39790 Federal Register / Vol. 83, No. 155 / Friday, August 10, 2018 / Notices 

1See Agencywide Documents Access and 

Management System (ADAMS) Accession No. 

ML17025A180. 

conducted in compliance with the 

Commission’s regulations; the issuance of the 

proposed license amendments will not be 

inimical to the common defense and security 

or to the health and safety of the public; and 

the issuance of the proposed amendments 

will be in accordance with 10 CFR part 51, 

‘‘Environmental Protection Regulations for 

Domestic Licensing and Related Regulatory 

Functions,’’ of the Commission’s regulations 

and all applicable requirements have been 

satisfied. The findings set forth above are 

supported by an NRC safety evaluation dated 

August 1, 2018. 

III. 

Accordingly, pursuant to Sections 161b, 

161i, and 184 of the Act; Title 42 of the 

United States Code Sections 2201(b), 2201(i), 

and 2234; and 10 CFR 50.80, IT IS HEREBY 

ORDERED that the application regarding the 

proposed license transfers is approved, 

subject to the following condition: 

1. Before completion of the proposed 

transaction, EOI shall provide the Director of 

the Office of Nuclear Reactor Regulation 

satisfactory documentary evidence that EAL 

has obtained the appropriate amount of 

insurance required of the licensees under 10 

CFR part 140 and 10 CFR part 50. 

IT IS FURTHER ORDERED that, consistent 

with 10 CFR 2.1315(b), the license 

amendments for ANO, Units 1 and 2, that 

make changes, as indicated in Enclosures 2 

and 3 to the cover letter forwarding this 

order, to conform the licenses to reflect the 

subject transfers, are approved. The 

amendments shall be issued and made 

effective at the time the proposed transfer 

actions are completed. 

IT IS FURTHER ORDERED that, after 

receipt of all required regulatory approvals of 

the proposed transfer actions, EOI shall 

inform the Director of the Office of Nuclear 

Reactor Regulation in writing of such receipt, 

and of the date of closing of the transfers, no 

later than 5 business days before the date of 

the closing of the transfers. Should the 

proposed transfers not be completed within 

1 year of this order’s date of issuance, this 

order shall become null and void; however, 

upon written application and for good cause 

shown, such date may be extended by order. 

This order is effective upon issuance. 

For further details with respect to this 

order, see the application dated September 

21, 2017 (Agencywide Documents Access 

and Management System (ADAMS) 

Accession No. ML17268A213) and the NRC’s 

safety evaluation dated August 1, 2018 

(ADAMS Accession No. ML18177A236), 

which are available for public inspection at 

the NRC’s Public Document Room located at 

One White Flint North, Public File Area 01– 

F21, 11555 Rockville Pike (first floor), 

Rockville, Maryland. Publicly available 

documents created or received at the NRC are 

accessible electronically through ADAMS in 

the NRC Library at http://www.nrc.gov/ 

reading-rm/adams.html. Persons who do not 

have access to ADAMS, or who encounter 

problems accessing the documents in 

ADAMS, should contact the NRC Public 

Document Room reference staff by telephone 

at 1–800–397–4209 or 301–415–4737, or by 

email to pdr.resource@nrc.gov. 

Dated at Rockville, Maryland, this 1st day 

of August, 2018. 

For the Nuclear Regulatory Commission. 

Brian E. Holian, 

Acting Director, Office of Nuclear Reactor 

Regulation. 

[FR Doc. 2018–17168 Filed 8–9–18; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[Docket Nos. 50–368, 50–334, 50–445, 50– 

302, 50–348, 50–364, 50–336, 50–338, 50– 

339, 50–282, 50–306, 50–327, 50–498, 50– 

499, 50–335, 50–280, 50–395, 50–390; NRC– 

2017–0188] 

Entergy Operations, Inc.; FirstEnergy 

Nuclear Operating Company; Vistra 

Operations Company, LLC; Duke 

Entergy Florida, Southern Nuclear 

Operating Company, Inc.; Dominion 

Nuclear Connecticut, Inc.; Virginia 

Electric and Power Company; Northern 

States Power Company—Minnesota; 

South Carolina Electric & Gas 

Company, Inc.; STP Nuclear Operating 

Company; Tennessee Valley Authority 

AGENCY: Nuclear Regulatory 

Commission. 

ACTION: Director’s decision under 10 

CFR 2.206; issuance. 

SUMMARY: The U.S. Nuclear Regulatory 

Commission (NRC) has issued a 

director’s decision in response to a 

petition dated January 24, 2017, filed by 

Mr. Paul Gunter on behalf of Beyond 

Nuclear, and representing numerous 

public interest groups (collectively, 

Beyond Nuclear, et al., or petitioners), 

requesting that the NRC take action with 

regard to licensees of plants that 

currently rely on potentially defective 

safety-related components and 

potentially falsified quality assurance 

documentation supplied by AREVA-Le 

Creusot Forge and Japan Casting and 

Forging Corporation. The petitioners’ 

requests are included in the 

SUPPLEMENTARY INFORMATION section of 

this document. 

DATES: The director’s decision was 

issued on August 2, 2018. 

ADDRESSES: Please refer to Docket ID 

NRC–2017–0188 when contacting the 

NRC about the availability of 

information regarding this document. 

You may obtain publicly-available 

information related to this document 

using any of the following methods: 

• Federal Rulemaking Website: Go to 

http://www.regulations.gov and search 

for Docket ID NRC–2017–0188. Address 

questions about NRC dockets to Jennifer 

Borges; telephone: 301–287–9127; 

email: Jennifer.Borges@nrc.gov. For 

technical questions, contact the 

individual listed in the FOR FURTHER

INFORMATION CONTACT section of this 

document. 

• NRC’s Agencywide Documents 

Access and Management System 

(ADAMS): You may obtain publiclyavailable documents online in the 

ADAMS Public Documents collection at 

http://www.nrc.gov/reading-rm/ 

adams.html. To begin the search, select 

‘‘Begin Web-based ADAMS Search.’’ For 

problems with ADAMS, please contact 

the NRC’s Public Document Room (PDR) 

reference staff at 1–800–397–4209, 301– 

415–4737, or by e-mail to pdr.resource@

nrc.gov. The ADAMS accession number 

for each document referenced (if it is 

available in ADAMS) is provided the 

first time that it is mentioned in this 

document. 

• NRC’s PDR: You may examine and 

purchase copies of public documents at 

the NRC’s PDR, Room O1–F21, One 

White Flint North, 11555 Rockville 

Pike, Rockville, Maryland 20852. 

FOR FURTHER INFORMATION CONTACT: 

Perry Buckberg, Office of Nuclear 

Reactor Regulation, U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001; telephone: 301–415– 

1383; email: Perry.Buckberg@nrc.gov. 

SUPPLEMENTARY INFORMATION: The text of 

the director’s decision is attached. 

Dated at Rockville, Maryland, this 7th day 

of August 2018. 

For the Nuclear Regulatory Commission. 

Perry H. Buckberg, 

Senior Project Manager, Special Projects and 

Process Branch, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

Attachment—Director’s Decision DD–18–03 

UNITED STATES OF AMERICA 

NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR 

REGULATION 

Brian E. Holian, Acting Director 

In the Matter of Power Reactor Licensees 

Docket Nos.: See Attached List 

License Nos.: See Attached List 

DIRECTOR’S DECISION UNDER 10 CFR 

2.206 

I. Introduction 

On January 24, 2017,1 Mr. Paul Gunter 

submitted a petition on behalf of Beyond 

Nuclear that represents numerous public 

interest groups (collectively referred to as the 

Petitioners) under Title 10 of the Code of 

Federal Regulations (10 CFR) 2.206, 

‘‘Requests for Action under This Subpart.’’ 


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2See ADAMS Accession No. ML17052A032. 

3See ADAMS Accession No. ML17068A061. 

4See ADAMS Accession No. ML17067A562. 

5See ADAMS Accession No. ML17174A087. 

6See ADAMS Accession No. ML17174A788. 

7See ADAMS Accession No. ML17179A288. 

8See ADAMS Accession No. ML17184A058. 

9See ADAMS Accession No. ML17187A026. 

10The petition incorrectly states that JCFC is a 

subcontractor to ACF. 

11See the report titled ‘‘Irregularities and 

Anomalies Relating to the Forged Components of Le 

Creusot Forge,’’ dated September 26, 2016, Large 

and Associates Consulting Engineers, London, 

England (available at http://

www.largeassociates.com/CZ3233/Note_

LargeAndAssociates_EN_26092016.pdf). 

12See ADAMS Accession No. ML17039A501. 

13See ADAMS Accession No. ML17081A418. 

14See ADAMS Accession No. ML17052A033. 

The Petitioners supplemented their petition 

by e-mails dated February 16,2 March 6,3,4

June 16,5 June 22,6 June 27,7 June 30,8 and 

July 5, 2017.9 The June 16 and June 22, 2017, 

supplements added the Crystal River Unit 3 

Nuclear Generating Plant (Crystal River Unit 

3) to the list of plants subject to the petition 

and requested slightly different enforcement 

actions. The rest of the supplements did not 

expand the scope of the petition or request 

additional actions that should be considered 

as a new petition. The Petitioners asked the 

U.S. Nuclear Regulatory Commission (NRC) 

to take emergency enforcement action at U.S. 

nuclear power plants that currently rely on 

potentially defective safety-related 

components and potentially falsified quality 

assurance documentation supplied by 

AREVA-Le Creusot Forge (ACF) and its 

subcontractor, Japan Casting and Forging 

Corporation (JCFC).10 Table 1 lists potentially 

affected components and the at-risk reactors 

identified in the petition. 

TABLE 1—LIST OF POTENTIALLY AFFECTED COMPONENTS AND REACTORS

Reactor pressure 

vessels 

Replacement reactor pressure 

vessel heads Steam generators Steam pressurizers 

Prairie Island, Units 1 and 2 (MN) Arkansas Nuclear One, Unit 2 

(AR).

Beaver Valley, Unit 1 (PA) ........... Millstone, Unit 2 (CT). 

Beaver Valley, Unit 1 (PA) ........... Comanche Peak, Unit 1 (TX) ....... Saint Lucie, Unit 1 (FL). 

North Anna, Units 1 and 2 (VA) ... V.C. Summer (SC) .......................

Surry, Unit 1 (VA) ......................... Farley, Units 1 and 2 (AL).

Crystal River, Unit 3 (FL) ............. South Texas, Units 1 and 2 (TX).

Sequoyah, Unit 1 (TN).

Watts Bar, Unit 1 (TN).

Specifically, the Petitioners asked the NRC 

to take enforcement actions consistent with 

the following: 

1. Suspend power operations of U.S. 

nuclear power plants that rely on ACF 

components and subcontractors pending a 

full inspection (including nondestructive 

examination by ultrasonic testing) and 

material testing. If carbon anomalies (‘‘carbon 

segregation’’ or ‘‘carbon macrosegregation’’ 

(CMAC)) in excess of the design-basis 

specifications for at-risk component parts are 

identified, require the licensee to do one of 

the following: 

a. Replace the degraded at-risk 

component(s) with quality-certified 

components. 

b. For those at-risk degraded components 

that a licensee seeks to allow to remain in 

service, apply through the license 

amendment request process to demonstrate 

that a revised design basis is achievable and 

will not render the inservice component 

unacceptably vulnerable to fast fracture 

failure at any time and in any credible 

service condition throughout the current 

license of the power reactor. 

2. Alternatively modify the licensees’ 

operating licenses to require the licensees to 

perform the requested emergency 

enforcement actions at the next scheduled 

outage. 

3. Issue a letter to all U.S. light-water 

reactor operators under 10 CFR 50.54(f) 

requiring licensees to provide the NRC with 

information under oath and affirming 

specifically how U.S. operators are reliably 

monitoring contractors and subcontractors 

for the potential carbon segmentation 

anomaly in the supply chain and the 

reliability of the quality assurance 

certification of those components, and 

publicly release the responses. 

The June 16 and June 22, 2017, 

supplements to the petitions added Crystal 

River Unit 3, which is currently shut down, 

and the licensee Duke Energy to the list of 

facilities for which the Petitioners requested 

the following fourth NRC action: 

a. Confirm the sale, delivery, quality 

control and quality assurance certification 

and installation of the replacement reactor 

pressure vessel head as supplied to Crystal 

River Unit 3 by then Framatome and now 

AREVA-Le Creusot Forge industrial facility 

in Charlon-St. Marcel, France and; 

b. With completion and confirmation [of 

the above Crystal River Unit 3 actions], the 

modification of Duke Energy’s current license 

for the permanently closed Crystal River Unit 

3 nuclear power station in Crystal River, 

Florida, to inspect and conduct the 

appropriate material test(s) for carbon 

macrosegregation on sufficient samples 

harvested from the installed and now 

inservice irradiated Le Creusot Forge reactor 

pressure vessel head [sic]. The Petitioners 

assert that the appropriate material testing 

include Optical Emissions Spectrometry 

(OES). 

As the basis of their requests, the 

Petitioners cited the expert review by Large 

and Associates Consulting Engineers that 

identified significant irregularities and 

anomalies in both the manufacturing process 

and quality assurance documentation of large 

reactor components manufactured by the 

ACF for French reactors and reactors in other 

countries.11

On February 2, 2017,12 the Office of 

Nuclear Reactor Regulation (NRR) petition 

manager acknowledged receipt of the petition 

and offered an opportunity for the Petitioners 

to address NRR’s 10 CFR 2.206 Petition 

Review Board (PRB) to discuss the petition. 

The Petitioners accepted the offer, and the 

meeting was held on March 8, 2017. The 

transcript 13 of that meeting is publicly 

available. 

On February 8, 2017, the PRB met 

internally to discuss the request for 

immediate actions and informed the 

Petitioners on February 13, 2017,14 that no 

actions were warranted at that time because 

the NRC has reasonable assurance of public 

health and safety and protection of the 

environment. The basis for the PRB’s 

determination included the following: 

• Extent of Condition. Internationally, 

CMAC has been found only in components 

produced by ACF using a specific 

processing route. Based on the staff’s 

knowledge as of February 2017, only a 

subset of the plants identified in the 

petition contain components that may have 

used the processing route that resulted in 

the excess CMAC found in international 

plants. 

• Degree of Condition. If CMAC is present 

in a component, it occurs in a localized 

region of the forged component. It is not a 

bulk material phenomenon, does not go 

through thickness, and is not expected to 

affect the structural integrity of the 

component. In addition, based on the 

staff’s knowledge as of February 2017, the 

highest levels of CMAC observed 

internationally, if present in the postulated 

regions of U.S. components, are not 

expected to alter the mechanical properties 

of the material enough to affect the 

structural integrity of the components. 

Destructive examinations of components 

containing regions of CMAC have been 

conducted internationally to determine 

how CMAC affects mechanical properties 

and such examinations confirm that 

structural integrity has not been impacted. 

A summary of the international 

investigation is summarized in II.A below, 

and details of the investigation and its 


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15See ADAMS Accession No. ML18017A441. 

16See ADAMS Accession No. ML17171A108. 

17See ADAMS Accession No. ML17171A106. 

18See ADAMS Accession No. ML17142A334. 

19See ADAMS Accession No. ML17198A329. 

20See ADAMS Accession No. ML18107A402. 

21See ADAMS Accession No. ML18012A156. 

22See ADAMS Accession No. ML12192A058. 

23See ADAMS Accession No. ML17040A100. 

24See ASN/Institut de Radioprotection et de 

Suˆrete ́ Nucle ́aire (IRSN) (Radioprotection and 

Nuclear Safety Institute) report CODEP–DEP–2015– 

037971, ‘‘Analysis of the Procedure Proposed by 

AREVA to Prove Adequate Toughness of the Dome 

of the Flamanville 3 EPR Reactor Pressure Vessel 

Lower Head and Closure Head,’’ English 

translation, dated September 16, 2015. http://

www.french-nuclear-safety.fr/Media/Files/00- 

Publications/Report-to-the-Advisory-Committee-ofExperts-for-Nuclear-Pressure-Equipment. 

impact on structural integrity are described 

in the staff’s evaluation dated February 22, 

2018.15

• Safety Significance. The staff’s 

preliminary safety assessment concluded 

that the safety significance of CMAC to the 

U.S. nuclear power reactor fleet appears to 

be negligible. The staff based its 

assessment on knowledge of the material 

processing, qualitative analysis, 

compliance of U.S. components with the 

American Society of Mechanical Engineers 

Boiler Pressure and Vessel Code (ASME 

Code), and the results of preliminary 

structural evaluations. The NRC 

subsequently presented the basis for this 

determination in a technical session, titled 

‘‘Carbon Macrosegregation in Large 

Nuclear Forgings,’’ at the NRC-sponsored 

Regulatory Information Conference on 

March 15, 2017.16 17

On April 11, 2017, the PRB met to discuss 

the petition with respect to the criteria for 

consideration under 10 CFR 2.206. Based on 

that review, the PRB determined that the 

petition request meets the criteria for 

consideration under 10 CFR 2.206. On May 

19, 2017, the petition manager informed the 

Petitioners that the initial recommendation 

was to accept the petition for review but to 

refer a portion of the petition (i.e., the 

concern of potentially falsified quality 

assurance documentation) to the NRC’s 

allegation process for appropriate action.18

The petition manager also offered the 

Petitioners an opportunity to comment on the 

PRB’s recommendations. On July 5, 2017, the 

petition manager clarified the initial 

recommendation and asked for a response as 

to whether the Petitioners wanted to address 

the PRB a second time to comment on its 

recommendations. The Petitioners did not 

request a second opportunity to address the 

PRB. Therefore, the PRB’s initial 

recommendations to accept part of the 

petition for review under 10 CFR 2.206 and 

to refer a part to another NRC process became 

final. On August 30, 2017, the petition 

manager issued an acknowledgment letter to 

the Petitioners.19

By a letter to the Petitioners which copied 

the licensees dated June 6, 2018,20 the NRC 

issued the proposed director’s decision for 

comment. The Petitioners were asked to 

provide comments within 14 days on any 

part of the proposed director’s decision 

considered to be erroneous or any issues in 

the petition that were not addressed. The 

NRC staff did not receive any comments on 

the proposed director’s decision. 

The petition and other references related to 

this petition are available for inspection in 

the NRC’s Public Document Room (PDR), 

located at O1F21, 11555 Rockville Pike (first 

floor), Rockville, MD 20852. Publicly 

available documents created or received at 

the NRC are accessible electronically through 

ADAMS in the NRC Library at http://

www.nrc.gov/reading-rm/adams.html. 

Persons who do not have access to ADAMS 

or who encounter problems in accessing the 

documents located in ADAMS should 

contact the NRC’s PDR reference staff by 

telephone at 1-800-397-4209 or 301-415-4737 

or by e-mail to pdr.resource@nrc.gov. 

II. Discussion 

Under the 10 CFR 2.206(b) petition review 

process, the Director of the NRC office with 

responsibility for the subject matter shall 

either institute the requested proceeding or 

shall advise the person who made the request 

in writing that no proceeding will be 

instituted, in whole or in part, with respect 

to the request and the reason for the decision. 

Accordingly, the decision of the NRR 

Director is provided below. As further 

discussed below, the petition is denied. 

The NRC’s policy is to have an effectively 

coordinated program to promptly and 

systematically review relevant domestic and 

applicable international operational 

experience (OpE) information. The program 

supplies the means for assessing the 

significance of OpE information, offering 

timely and effective communication to 

stakeholders, and applying the lessons 

learned to regulatory decisions and programs 

affecting nuclear reactors. The NRC 

Management Directive 8.7, ‘‘Reactor 

Operating Experience Program,’’ dated 

February 1, 2018, describes the Reactor OpE 

Program.21 The NRR Office Instruction (OI) 

LIC-401, ‘‘NRR-NRO Reactor Operating 

Experience Program,’’ Revision 3, addresses 

the specific implementation of the Reactor 

OpE Program.22

As reported in internal NRC 

communications, AREVA notified France’s 

nuclear safety authority, Autorite ́ de Suˆrete ́

Nucle ́aire (ASN), of an anomaly in the 

composition of the steel in certain zones of 

the reactor pressure vessel (RPV) upper and 

lower heads of the Flamanville Nuclear 

Power Plant (Flamanville), Unit 3, in 

Manche, France. Both the upper and lower 

vessel heads were manufactured by ACF. 

According to ASN, chemical and mechanical 

property testing performed by AREVA in late 

2014 (on a vessel head similar to that of the 

Flamanville European Pressurized Reactor 

(EPR)) revealed a zone of high carbon 

concentration (0.30 percent as opposed to a 

target value of 0.22 percent), which led to 

lower than expected mechanical toughness 

values in that area. Initial measurements 

confirmed the presence of this anomaly in 

the Flamanville, Unit 3, RPV upper and 

bottom heads. 

In accordance with the process described 

in NRR OI LIC-401, the NRC’s Reactor OpE 

Program staff ensured that the appropriate 

technical experts within the NRC were aware 

of the issue and were evaluating these issues 

for relevance to the U.S. industry. In 

addition, the NRC has strong collaboration 

with the international community and was 

separately in contact with ASN to discuss 

this issue. 

A. Description of the Issue 

The CMAC is a known phenomenon that 

takes place during the casting of large ingots. 

The CMAC is a material heterogeneity in the 

form of a chemical (i.e., carbon) gradient that 

deviates from the nominal composition and 

may exceed specification limits. Portions of 

the ingot containing CMAC that exceed 

specification limits (positive CMAC) are 

purposefully removed and discarded as part 

of the material processing. Regions of 

positive CMAC that are not appropriately 

removed result in localized regions near the 

surface of the final component with higher 

strength and lower toughness relative to the 

bulk material. 

In April 2015, regions of positive CMAC 

were discovered in EPR RPV heads that were 

manufactured for the Flamanville plant. The 

ACF had produced the forgings for the 

Flamanville upper and lower RPV heads. The 

discovery of the CMAC in the heads 

prompted ASN to ask the operator, E ́lectricite ́

de France S.A. (EDF) (Electricity of France), 

to review inservice forged components at all 

of its plants to determine the potential extent 

of the condition. The review identified steam 

generator (SG) channel heads (also 

commonly referred to as SG primary heads) 

produced by ACF and JCFC as the 

components most likely to contain a region 

of CMAC. The ASN requested that 

nondestructive testing be performed on these 

SG channel heads to characterize the carbon 

content and confirm the absence of 

unacceptable flaws. 

On October 18, 2016, ASN ordered the 

acceleration of the nondestructive testing of 

the potentially affected ACF and JCFC SG 

channel heads, which required completion of 

the remaining nondestructive testing within 

3 months. The discovery of higher than 

expected carbon values measured on an 

inservice SG channel head produced by JCFC 

prompted the accelerated schedule. As a 

result, to perform the required 

nondestructive tests, EDF had to shut down 

its plants before their scheduled outages. 

AREVA Inc. (AREVA Inc. or AREVA), 

located in Lynchburg, VA, provides 

safety-related products and services for U.S. 

operating nuclear power plants, including 

replacements for reactor coolant pressure 

boundary components. On February 3, 

2017,23 AREVA Inc. submitted a list to the 

NRC of the U.S. reactors that have received 

components fabricated with forgings from 

ACF. Operating U.S. plants have no known 

components from JCFC. 

In September 2015, June 2016, and June 

2017, ASN convened an Advisory Committee 

of Experts for Nuclear Pressure Equipment to 

obtain its technical opinion on the 

consequences of CMAC for the serviceability 

of the Flamanville EPR reactor vessel domes. 

The resulting series of publicly available 

reports (CODEP–DEP–2015–037971,24


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25See ASN/IRSN report CODEP–DEP–2016– 

019209, ‘‘Procedure Proposed by AREVA to Prove 

Adequate Toughness of the Domes of the 

Flamanville 3 EPR Reactor Pressure Vessel Bottom 

Head and Closure Head,’’ English translation, dated 

June 17, 2016. https://www.asn.fr/content/ 

download/106732/811356/version/6/file/CODEPDEP-2016-019209-advisorycommitte24june2016- 

summaryreport.pdf. 

26See ASN/IRSN report CODEP–DEP–2017– 

019368, ‘‘Analysis of the Consequences of the 

Anomaly in the Flamanville EPR Reactor Pressure 

Vessel Head Domes on Their Serviceability,’’ 

English translation, dated June 15, 2017. http://

www.irsn.fr/FR/expertise/rapports_gp/Documents/ 

GPESPN/IRSN-ASNDEP_GPESPN-Report_pressurevessel-FA3_201706.pdf. 

27See ADAMS Accession No. ML072830076. 

28See ADAMS Accession No. ML072820691. 

29EPRI Report No. 3002010331, ‘‘Materials 

Reliability Program: Evaluation of Risk from Carbon 

Macrosegregation in Reactor Pressure Vessels and 

Other Large Nuclear Forgings (MRP–417),’’ issued 

June 2017 (available at ADAMS Accession No. 

ML18054A862). 

30See ADAMS Accession No. ML072830076. 

31See ADAMS Accession No. ML072820691. 

32See ADAMS Accession No. ML17124A575. 

33See ADAMS Accession No. ML13148A361. 

34See ADAMS Accession No. ML113190538. 

CODEP–DEP–2016–019209,25 and CODEP– 

DEP–2017–019368 26) justified the continued 

use of the Flamanville heads. In this effort, 

AREVA conducted hundreds of mechanical 

and chemical property experiments on three 

full-scale replica heads that were 

manufactured by ACF using the same process 

as that used for the Flamanville heads. Using 

these experimental results, AREVA 

conducted a variety of code-related fracture 

and strength analyses that demonstrated that 

the risk of fast fracture from CMAC was 

extremely low. Through this effort, ASN 

concluded that the serviceability of the heads 

is acceptable as long as EDF conducts the 

required inservice inspections. However, 

because of its inability to conduct an 

adequate inservice inspection on the 

Flamanville upper head, ASN concluded that 

the upper head long-term serviceability could 

not be confirmed and that the head should 

be replaced after a few years of operation. 

B. Initial Actions by the NRC and the U.S. 

Nuclear Industry 

Beginning in December 2016, the NRC staff 

conducted a preliminary safety assessment to 

determine the potential safety significance 

posed to the U.S. nuclear power reactor fleet 

by the CMAC observed in reactor coolant 

system (RCS) components overseas and 

concluded that the failure of an RPV/SG head 

component has a very low probability, even 

if the worst practical degree of CMAC occurs 

within that component. The NRC staff used 

a qualitative failure comparison to assess the 

relative likelihood of failure of an RPV shell 

(which is not expected to be subject to 

positive CMAC) with RPV/SG head 

component types that could be affected by 

CMAC. Based on this comparison, the NRC 

determined the following: 

• The RPV shell experiences higher stresses 

under both normal operations and 

postulated accident scenarios. 

• The weld region of an RPV shell has a 

greater likelihood of having more flaws and 

larger fabrication flaws. The larger 

fabrication flaws typically have the higher 

potential to result in component failure. 

• Although the initial toughness of an RPV 

shell material may be greater than an RPV/ 

SG head with postulated positive CMAC, 

the shell toughness decreases as the result 

of radiation embrittlement after several 

years of operation. As a result, the current 

as-operated toughness of RPV shell 

material is expected to be lower than the 

toughness of RPV/SG head material with 

postulated CMAC. The RPV shell material 

is known to have adequate toughness for 

safe operation. 

When combining all these individual 

attributes, an RPV/SG head component with 

postulated CMAC is much less likely to fail 

than an RPV shell. Past research and 

operating experience has demonstrated that 

failure of an RPV shell under normal 

operations or postulated accident scenarios 

has a very low probability of occurrence.27 28

Therefore, the failure of an RPV/SG head 

component also has a very low probability, 

even if the worst practical degree of CMAC 

occurs within that component. The NRC 

presented the basis for this preliminary 

determination in a technical session titled 

‘‘Carbon Macrosegregation in Large Nuclear 

Forgings’’ (cited above) at the March 15, 

2017, NRC-sponsored Regulatory Information 

Conference. 

Concurrent with the NRC analyses, the 

U.S. industry initiated a research program in 

early 2017, conducted by the Electric Power 

Research Institute (EPRI), to address the 

generic safety significance of elevated carbon 

levels caused by CMAC in the components of 

interest. This program was divided into the 

following four main tasks, each aimed at 

developing both qualitative and quantitative 

information to make a safety determination: 

1. extension of RPV probabilistic fracture 

mechanics (PFM) analyses to 

qualitatively bound other components 

2. development of a robust technical basis to 

support the hypothesis that RPV 

integrity bounds other components 

3. quantitative structural analyses to assess 

whether the results of the PFM analyses 

of the RPV beltline (Task 1) bound the 

other forged components 

4. a white paper assessing the effect of CMAC 

on SG tubesheets based on expert 

judgment and experience with the 

fabrication of the tubesheets as large 

forgings 

As of the writing of this document, Task 

1 has been completed and has been publicly 

released as Materials Reliability Program 

(MRP)-417.29 The other tasks are still under 

development with the expected release of the 

report(s) in 2018. 

The MRP-417 addresses the structural 

significance of the potential presence of 

CMAC in large, forged pressurized-water 

reactor pressure-retaining components, 

including the RPV head, beltline and nozzle 

shell forgings, and the SG and pressurizer 

ring and head forgings through the end of an 

80-year operating interval. The assessment 

was made using the NRC risk safety criterion 

of a 95th percentile through-wall crack 

frequency (TWCF) of less than 1×10¥6 per 

year (yr¥1) (10 CFR 50.61a, ‘‘Alternative 

Fracture Toughness Requirements for 

Protection against Pressurized Thermal 

Shock Events’’) for pressurized thermal shock 

(PTS) events and a conditional probability of 

failure (CPF) of less than 1×10¥6 for normal 

operating transients. These analyses used 

many of the same assumptions and inputs as 

those used in the basis for the 10 CFR 50.61a 

alternate PTS rule.30 31 In addition, the 

analysts approximated the effect of carbon 

content on the fracture toughness of the steel 

through a review of the available literature. 

The MRP-417 describes the analyses and 

results for bounding values for the RPV shell, 

RPV upper head, SG channel head, 

pressurizer shell, and pressurizer head 

components based on the analyses 

assumptions from the alternate PTS rule in 

conjunction with the effect of the CMAC on 

the material toughness. The report’s 

deterministic results suggest that the RPV 

vessel behavior bounds the behavior of the 

pressurizer components. In addition, the 

probabilistic results suggest that in all cases, 

assuming the maximum carbon content 

observed in the field, the calculated TWCF 

and CPF were below the NRC risk safety 

criterion of the 95th percentile TWCF of less 

than 1×10¥6 yr¥1 for PTS events and a CPF 

of less than 1×10¥6 for normal operating 

transients. MRP-417 concludes that there is 

substantial margin against failure through an 

80-year operating interval using the assumed 

CMAC distributions in the RPV, SG, and 

pressurizer rings and head forgings in 

pressurized-water reactors. 

In March 2017, an NRC inspection team 

performed a limited-scope vendor inspection 

at the AREVA facility in Lynchburg, Virginia, 

to review documentation from ACF and 

assess AREVA’s compliance with the 

provisions of selected portions of Appendix 

B, ‘‘Quality Assurance Criteria for Nuclear 

Power Plants and Fuel Reprocessing Plants,’’ 

to 10 CFR Part 50, and 10 CFR Part 21, 

‘‘Reporting of Defects and Noncompliance.’’ 

This inspection focused on AREVA’s 

documentation and evaluation of potential 

carbon macrosegregation issues in forgings 

supplied by AREVA for U.S. operating 

nuclear power plants. Specifically, the NRC 

inspection reviewed documentation to verify 

that forgings met the ASME Code 

requirements for carbon content and 

mechanical properties. The NRC issued the 

inspection report on May 10, 2017.32 The 

limited-scope inspection reviewed policies 

and procedures that govern implementation 

of AREVA’s 10 CFR Part 21 program, and 

nonconformance and corrective action 

policies and procedures under its approved 

quality assurance program related to the 

manufacturing processes used by ACF to 

fabricate inservice U.S. components and the 

resulting mechanical properties. The NRC 

inspection team used Inspection Procedure 

(IP) 43002, ‘‘Routine Inspections of Nuclear 

Vendors,’’ 33 and IP 36100, ‘‘Inspection of 10 

CFR Part 21 and Programs for Reporting 

Defects and Noncompliance.’’ 34 The 

inspection team did not identify any 

violations or nonconformances during the 

inspection. 


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35See ADAMS Accession No. ML14035A143. 

The inspection report contains the 

following primary material processing and 

property observations: 

• A population of the components produced 

by ACF has a low or no possibility of 

containing regions of CMAC. 

• Carbon levels and mechanical properties 

for the components reviewed conformed to 

ASME Code requirements. 

• The information reviewed did not 

challenge the NRC’s preliminary 

determination on the CMAC topic (i.e., that 

the safety significance to the U.S. nuclear 

power reactor fleet appears to be 

negligible). 

The NRC staff also documented its 

risk-informed evaluation of the potential 

safety significance of CMAC in components 

produced by ACF, as it relates to the safe 

operation of U.S. plants, and options for 

addressing the topic using its risk-informed 

decision-making process in NRR OI LIC-504, 

‘‘Integrated Risk-Informed Decision-Making 

Process for Emergent Issues,’’ Revision 4, 

dated June 2, 2014,35 to evaluate this issue. 

C. Applicable NRC Regulatory Requirements 

and Guidance 

The NRC requires U.S. nuclear reactor 

components fabricated with forgings from 

ACF to be manufactured and procured in 

accordance with all applicable regulations, as 

well as the ASME Code requirements that are 

incorporated by reference. The regulations 

most pertinent to the prevention and 

identification of CMAC in regions of RCS 

components are the ASME Code 

requirements incorporated by reference in 10 

CFR 50.55a, ‘‘Codes and Standards,’’ and 

quality assurance requirements in 10 CFR 

part 50, Appendix B. In addition to the NRC 

regulations and ASME Code requirements 

that are focused on the process and quality 

controls for addressing CMAC, there are also 

regulations that focus on performance and 

design criteria that may be impacted by 

regions of CMAC. These regulations include: 

10 CFR 50.60, ‘‘Acceptance criteria for 

fracture prevention measures for lightwater 

nuclear power reactors for normal 

operation,’’ Appendix A to 10 CFR part 50, 

‘‘General Design Criteria for Nuclear Power 

Plants,’’ and Appendix G to 10 CFR part 50, 

‘‘Fracture Toughness Requirements.’’ The 

applicability of specific NRC regulations and 

ASME Code requirements will, in part, 

depend on the dates that the regulations or 

requirements became effective relative to a 

component being put into operation. The 

plant-specific design basis and current 

licensing basis address the fundamental 

regulatory requirements pertaining to the 

integrity of the components of interest. 

Appendix B to 10 CFR part 50 establishes 

quality assurance requirements for the 

design, manufacture, construction, and 

operation of the structures, systems, and 

components (SSCs) for nuclear facilities. 

Appendix B requirements apply to all 

activities affecting the safety-related 

functions of those SSCs. These activities 

include designing, purchasing, fabricating, 

handling, installing, inspecting, testing, 

operating, maintaining, repairing, and 

modifying SSCs. To accomplish these 

activities, licensees must contractually pass 

down the requirements of Appendix B 

through procurement documentation to 

suppliers of SSCs, as stated in the Appendix 

B criteria below. 

Criterion IV, ‘‘Procurement Document 

Control,’’ of 10 CFR Part 50, Appendix B, 

states the following: 

Measures shall be established to assure that 

applicable regulatory requirements, design 

bases, and other requirements which are 

necessary to assure adequate quality are 

suitably included or referenced in the 

documents for procurement of material, 

equipment, and services, whether 

purchased by the applicant or by its 

contractors or subcontractors. To the extent 

necessary, procurement documents shall 

require contractors or subcontractors to 

provide a quality assurance program 

consistent with the pertinent provisions of 

this appendix. 

Criterion VII, ‘‘Control of Purchased 

Material, Equipment, and Services,’’ of 10 

CFR Part 50, Appendix B, in part, states, the 

following: 

Documentary evidence that material and 

equipment conform to the procurement 

requirements shall be available at the 

nuclear power plant or fuel reprocessing 

plant site prior to installation or use of 

such material and equipment. This 

documentary evidence shall be retained at 

the nuclear power plant or fuel 

reprocessing plant site and shall be 

sufficient to identify the specific 

requirements, such as codes, standards, or 

specifications, met by the purchased 

material and equipment. 

The licensee is responsible for ensuring 

that the procurement documentation 

appropriately identifies the applicable 

regulatory and technical requirements and 

for determining whether the purchased items 

conform to the procurement documentation. 

Criterion XV, ‘‘Nonconforming Materials, 

Parts, or Components,’’ of 10 CFR Part 50, 

Appendix B, states the following: 

Measures shall be established to control 

materials, parts, or components which do 

not conform to requirements in order to 

prevent their inadvertent use or 

installation. These measures shall include, 

as appropriate, procedures for 

identification, documentation, segregation, 

disposition, and notification to affected 

organizations. Nonconforming items shall 

be reviewed and accepted, rejected, 

repaired or reworked in accordance with 

documented procedures. 

Nonconformances identified by the 

supplier during manufacturing must be 

technically evaluated and dispositioned 

accordingly. If the supplier identifies a 

nonconformance, such as the presence of 

CMAC in the final product, it must perform 

an engineering evaluation and document the 

nonconformance on the associated certificate 

of conformance. The licensee is responsible 

for reviewing the certificate of conformance 

during receipt inspection for acceptance of 

the final product upon delivery. 

Under 10 CFR Part 21, the NRC requires 

both licensees and their suppliers to evaluate 

any condition or defect in a component that 

could create a substantial safety hazard. 

Regions of CMAC in RCS components 

suspected of having the potential to create a 

substantial safety hazard would be an 

example of a condition that licensees and 

their suppliers must evaluate. In addition, 10 

CFR Part 21 requires the entity to notify the 

NRC if it becomes aware of information that 

reasonably indicates that a basic component 

contains defects that could create substantial 

safety hazard. 

D. Summary of the NRC’s Evaluation 

The NRC’s evaluation of this issue 

consisted of conducting preliminary safety 

analyses as described above, reviewing the 

testing and analyses performed by the French 

licensee, meeting with French and Japanese 

regulators to discuss their evaluation, 

reviewing the nuclear industry’s evaluation 

of the issue, conducting an onsite inspection 

of manufacturing and procurement records, 

and determining the final safety assessment 

using a risk-informed decision-making 

process. The staff’s evaluation dated 

February 22, 2018, documents the NRC’s full 

evaluation of the CMAC topics as it relates 

to plants operating in the United States. 

The staff reviewed the publicly available 

ASN documentation on this issue (CODEP– 

DEP–2015–037971, CODEP–DEP–2016– 

019209, and CODEP–DEP–2017–019368) and 

concluded that, although ASN’s decisions 

and actions are based solely on French 

nuclear regulations which do not directly 

correlate to U.S. regulations, the 

experimental results and the fast fracture 

analyses can provide direct insight into the 

expected behavior of postulated CMAC in 

U.S.-forged components. As concluded by 

ASN, the analyses demonstrate that the fast 

fracture of the Flamanville heads from the 

impacts of CMAC can be ruled out in view 

of the margins determined by the analyses. 

The NRC staff reviewed the technical 

information in MRP-417 and concluded that 

it was credible for use in this assessment for 

the following reasons: 

• The risk criteria used for the CPF and 95th

percentile TWCF were identical to those 

used in the development of 10 CFR 50.61a. 

• Major probabilistic inputs, such as flaw 

distribution, standard material properties, 

transients, and normal operating 

conditions were identical to those used in 

the development of 10 CFR 50.61a. 

• The CMAC distribution and toughness 

relationships used were based on historical 

literature and empirical data. 

• The assumptions made using the 

computational model were consistent with, 

or were conservative as compared to those 

used in the analyses for the development 

of 10 CFR 50.61a. 

The NRC assessment of MRP–417 for this 

report does not constitute a regulatory 

endorsement of its full contents. The NRC 

staff will assess the other industry reports on 

the CMAC topic in the same manner as such 

reports become available. 

Although these evaluations provide useful 

information to address the impacts of 

postulated CMAC in forged components in 

service at U.S. operating reactors, the NRC 

staff used an analysis approach, leveraging 


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36See ADAMS Accession No. ML053490187. 

existing PFM results and examining them in 

the context of the NRC’s approach to the 

risk-informed decision-making process 

described in NRR OI LIC–504. 

Consistent with LIC–504, for this review, 

the NRC staff considered the following five 

principles of risk-informed decision-making 

when considering options for addressing this 

issue: 

• Principle 1. The proposed change must 

meet the current regulations unless it is 

explicitly related to a requested exemption 

or rule change. 

• Principle 2. The proposed change shall be 

consistent with the defense-in-depth 

philosophy. 

• Principle 3. The proposed change shall 

maintain sufficient safety margins. 

• Principle 4. When the proposed change 

results in an increase in core damage 

frequency or risk, the increases should be 

small and consistent with the intent of the 

Commission’s safety goals. 

• Principle 5. Monitoring programs should 

be in place. 

The NRC staff considered the following 

four options to address the potential impact 

of the international CMAC OpE on the U.S. 

nuclear power reactor fleet. Options 2, 3, and 

4 align with the Petitioners’ requests. 

• Option 1: Evaluate and Monitor 

• Option 2: Issue a Generic Communication 

• Option 3: Issue Orders Requiring 

Inspections 

• Option 4: Issue Orders Suspending 

Operation 

Option 1 

This option consists of the NRC staff 

continuing to monitor all domestic and 

international information associated with the 

CMAC topic. The staff will evaluate new 

information, as it becomes available, to 

ensure that conservatism in the staff’s final 

safety determination is maintained. Aspects 

of the staff’s safety determination that may be 

evaluated against new information includes 

the extent of condition in the U.S., potential 

degree of CMAC on a generic basis, or data 

affecting the relationship between CMAC and 

mechanical performance. This information is 

to be evaluated to determine if there is 

reasonable assurance that adequate 

defense-in-depth, sufficient safety margin, 

and an acceptable level of risk is maintained 

with an appropriate degree of conservatism. 

If new information becomes available that 

warrants evaluation and it is concluded that 

the staff’s safety determination remain 

appropriately conservative, then no 

additional actions will be taken. 

Alternatively, if the staff cannot conclude 

that there is reasonable assurance of 

structural integrity, additional action(s) will 

be considered. The NRC will communicate 

with applicable stakeholders, as appropriate. 

Option 2 

The second option involves issuing a 

generic letter (GL) to the licensees operating 

with components forged by ACF. The 

objective of the GL would be to confirm that 

the licensees’ 10 CFR Part 50, Appendix B, 

quality assurance programs have verified that 

the components produced by ACF comply 

with the applicable NRC regulations and 

ASME Code requirements. The GL would 

request that the licensees (1) provide the 

documentation necessary to confirm that the 

components in question meet all applicable 

NRC regulations and ASME Code 

requirements and (2) describe how their 10 

CFR Part 50, Appendix B, quality assurance 

programs verified that the components 

complied with all applicable NRC regulations 

and ASME Code requirements, specifically, 

those related to the manufacturing of the 

components relevant to the CMAC topic. 

Section II.C of this Director’s Decision 

provides the regulatory requirements and the 

10 CFR Part 50, Appendix B, quality 

assurance program, as they relate to the 

CMAC topic. A GL can require a written 

response in accordance with 10 CFR 50.54(f). 

Option 3 

The third option involves issuing an order 

to the licensees operating with inservice 

components produced by ACF. The order 

would require licensees with components 

from ACF to conduct nondestructive 

examinations of these inservice components 

during the next scheduled outage. The 

objective of the examination would be to 

verify the condition of the components (e.g., 

no unacceptable flaw or indications) and to 

verify carbon levels. If the nondestructive 

examinations reveal a condition that is 

adverse to safety or does not conform to 

requirements, the plant would not be allowed 

to restart until the issue is addressed and 

until the NRC grants its approval. 

Option 4 

Option 4 is identical to Option 3, except 

that the NRC orders would require immediate 

plant shutdowns to perform the inspections. 

This Option would be preferable in the case 

of an immediate safety issue posing a clearly 

demonstrated significant and immediate risk 

to an operating plant. NRR OI LIC–504 

defines a risk significant condition as 

significant enough to warrant immediate 

action if the calculated large early release 

frequency (LERF) is on the order of 1×10¥4

yr¥1. 

Assessment of Options 

The NRC staff evaluated the relative merits 

of the four options discussed in the 

preceding section. The staff has concluded 

that any of the four options proposed will 

adequately address the possible safety impact 

to the U.S. nuclear power reactor fleet posed 

by potential regions of CMAC in components 

produced by ACF. However, all four options 

are not equivalent or warranted, as discussed 

below. 

Option 1: Evaluate and Monitor 

To properly assess this option, the NRC 

assessed each of the five principles of the 

risk-informed decision-making process 

within the context of this option. 

Principle 1—Compliance with Existing 

Regulations 

A licensee is responsible for ensuring that 

the applicable regulatory and technical 

requirements are appropriately identified in 

the procurement documentation and for 

evaluating whether the purchased items, 

upon receipt, conform to the procurement 

documentation, in accordance with 10 CFR 

part 50, Appendix B. The NRC expects that 

licensees and vendors subject to NRC 

jurisdiction affected by the potential 

presence of CMAC have verified compliance 

with applicable NRC requirements and 

regulations for each potentially affected 

component or, alternatively, performed an 

appropriate evaluation that concludes that 

the condition is not adverse to safety. The 

NRC has not received a 10 CFR part 21 

notification from a component supplier or 

licensee associated with CMAC. The ongoing 

evaluations have not yet determined that a 

deviation exists under 10 CFR part 21. The 

NRC confirms licensee and vendor 

compliance with NRC requirements through 

submitted reports, routine inspections, and 

continuous oversight provided by the plant 

resident inspector. For example, the NRC 

reviews 10 CFR part 21 evaluations and the 

response to operational experience routinely 

as part of the Reactor Oversight Process 

(ROP). Specifically, IP 71152,36 ‘‘Problem 

Identification and Resolution,’’ provides 

guidance on reviewing licensee evaluations 

to ensure that potential supplier deviations 

are adequately captured to identify and 

address potential defects. A review of the 10 

CFR part 21 process is also part of the vendor 

inspection program. Any non-compliances 

identified through NRC oversight activities 

are addressed through the enforcement 

program to ensure compliance is restored. In 

addition, safety concerns identified through 

NRC’s oversight activities may be escalated, 

such as to conduct a reactive inspection or 

to issue a Confirmatory Action Letter or 

Safety Order. Therefore, Principle 1 is 

satisfied for Option 1. 

Principle 2—Consistency with the 

Defense-in-Depth Philosophy 

The aspect of defense-in-depth of relevance 

to the potential presence of CMAC in regions 

of RCS components is ‘‘barrier integrity.’’ The 

reactor coolant pressure boundary is one of 

the three principal fission-product release 

barriers in a U.S. plant. Under 10 CFR 50.61a, 

the NRC established a 95th percentile TWCF 

of less than 1×10¥6 yr¥1 and a CDF of less 

than 1×10¥6 as acceptable RPV failure 

probabilities. The conservative assessment 

performed by the industry and described 

earlier showed that the probability of 

compromising the barrier integrity function 

for the inservice U.S. components of interest 

are significantly below these acceptance 

levels. If a design-basis accident were to 

compromise the pressure boundary, the 

remaining two independent fission-product 

release barriers (i.e., fuel cladding and 

containment) would still provide adequate 

defense-in-depth. The NRC has reasonable 

assurance that U.S. plants with components 

produced by ACF maintain adequate 

defense-in-depth. Therefore, Principle 2 is 

satisfied for Option 1. 

Principle 3—Maintenance of Adequate 

Safety Margins 

A region of CMAC in a component could 

reduce the margin against fracture. However, 

it has been shown that this reduction in 


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margin does not affect the safe operation of 

the inservice components being evaluated. 

The ASN evaluation described earlier 

determined that the safety margin against fast 

fracture is maintained in all conditions 

analyzed. Industry determined in MRP-417 

that the CMAC levels necessary to be 

considered significant to safety are more than 

200 percent of those observed in 

components. Based on its review of these 

evaluations, the NRC has reasonable 

assurance that U.S. plants with components 

produced by ACF maintain sufficient safety 

margins. Therefore, Principle 3 is satisfied for 

Option 1. 

Principle 4—Demonstration of 

Acceptable Levels of Risk 

If it is conservatively assumed that the 

TWCF equates to the LERF (neglecting 

mitigating factors), the calculated 95th

percentile TWCF for components with CMAC 

and thus the LERF is less than 1×10¥6 yr¥1. 

Because this is below the immediate safety 

determination limit, there is no immediate 

safety concern. Therefore, Principle 4 is 

satisfied for Option 1. 

Principle 5—Implementation of Defined 

Performance Measurement Strategies 

Because there is no indication that the U.S. 

inservice components produced by ACF are 

noncompliant with the applicable regulations 

and because the NRC has reasonable 

assurance that defense-in-depth, safety 

margins, and risk levels are adequately 

maintained, the current monitoring programs 

at the plants are adequate, and additional 

performance measurement strategies are not 

warranted. However, the NRC staff would 

continue to monitor the U.S. nuclear industry 

and international activities related to the 

CMAC topic to analyze any new information 

to determine whether additional performance 

measurement strategies are necessary. 

Therefore, Principle 5 is satisfied for Option 

1. 

Option 2: Issue a Generic Communication 

This option reinforces the regulatory 

determination made in Option 1 by issuing 

a GL requesting that the documentation and 

evaluations performed by licensees and their 

component suppliers conclude that the 

components produced by ACF do not have 

defects or deviations that pose a substantial 

safety hazard. The NRC would not expect the 

information collected in the response to a GL 

to change any of the conclusions reached in 

Option 1, including those related to 

defense-in-depth, safety margins, or risk-level 

determinations. Therefore, all five principles 

of risk-informed decision-making would also 

be satisfied for Option 2. Additionally, the 

relevant vendors have informed the affected 

licensees of the CMAC topic. Vendors and 

licensees must meet their 10 CFR part 21 

evaluation and reporting responsibilities if 

the condition warrants such action. As part 

of the ROP and vendor inspection program, 

the NRC reviews these evaluations for 

adequacy. 

Option 3: Issue Orders Requiring Inspections 

This option reinforces the determinations 

made in Option 1 by performing inspections 

to confirm that an appropriate degree of 

conservatism was used in the evaluations of 

the potential impact of CMAC on U.S. 

components produced by AFC. The NRC 

would not expect the information collected 

by performing nondestructive examinations 

of the inservice components to significantly 

affect the defense-in-depth, safety margins, or 

risk-level determinations made in Option 1. 

Therefore, all five principles of risk-informed 

decision-making would also be satisfied for 

Option 3. 

Option 4: Issue Orders Suspending 

Operation 

In evaluating the international, U.S. 

industry, and NRC safety assessments, the 

NRC determined that the impact of CMAC on 

the integrity of the U.S.-forged components 

in question is small and that the calculated 

95th percentile TWCF for PTS and the CPF for 

normal operating conditions fall below the 

NRC’s safety criteria of 1×10¥6 yr¥1 and 

1×10¥6, respectively. Because the 

assumption that the TWCF is equivalent to 

the LERF because of mitigating factors is 

extremely conservative, the results indicate 

that the impacts of CMAC would result in a 

risk of LERF less than 1×10¥4 yr¥1. 

Therefore, because the NRC’s risk criterion to 

shut down a plant is not met, the agency 

dismissed Option 4 without an evaluation of 

the five principles of risk-informed 

decision-making. 

Final Assessment 

The staff determined that Option 1 was the 

most appropriate action based on the 

material and processing information 

reviewed by the staff during the vender 

inspection of AREVA, experimental data and 

evaluation reported by ASN, PFM analyses 

conducted by the industry, the staff’s review 

of the open literature on CMAC in steel 

ingots and its effect on performance, and an 

evaluation demonstrating that Option 1 

satisfies all five key principles of 

risk-informed decision-making. Additionally, 

this compilation of information reviewed 

affirms the staff’s preliminary safety 

assessment that the safety significance of 

CMAC to U.S. plants appears to be negligible 

and does not warrant immediate action. If 

new information becomes available that calls 

into question the conservatism of the 

evaluations supporting Option 1 or the 

regulatory compliance of the plants with 

inservice components from ACF, the NRC 

staff will reevaluate the need for additional 

actions. The staff’s evaluation dated February 

22, 2018, documents the NRC’s full 

evaluation of the CMAC topics as it relates 

to plants operating in the United States. 

E. Evaluation of the Petitioners’ Requests 

Petitioners’ Request 1: Suspend power 

operations of U.S. nuclear power plants that 

rely on ACF components and subcontractors 

pending a full inspection (including 

nondestructive examination by ultrasonic 

testing) and material testing. If carbon 

anomalies (‘‘carbon segregation’’ or ‘‘carbon 

macrosegregation’’) in excess of the 

design-basis specifications for at-risk 

component parts are identified, require the 

licensee to do one of the following: 

a. replace the degraded at-risk 

component(s) with quality certified 

components, or 

b. for those at-risk degraded components 

that a licensee seeks to allow to remain 

in-service, make application through the 

license amendment request process to 

demonstrate that a revised design-basis 

is achievable and will not render the 

in-service component unacceptably 

vulnerable to fast fracture failure at any 

time, and in any credible service 

condition, throughout the current 

license of the power reactor. 

NRC Response: 

This request is essentially identical to 

Option 4 described above. The NRC has 

determined, through its PFM analyses, that 

the expected impact of CMAC on the LERF 

is less than 1×10¥6 yr¥1. Therefore, the risk 

criterion to shut down a plant is not met. 

Petitioners’ Request 2: Alternatively 

modify the operating licenses to require the 

affected operators to perform the requested 

emergency enforcement actions at the next 

scheduled outage. 

NRC Response: 

This request is essentially identical to 

Option 3 described above. As discussed 

above, performing nondestructive 

examinations of the inservice components is 

not expected to provide information that 

would significantly affect the defense-indepth, safety margins, or risk-level 

determinations that would be provided by 

continued monitoring and evaluation of new 

information. 

Petitioners’ Request 3: Issue a letter to all 

U.S. light-water reactor operators under 10 

CFR 50.54(f) requiring licensees to provide 

the NRC with information under oath and 

affirming specifically how U.S. operators are 

reliably monitoring contractors and 

subcontractors for the potential carbon 

segmentation anomaly in the supply chain 

and the reliability of the quality assurance 

certification of those components, and 

publicly release the responses. 

NRC Response: 

This request is essentially identical to 

Option 2 described above. As discussed 

above, the information collected through a 10 

CFR 50.54(f) request for information or a GL 

is not expected to change any of 

defense-in-depth, safety margins, or risk-level 

determinations that would be provided by 

continued monitoring and evaluation of new 

information. In addition, the relevant 

vendors and licensees must meet their 10 

CFR Part 21 evaluation and reporting 

responsibilities if the condition warrants 

such action. As part of the ROP and vendor 

inspection program, the NRC reviews these 

evaluations for adequacy. 

Petitioners’ Request 4: [The Petitioners 

added Crystal River Unit 3 to the plants for 

which they requested actions, which include 

the following]: 

a. Confirm the sale, delivery, quality 

control and quality assurance 

certification and installation of the 

replacement reactor pressure vessel 

head as supplied to Crystal River Unit 


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37See ADAMS Accession No. ML17040A100. 38See ADAMS Accession No. ML17009A278. 

3 by then Framatome and now 

AREVA-Le Creusot Forge industrial 

facility in Charlon-St. Marcel, France 

and; 

b. With completion and confirmation [of 

the above Crystal River Unit 3 actions], 

the modification of Duke Energy’s 

current license for the permanently 

closed Crystal River Unit 3 nuclear 

power station in Crystal River, Florida, 

to inspect and conduct the appropriate 

material test(s) for carbon 

macrosegregation on sufficient samples 

harvested from the installed and now in 

service irradiated Le Creusot Forge 

reactor pressure vessel head [sic]. The 

Petitioners assert that the appropriate 

material testing include OES. 

NRC Response: 

AREVA did not identify Crystal River Unit 

3 as a plant that contained components from 

ACF,37 38 and the staff has not confirmed that 

this unit contained any forgings 

manufactured from ingots produced by ACF. 

In addition, Crystal River Unit 3 is currently 

shut down and in the process of 

decommissioning. Therefore, the Petitioners’ 

requests 1, 2, 3, and 4(a) do not apply to this 

plant. However, the acquisition and 

subsequent testing of irradiated and aged 

plant material from decommissioned plants 

could be a valuable research activity that 

might offer useful scientific information on 

the progress of aging mechanisms. The 

harvesting of reactor vessel material from 

plants that have been permanently shut 

down can be a complex and 

radiation-dose-intensive effort. The NRC’s 

Office of Nuclear Regulatory Research has 

previously obtained samples appropriate for 

testing from shutdown plants. In regard to 

this request, the NRC may, in the future, seek 

to purchase samples. However, the identified 

facility has ceased operations, and there is no 

safety concern at those facilities that justifies 

enforcement-related action (i.e., to modify, 

suspend, or revoke the license) to give the 

NRC reasonable assurance of the adequate 

protection of public health and safety. 

III. Conclusion 

Based on the evaluations provided above, 

and documented in the February 22, 2018, 

NRC memorandum, the NRR Director has 

determined that the actions requested by the 

Petitioners, will not be granted in whole or 

in part. 

As provided for in 10 CFR 2.206(c), a copy 

of this Director’s Decision will be filed with 

the Secretary of the Commission for the 

Commission to review. As provided for by 

this regulation, the decision will constitute 

the final action of the Commission 25 days 

after the date of the decision unless the 

Commission, on its own motion, institutes a 

review of the decision within that time. 

Dated at Rockville, MD, this 2nd day of 

August 2018. 

For the Nuclear Regulatory Commission. 

Brian E. Holian, 

Acting Director, Office of Nuclear Reactor 

Regulation. 

Attachment: 

List of Affected Reactors 

LIST OF POWER REACTORS AFFECTED BY THE PETITION

Plant Docket No. 

Facility 

operating 

license No. 

Prairie Island Nuclear Generating Plant, Unit 1 ...................................................................................................... 05000282 DPR–42 

Prairie Island Nuclear Generating Plant, Unit 2 ...................................................................................................... 05000306 DPR–60 

Arkansas Nuclear One, Unit 2 ................................................................................................................................ 05000368 NPF–6 

Beaver Valley Power Station, Unit 1 ....................................................................................................................... 05000334 DPR–66 

North Anna Power Station, Unit 1 ........................................................................................................................... 05000338 NPF–4 

North Anna Power Station, Unit 2 ........................................................................................................................... 05000339 NPF–7 

Surry Power Station, Unit 1 ..................................................................................................................................... 05000280 DPR–32 

Comanche Peak Nuclear Power Plant, Unit 1 ........................................................................................................ 05000445 NPF–87 

V.C. Summer Nuclear Station, Unit 1 ..................................................................................................................... 05000395 NPF–12 

Joseph M. Farley Nuclear Plant, Unit 1 .................................................................................................................. 05000348 NPF–2 

Joseph M. Farley Nuclear Plant, Unit 2 .................................................................................................................. 05000364 NPF–8 

South Texas Project, Unit 1 .................................................................................................................................... 05000498 NPF–76 

South Texas Project, Unit 2 .................................................................................................................................... 05000499 NPF–80 

Sequoyah Nuclear Plant, Unit 1 .............................................................................................................................. 05000327 DPR–77 

Watts Bar Nuclear Plant, Unit 1 .............................................................................................................................. 05000390 NPF–90 

Millstone Power Station, Unit 2 ............................................................................................................................... 05000336 NPF–65 

Saint Lucie Plant, Unit 1 .......................................................................................................................................... 05000335 DPR–67 

Crystal River Unit 3 Nuclear Generating Plant ....................................................................................................... 05000302 DPR–72 

[FR Doc. 2018–17131 Filed 8–9–18; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[Docket Nos. 50–382 and 72–75; NRC–2017– 

0239] 

In the Matter of Entergy Louisiana, LLC 

and Entergy Operations, Inc. Waterford 

Steam Electric Station, Unit 3, and 

Independent Spent Fuel Storage 

Installation Facility 

AGENCY: Nuclear Regulatory 

Commission. 

ACTION: Indirect transfer of license; 

order. 

SUMMARY: The U.S. Nuclear Regulatory 

Commission (NRC) is issuing an order to 

permit the indirect transfer of 

membership interests in Entergy 

Louisiana, LLC (ELL; the owner of 

Waterford Steam Electric Station, Unit 

3, and the independent spent fuel 

storage installation facility) to the extent 

ELL is affected by the addition of 

Entergy Arkansas, LLC; Entergy 

Mississippi, LLC; and Entergy New 

Orleans, LLC to Entergy Utility Holding 

Company, LLC (EUHC). Upon execution 

of the transfer, these changes will result 

in additional members of EUHC that 

may dilute the resources and voting 

power of its members, including ELL. 

DATES: The order was issued on August 

1, 2018, and is effective for one year. 

ADDRESSES: Please refer to Docket ID 

NRC–2017–0239 when contacting the 

NRC about the availability of 

information regarding this document. 

You may obtain publicly-available 

information related to this document 

using any of the following methods: 

• Federal Rulemaking Website: Go to 

http://www.regulations.gov and search 

for Docket ID NRC–2017–0239. Address 

questions about NRC dockets to Jennifer 

Borges; telephone: 301–287–9127; 

email: Jennifer.Borges@nrc.gov. For 


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