[Federal Register Volume 83, Number 155 (Friday, August 10, 2018)]
[Notices]
[Pages 39790-39797]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2018-17131]


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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.

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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[dash]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: [email protected]. 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 publicly-available 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 [email protected]. 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: [email protected].

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.''

[[Page 39791]]

The Petitioners supplemented their petition by e[dash]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[dash]related 
components and potentially falsified quality assurance documentation 
supplied by AREVA[dash]Le Creusot Forge (ACF) and its subcontractor, 
Japan Casting and Forging Corporation (JCFC).\10\ Table 1 lists 
potentially affected components and the at[dash]risk reactors 
identified in the petition.
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    \1\ See Agencywide Documents Access and Management System 
(ADAMS) Accession No. ML17025A180.
    \2\ See ADAMS Accession No. ML17052A032.
    \3\ See ADAMS Accession No. ML17068A061.
    \4\ See ADAMS Accession No. ML17067A562.
    \5\ See ADAMS Accession No. ML17174A087.
    \6\ See ADAMS Accession No. ML17174A788.
    \7\ See ADAMS Accession No. ML17179A288.
    \8\ See ADAMS Accession No. ML17184A058.
    \9\ See ADAMS Accession No. ML17187A026.
    \10\ The petition incorrectly states that JCFC is a 
subcontractor to ACF.

                          Table 1--List of Potentially Affected Components and Reactors
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                                         Replacement reactor
       Reactor pressure vessels         pressure vessel heads       Steam generators        Steam pressurizers
----------------------------------------------------------------------------------------------------------------
Prairie Island, Units 1 and 2 (MN)...  Arkansas Nuclear One,    Beaver Valley, Unit 1    Millstone, Unit 2 (CT).
                                        Unit 2 (AR).             (PA).
                                       Beaver Valley, Unit 1    Comanche Peak, Unit 1    Saint Lucie, Unit 1
                                        (PA).                    (TX).                    (FL).
                                       North Anna, Units 1 and  V.C. Summer (SC).......  .......................
                                        2 (VA).
                                       Surry, Unit 1 (VA).....  Farley, Units 1 and 2
                                                                 (AL).
                                       Crystal River, Unit 3    South Texas, Units 1
                                        (FL).                    and 2 (TX).
                                                                Sequoyah, Unit 1 (TN)..
                                                                Watts Bar, Unit 1 (TN).
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    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[dash]basis specifications for at[dash]risk component parts 
are identified, require the licensee to do one of the following:
    a. Replace the degraded at[dash]risk component(s) with 
quality[dash]certified components.
    b. For those at[dash]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[dash]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[dash]Le Creusot Forge industrial facility in 
Charlon[dash]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\
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    \11\ See 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).
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    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.
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    \12\ See ADAMS Accession No. ML17039A501.
    \13\ See ADAMS Accession No. ML17081A418.
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    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:
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    \14\ See ADAMS Accession No. ML17052A033.

 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

[[Page 39792]]

impact on structural integrity are described in the staff's 
evaluation dated February 22, 2018.\15\
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    \15\ See ADAMS Accession No. ML18017A441.
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 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[dash]sponsored Regulatory Information Conference on March 15, 
2017.16 17

    \16\ See ADAMS Accession No. ML17171A108.
    \17\ See ADAMS Accession No. ML17171A106.
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    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\
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    \18\ See ADAMS Accession No. ML17142A334.
    \19\ See ADAMS Accession No. ML17198A329.
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    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.
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    \20\ See ADAMS Accession No. ML18107A402.
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    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[dash]800[dash]397[dash]4209 or 301[dash]415[dash]4737 
or by e[dash]mail to [email protected].

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[dash]401, ``NRR[dash]NRO Reactor 
Operating Experience Program,'' Revision 3, addresses the specific 
implementation of the Reactor OpE Program.\22\
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    \21\ See ADAMS Accession No. ML18012A156.
    \22\ See ADAMS Accession No. ML12192A058.
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    As reported in internal NRC communications, AREVA notified 
France's nuclear safety authority, Autorit[eacute] de 
S[ucirc]ret[eacute] Nucl[eacute]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[dash]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, [Eacute]lectricit[eacute] 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[dash]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.
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    \23\ See ADAMS Accession No. ML17040A100.
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    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\

[[Page 39793]]

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[dash]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[dash]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[dash]term serviceability could 
not be confirmed and that the head should be replaced after a few 
years of operation.
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    \24\ See ASN/Institut de Radioprotection et de 
S[ucirc]ret[eacute] Nucl[eacute]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-of-Experts-for-Nuclear-Pressure-Equipment.
    \25\ See 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/CODEP-DEP-2016-019209-advisorycommitte24june2016-summaryreport.pdf.
    \26\ See 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_pressure-vessel-FA3_201706.pdf.
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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[dash]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[dash]sponsored 
Regulatory Information Conference.
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    \27\ See ADAMS Accession No. ML072830076.
    \28\ See ADAMS Accession No. ML072820691.
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    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)[dash]417.\29\ The other tasks are still under development with 
the expected release of the report(s) in 2018.
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    \29\ EPRI 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).
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    The MRP[dash]417 addresses the structural significance of the 
potential presence of CMAC in large, forged pressurized[dash]water 
reactor pressure[dash]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[dash]year operating 
interval. The assessment was made using the NRC risk safety 
criterion of a 95\th\ percentile through[dash]wall crack frequency 
(TWCF) of less than 1x10-\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 1x10-\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.
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    \30\ See ADAMS Accession No. ML072830076.
    \31\ See ADAMS Accession No. ML072820691.
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    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 95\th\ 
percentile TWCF of less than 1x10-\6\ yr-\1\ 
for PTS events and a CPF of less than 1x10-\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[dash]water reactors.
    In March 2017, an NRC inspection team performed a 
limited[dash]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[dash]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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    \32\ See ADAMS Accession No. ML17124A575.
    \33\ See ADAMS Accession No. ML13148A361.
    \34\ See ADAMS Accession No. ML113190538.

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[[Page 39794]]

    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[dash]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[dash]informed 
decision[dash]making process in NRR OI LIC[dash]504, ``Integrated 
Risk[dash]Informed Decision[dash]Making Process for Emergent 
Issues,'' Revision 4, dated June 2, 2014,\35\ to evaluate this 
issue.
---------------------------------------------------------------------------

    \35\ See ADAMS Accession No. ML14035A143.
---------------------------------------------------------------------------

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[dash]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[dash]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[dash]informed 
decision[dash]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.[dash]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[dash]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

[[Page 39795]]

existing PFM results and examining them in the context of the NRC's 
approach to the risk[dash]informed decision[dash]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[dash]informed 
decision[dash]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[dash]in[dash]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[dash]in[dash]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 
1x10-\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.
---------------------------------------------------------------------------

    \36\ See ADAMS Accession No. ML053490187.
---------------------------------------------------------------------------

Principle 2--Consistency with the Defense-in[dash]Depth Philosophy

    The aspect of defense[dash]in[dash]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[dash]product release barriers in a U.S. 
plant. Under 10 CFR 50.61a, the NRC established a 95\th\ percentile 
TWCF of less than 1x10-6 yr-1 and a CDF of 
less than 1x10-6 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[dash]basis 
accident were to compromise the pressure boundary, the remaining two 
independent fission[dash]product release barriers (i.e., fuel 
cladding and containment) would still provide adequate 
defense[dash]in[dash]depth. The NRC has reasonable assurance that 
U.S. plants with components produced by ACF maintain adequate 
defense[dash]in[dash]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

[[Page 39796]]

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[dash]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 95\th\ 
percentile TWCF for components with CMAC and thus the LERF is less 
than 1x10-6 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[dash]in[dash]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[dash]in[dash]depth, safety margins, or risk[dash]level 
determinations. Therefore, all five principles of risk[dash]informed 
decision[dash]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[dash]in[dash]depth, safety margins, or risk[dash]level 
determinations made in Option 1. Therefore, all five principles of 
risk[dash]informed decision[dash]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.[dash]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 1x10-6 yr-1 and 1x10-6, 
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 1x10-4 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[dash]informed decision[dash]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[dash]informed decision[dash]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[dash]basis specifications for at[dash]risk component 
parts are identified, require the licensee to do one of the 
following:

    a. replace the degraded at[dash]risk component(s) with quality 
certified components, or
    b. for those at[dash]risk degraded components that a licensee 
seeks to allow to remain in[dash]service, make application through 
the license amendment request process to demonstrate that a revised 
design[dash]basis is achievable and will not render the 
in[dash]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 1x10-6 
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[dash]in-depth, safety 
margins, or risk[dash]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[dash]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[dash]in[dash]depth, safety margins, or 
risk[dash]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

[[Page 39797]]

3 by then Framatome and now AREVA[dash]Le Creusot Forge industrial 
facility in Charlon[dash]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[dash]dose[dash]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[dash]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.
---------------------------------------------------------------------------

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

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