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, 39790-39797 [2018-17131]
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Federal Register / Vol. 83, No. 155 / Friday, August 10, 2018 / Notices
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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 https://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.
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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
Nuclear Regulatory
Commission.
ACTION: Director’s decision under 10
CFR 2.206; issuance.
AGENCY:
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
https://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
SUMMARY:
PO 00000
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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
https://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.’’
1 See Agencywide Documents Access and
Management System (ADAMS) Accession No.
ML17025A180.
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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
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Prairie Island, Units 1 and 2 (MN)
Steam generators
Arkansas Nuclear One, Unit 2
(AR).
Beaver Valley, Unit 1 (PA) ...........
North Anna, Units 1 and 2 (VA) ...
Surry, Unit 1 (VA) .........................
Crystal River, Unit 3 (FL) .............
Beaver Valley, Unit 1 (PA) ...........
Millstone, Unit 2 (CT).
Comanche Peak, Unit 1 (TX) .......
V.C. Summer (SC) .......................
Farley, Units 1 and 2 (AL).
South Texas, Units 1 and 2 (TX).
Sequoyah, Unit 1 (TN).
Watts Bar, Unit 1 (TN).
Saint Lucie, Unit 1 (FL).
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.
2 See
ADAMS Accession No. ML17052A032.
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.
3 See
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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
8 See
ADAMS Accession No. ML17184A058.
ADAMS Accession No. ML17187A026.
10 The petition incorrectly states that JCFC is a
subcontractor to ACF.
11 See the report titled ‘‘Irregularities and
Anomalies Relating to the Forged Components of Le
Creusot Forge,’’ dated September 26, 2016, Large
9 See
PO 00000
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Steam pressurizers
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
and Associates Consulting Engineers, London,
England (available at https://
www.largeassociates.com/CZ3233/Note_
LargeAndAssociates_EN_26092016.pdf).
12 See ADAMS Accession No. ML17039A501.
13 See ADAMS Accession No. ML17081A418.
14 See ADAMS Accession No. ML17052A033.
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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 https://
www.nrc.gov/reading-rm/adams.html.
15 See
ADAMS Accession No. ML18017A441.
16 See
ADAMS Accession No. ML17171A108.
ADAMS Accession No. ML17171A106.
18 See ADAMS Accession No. ML17142A334.
19 See ADAMS Accession No. ML17198A329.
20 See ADAMS Accession No. ML18107A402.
17 See
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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
´
ˆ ´
nuclear safety authority, Autorite de Surete
´
Nucleaire (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.
21 See
22 See
PO 00000
ADAMS Accession No. ML18012A156.
ADAMS Accession No. ML12192A058.
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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, Electricite
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
23 See
ADAMS Accession No. ML17040A100.
ASN/Institut de Radioprotection et de
´
ˆ ´
Surete Nucleaire (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. https://
www.french-nuclear-safety.fr/Media/Files/00Publications/Report-to-the-Advisory-Committee-ofExperts-for-Nuclear-Pressure-Equipment.
24 See
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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
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/CODEPDEP-2016-019209-advisorycommitte24june2016summaryreport.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. https://
www.irsn.fr/FR/expertise/rapports_gp/Documents/
GPESPN/IRSN-ASNDEP_GPESPN-Report_pressurevessel-FA3_201706.pdf.
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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
27 See
ADAMS Accession No. ML072830076.
ADAMS Accession No. ML072820691.
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).
39793
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.
28 See
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30 See
ADAMS Accession No. ML072830076.
ADAMS Accession No. ML072820691.
32 See ADAMS Accession No. ML17124A575.
33 See ADAMS Accession No. ML13148A361.
34 See ADAMS Accession No. ML113190538.
31 See
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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
35 See
ADAMS Accession No. ML14035A143.
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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
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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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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
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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
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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
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39795
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
36 See
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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.
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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
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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:
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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
E:\FR\FM\10AUN1.SGM
10AUN1
39797
Federal Register / Vol. 83, No. 155 / Friday, August 10, 2018 / Notices
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.
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.
III. Conclusion
Based on the evaluations provided above,
and documented in the February 22, 2018,
Attachment:
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.
List of Affected Reactors
LIST OF POWER REACTORS AFFECTED BY THE PETITION
Plant
Docket No.
Prairie Island Nuclear Generating Plant, Unit 1 ......................................................................................................
Prairie Island Nuclear Generating Plant, Unit 2 ......................................................................................................
Arkansas Nuclear One, Unit 2 ................................................................................................................................
Beaver Valley Power Station, Unit 1 .......................................................................................................................
North Anna Power Station, Unit 1 ...........................................................................................................................
North Anna Power Station, Unit 2 ...........................................................................................................................
Surry Power Station, Unit 1 .....................................................................................................................................
Comanche Peak Nuclear Power Plant, Unit 1 ........................................................................................................
V.C. Summer Nuclear Station, Unit 1 .....................................................................................................................
Joseph M. Farley Nuclear Plant, Unit 1 ..................................................................................................................
Joseph M. Farley Nuclear Plant, Unit 2 ..................................................................................................................
South Texas Project, Unit 1 ....................................................................................................................................
South Texas Project, Unit 2 ....................................................................................................................................
Sequoyah Nuclear Plant, Unit 1 ..............................................................................................................................
Watts Bar Nuclear Plant, Unit 1 ..............................................................................................................................
Millstone Power Station, Unit 2 ...............................................................................................................................
Saint Lucie Plant, Unit 1 ..........................................................................................................................................
Crystal River Unit 3 Nuclear Generating Plant .......................................................................................................
[FR Doc. 2018–17131 Filed 8–9–18; 8:45 am]
ACTION:
BILLING CODE 7590–01–P
Indirect transfer of license;
order.
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
SUMMARY:
NUCLEAR REGULATORY
COMMISSION
daltland on DSKBBV9HB2PROD with NOTICES
[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
Nuclear Regulatory
Commission.
AGENCY:
37 See
ADAMS Accession No. ML17040A100.
VerDate Sep<11>2014
19:03 Aug 09, 2018
Jkt 244001
38 See
PO 00000
Fmt 4703
Sfmt 4703
DPR–42
DPR–60
NPF–6
DPR–66
NPF–4
NPF–7
DPR–32
NPF–87
NPF–12
NPF–2
NPF–8
NPF–76
NPF–80
DPR–77
NPF–90
NPF–65
DPR–67
DPR–72
may dilute the resources and voting
power of its members, including ELL.
The order was issued on August
1, 2018, and is effective for one year.
DATES:
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
https://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
ADDRESSES:
ADAMS Accession No. ML17009A278.
Frm 00141
05000282
05000306
05000368
05000334
05000338
05000339
05000280
05000445
05000395
05000348
05000364
05000498
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05000327
05000390
05000336
05000335
05000302
Facility
operating
license No.
E:\FR\FM\10AUN1.SGM
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]]>Agencies
[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]
-----------------------------------------------------------------------
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[dash]available information related to this document
using any of the following methods:
Federal Rulemaking Website: Go to https://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 https://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.
---------------------------------------------------------------------------
\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
----------------------------------------------------------------------------------------------------------------
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).
----------------------------------------------------------------------------------------------------------------
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\
---------------------------------------------------------------------------
\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 https://www.largeassociates.com/CZ3233/Note_LargeAndAssociates_EN_26092016.pdf).
---------------------------------------------------------------------------
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.
---------------------------------------------------------------------------
\12\ See ADAMS Accession No. ML17039A501.
\13\ See ADAMS Accession No. ML17081A418.
---------------------------------------------------------------------------
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:
---------------------------------------------------------------------------
\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\
---------------------------------------------------------------------------
\15\ See ADAMS Accession No. ML18017A441.
---------------------------------------------------------------------------
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.
---------------------------------------------------------------------------
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\
---------------------------------------------------------------------------
\18\ See ADAMS Accession No. ML17142A334.
\19\ See ADAMS Accession No. ML17198A329.
---------------------------------------------------------------------------
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.
---------------------------------------------------------------------------
\20\ See ADAMS Accession No. ML18107A402.
---------------------------------------------------------------------------
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
https://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\
---------------------------------------------------------------------------
\21\ See ADAMS Accession No. ML18012A156.
\22\ See ADAMS Accession No. ML12192A058.
---------------------------------------------------------------------------
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.
---------------------------------------------------------------------------
\23\ See ADAMS Accession No. ML17040A100.
---------------------------------------------------------------------------
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.
---------------------------------------------------------------------------
\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.
https://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. https://www.irsn.fr/FR/expertise/rapports_gp/Documents/GPESPN/IRSN-ASNDEP_GPESPN-Report_pressure-vessel-FA3_201706.pdf.
---------------------------------------------------------------------------
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.
---------------------------------------------------------------------------
\27\ See ADAMS Accession No. ML072830076.
\28\ See ADAMS Accession No. ML072820691.
---------------------------------------------------------------------------
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.
---------------------------------------------------------------------------
\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).
---------------------------------------------------------------------------
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.
---------------------------------------------------------------------------
\30\ See ADAMS Accession No. ML072830076.
\31\ See ADAMS Accession No. ML072820691.
---------------------------------------------------------------------------
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.
---------------------------------------------------------------------------
\32\ See ADAMS Accession No. ML17124A575.
\33\ See ADAMS Accession No. ML13148A361.
\34\ See ADAMS Accession No. ML113190538.
---------------------------------------------------------------------------
[[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.
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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
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[FR Doc. 2018-17131 Filed 8-9-18; 8:45 am]
BILLING CODE 7590-01-P
