[Federal Register Volume 83, Number 155 (Friday, August 10, 2018)]
[Notices]
[Pages 39790-39797]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2018-17131]
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NUCLEAR REGULATORY COMMISSION
[Docket Nos. 50-368, 50-334, 50-445, 50-302, 50-348, 50-364, 50-336,
50-338, 50-339, 50-282, 50-306, 50-327, 50-498, 50-499, 50-335, 50-280,
50-395, 50-390; NRC-2017-0188]
Entergy Operations, Inc.; FirstEnergy Nuclear Operating Company;
Vistra Operations Company, LLC; Duke Entergy Florida, Southern Nuclear
Operating Company, Inc.; Dominion Nuclear Connecticut, Inc.; Virginia
Electric and Power Company; Northern States Power Company--Minnesota;
South Carolina Electric & Gas Company, Inc.; STP Nuclear Operating
Company; Tennessee Valley Authority
AGENCY: Nuclear Regulatory Commission.
ACTION: Director's decision under 10 CFR 2.206; issuance.
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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) has issued a
director's decision in response to a petition dated January 24, 2017,
filed by Mr. Paul Gunter on behalf of Beyond Nuclear, and representing
numerous public interest groups (collectively, Beyond Nuclear, et al.,
or petitioners), requesting that the NRC take action with regard to
licensees of plants that currently rely on potentially defective
safety-related components and potentially falsified quality assurance
documentation supplied by AREVA-Le Creusot Forge and Japan Casting and
Forging Corporation. The petitioners' requests are included in the
SUPPLEMENTARY INFORMATION section of this document.
DATES: The director's decision was issued on August 2, 2018.
ADDRESSES: Please refer to Docket ID NRC-2017-0188 when contacting the
NRC about the availability of information regarding this document. You
may obtain publicly[dash]available information related to this document
using any of the following methods:
Federal Rulemaking Website: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0188. Address
questions about NRC dockets to Jennifer Borges; telephone: 301-287-
9127; email: [email protected]. For technical questions, contact
the individual listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``Begin Web-based ADAMS
Search.'' For problems with ADAMS, please contact the NRC's Public
Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or
by e-mail to [email protected]. The ADAMS accession number for each
document referenced (if it is available in ADAMS) is provided the first
time that it is mentioned in this document.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
FOR FURTHER INFORMATION CONTACT: Perry Buckberg, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; telephone: 301-415-1383; email: [email protected].
SUPPLEMENTARY INFORMATION: The text of the director's decision is
attached.
Dated at Rockville, Maryland, this 7th day of August 2018.
For the Nuclear Regulatory Commission.
Perry H. Buckberg,
Senior Project Manager, Special Projects and Process Branch, Division
of Operating Reactor Licensing, Office of Nuclear Reactor Regulation.
Attachment--Director's Decision DD-18-03
UNITED STATES OF AMERICA
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
Brian E. Holian, Acting Director
In the Matter of Power Reactor Licensees
Docket Nos.: See Attached List
License Nos.: See Attached List
DIRECTOR'S DECISION UNDER 10 CFR 2.206
I. Introduction
On January 24, 2017,\1\ Mr. Paul Gunter submitted a petition on
behalf of Beyond Nuclear that represents numerous public interest
groups (collectively referred to as the Petitioners) under Title 10
of the Code of Federal Regulations (10 CFR) 2.206, ``Requests for
Action under This Subpart.''
[[Page 39791]]
The Petitioners supplemented their petition by e[dash]mails dated
February 16,\2\ March 6,\3,4\ June 16,\5\ June 22,\6\ June 27,\7\
June 30,\8\ and July 5, 2017.\9\ The June 16 and June 22, 2017,
supplements added the Crystal River Unit 3 Nuclear Generating Plant
(Crystal River Unit 3) to the list of plants subject to the petition
and requested slightly different enforcement actions. The rest of
the supplements did not expand the scope of the petition or request
additional actions that should be considered as a new petition. The
Petitioners asked the U.S. Nuclear Regulatory Commission (NRC) to
take emergency enforcement action at U.S. nuclear power plants that
currently rely on potentially defective safety[dash]related
components and potentially falsified quality assurance documentation
supplied by AREVA[dash]Le Creusot Forge (ACF) and its subcontractor,
Japan Casting and Forging Corporation (JCFC).\10\ Table 1 lists
potentially affected components and the at[dash]risk reactors
identified in the petition.
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\1\ See Agencywide Documents Access and Management System
(ADAMS) Accession No. ML17025A180.
\2\ See ADAMS Accession No. ML17052A032.
\3\ See ADAMS Accession No. ML17068A061.
\4\ See ADAMS Accession No. ML17067A562.
\5\ See ADAMS Accession No. ML17174A087.
\6\ See ADAMS Accession No. ML17174A788.
\7\ See ADAMS Accession No. ML17179A288.
\8\ See ADAMS Accession No. ML17184A058.
\9\ See ADAMS Accession No. ML17187A026.
\10\ The petition incorrectly states that JCFC is a
subcontractor to ACF.
Table 1--List of Potentially Affected Components and Reactors
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Replacement reactor
Reactor pressure vessels pressure vessel heads Steam generators Steam pressurizers
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Prairie Island, Units 1 and 2 (MN)... Arkansas Nuclear One, Beaver Valley, Unit 1 Millstone, Unit 2 (CT).
Unit 2 (AR). (PA).
Beaver Valley, Unit 1 Comanche Peak, Unit 1 Saint Lucie, Unit 1
(PA). (TX). (FL).
North Anna, Units 1 and V.C. Summer (SC)....... .......................
2 (VA).
Surry, Unit 1 (VA)..... Farley, Units 1 and 2
(AL).
Crystal River, Unit 3 South Texas, Units 1
(FL). and 2 (TX).
Sequoyah, Unit 1 (TN)..
Watts Bar, Unit 1 (TN).
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Specifically, the Petitioners asked the NRC to take enforcement
actions consistent with the following:
1. Suspend power operations of U.S. nuclear power plants that
rely on ACF components and subcontractors pending a full inspection
(including nondestructive examination by ultrasonic testing) and
material testing. If carbon anomalies (``carbon segregation'' or
``carbon macrosegregation'' (CMAC)) in excess of the
design[dash]basis specifications for at[dash]risk component parts
are identified, require the licensee to do one of the following:
a. Replace the degraded at[dash]risk component(s) with
quality[dash]certified components.
b. For those at[dash]risk degraded components that a licensee
seeks to allow to remain in service, apply through the license
amendment request process to demonstrate that a revised design basis
is achievable and will not render the inservice component
unacceptably vulnerable to fast fracture failure at any time and in
any credible service condition throughout the current license of the
power reactor.
2. Alternatively modify the licensees' operating licenses to
require the licensees to perform the requested emergency enforcement
actions at the next scheduled outage.
3. Issue a letter to all U.S. light[dash]water reactor operators
under 10 CFR 50.54(f) requiring licensees to provide the NRC with
information under oath and affirming specifically how U.S. operators
are reliably monitoring contractors and subcontractors for the
potential carbon segmentation anomaly in the supply chain and the
reliability of the quality assurance certification of those
components, and publicly release the responses.
The June 16 and June 22, 2017, supplements to the petitions
added Crystal River Unit 3, which is currently shut down, and the
licensee Duke Energy to the list of facilities for which the
Petitioners requested the following fourth NRC action:
a. Confirm the sale, delivery, quality control and quality
assurance certification and installation of the replacement reactor
pressure vessel head as supplied to Crystal River Unit 3 by then
Framatome and now AREVA[dash]Le Creusot Forge industrial facility in
Charlon[dash]St. Marcel, France and;
b. With completion and confirmation [of the above Crystal River
Unit 3 actions], the modification of Duke Energy's current license
for the permanently closed Crystal River Unit 3 nuclear power
station in Crystal River, Florida, to inspect and conduct the
appropriate material test(s) for carbon macrosegregation on
sufficient samples harvested from the installed and now inservice
irradiated Le Creusot Forge reactor pressure vessel head [sic]. The
Petitioners assert that the appropriate material testing include
Optical Emissions Spectrometry (OES).
As the basis of their requests, the Petitioners cited the expert
review by Large and Associates Consulting Engineers that identified
significant irregularities and anomalies in both the manufacturing
process and quality assurance documentation of large reactor
components manufactured by the ACF for French reactors and reactors
in other countries.\11\
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\11\ See the report titled ``Irregularities and Anomalies
Relating to the Forged Components of Le Creusot Forge,'' dated
September 26, 2016, Large and Associates Consulting Engineers,
London, England (available at http://www.largeassociates.com/CZ3233/Note_LargeAndAssociates_EN_26092016.pdf).
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On February 2, 2017,\12\ the Office of Nuclear Reactor
Regulation (NRR) petition manager acknowledged receipt of the
petition and offered an opportunity for the Petitioners to address
NRR's 10 CFR 2.206 Petition Review Board (PRB) to discuss the
petition. The Petitioners accepted the offer, and the meeting was
held on March 8, 2017. The transcript \13\ of that meeting is
publicly available.
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\12\ See ADAMS Accession No. ML17039A501.
\13\ See ADAMS Accession No. ML17081A418.
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On February 8, 2017, the PRB met internally to discuss the
request for immediate actions and informed the Petitioners on
February 13, 2017,\14\ that no actions were warranted at that time
because the NRC has reasonable assurance of public health and safety
and protection of the environment. The basis for the PRB's
determination included the following:
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\14\ See ADAMS Accession No. ML17052A033.
Extent of Condition. Internationally, CMAC has been found
only in components produced by ACF using a specific processing
route. Based on the staff's knowledge as of February 2017, only a
subset of the plants identified in the petition contain components
that may have used the processing route that resulted in the excess
CMAC found in international plants.
Degree of Condition. If CMAC is present in a component, it
occurs in a localized region of the forged component. It is not a
bulk material phenomenon, does not go through thickness, and is not
expected to affect the structural integrity of the component. In
addition, based on the staff's knowledge as of February 2017, the
highest levels of CMAC observed internationally, if present in the
postulated regions of U.S. components, are not expected to alter the
mechanical properties of the material enough to affect the
structural integrity of the components. Destructive examinations of
components containing regions of CMAC have been conducted
internationally to determine how CMAC affects mechanical properties
and such examinations confirm that structural integrity has not been
impacted. A summary of the international investigation is summarized
in II.A below, and details of the investigation and its
[[Page 39792]]
impact on structural integrity are described in the staff's
evaluation dated February 22, 2018.\15\
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\15\ See ADAMS Accession No. ML18017A441.
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Safety Significance. The staff's preliminary safety
assessment concluded that the safety significance of CMAC to the
U.S. nuclear power reactor fleet appears to be negligible. The staff
based its assessment on knowledge of the material processing,
qualitative analysis, compliance of U.S. components with the
American Society of Mechanical Engineers Boiler Pressure and Vessel
Code (ASME Code), and the results of preliminary structural
evaluations. The NRC subsequently presented the basis for this
determination in a technical session, titled ``Carbon
Macrosegregation in Large Nuclear Forgings,'' at the
NRC[dash]sponsored Regulatory Information Conference on March 15,
2017.16 17
\16\ See ADAMS Accession No. ML17171A108.
\17\ See ADAMS Accession No. ML17171A106.
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On April 11, 2017, the PRB met to discuss the petition with
respect to the criteria for consideration under 10 CFR 2.206. Based
on that review, the PRB determined that the petition request meets
the criteria for consideration under 10 CFR 2.206. On May 19, 2017,
the petition manager informed the Petitioners that the initial
recommendation was to accept the petition for review but to refer a
portion of the petition (i.e., the concern of potentially falsified
quality assurance documentation) to the NRC's allegation process for
appropriate action.\18\ The petition manager also offered the
Petitioners an opportunity to comment on the PRB's recommendations.
On July 5, 2017, the petition manager clarified the initial
recommendation and asked for a response as to whether the
Petitioners wanted to address the PRB a second time to comment on
its recommendations. The Petitioners did not request a second
opportunity to address the PRB. Therefore, the PRB's initial
recommendations to accept part of the petition for review under 10
CFR 2.206 and to refer a part to another NRC process became final.
On August 30, 2017, the petition manager issued an acknowledgment
letter to the Petitioners.\19\
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\18\ See ADAMS Accession No. ML17142A334.
\19\ See ADAMS Accession No. ML17198A329.
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By a letter to the Petitioners which copied the licensees dated
June 6, 2018,\20\ the NRC issued the proposed director's decision
for comment. The Petitioners were asked to provide comments within
14 days on any part of the proposed director's decision considered
to be erroneous or any issues in the petition that were not
addressed. The NRC staff did not receive any comments on the
proposed director's decision.
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\20\ See ADAMS Accession No. ML18107A402.
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The petition and other references related to this petition are
available for inspection in the NRC's Public Document Room (PDR),
located at O1F21, 11555 Rockville Pike (first floor), Rockville, MD
20852. Publicly available documents created or received at the NRC
are accessible electronically through ADAMS in the NRC Library at
http://www.nrc.gov/reading-rm/adams.html. Persons who do not have
access to ADAMS or who encounter problems in accessing the documents
located in ADAMS should contact the NRC's PDR reference staff by
telephone at 1[dash]800[dash]397[dash]4209 or 301[dash]415[dash]4737
or by e[dash]mail to [email protected].
II. Discussion
Under the 10 CFR 2.206(b) petition review process, the Director
of the NRC office with responsibility for the subject matter shall
either institute the requested proceeding or shall advise the person
who made the request in writing that no proceeding will be
instituted, in whole or in part, with respect to the request and the
reason for the decision. Accordingly, the decision of the NRR
Director is provided below. As further discussed below, the petition
is denied.
The NRC's policy is to have an effectively coordinated program
to promptly and systematically review relevant domestic and
applicable international operational experience (OpE) information.
The program supplies the means for assessing the significance of OpE
information, offering timely and effective communication to
stakeholders, and applying the lessons learned to regulatory
decisions and programs affecting nuclear reactors. The NRC
Management Directive 8.7, ``Reactor Operating Experience Program,''
dated February 1, 2018, describes the Reactor OpE Program.\21\ The
NRR Office Instruction (OI) LIC[dash]401, ``NRR[dash]NRO Reactor
Operating Experience Program,'' Revision 3, addresses the specific
implementation of the Reactor OpE Program.\22\
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\21\ See ADAMS Accession No. ML18012A156.
\22\ See ADAMS Accession No. ML12192A058.
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As reported in internal NRC communications, AREVA notified
France's nuclear safety authority, Autorit[eacute] de
S[ucirc]ret[eacute] Nucl[eacute]aire (ASN), of an anomaly in the
composition of the steel in certain zones of the reactor pressure
vessel (RPV) upper and lower heads of the Flamanville Nuclear Power
Plant (Flamanville), Unit 3, in Manche, France. Both the upper and
lower vessel heads were manufactured by ACF. According to ASN,
chemical and mechanical property testing performed by AREVA in late
2014 (on a vessel head similar to that of the Flamanville European
Pressurized Reactor (EPR)) revealed a zone of high carbon
concentration (0.30 percent as opposed to a target value of 0.22
percent), which led to lower than expected mechanical toughness
values in that area. Initial measurements confirmed the presence of
this anomaly in the Flamanville, Unit 3, RPV upper and bottom heads.
In accordance with the process described in NRR OI LIC[dash]401,
the NRC's Reactor OpE Program staff ensured that the appropriate
technical experts within the NRC were aware of the issue and were
evaluating these issues for relevance to the U.S. industry. In
addition, the NRC has strong collaboration with the international
community and was separately in contact with ASN to discuss this
issue.
A. Description of the Issue
The CMAC is a known phenomenon that takes place during the
casting of large ingots. The CMAC is a material heterogeneity in the
form of a chemical (i.e., carbon) gradient that deviates from the
nominal composition and may exceed specification limits. Portions of
the ingot containing CMAC that exceed specification limits (positive
CMAC) are purposefully removed and discarded as part of the material
processing. Regions of positive CMAC that are not appropriately
removed result in localized regions near the surface of the final
component with higher strength and lower toughness relative to the
bulk material.
In April 2015, regions of positive CMAC were discovered in EPR
RPV heads that were manufactured for the Flamanville plant. The ACF
had produced the forgings for the Flamanville upper and lower RPV
heads. The discovery of the CMAC in the heads prompted ASN to ask
the operator, [Eacute]lectricit[eacute] de France S.A. (EDF)
(Electricity of France), to review inservice forged components at
all of its plants to determine the potential extent of the
condition. The review identified steam generator (SG) channel heads
(also commonly referred to as SG primary heads) produced by ACF and
JCFC as the components most likely to contain a region of CMAC. The
ASN requested that nondestructive testing be performed on these SG
channel heads to characterize the carbon content and confirm the
absence of unacceptable flaws.
On October 18, 2016, ASN ordered the acceleration of the
nondestructive testing of the potentially affected ACF and JCFC SG
channel heads, which required completion of the remaining
nondestructive testing within 3 months. The discovery of higher than
expected carbon values measured on an inservice SG channel head
produced by JCFC prompted the accelerated schedule. As a result, to
perform the required nondestructive tests, EDF had to shut down its
plants before their scheduled outages.
AREVA Inc. (AREVA Inc. or AREVA), located in Lynchburg, VA,
provides safety[dash]related products and services for U.S.
operating nuclear power plants, including replacements for reactor
coolant pressure boundary components. On February 3, 2017,\23\ AREVA
Inc. submitted a list to the NRC of the U.S. reactors that have
received components fabricated with forgings from ACF. Operating
U.S. plants have no known components from JCFC.
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\23\ See ADAMS Accession No. ML17040A100.
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In September 2015, June 2016, and June 2017, ASN convened an
Advisory Committee of Experts for Nuclear Pressure Equipment to
obtain its technical opinion on the consequences of CMAC for the
serviceability of the Flamanville EPR reactor vessel domes. The
resulting series of publicly available reports (CODEP-DEP-2015-
037971,\24\
[[Page 39793]]
CODEP-DEP-2016-019209,\25\ and CODEP-DEP-2017-019368 \26\) justified
the continued use of the Flamanville heads. In this effort, AREVA
conducted hundreds of mechanical and chemical property experiments
on three full[dash]scale replica heads that were manufactured by ACF
using the same process as that used for the Flamanville heads. Using
these experimental results, AREVA conducted a variety of
code[dash]related fracture and strength analyses that demonstrated
that the risk of fast fracture from CMAC was extremely low. Through
this effort, ASN concluded that the serviceability of the heads is
acceptable as long as EDF conducts the required inservice
inspections. However, because of its inability to conduct an
adequate inservice inspection on the Flamanville upper head, ASN
concluded that the upper head long[dash]term serviceability could
not be confirmed and that the head should be replaced after a few
years of operation.
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\24\ See ASN/Institut de Radioprotection et de
S[ucirc]ret[eacute] Nucl[eacute]aire (IRSN) (Radioprotection and
Nuclear Safety Institute) report CODEP-DEP-2015-037971, ``Analysis
of the Procedure Proposed by AREVA to Prove Adequate Toughness of
the Dome of the Flamanville 3 EPR Reactor Pressure Vessel Lower Head
and Closure Head,'' English translation, dated September 16, 2015.
http://www.french-nuclear-safety.fr/Media/Files/00-Publications/Report-to-the-Advisory-Committee-of-Experts-for-Nuclear-Pressure-Equipment.
\25\ See ASN/IRSN report CODEP-DEP-2016-019209, ``Procedure
Proposed by AREVA to Prove Adequate Toughness of the Domes of the
Flamanville 3 EPR Reactor Pressure Vessel Bottom Head and Closure
Head,'' English translation, dated June 17, 2016. https://www.asn.fr/content/download/106732/811356/version/6/file/CODEP-DEP-2016-019209-advisorycommitte24june2016-summaryreport.pdf.
\26\ See ASN/IRSN report CODEP-DEP-2017-019368, ``Analysis of
the Consequences of the Anomaly in the Flamanville EPR Reactor
Pressure Vessel Head Domes on Their Serviceability,'' English
translation, dated June 15, 2017. http://www.irsn.fr/FR/expertise/rapports_gp/Documents/GPESPN/IRSN-ASNDEP_GPESPN-Report_pressure-vessel-FA3_201706.pdf.
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B. Initial Actions by the NRC and the U.S. Nuclear Industry
Beginning in December 2016, the NRC staff conducted a
preliminary safety assessment to determine the potential safety
significance posed to the U.S. nuclear power reactor fleet by the
CMAC observed in reactor coolant system (RCS) components overseas
and concluded that the failure of an RPV/SG head component has a
very low probability, even if the worst practical degree of CMAC
occurs within that component. The NRC staff used a qualitative
failure comparison to assess the relative likelihood of failure of
an RPV shell (which is not expected to be subject to positive CMAC)
with RPV/SG head component types that could be affected by CMAC.
Based on this comparison, the NRC determined the following:
The RPV shell experiences higher stresses under both normal
operations and postulated accident scenarios.
The weld region of an RPV shell has a greater likelihood of
having more flaws and larger fabrication flaws. The larger
fabrication flaws typically have the higher potential to result in
component failure.
Although the initial toughness of an RPV shell material may
be greater than an RPV/SG head with postulated positive CMAC, the
shell toughness decreases as the result of radiation embrittlement
after several years of operation. As a result, the current
as[dash]operated toughness of RPV shell material is expected to be
lower than the toughness of RPV/SG head material with postulated
CMAC. The RPV shell material is known to have adequate toughness for
safe operation.
When combining all these individual attributes, an RPV/SG head
component with postulated CMAC is much less likely to fail than an
RPV shell. Past research and operating experience has demonstrated
that failure of an RPV shell under normal operations or postulated
accident scenarios has a very low probability of
occurrence.27 28 Therefore, the failure of an RPV/SG head
component also has a very low probability, even if the worst
practical degree of CMAC occurs within that component. The NRC
presented the basis for this preliminary determination in a
technical session titled ``Carbon Macrosegregation in Large Nuclear
Forgings'' (cited above) at the March 15, 2017, NRC[dash]sponsored
Regulatory Information Conference.
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\27\ See ADAMS Accession No. ML072830076.
\28\ See ADAMS Accession No. ML072820691.
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Concurrent with the NRC analyses, the U.S. industry initiated a
research program in early 2017, conducted by the Electric Power
Research Institute (EPRI), to address the generic safety
significance of elevated carbon levels caused by CMAC in the
components of interest. This program was divided into the following
four main tasks, each aimed at developing both qualitative and
quantitative information to make a safety determination:
1. extension of RPV probabilistic fracture mechanics (PFM) analyses
to qualitatively bound other components
2. development of a robust technical basis to support the hypothesis
that RPV integrity bounds other components
3. quantitative structural analyses to assess whether the results of
the PFM analyses of the RPV beltline (Task 1) bound the other forged
components
4. a white paper assessing the effect of CMAC on SG tubesheets based
on expert judgment and experience with the fabrication of the
tubesheets as large forgings
As of the writing of this document, Task 1 has been completed
and has been publicly released as Materials Reliability Program
(MRP)[dash]417.\29\ The other tasks are still under development with
the expected release of the report(s) in 2018.
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\29\ EPRI Report No. 3002010331, ``Materials Reliability
Program: Evaluation of Risk from Carbon Macrosegregation in Reactor
Pressure Vessels and Other Large Nuclear Forgings (MRP-417),''
issued June 2017 (available at ADAMS Accession No. ML18054A862).
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The MRP[dash]417 addresses the structural significance of the
potential presence of CMAC in large, forged pressurized[dash]water
reactor pressure[dash]retaining components, including the RPV head,
beltline and nozzle shell forgings, and the SG and pressurizer ring
and head forgings through the end of an 80[dash]year operating
interval. The assessment was made using the NRC risk safety
criterion of a 95\th\ percentile through[dash]wall crack frequency
(TWCF) of less than 1x10-\6\ per year (yr-\1\)
(10 CFR 50.61a, ``Alternative Fracture Toughness Requirements for
Protection against Pressurized Thermal Shock Events'') for
pressurized thermal shock (PTS) events and a conditional probability
of failure (CPF) of less than 1x10-\6\ for normal
operating transients. These analyses used many of the same
assumptions and inputs as those used in the basis for the 10 CFR
50.61a alternate PTS rule.30 31 In addition, the analysts
approximated the effect of carbon content on the fracture toughness
of the steel through a review of the available literature.
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\30\ See ADAMS Accession No. ML072830076.
\31\ See ADAMS Accession No. ML072820691.
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The MRP-417 describes the analyses and results for bounding
values for the RPV shell, RPV upper head, SG channel head,
pressurizer shell, and pressurizer head components based on the
analyses assumptions from the alternate PTS rule in conjunction with
the effect of the CMAC on the material toughness. The report's
deterministic results suggest that the RPV vessel behavior bounds
the behavior of the pressurizer components. In addition, the
probabilistic results suggest that in all cases, assuming the
maximum carbon content observed in the field, the calculated TWCF
and CPF were below the NRC risk safety criterion of the 95\th\
percentile TWCF of less than 1x10-\6\ yr-\1\
for PTS events and a CPF of less than 1x10-\6\ for normal
operating transients. MRP-417 concludes that there is substantial
margin against failure through an 80-year operating interval using
the assumed CMAC distributions in the RPV, SG, and pressurizer rings
and head forgings in pressurized[dash]water reactors.
In March 2017, an NRC inspection team performed a
limited[dash]scope vendor inspection at the AREVA facility in
Lynchburg, Virginia, to review documentation from ACF and assess
AREVA's compliance with the provisions of selected portions of
Appendix B, ``Quality Assurance Criteria for Nuclear Power Plants
and Fuel Reprocessing Plants,'' to 10 CFR Part 50, and 10 CFR Part
21, ``Reporting of Defects and Noncompliance.'' This inspection
focused on AREVA's documentation and evaluation of potential carbon
macrosegregation issues in forgings supplied by AREVA for U.S.
operating nuclear power plants. Specifically, the NRC inspection
reviewed documentation to verify that forgings met the ASME Code
requirements for carbon content and mechanical properties. The NRC
issued the inspection report on May 10, 2017.\32\ The
limited[dash]scope inspection reviewed policies and procedures that
govern implementation of AREVA's 10 CFR Part 21 program, and
nonconformance and corrective action policies and procedures under
its approved quality assurance program related to the manufacturing
processes used by ACF to fabricate inservice U.S. components and the
resulting mechanical properties. The NRC inspection team used
Inspection Procedure (IP) 43002, ``Routine Inspections of Nuclear
Vendors,'' \33\ and IP 36100, ``Inspection of 10 CFR Part 21 and
Programs for Reporting Defects and Noncompliance.'' \34\ The
inspection team did not identify any violations or nonconformances
during the inspection.
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\32\ See ADAMS Accession No. ML17124A575.
\33\ See ADAMS Accession No. ML13148A361.
\34\ See ADAMS Accession No. ML113190538.
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[[Page 39794]]
The inspection report contains the following primary material
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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.
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\35\ See ADAMS Accession No. ML14035A143.
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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.
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Principle 2--Consistency with the Defense-in[dash]Depth Philosophy
The aspect of defense[dash]in[dash]depth of relevance to the
potential presence of CMAC in regions of RCS components is ``barrier
integrity.'' The reactor coolant pressure boundary is one of the
three principal fission[dash]product release barriers in a U.S.
plant. Under 10 CFR 50.61a, the NRC established a 95\th\ percentile
TWCF of less than 1x10-6 yr-1 and a CDF of
less than 1x10-6 as acceptable RPV failure probabilities.
The conservative assessment performed by the industry and described
earlier showed that the probability of compromising the barrier
integrity function for the inservice U.S. components of interest are
significantly below these acceptance levels. If a design[dash]basis
accident were to compromise the pressure boundary, the remaining two
independent fission[dash]product release barriers (i.e., fuel
cladding and containment) would still provide adequate
defense[dash]in[dash]depth. The NRC has reasonable assurance that
U.S. plants with components produced by ACF maintain adequate
defense[dash]in[dash]depth. Therefore, Principle 2 is satisfied for
Option 1.
Principle 3--Maintenance of Adequate Safety Margins
A region of CMAC in a component could reduce the margin against
fracture. However, it has been shown that this reduction in
[[Page 39796]]
margin does not affect the safe operation of the inservice
components being evaluated. The ASN evaluation described earlier
determined that the safety margin against fast fracture is
maintained in all conditions analyzed. Industry determined in
MRP[dash]417 that the CMAC levels necessary to be considered
significant to safety are more than 200 percent of those observed in
components. Based on its review of these evaluations, the NRC has
reasonable assurance that U.S. plants with components produced by
ACF maintain sufficient safety margins. Therefore, Principle 3 is
satisfied for Option 1.
Principle 4--Demonstration of Acceptable Levels of Risk
If it is conservatively assumed that the TWCF equates to the
LERF (neglecting mitigating factors), the calculated 95\th\
percentile TWCF for components with CMAC and thus the LERF is less
than 1x10-6 yr-1. Because this is below the
immediate safety determination limit, there is no immediate safety
concern. Therefore, Principle 4 is satisfied for Option 1.
Principle 5--Implementation of Defined Performance Measurement
Strategies
Because there is no indication that the U.S. inservice
components produced by ACF are noncompliant with the applicable
regulations and because the NRC has reasonable assurance that
defense[dash]in[dash]depth, safety margins, and risk levels are
adequately maintained, the current monitoring programs at the plants
are adequate, and additional performance measurement strategies are
not warranted. However, the NRC staff would continue to monitor the
U.S. nuclear industry and international activities related to the
CMAC topic to analyze any new information to determine whether
additional performance measurement strategies are necessary.
Therefore, Principle 5 is satisfied for Option 1.
Option 2: Issue a Generic Communication
This option reinforces the regulatory determination made in
Option 1 by issuing a GL requesting that the documentation and
evaluations performed by licensees and their component suppliers
conclude that the components produced by ACF do not have defects or
deviations that pose a substantial safety hazard. The NRC would not
expect the information collected in the response to a GL to change
any of the conclusions reached in Option 1, including those related
to defense[dash]in[dash]depth, safety margins, or risk[dash]level
determinations. Therefore, all five principles of risk[dash]informed
decision[dash]making would also be satisfied for Option 2.
Additionally, the relevant vendors have informed the affected
licensees of the CMAC topic. Vendors and licensees must meet their
10 CFR part 21 evaluation and reporting responsibilities if the
condition warrants such action. As part of the ROP and vendor
inspection program, the NRC reviews these evaluations for adequacy.
Option 3: Issue Orders Requiring Inspections
This option reinforces the determinations made in Option 1 by
performing inspections to confirm that an appropriate degree of
conservatism was used in the evaluations of the potential impact of
CMAC on U.S. components produced by AFC. The NRC would not expect
the information collected by performing nondestructive examinations
of the inservice components to significantly affect the
defense[dash]in[dash]depth, safety margins, or risk[dash]level
determinations made in Option 1. Therefore, all five principles of
risk[dash]informed decision[dash]making would also be satisfied for
Option 3.
Option 4: Issue Orders Suspending Operation
In evaluating the international, U.S. industry, and NRC safety
assessments, the NRC determined that the impact of CMAC on the
integrity of the U.S.[dash]forged components in question is small
and that the calculated 95th percentile TWCF for PTS and
the CPF for normal operating conditions fall below the NRC's safety
criteria of 1x10-6 yr-1 and 1x10-6,
respectively. Because the assumption that the TWCF is equivalent to
the LERF because of mitigating factors is extremely conservative,
the results indicate that the impacts of CMAC would result in a risk
of LERF less than 1x10-4 yr-1. Therefore,
because the NRC's risk criterion to shut down a plant is not met,
the agency dismissed Option 4 without an evaluation of the five
principles of risk[dash]informed decision[dash]making.
Final Assessment
The staff determined that Option 1 was the most appropriate
action based on the material and processing information reviewed by
the staff during the vender inspection of AREVA, experimental data
and evaluation reported by ASN, PFM analyses conducted by the
industry, the staff's review of the open literature on CMAC in steel
ingots and its effect on performance, and an evaluation
demonstrating that Option 1 satisfies all five key principles of
risk[dash]informed decision[dash]making. Additionally, this
compilation of information reviewed affirms the staff's preliminary
safety assessment that the safety significance of CMAC to U.S.
plants appears to be negligible and does not warrant immediate
action. If new information becomes available that calls into
question the conservatism of the evaluations supporting Option 1 or
the regulatory compliance of the plants with inservice components
from ACF, the NRC staff will reevaluate the need for additional
actions. The staff's evaluation dated February 22, 2018, documents
the NRC's full evaluation of the CMAC topics as it relates to plants
operating in the United States.
E. Evaluation of the Petitioners' Requests
Petitioners' Request 1: Suspend power operations of U.S. nuclear
power plants that rely on ACF components and subcontractors pending
a full inspection (including nondestructive examination by
ultrasonic testing) and material testing. If carbon anomalies
(``carbon segregation'' or ``carbon macrosegregation'') in excess of
the design[dash]basis specifications for at[dash]risk component
parts are identified, require the licensee to do one of the
following:
a. replace the degraded at[dash]risk component(s) with quality
certified components, or
b. for those at[dash]risk degraded components that a licensee
seeks to allow to remain in[dash]service, make application through
the license amendment request process to demonstrate that a revised
design[dash]basis is achievable and will not render the
in[dash]service component unacceptably vulnerable to fast fracture
failure at any time, and in any credible service condition,
throughout the current license of the power reactor.
NRC Response:
This request is essentially identical to Option 4 described
above. The NRC has determined, through its PFM analyses, that the
expected impact of CMAC on the LERF is less than 1x10-6
yr-1. Therefore, the risk criterion to shut down a plant
is not met.
Petitioners' Request 2: Alternatively modify the operating licenses
to require the affected operators to perform the requested emergency
enforcement actions at the next scheduled outage.
NRC Response:
This request is essentially identical to Option 3 described
above. As discussed above, performing nondestructive examinations of
the inservice components is not expected to provide information that
would significantly affect the defense[dash]in-depth, safety
margins, or risk[dash]level determinations that would be provided by
continued monitoring and evaluation of new information.
Petitioners' Request 3: Issue a letter to all U.S. light[dash]water
reactor operators under 10 CFR 50.54(f) requiring licensees to
provide the NRC with information under oath and affirming
specifically how U.S. operators are reliably monitoring contractors
and subcontractors for the potential carbon segmentation anomaly in
the supply chain and the reliability of the quality assurance
certification of those components, and publicly release the
responses.
NRC Response:
This request is essentially identical to Option 2 described
above. As discussed above, the information collected through a 10
CFR 50.54(f) request for information or a GL is not expected to
change any of defense[dash]in[dash]depth, safety margins, or
risk[dash]level determinations that would be provided by continued
monitoring and evaluation of new information. In addition, the
relevant vendors and licensees must meet their 10 CFR Part 21
evaluation and reporting responsibilities if the condition warrants
such action. As part of the ROP and vendor inspection program, the
NRC reviews these evaluations for adequacy.
Petitioners' Request 4: [The Petitioners added Crystal River Unit 3
to the plants for which they requested actions, which include the
following]:
a. Confirm the sale, delivery, quality control and quality
assurance certification and installation of the replacement reactor
pressure vessel head as supplied to Crystal River Unit
[[Page 39797]]
3 by then Framatome and now AREVA[dash]Le Creusot Forge industrial
facility in Charlon[dash]St. Marcel, France and;
b. With completion and confirmation [of the above Crystal River
Unit 3 actions], the modification of Duke Energy's current license
for the permanently closed Crystal River Unit 3 nuclear power
station in Crystal River, Florida, to inspect and conduct the
appropriate material test(s) for carbon macrosegregation on
sufficient samples harvested from the installed and now in service
irradiated Le Creusot Forge reactor pressure vessel head [sic]. The
Petitioners assert that the appropriate material testing include
OES.
NRC Response:
AREVA did not identify Crystal River Unit 3 as a plant that
contained components from ACF,\37 38\ and the staff has not
confirmed that this unit contained any forgings manufactured from
ingots produced by ACF. In addition, Crystal River Unit 3 is
currently shut down and in the process of decommissioning.
Therefore, the Petitioners' requests 1, 2, 3, and 4(a) do not apply
to this plant. However, the acquisition and subsequent testing of
irradiated and aged plant material from decommissioned plants could
be a valuable research activity that might offer useful scientific
information on the progress of aging mechanisms. The harvesting of
reactor vessel material from plants that have been permanently shut
down can be a complex and radiation[dash]dose[dash]intensive effort.
The NRC's Office of Nuclear Regulatory Research has previously
obtained samples appropriate for testing from shutdown plants. In
regard to this request, the NRC may, in the future, seek to purchase
samples. However, the identified facility has ceased operations, and
there is no safety concern at those facilities that justifies
enforcement[dash]related action (i.e., to modify, suspend, or revoke
the license) to give the NRC reasonable assurance of the adequate
protection of public health and safety.
---------------------------------------------------------------------------
\37\ See ADAMS Accession No. ML17040A100.
\38\ See ADAMS Accession No. ML17009A278.
---------------------------------------------------------------------------
III. Conclusion
Based on the evaluations provided above, and documented in the
February 22, 2018, NRC memorandum, the NRR Director has determined
that the actions requested by the Petitioners, will not be granted
in whole or in part.
As provided for in 10 CFR 2.206(c), a copy of this Director's
Decision will be filed with the Secretary of the Commission for the
Commission to review. As provided for by this regulation, the
decision will constitute the final action of the Commission 25 days
after the date of the decision unless the Commission, on its own
motion, institutes a review of the decision within that time.
Dated at Rockville, MD, this 2nd day of August 2018.
For the Nuclear Regulatory Commission.
Brian E. Holian,
Acting Director, Office of Nuclear Reactor Regulation.
Attachment:
List of Affected Reactors
List of Power Reactors Affected by the Petition
----------------------------------------------------------------------------------------------------------------
Plant Docket No. Facility operating license No.
----------------------------------------------------------------------------------------------------------------
Prairie Island Nuclear Generating Plant, Unit 1........... 05000282 DPR-42
Prairie Island Nuclear Generating Plant, Unit 2........... 05000306 DPR-60
Arkansas Nuclear One, Unit 2.............................. 05000368 NPF-6
Beaver Valley Power Station, Unit 1....................... 05000334 DPR-66
North Anna Power Station, Unit 1.......................... 05000338 NPF-4
North Anna Power Station, Unit 2.......................... 05000339 NPF-7
Surry Power Station, Unit 1............................... 05000280 DPR-32
Comanche Peak Nuclear Power Plant, Unit 1................. 05000445 NPF-87
V.C. Summer Nuclear Station, Unit 1....................... 05000395 NPF-12
Joseph M. Farley Nuclear Plant, Unit 1.................... 05000348 NPF-2
Joseph M. Farley Nuclear Plant, Unit 2.................... 05000364 NPF-8
South Texas Project, Unit 1............................... 05000498 NPF-76
South Texas Project, Unit 2............................... 05000499 NPF-80
Sequoyah Nuclear Plant, Unit 1............................ 05000327 DPR-77
Watts Bar Nuclear Plant, Unit 1........................... 05000390 NPF-90
Millstone Power Station, Unit 2........................... 05000336 NPF-65
Saint Lucie Plant, Unit 1................................. 05000335 DPR-67
Crystal River Unit 3 Nuclear Generating Plant............. 05000302 DPR-72
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[FR Doc. 2018-17131 Filed 8-9-18; 8:45 am]
BILLING CODE 7590-01-P