclic stress from the fatigue curve. Therefore, the FIV integrity of the Oconee RV replacement bolting is deemed acceptable for SLR for 80 years.

4.5 References 4-1. BAW-2248A, Demonstration of the Management of Aging Effects for the Reactor Vessel Internals, March 2000, ADAMS Accession Number ML003708443.

4-2. BAW-10051, Revision 1, Design of Reactor Internals and Incore Instrumentation Nozzles for Flow-Induced Vibration, September 1972, revised in November 1972, Acceptability of BAW-10051, Revision 1, ADAMS Accession Number 19316A566 4-3. BAW-10051A, Supplement 1, Structural Analysis of 177-Fuel Assembly Redesigned Surveillance Specimen Holder Tube," ADAMS Accession Number ML19248D133, 7908020516 (legacy) 4-4. BAW-1843PA, Evaluation of Internals Bolting Concerns in 177 FA Plants, January 1986, ADAMS Accession Number ML20197G640.

4-5. S. H. Crandall and W. D. Mark, Random Vibration in Mechanical Systems, 1963.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 5-1 5.0 REACTOR VESSEL INTERNALS FLUENCE/DPA AT 72 EFPY (SLRA SECTION 4.7.1.3 AND XI.M16A) 5.1 Introduction In accordance with Section 2.0 above, 72 EFPY Oconee-specific fluence/dpa estimates were used to assess susceptibility to irradiation embrittlement relative to the BAW-10008, Part 1, Revision 1, TLAA evaluation for the following reactor vessel internals component items.

  • Baffle plates (Category 2 item)
  • Plenum cover (Category 3 item)
  • Plenum cylinder reinforcing plate (Category 3 item)
  • Core support shield top flange (Category 3 item)

The 72 EFPY fluence/dpa values used to evaluate these items are reported in Table 5-1. The fluence/dpa values used for the Oconee gap analysis (GALL SLR XI.M16A), which is based on MRP-189 Revision 3 and was developed to bound the B&W fleet of reactors, is reported in Table 5-2. Oconee-specific DORT and MCNP 72 EFPY evaluations reported herein were developed to demonstrate that the fluence/dpa input to the TLAA evaluation reported in Section 2.0 above and the fluence/dpa input to MRP-189 Revision 3 used to develop the Oconee-specific gap analysis are acceptable for the subsequent period of extended operation.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 5-2 5.2 Methodology Framatomes Oconee-specific evaluation of the RVI included both deterministic discrete ordinate transport (DORT) and Monte Carlo N-Particle (MCNP) codes. The purpose of these calculations is to verify that the projected 72 EFPY fluence values developed for the update of BAW-10008, Part 1, Revision 1 (Section 2) and the fluence/dpa input to MRP-189 Revision 3 used to develop the Oconee-specific gap analysis are acceptable.

The DORT evaluation followed the modeling procedures defined in Framatomes NRC approved methodology described in BAW-2241PA, Revision 2, Fluence and Uncertainty Methodologies. There were no additional modeling considerations to address the inherent limitations of the DORT code. Instead, an additional analysis was performed using Framatomes hybrid 3D Deterministic-Monte Carlo methodology, called SVAM (Reference 5-5). The hybrid model include full three dimensional detail of the reactor internals components of interest. [

]

5.3 Oconee Specific 72 EFPY Reactor Internals Fluence/dpa This section lists the key inputs and modeling approximations in the Oconee-specific DORT and MCNP RVI fluence calculations. Both analyses followed a best-estimate, rather than a bounding or conservative fluence methodology. Oconee-specific values were used for all of the key inputs, including the power level, axial and radial power distributions, materials, geometry definition, and operating history.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 5-3

[

]

5.3.1 DORT 2-D Calculation of RVI Neutron Exposure to 72 EFPY The Oconee-specific DORT calculation utilized Framatomes NRC approved fluence methodology. To determine the three-dimensional (3D) fluence rate (flux) and fluence, two separate 2-dimensional (2D) models are combined in accordance with the methodology presented in BAW-2241PA. In a manner consistent with previous fluence analyses, the Oconee reactor is modeled in 2D planar cylindrical coordinates (R) and 2D axial cylindrical coordinates (RZ).

[

]

[

]

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 5-4 The synthesized fluence rates provide the basis for calculating the fluence and dpa results based on appropriate response functions. For fluence, the response function includes only fluence rates with energies > 1.0 MeV. However, the dpa response function extends to the thermal region. It is assumed that the fluence impact of a 1.64%

MUR will be bounded by a 2% increase in fluence rate. This is accounted for by multiplying the resultant Cycle 26 flux and dpa rate values by 1.02.

A cross sectional view of the DORT model is provided in Figure 5-1. [

]

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 5-5 Figure 5-1 Sketch of the Oconee RVI DORT model

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 5-6 5.3.2 MCNP Calculation of RVI Neutron Exposure to 72 EFPY The Oconee-specific MCNP model was developed to address known deficiencies in DORT RVI fluence calculations. For example,

1. The reactor internals are homogenized in the DORT model. A study performed by Oak Ridge National Lab found that RVI homogenization leads to a systemic under prediction of the fast flux, particularly at locations above and below the core.
2. The neutron flux spectrum used to collapse the iron isotopes in steel in the BUGLE-96 library is based on a fine-group (VITAMIN-B6) solution at 1/4 of the RV thickness, effectively at the core midplane. The spectrum in the internals at locations above and below the core is likely to be different. Therefore, the BUGLE-96 cross section library may not be ideal for flux calculations outside of the beltline.
3. The BUGLE-96 cross section library only has 4 energy groups in the thermal range. This resolution may is not sufficient for dpa calculations. Particularly in locations far away from the core.
4. Symmetric quadratures may not be accurate in regions above and below the core. Higher order or biased quadrature sets may be more appropriate.

The Oconee MCNP model is three dimensional (3-D), with a heterogeneous representation of the reactor core and internals components. The inputs were selected to address the known deficiencies in the DORT model. [

]

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 5-7 Figure 5-2 through Figure 5-4 provide an example of the detail included in the MCNP model. [

]

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 5-8 Figure 5-2 View of Lower Grid Assembly and Core Barrel

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 5-9 Figure 5-3 View of Upper Vessel Base Metal

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 5-10 Figure 5-4 View of Incore Guide Support Plate

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 5-11 Although the MCNP model contains extremely precise representations of the reactor core and internals, model precision does not guarantee accurate results. Accuracy is a measure of how close the MCNP estimate of flux is to the truth. The difference between this true value and MCNPs flux estimate is called the systematic error (or bias), which is seldom known. Error or uncertainty estimates for the results of Monte Carlo calculations refer only to the precision of the result and not to the accuracy. It is possible to calculate a highly precise result that is far from the physical truth because nature has not been modeled faithfully. For example, the exact coolant temperature is not known at all locations. Therefore, the MCNP model assumes that the temperature is constant over large regions of the RVI. Additionally, no reactor component is manufactured with perfect symmetry. However, the MCNP model assumes perfect symmetry of the reactor vessel and all internals component items. Typically, the error associated with modeling approximations is quantified by comparing the calculations to measurements that have a well-defined uncertainty. At the time this document was prepared, uncertainty and sensitivity calculations have not been performed for the MCNP (or the DORT) RVI fluence methodology.

5.3.3 Estimate of Uncertainty The uncertainty for the RV internals is currently being evaluated through the PWROG (PA-MSC-1403R1). It is expected that consideration of uncertainty in the RV internals fluence will not impact the RV internals TLAA (BAW-10008, Part 1, Revision 1) or the gap analysis (XI.M16A). This is because there is a large margin between the fluence values used to support the ONS RV internals (TLAA BAW-10008, Part 1, Revision 1 and the XI.M16A gap analysis) and the ONS best estimate fluence values (DORT and MCNP). This margin is discussed in Sections 5.4 and 5.5.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 5-12 5.4 Oconee TLAA (BAW-10008, Part 1, Revision 1) Fluence Margin The fluence values used to support the TLAA in Section 2.0 were intended to be

[ ] These values have significant margin relative to the Oconee-specific best estimate fluences calculated for 80 years (72 EFPY) by the DORT and MCNP methodologies, which are discussed above in Sections 5.3.1 and 5.3.2, respectively.

The margin between the Oconee-specific best estimate fluence values and the values used to support the TLAA in Section 2.0 are shown in Table 5-3. The ONS 72 EFPY values reported in Table 5-3 are the maximum considering the DORT and MCNP best estimate values. The margin (or the allowed percent difference) is very large far from the core and is lower close to the core. This is because the uncertainty in the best estimate fluence values is expected to increase with increasing distance from the core.

Therefore, more conservative fluence values were used in the TLAA for component items farther from the core. For component items like the baffle plates and the bottom of the plenum cylinder that are closer to the core, the uncertainty in the Oconee-specific best estimate fluence values is expected to be low, so the margins allowed for these component items should remain adequate.

5.5 Oconee Gap Analysis (XI.M16A) Fluence Margin The fluence values used to support the gap analysis (XI.M16A) were intended to be

[ ] These values have significant margin relative to the Oconee-specific best estimate fluences calculated for 80 years (72 EFPY Table 5-2) by the DORT and MCNP methodologies, which are discussed above in Sections 5.3.1 and 5.3.2, respectively.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 5-13 The margin between the Oconee-specific best estimate fluence values and the values used to support the gap analysis (GALL SLR XI.M16A) are shown in Table 5-4. Note that, since there are hundreds of component items in the RV internals, only the lowest margin component item for each assembly are included in the table. The margin (or the allowed percent difference) is very large far from the core and is lower close to the core.

This is because the uncertainty in the best estimate fluence values is expected to increase with increasing distance from the core. Therefore, more conservative fluence values were used in the gap analysis for component items farther from the core. For component items close to the core, the uncertainty in the Oconee-specific best estimate fluence values is expected to be low, so the margins allowed for these component items should remain adequate. In addition, while exceeding the screening criteria may change the preliminary categorization, it may not actually impact the inspection recommendations in the gap analysis (GALL SLR XI.M16A).

5.6 Summary and Conclusions The 72 EFPY fluence/dpa is reported in Table 5-1. The fluence/dpa used for the Oconee gap analysis (GALL SLR XI.M16A), which is based on MRP-189 Revision 3 and was developed to bound the B&W fleet of reactors, is reported in Table 5-2.

Oconee-specific DORT and MCNP 72 EFPY evaluations reported herein demonstrate that the fluence/dpa input to the TLAA evaluation reported in Chapter 2 above and the fluence/dpa input to MRP-189 Revision 3 used to develop the Oconee-specific gap analysis are acceptable for the subsequent period of extended operation. Reactor vessel internals fluence shall be monitored for the Baffle plates (Category 2 item),

Plenum cover (Category 3 item), Plenum cylinder reinforcing plate (Category 3 item),

and Core support shield top flange (Category 3 item) in accordance NUREG-2191, X.M2, Neutron Fluence Monitoring to ensure that the TLAA reported in Section 2.0 remains valid.

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 5-14 Table 5-1 Projected Fast Fluence (E > 1.0 MeV) of Reactor Vessel Internals Component Items in Support of TLAA in Section 2.0

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 5-15 Table 5-2 Projected 72 EFPY, 80 EFPY Fluence and dpa, MRP-189 Revision 3, used to Support the Gap Analysis

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 5-16 Table 5-3 Margin between Projected Fast Fluence (E > 1.0 MeV) of Reactor Vessel Internals Component Items in Support of TLAA in Section 2.0 and the ONS-Specific Best-Estimate 72 EFPY Fluence Values

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 5-17 Table 5-4 Margin between Projected Fast Fluence (E > 1.0 MeV) of Reactor Vessel Internals Component Items in Support of the Gap Analysis and the ONS-Specific Best-Estimate 72 EFPY Fluence Values

Framatome Inc. ANP-3899NP Revision 0 Framatome Reactor Vessel Internals TLAA Input to the ONS SLRA Page 5-18 5.7 References 5-1. BAW-10008, Part 1, Revision 1, Reactor Internals Stress and Deflection Due to Loss-of-Coolant Accident and Maximum Hypothetical Earthquake, Duke Energy Letter Incorporating BAW-10008, Part 1, Revision 1 into the Oconee Current Licensing Basis (CLB), ADAMS Accession Number ML19312B713. AEC letter to Duke regarding approved SER for BAW-10008, Part 1, Revision 1, ML19319B162.

5-2. BAW-2241PA, Revision 2, Fluence and Uncertainty Methodologies, ADAMS Accession Number ML073310655 (Proprietary) / ML073310660 (Non-Proprietary) 5-3. Materials Reliability Program: Screening, Categorization, and Ranking of Babcock & Wilcox-Designed Pressurized Water Reactor Internals Component Items and Welds (MRP-189, Revision 3). EPRI, Palo Alto, CA: 2019.

3002013218. ADAMS Accession Number ML20091K286, February 2020.

5-4. Materials Reliability Program: PWR (Pressurized Water Reactor) Internals Material Aging Degradation Mechanism Screening and Threshold Values (MRP-175, Revision 1). EPRI, Palo Alto, CA: 2017. 3002010268. ADAMS Accession Number ML17361A187, December 28, 2017.

5-5. ANP-10348P, Fluence Methodologies in SLR, 2020, ADAMS Accession Number ML20223A019.

ENCLOSURE 4 ATTACHMENT 3 OCONEE NUCLEAR STATION SLR-ONS-TLAA-0306NP, Revision 0, Environmentally-Assisted Fatigue Oconee Subsequent License Renewal Application Supplemental Section 4.3.4

DUKE ENERGY OCONEE NUCLEAR STATION UNITS 1, 2, and 3 SLR-ONS-TLAA-0306NP Environmentally-Assisted Fatigue Oconee Subsequent License Renewal Application Supplemental Section 4.3.4 Revision 0

[NON-PROPRIETARY]

NOTE: Text that is within brackets is proprietary to Duke Energy or Framatome.

The subscripts D or F respectively are used to identify the appropriate company.

LIST OF ABBREVIATIONS ABBREVIATION DEFINITION ANSI American National Standard Institute AOR Analysis of Record ASME American Society of Mechanical Engineers CFR Code of Federal Regulations CLB Current Licensing Basis CUF Cumulative Usage Factor DO Dissolved Oxygen EAF Environmentally-Assisted Fatigue EPRI Electric Power Research Institute FAD Functional Area Document IPA Integrated Plant Assessment NUREG U.S Nuclear Regulatory Commission technical report designation NUREG/CR Contractor-prepared NUREG OSC Oconee Calculation PWR Pressurized Water Reactor PZR Pressurizer RC Reactor Coolant RCP Reactor Coolant Pump RCS Reactor Coolant System ROTSG Replacement Once-Through Steam Generator RV Reactor Vessel SLR Subsequent License Renewal TLAA Time Limited Aging Analysis UEN Usage considering Environmental Effects UFSAR Updated Final Safety Analysis Report USAS United States of America Standards Page 1 of 13

INTRODUCTION The purpose of this document is to provide supplemental information to the non-proprietary information reported in the subsequent license renewal application for Oconee Nuclear Station Units 1, 2, and 3 for environmentally-assisted fatigue (EAF). Table 4.1-2 of NUREG-2192, Standard Review Plan of Subsequent License Renewal Applications for Nuclear Power Plants, provides the generic time-limited aging analyses (TLAA) and metal fatigue (Section 4.3) contains environmentally-assisted fatigue as a subsection. This report includes information that will assist the NRC review of environmentally-assisted fatigue, but could not be included in the Oconee subsequent license renewal application, which is non-proprietary.

NOTE: The following section supplements Section 4.3.4 of the Oconee Subsequent License Renewal Application.

ENVIRONMENTALLY-ASSISTED FATIGUE (EAF)

TLAA

Description:

As outlined in Section X.M1 of NUREG-2191 (Reference 7) and Section 4.3 of NUREG-2192 (Reference 8) the effects of the reactor water environment on cumulative usage factor (CUF) must be examined for a set of sample critical components for the plant. This sample set includes the locations identified in NUREG/CR-6260, Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components (Reference 3) and additional plant-specific component locations in the reactor coolant pressure boundary if they may be more limiting than those considered in NUREG/CR-6260. These additional limiting locations are identified through an environmental fatigue screening evaluation. The environmentally-assisted fatigue (EAF) screening process made use of existing fatigue usage values for the ASME Code,Section III components. The EAF screening evaluation reviewed the CLB fatigue evaluations for all ASME Code,Section III reactor coolant pressure boundary components and piping and ANSI B31.7 piping, including the NUREG/CR-6260 locations, to determine the lead EAF indicator (also referred to as Sentinel) locations for EAF. As a result of the EAF screening evaluation, there were other locations found that could potentially be more limiting than the NUREG/CR-6260 locations.

TLAA Evaluation:

To support subsequent license renewal, calculations were prepared to document the evaluations of environmentally-assisted fatigue for ASME Code,Section III pressure boundary components and piping and ANSI B31.7 piping that contact the reactor coolant and determine fatigue-sensitive locations for comparison and ranking. These evaluations are for subsequent license renewal purposes and do not amend the existing design reports. Discussion of the screening approach used for the ASME Code,Section III components and piping and ANSI B31.7 piping are provided below. A consolidated tabulation for pressure boundary components and piping is presented for the Sentinel locations in Table 4.3.4-1.

EAF Screening Methodology As background, for initial License Renewal, EAF was evaluated only for the NUREG/CR-6260 locations applicable to ONS. To ensure that any additional plant-specific component locations in the reactor coolant pressure boundary that may be more limiting than those considered in NUREG/CR-6260 are addressed, EAF screening was performed for all ASME Code,Section III components with existing fatigue usage values, including B31.7 piping which was reanalyzed Page 2 of 13

using ASME Code Section III during the steam generator replacement project. These were reviewed and categorized into the following five groups of major components listed below.

6. Reactor Vessel (RV) and sub-components
7. Reactor Coolant Pumps (RCPs) and sub-components
8. Steam Generators (ROTSGs) - primary side
9. Pressurizer (PZR) and sub-components
10. RCS Loop and Branch Piping Screening Fen factors were developed for each component so that EAF usage (Uen) can be calculated and compared. NUREG/CR-6583 (ANL-97/18), Effects of LWR Coolant Environments on Fatigue Design Curves of Carbon and Low-Alloy Steels [Reference 9] was used for carbon and low-alloy steels. NUREG/CR-5704, Effects of LWR Coolant Environments on Fatigue Design Curves of Austenitic Stainless Steels, [Reference 10] was used for stainless steels and NUREG/CR-6909, Revision 0, Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials, Final Report [Reference 11] was used for Ni-Cr-Fe alloys.

Conservative values are chosen for each of the Fen input parameters: sulfur content, service temperature, strain rate, and dissolved oxygen (DO). EPRI Technical Report, Pressurized Water Reactor Primary Water Chemistry Guidelines [Reference 12] is followed. A value of 0.005 ppm for dissolved oxygen (DO) content when the system temperatures are elevated is the action level 1 value for pressurized water reactors (PWR) environmental conditions under normal operation. Therefore, the Fen calculations use a value below the lowest threshold value for the dissolved oxygen (DO) content. For NUREG/CR-6583 and NUREG/CR-5704 this value is 0.05 ppm and for NUREG/CR-6909 this value is 0.04 ppm. Sample dissolved oxygen (DO) information from Oconee plant operation provides a maximum DO less than 10 ppb (0.010 ppm) during heatups and cooldowns.

For the periods during heatup/cooldown operations when DO may be elevated, pressurizer temperature is 250°F (121°C). During these times, the system T between the pressurizer and RCS is low ( 80°F), and pressurizer insurge/outsurge transient events with this magnitude of system T are not significant contributors to fatigue. For pressurizer temperatures above 250°F, oxygen scavenger addition is used to control dissolved oxygen. For cooldowns, H2 concentration in the RCS is used to minimize dissolved oxygen in the RCS through the pressurizer steam bubble collapse. Furthermore, fluid temperatures during the elevated DO times are in the range where service temperature values are the lowest. Therefore, the use of 0.05 ppm and 0.04 ppm for DO content is acceptable for the Fen evaluations.

Any location that was not part of the Class 1 reactor coolant boundary was removed from consideration. Other locations were also excluded during this step. These locations included locations not in contact with primary coolant, locations excluded fatigue usage factor calculation based on fatigue waivers (ONS SLRA, Section 4.3.2.9) and locations where the CUF was so low that applying the maximum possible Fen factor results in a Uen less than 0.8.

For the components and subcomponents where the NUREG/CR-6260 locations have the highest screening Uen no additional locations were considered for EAF. For those components and subcomponents where the NUREG/CR-6260 locations do not have the highest screening Page 3 of 13

Uen or a NUREG/CR-6260 location does not exist, the locations within the components and subcomponents that have the highest screening Uen are the Sentinel locations. The final set of sentinel locations are meant to supplement those identified in NUREG/CR-6260, resulting in a comprehensive list of plant-specific sentinel locations for EAF consideration.

The screening process for the sentinel locations were identified as follows:

  • Prescreening was performed where a location was eliminated if the location has a usage factor that is so low that, even if the maximum possible environmental fatigue correction factor (Fen) were applied, the resulting EAF usage (Uen) would be less than 0.8. The location is not exposed to reactor water, and therefore EAF is not applicable. The location is not part of the primary pressure boundary and is therefore excluded from the EPRI screening process.
  • Piping and/or vessel components that undergo essentially the same thermal and pressure transients during plant operations are grouped into thermal zones. Within each material type, the location with the highest Uen* (estimated Uen) is selected during initial screening; the second location with the second highest Uen* is also selected if the top two Uen* values are within a factor of two. However, if the third highest Uen* value is within 25% of the highest Uen* value within a thermal zone, then the top three locations in that thermal zone are selected.
  • To reduce excess conservatism for stainless steel location due to the very large maximum Fen, an estimated Fen (Fen*) is calculated as the average of the value based on a qualitative estimate of strain rate, and the value based on the worst possible strain rate, using the same values of DO and estimated upper bound temperature for design transients in both cases.
  • Additional screening was performed to determine that, within any system, a location can be considered bounded if all of the following criteria are true:

o The Fen for the bounded location is smaller than that for the bounding location.

o The U for the bounded location is smaller than that for the bounding location.

o The Uen for the bounded location is less than half of that for the bounding location.

o The bounded location is not a NUREG/CR-6260 location.

Initial Oconee screening was completed in 2014 using NUREG/CR-6909, Revision 0. This work was done as part of first steps in expanding the initial LR NUREG/CR-6260 work. In 2019 for SLR, the Fen values calculated as part of the initial screening were compared to the calculation methods proposed in Revision 1 of NUREG/CR-6909. It was concluded that revising the screening methodology would result in a similar ranking. Furthermore, conservatism added as part of the initial screening for Fen values for stainless steel were removed as part of the analysis for 80-years of operation.

Assessment of Sentinel Locations For locations where the screening Uen exceeded 1.0, further evaluations were performed using NUREG/CR-6909, Revision 1, Effect of LWR Water Environments on the Fatigue Life of Page 4 of 13

Materials, Final Report [Reference 11] was used. This approach was applied to the following components:

  • Bottom Mounted Instrumentation (BMI) Weld
  • Core Flood Nozzles
  • Loop Drain 2B2/3B2 piping
  • Pressurizer Surge Line Piping
  • Pressurizer Spray Piping, pt. 190
  • HPI Piping, Stop Vlv-to-Check (bounds HPI Nozzles)
  • Decay Heat Drop Lines
  • Core Flood Piping Sentinel locations are those locations chosen for more detailed analysis, monitoring, inspection, or replacement. These locations are chosen to have bounding Uen values compared with other locations. Except for the HPI Nozzles where the HPI piping bounded the nozzle, NUREG/CR-6260 locations were retained, regardless of the revised calculated Uen. The results of the EAF calculations for the sentinel locations are summarized in Table 4.3.4-1. In addition to the Uen values, the original analysis of record (AOR) CUF, the reduced CUF developed for SLR, and a brief summary of the conservatisms removed from the AOR for Safety Class 1 components are also provided in Table 4.3.4-1. Components with a Uen less than unity do not require fatigue management per GALL-SLR. Transients associated with components with a Uen less than unity will be monitored by the Fatigue Monitoring AMP (ONS SLRA, Appendix B3.1).

Appendix L Flaw Tolerance Evaluations For two of the sentinel piping locations (Pressurizer Surge Line and HPI piping), the effects of fatigue will be managed by application of the Inservice Inspection Program (ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program (ONS SLRA, Appendix B2.1.1))

during the subsequent period of extended operation based on results of flaw tolerance evaluations conducted per the guidance of ASME Code,Section XI [Reference 13],

Nonmandatory, Appendix L. NUREG-2192 permits inspections as a management method for fatigue as long as a flaw tolerance evaluation is performed to determine the acceptable time between inspections. The ASME Code,Section XI, Appendix L crack growth evaluation is used in conjunction with calculated allowable flaw sizes to determine the required inspection interval for a postulated flaw in the piping at the bounding location. For a postulated initial flaw, crack growth is simulated until the flaw has reached the allowable flaw depth or the end of the subsequent period of extended operation, whichever comes first.

The purpose of the flaw tolerance evaluation is to establish an appropriate inspection frequency that is consistent with the typical 10-year inservice inspection program. The input used from each of the piping systems (geometry, transient cycles and definitions, material properties, piping loads, etc.) bounds Units 1, 2 and 3. The ASME Code,Section XI, Appendix L flaw tolerance evaluation consists of postulating a hypothetical inside surface axial and Page 5 of 13

circumferential flaw. There are two Sentinel locations analyzed using ASME Code,Section XI, Appendix L presented in Table 4.3.4-2.

For the branch line piping, the transients used to simulate growth of the postulated flaws in the ASME Code,Section XI, Appendix L evaluations used ten years of projected cycles. For the pressurizer surge line elbow, as discussed in Section 4.3.1, a reduced set of allowable transient cycles were used for the ten years of projected cycles. Following re-inspection, the cycle counts used in the ASME Code,Section XI, Appendix L evaluation are set to zero when no flaw is disclosed for each new inspection interval. The ASME Code,Section XI, Appendix L evaluations used ten years of projected cycles and the branch line piping will be inspected on a ten-year inspection frequency. The selection of transients is conservative, thereby ensuring the inspection frequencies remain adequate. The Fatigue Monitoring AMP (ONS SLRA, Appendix B3.1) tracks significant transient cycles for the Safety Class 1 components. The Corrective Action Program tracks specific activities such as HPI injection and insurge and outsurge cycles pertaining to the surge line.

The maximum allowable end-of-evaluation period flaw size was determined based on the acceptance criteria and evaluation procedures in ASME Code,Section XI, Appendix C, 2007 Edition with 2008 Addenda which is the current code of record. Based on previous inspection records, there are no detected indications at these locations; therefore, the methodology of ASME Code,Section XI, Appendix L can be used. As per ASME Code,Section XI, Appendix L, a postulated flaw size larger than the ASME Code,Section XI acceptance standards in Table IWB-3514-2 was used in the fatigue crack growth (FCG) analysis. The piping systems evaluated here are constructed from stainless steel material, where the only significant crack growth mechanism of consideration is fatigue crack growth.

Based on the fatigue crack growth evaluation, the allowable operating period was determined as the length of time it takes for the postulated initial flaw size to grow to the maximum allowable end-of-evaluation period flaw size. The fatigue crack growth analysis was completed using the crack growth rates from ASME Code,Section XI, Code Case N-809, Reference Fatigue Crack Growth Rate Curves for Austenitic Stainless Steels in Pressurized Water Reactor Environments [Reference 14]. The results of the ASME Code,Section XI, Appendix L evaluations are provided in Table 4.3.4-2.

The ASME Code,Section XI, Appendix L inspections will be conducted by the Inservice Inspection (ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program (ONS SLRA, Appendix B2.1.1)) program. The ASME Code,Section XI, Appendix L inspections are identified in Table 4.3.4-3. Each weld in the inspection population will be volumetrically inspected using the code required techniques once prior to establishing the Inspection Interval schedule for Units 1, 2, and 3 ASME Code,Section XI, Appendix L locations. Going forward after the first ultrasonic inspection, one weld in each of the piping lines will be ultrasonically inspected every 10 years.

Loop Drain Piping As shown in Table 4.3.4-1, the loop drain 2B2/3B2 piping contains a Uen that is less than unity.

The alternating stress associated with the MRP-146 loading are low stress high cycle fatigue and do not result in an EAF penalty factor provided the alternating is at or below the threshold of 28.3 ksi. This is discussed in MRP-146, Revision 2, Appendix F and is defined in NUREG/CR-6909, Revision 1. In Appendix F of MRP-146, a sample B&W 2.5 drain line was evaluated and Page 6 of 13

shown to have an alternating stress of 22.37 ksi (< 28.3 ksi). Therefore, there is no EAF penalty expected for the drain lines.

Oconee currently has these drain lines scoped in for MRP-146 and will continue inspection under the MRP-146 management program for the subsequent period of operation. These inspections are currently part of the Augmented Inservice Inspection (AISI) program.

TLAA Disposition: 10 CFR 54.21(c)(1)(iii)

The effects of fatigue on the intended function(s) of ASME Code,Section III pressure boundary components and piping and ANSI B31.7 piping that contact reactor coolant will be adequately managed by the Fatigue Monitoring aging management program (ONS SLRA, Appendix B3.1) and the ASME Section XI Inservice Inspection, Subsection IWB, IWC, and IWD aging management program (ONS SLRA, Appendix B2.1.1) through the subsequent period of extended operation.

Page 7 of 13

Table 4.3.4-1 Sentinel Locations System / AOR4 Fatigue SLR Thermal Location Material Fen Uen Analysis Method Management CUF CUF Zone Method The CRDM weld is part of the RVCH replacement. The SLR CUF of [ ]D:a,b,c is based on reduced Power Loading/Unloading cycles.

Ni-Cr-Fe CRDM weld(1) [ ]D:a,b,c [ ]D:a,b,c [ ]D:a,b,c 0.772 Power loading/unloading cycles None Required(5)

Alloy is excluded from the Fatigue Monitoring program, which will require reconsideration if ONS implements flexible power operation.

Recalculated CUFin-air using appropriate stress ranges and Bottom Mounted Ni-CR-Fe number of operating transient

[ ]F [ ]F [ ]F 0.744 None Required(5)

Instrumentation (BMI)(1) Alloy cycles using NUREG/CR-6909 RC / RC fatigue curves and utilizing transients maximum Fen.

only Rx Vessel Shell Lower Used the bounding CUFin-air and LAS [ ]F [ ]F [ ]F 0.756 None Required(5)

Head(2) applied the maximum Fen.

Recalculated CUFin-air using appropriate stress ranges and Rx Vessel Inlet and Outlet number of operating transient LAS [ ]F [ ]F [ ]F 0.832 None Required(5)

Nozzles(2) cycles using NUREG/CR-6909 fatigue curves and utilizing maximum Fen Recalculated CUFin-air using appropriate stress ranges and number of operating transient Core Flood Nozzles(2) LAS [ ]F [ ]F [ ]F 0.882 None Required(5) cycles using NUREG/CR-6909 fatigue curves and utilizing maximum Fen Page 8 of 13

System / AOR4 Fatigue SLR Thermal Location Material Fen Uen Analysis Method Management CUF CUF Zone Method Usage is due to thermal ASME Section XI stratification caused by turbulent Inservice penetration, which has an Inspection, Loop drain 2B2/3B2 piping(3) SS [ ]D:a,b,c [ ]D:a,b,c [ ]D:a,b,c 0.51 alternating stress below the Subsection IWB, threshold value of 28.3 ksi giving IWC, and IWD a Fen = [ ]D:a,b,c program (B2.1.1)

Used service temperature of 550°F to calculate Fen with AOR Pressurizer heater SS [ ]F [ ]D:a,b,c [ ]D:a,b,c 0.723 alternating stress of 68.83 ksi to None Required(5) penetration weld calculate usage from NUREG/CR-6909 fatigue curve.

ASME Section XI Inservice Pressurizer surge line Inspection, SS [ ]D:a,b,c [ ]D:a,b,c [ ]D:a,b,c 1.28 Appendix L piping(2) Subsection IWB, IWC, and IWD RC / PZR program (B2.1.1) lower head and surge Refined evaluation to develop line updated alternating stresses and Pressurizer Hot Leg surge apply a Fen of 1 to load pairs nozzle weld overlay (path SS [ ]D:a,b,c [ ]D:a,b,c [ ]D:a,b,c 0.702 None Required(5) where alternating stress is below 17) the threshold value of 28.3 ksi.

Max Fen = [ ]D:a,b,c Refined evaluation to develop updated alternating stresses and Pressurizer surge nozzle apply a Fen of 1 to load pairs SS [ ]D:a,b,c [ ]D:a,b,c [ ]D:a,b,c 0.979 None Required(5) weld overlay (Path 3) where alternating stress is below the threshold value of 28.3 ksi.

Max Fen = [ ]D:a,b,c Used service temperature of 557°F to calculate Fen with AOR RC/PZR Pressurizer spray piping, pt.

SS [ ]D:a,b,c [ ]D:a,b,c [ ]D:a,b,c 0.816 alternating stresses to calculate None Required(5) spray 190 (Aux. Spray Tee) usage from NUREG/CR-6909 fatigue curve.

Page 9 of 13

System / AOR4 Fatigue SLR Thermal Location Material Fen Uen Analysis Method Management CUF CUF Zone Method ASME Section XI Inservice High HPI piping, stop valve-to-Inspection, Pressure check (usage bounds HPI SS [ ]D:a,b,c [ ]D:a,b,c [ ]D:a,b,c 3.38 Appendix L Subsections IWB, Injection(2) Nozzle)

IWC, and IWD program (B2.1.1)

Refined evaluation using calculated strain rate and a Decay Heat service temperature of 500°F to Removal Decay Heat Drop Lines(2) SS [ ]D:a,b,c [ ]D:a,b,c [ ]D:a,b,c 0.755 calculate Fen with AOR None Required(5)

System alternating stresses to calculate usage from NUREG/CR-6909 fatigue curve.

Refined evaluation using a service temperature of 450°F to calculate Fen with AOR Core Flood Core Flood Piping SS [ ]D:a,b,c [ ]D:a,b,c [ ]D:a,b,c 0.752 None Required(5) alternating stresses to calculate usage from NUREG/CR-6909 fatigue curve.

Notes:

1. Installed equipment is not original. Since the CRDM Weld is not original equipment the next highest location was the RV Instrumentation (BMI) weld.

These locations also bound steam generator Ni-Cr-Fe alloy locations.

2. NUREG/CR-6260 location.
3. The usage presented for drain line is based on MRP-146 turbulence penetration behavior which is a low stress amplitude/high cycle loading. Per NUREG/CR-6909 Rev. 1, no EAF penalty would apply to alternating stress < 28.3 ksi. The fatigue damage occurs under normal operation, therefore, the 60-year usage of [ ]D:a,b,c is adjusted by 80/60 x [ ]D:a,b,c = 0.51.
4. Analysis of Record.
5. Uen less than unity. No fatigue management required per GALL-SLR. Transients for this location are monitored in the Fatigue Monitoring program.

D. Duke Energy Proprietary F. Framatome Proprietary Page 10 of 13

Table 4.3.4-2 ASME Section XI Appendix L Results Acceptable Maximum Allowable Flaw Interval Standards Flaw Final Flaw Allowable End- Operating Location Size Table Configuration (Years) Size (a/t)(3) of-Evaluation Period IWB-3514-2 (a/t)(2) Flaw Size (a/t)(4) (Years)

Axial 10 0.15 0.383 0.623 10(5)

Pressurizer Surge Line Piping (elbow cheek)

Circumferential(1) 10 0.15 0.187 0.750 10(5)

Axial 10 0.15 0.157 0.623 10(5)

Pressurizer Surge Line Piping (weld)

Circumferential(1) 10 0.15 0.192 0.407 10(5)

HPI piping, stop valve-to-check valve (usage Axial 10 0.15 0.2322 0.75 69 bounds HPI nozzle) Circumferential(1) 10 0.15 0.1894 0.56 178 Notes:

1. Aspect ratio (AR) is limited to 1/8 for semi-elliptical circumferential flaw, so a full 360 degree circumferential flaw model is conservatively used when the initial aspect ratio is less than 1/8 limits = constant as the crack grows through the wall thickness.
2. Initial postulated flaw size which is based on ASME Code,Section XI Table IWB-3514-2 for an aspect ratio of 6. The methodology of the initial flaw size is based on ASME Code,Section XI Appendix L-3210.
3. The final flaw size based on fatigue crack growth per ASME Code Case N-809 with a constant aspect ratio. The aspect ratio for the FCG is determined per ASME Code,Section XI Appendix L.
4. The maximum allowable end-of-evaluation flaw size is determined per ASME Code,Section XI Appendix C. The final flaw size after fatigue crack growth should be less than the maximum allowable end-of-evaluation flaw size.
5. The Appendix L interval of 10 years was evaluated as the allowable operating period.

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Table 4.3.4-3 Appendix L Inspections Location Line Weld No. Last ISI Inspection

  • Butt weld PSL-6 Pressurizer Surge Line
  • Butt weld PSL-7 Pressurizer Surge Line O1R29 (F2016)(1)

Piping - U1

  • Cheeks of base metal Piping elbow B (bounded by welds PSL-6 and PSL-7)

High Pressure Injection Stop valve-to-check valve HPI Piping, stop valve-to- butt weld Piping O1R30 (F2018)(2) check valve - U1

  • Loop 1B1
  • Loop 1B1: Weld 165
  • Loop 1B2
  • Loop 1B2: Weld 285
  • Butt weld PSL-6 Pressurizer Surge Line
  • Butt weld PSL-7 Pressurizer Surge Line O2R28 (F2017)(1)

Piping - U2

  • Cheeks of base metal Piping elbow B (bounded by welds PSL-6 and PSL-7)

High Pressure Injection Stop valve-to-check valve HPI Piping, stop valve-to- butt weld Piping O2R29 (F2019)(2) check valve - U2

  • Loop 1B1
  • Loop 2B1: Weld 165
  • Loop 1B2
  • Loop 2B2: Weld 285
  • Butt weld PSL-6 Pressurizer Surge Line Pressurizer Surge Line
  • Butt weld PSL-7 O3R29 (S2018)(1)

Piping - U3 Piping

  • Cheeks of base metal elbow B (bounded by welds PSL-6 and PSL-7)

High Pressure Injection Stop valve-to-check valve HPI Piping, stop valve-to- butt weld Piping O3R30 (S2020)(2) check valve - U3

  • Loop 1B1
  • Loop 3B1: Weld 155
  • Loop 1B2
  • Loop 2B2: Weld 250 Note(s):
1. The Pressurizer Surge Line inspections were previously established to meet GL 88-11.
2. The HPI Piping stop valve-to-check valve weld inspection was previously established to meet GL 88-08 commitments and supplemented by RT.

Page 12 of 13

REFERENCES:

1. 10 CFR 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants
2. OSC-2077-03-SLR-0500, Time-Limited Aging Analysis Identification for Subsequent License Renewal
3. NUREG/CR-6260 (INEL-95/0045), Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components, March 1995. (ML031480219)
4. OSS-0254.00-00-4015, Rev. 3, (MECH) Design Basis Specification for Oconee Piping Classification
5. ONS UFSAR, Revision 28
6. SLR-ONS-TLAA-0300, Rev. 1, Oconee Nuclear Station TLAA of Mechanical System Thermal Fatigue for SLR
7. NUREG-2191, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report, Revision 0.
8. NUREG-2192, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants, Revision 0.
9. NUREG/CR-6583 (ANL-97/18), Effects of LWR Coolant Environments on Fatigue Design Curves of Carbon and Low-Alloy Steels, February 1998.
10. NUREG/CR-5704, Effects of LWR Coolant Environments on Fatigue Design Curves of Austenitic Stainless Steels, April 1999.
11. NUREG/CR-6909, Rev. 0, Effect of LWR Water Environments on the Fatigue Life of Materials, Final Report.
12. EPRI Technical Report TR3002000505, Pressurized Water Reactor Primary Water Chemistry Guidelines, Volume 1, Revision 7, Palo Alto, CA, April 2014.
13. ASME Boiler and Pressure Vessel Code,Section XI, Division 1, Nonmandatory Appendix L, Operating Plant Fatigue Assessment.
14. ASME Code Case N-809, Reference Fatigue Crack Growth Rate Curves for Austenitic Stainless Steel in Pressurized Water Reactor Environments,Section XI, Division 1, ASME International, dated June 23, 2015
15. NUREG/CR-6909, Rev. 1, Effect of LWR Water Environments on the Fatigue Life of Materials, Final Report.
16. Framatome document ANP-3898P/NP, Framatome Reactor Vessel and RCP TLAA and Aging Management Review Input to the ONS SLRA Page 13 of 13