[Federal Register Volume 83, Number 108 (Tuesday, June 5, 2018)]
[Notices]
[Pages 26098-26109]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2018-11843]
=======================================================================
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
[NRC-2018-0105]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
AGENCY: Nuclear Regulatory Commission.
ACTION: Biweekly notice.
-----------------------------------------------------------------------
SUMMARY: Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954,
as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is
publishing this regular biweekly notice. The Act requires the
Commission to publish notice of any amendments issued, or proposed to
be issued, and grants the Commission the authority to issue and make
immediately effective any amendment to an operating license or combined
license, as applicable, upon a determination by the Commission that
such amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued, from May 8, 2018, to May 21, 2018. The last
biweekly notice was published on May 22, 2018.
DATES: Comments must be filed by July 5, 2018. A request for a hearing
must be filed by August 6, 2018.
ADDRESSES: You may submit comments by any of the following methods
(unless this document describes a different method for submitting
comments on a specific subject):
Federal Rulemaking website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0105. 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.
Mail comments to: May Ma, Office of Administration, Mail
Stop: TWFN-7-A60M, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001.
For additional direction on obtaining information and submitting
comments, see ``Obtaining Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Janet Burkhardt, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC
20555-0001; telephone: 301-415-1384, email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2018-0105, facility name, unit
number(s), plant docket number, application date, and subject when
contacting the NRC about the availability of information for this
action. You may obtain publicly-available information related to this
action by any of the following methods:
Federal Rulemaking website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0105.
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 ``ADAMS Public Documents'' and
then 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 email 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.
B. Submitting Comments
Please include Docket ID NRC-2018-0105, facility name, unit
number(s), plant docket number, application date, and subject in your
comment submission.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment into ADAMS.
II. Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses and Combined Licenses and Proposed No Significant
Hazards Consideration Determination
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Sec. 50.92 of title 10 of the Code of
Federal Regulations (10 CFR), this means that operation of the facility
in accordance with the proposed amendment would not (1) involve a
significant increase in the probability or consequences of an accident
previously evaluated; or (2) create the possibility of a new or
different kind of accident from any accident previously evaluated; or
(3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment
[[Page 26099]]
prior to the expiration of the 30-day comment period if circumstances
change during the 30-day comment period such that failure to act in a
timely way would result, for example in derating or shutdown of the
facility. If the Commission takes action prior to the expiration of
either the comment period or the notice period, it will publish in the
Federal Register a notice of issuance. If the Commission makes a final
no significant hazards consideration determination, any hearing will
take place after issuance. The Commission expects that the need to take
this action will occur very infrequently.
A. Opportunity To Request a Hearing and Petition for Leave To Intervene
Within 60 days after the date of publication of this notice, any
persons (petitioner) whose interest may be affected by this action may
file a request for a hearing and petition for leave to intervene
(petition) with respect to the action. Petitions shall be filed in
accordance with the Commission's ``Agency Rules of Practice and
Procedure'' in 10 CFR part 2. Interested persons should consult a
current copy of 10 CFR 2.309. The NRC's regulations are accessible
electronically from the NRC Library on the NRC's website at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of
the regulations is available at the NRC's Public Document Room, located
at One White Flint North, Room O1-F21, 11555 Rockville Pike (first
floor), Rockville, Maryland 20852. If a petition is filed, the
Commission or a presiding officer will rule on the petition and, if
appropriate, a notice of a hearing will be issued.
As required by 10 CFR 2.309(d) the petition should specifically
explain the reasons why intervention should be permitted with
particular reference to the following general requirements for
standing: (1) The name, address, and telephone number of the
petitioner; (2) the nature of the petitioner's right under the Act to
be made a party to the proceeding; (3) the nature and extent of the
petitioner's property, financial, or other interest in the proceeding;
and (4) the possible effect of any decision or order which may be
entered in the proceeding on the petitioner's interest.
In accordance with 10 CFR 2.309(f), the petition must also set
forth the specific contentions which the petitioner seeks to have
litigated in the proceeding. Each contention must consist of a specific
statement of the issue of law or fact to be raised or controverted. In
addition, the petitioner must provide a brief explanation of the bases
for the contention and a concise statement of the alleged facts or
expert opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to the specific sources and
documents on which the petitioner intends to rely to support its
position on the issue. The petition must include sufficient information
to show that a genuine dispute exists with the applicant or licensee on
a material issue of law or fact. Contentions must be limited to matters
within the scope of the proceeding. The contention must be one which,
if proven, would entitle the petitioner to relief. A petitioner who
fails to satisfy the requirements at 10 CFR 2.309(f) with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene.
Parties have the opportunity to participate fully in the conduct of the
hearing with respect to resolution of that party's admitted
contentions, including the opportunity to present evidence, consistent
with the NRC's regulations, policies, and procedures.
Petitions must be filed no later than 60 days from the date of
publication of this notice. Petitions and motions for leave to file new
or amended contentions that are filed after the deadline will not be
entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the three factors in 10
CFR 2.309(c)(1)(i) through (iii). The petition must be filed in
accordance with the filing instructions in the ``Electronic Submissions
(E-Filing)'' section of this document.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to establish when the hearing is held. If the final determination is
that the amendment request involves no significant hazards
consideration, the Commission may issue the amendment and make it
immediately effective, notwithstanding the request for a hearing. Any
hearing would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, then any hearing held would take place before
the issuance of the amendment unless the Commission finds an imminent
danger to the health or safety of the public, in which case it will
issue an appropriate order or rule under 10 CFR part 2.
A State, local governmental body, Federally-recognized Indian
Tribe, or agency thereof, may submit a petition to the Commission to
participate as a party under 10 CFR 2.309(h)(1). The petition should
state the nature and extent of the petitioner's interest in the
proceeding. The petition should be submitted to the Commission no later
than 60 days from the date of publication of this notice August 6,
2018. The petition must be filed in accordance with the filing
instructions in the ``Electronic Submissions (E-Filing)'' section of
this document, and should meet the requirements for petitions set forth
in this section, except that under 10 CFR 2.309(h)(2) a State, local
governmental body, or Federally-recognized Indian Tribe, or agency
thereof does not need to address the standing requirements in 10 CFR
2.309(d) if the facility is located within its boundaries.
Alternatively, a State, local governmental body, Federally-recognized
Indian Tribe, or agency thereof may participate as a non-party under 10
CFR 2.315(c).
If a hearing is granted, any person who is not a party to the
proceeding and is not affiliated with or represented by a party may, at
the discretion of the presiding officer, be permitted to make a limited
appearance pursuant to the provisions of 10 CFR 2.315(a). A person
making a limited appearance may make an oral or written statement of
his or her position on the issues but may not otherwise participate in
the proceeding. A limited appearance may be made at any session of the
hearing or at any prehearing conference, subject to the limits and
conditions as may be imposed by the presiding officer. Details
regarding the opportunity to make a limited appearance will be provided
by the presiding officer if such sessions are scheduled.
B. Electronic Submissions (E-Filing)
All documents filed in NRC adjudicatory proceedings, including a
request for hearing and petition for leave to intervene (petition), any
motion or other document filed in the proceeding prior to the
submission of a request for hearing or petition to intervene, and
documents filed by interested governmental entities that request to
participate under 10 CFR 2.315(c), must be filed in accordance with the
NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR
46562; August 3, 2012). The E-Filing process requires participants to
submit and serve all adjudicatory
[[Page 26100]]
documents over the internet, or in some cases to mail copies on
electronic storage media. Detailed guidance on making electronic
submissions may be found in the Guidance for Electronic Submissions to
the NRC and on the NRC website at http://www.nrc.gov/site-help/e-submittals.html. Participants may not submit paper copies of their
filings unless they seek an exemption in accordance with the procedures
described below.
To comply with the procedural requirements of E-Filing, at least 10
days prior to the filing deadline, the participant should contact the
Office of the Secretary by email at [email protected], or by
telephone at 301-415-1677, to (1) request a digital identification (ID)
certificate, which allows the participant (or its counsel or
representative) to digitally sign submissions and access the E-Filing
system for any proceeding in which it is participating; and (2) advise
the Secretary that the participant will be submitting a petition or
other adjudicatory document (even in instances in which the
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the
Secretary will establish an electronic docket for the hearing in this
proceeding if the Secretary has not already established an electronic
docket.
Information about applying for a digital ID certificate is
available on the NRC's public website at http://www.nrc.gov/site-help/e-submittals/getting-started.html. Once a participant has obtained a
digital ID certificate and a docket has been created, the participant
can then submit adjudicatory documents. Submissions must be in Portable
Document Format (PDF). Additional guidance on PDF submissions is
available on the NRC's public website at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the
time the document is submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an email notice confirming receipt of the document. The
E-Filing system also distributes an email notice that provides access
to the document to the NRC's Office of the General Counsel and any
others who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
document on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before adjudicatory documents are
filed so that they can obtain access to the documents via the E-Filing
system.
A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC's Electronic
Filing Help Desk through the ``Contact Us'' link located on the NRC's
public website at http://www.nrc.gov/site-help/e-submittals.html, by
email to [email protected], or by a toll-free call at 1-866-672-
7640. The NRC Electronic Filing Help Desk is available between 9 a.m.
and 6 p.m., Eastern Time, Monday through Friday, excluding government
holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
stating why there is good cause for not filing electronically and
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, 11555
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and
Adjudications Staff. Participants filing adjudicatory documents in this
manner are responsible for serving the document on all other
participants. Filing is considered complete by first-class mail as of
the time of deposit in the mail, or by courier, express mail, or
expedited delivery service upon depositing the document with the
provider of the service. A presiding officer, having granted an
exemption request from using E-Filing, may require a participant or
party to use E-Filing if the presiding officer subsequently determines
that the reason for granting the exemption from use of E-Filing no
longer exists.
Documents submitted in adjudicatory proceedings will appear in the
NRC's electronic hearing docket which is available to the public at
https://adams.nrc.gov/ehd, unless excluded pursuant to an order of the
Commission or the presiding officer. If you do not have an NRC-issued
digital ID certificate as described above, click cancel when the link
requests certificates and you will be automatically directed to the
NRC's electronic hearing dockets where you will be able to access any
publicly available documents in a particular hearing docket.
Participants are requested not to include personal privacy information,
such as social security numbers, home addresses, or personal phone
numbers in their filings, unless an NRC regulation or other law
requires submission of such information. For example, in some
instances, individuals provide home addresses in order to demonstrate
proximity to a facility or site. With respect to copyrighted works,
except for limited excerpts that serve the purpose of the adjudicatory
filings and would constitute a Fair Use application, participants are
requested not to include copyrighted materials in their submission.
For further details with respect to these license amendment
applications, see the application for amendment which is available for
public inspection in ADAMS and at the NRC's PDR. For additional
direction on accessing information related to this document, see the
``Obtaining Information and Submitting Comments'' section of this
document.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2 (MNS), Mecklenburg County, North
Carolina
Date of amendment request: December 8, 2017. A publicly-available
version is in ADAMS under Accession No. ML17352A404.
Description of amendment request: The amendments would modify the
MNS, Unit Nos. 1 and 2 Updated Final Safety Analysis Report (UFSAR) to
describe the methodology and results of the analyses performed to
evaluate the protection of the plant's structures, systems, and
components from tornado-generated missiles.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to the MNS UFSAR constitutes a license
amendment to incorporate use of a Nuclear Regulatory Commission
(NRC) approved probabilistic methodology to assess the need for
additional positive (physical) tornado missile protection of
specific features at the MNS site. The UFSAR changes will reflect
use of the Electric Power Research Institute (EPRI) Topical Report
``Tornado Missile Risk
[[Page 26101]]
Evaluation Methodology'' (EPRI NP-2005), Volumes I and II. As noted
in the NRC Safety Evaluation Report on this topic dated October 26,
1983, the current licensing criteria governing tornado missile
protection are contained in NUREG-0800, Sections 3.5.1.4 and 3.5.2.
These criteria generally specify that safety-related systems,
structures and components be provided positive tornado missile
protection (barriers) from the maximum credible tornado threat.
However, NUREG-0800 includes acceptance criteria permitting
relaxation of the above deterministic guidance, if it can be
demonstrated that the probability of damage to unprotected essential
safety-related features is sufficiently small.
As permitted in NUREG-0800 sections, the combined probability
will be maintained below an allowable level, i.e., an acceptance
criterion threshold, which reflects an extremely low probability of
occurrence. The approach assumes that if the sum of the individual
probabilities calculated for tornado missiles striking and damaging
portions of important systems, structures or components is greater
than or equal to 1 x 10-6 per year per unit, then
installation of unique missile barriers would be needed to lower the
total cumulative probability below the acceptance criterion of 1 x
10-6 per year per unit.
With respect to the probability of occurrence or the
consequences of an accident previously evaluated in the UFSAR, the
possibility of a tornado reaching the site and causing damage to
plant structures, systems and components is considered in the MNS
UFSAR.
The change being proposed does not affect the probability that
the natural phenomenon (a tornado) will reach the plant, but from a
licensing basis perspective, the change does affect the probability
that missiles generated by the winds of the tornado might strike and
damage certain plant structures, systems and components. There are a
limited number of safety-related components that could theoretically
be struck and damaged by tornadogenerated missiles. The probability
of tornado-generated missile strikes on important to safety
structures, systems and components is what was analyzed using the
probabilistic methods discussed above. The combined probability of
damage will be maintained below an extremely low acceptance
criterion to ensure overall plant safety. The proposed change is not
considered to constitute a significant increase in the probability
of occurrence or the consequences of an accident, due to the
extremely low probability of damage due to tornado-generated
missiles and thus an extremely low probability of a radiological
release.
The results of the analysis documented in this [license
amendment request (LAR)] are below the acceptance criterion of 1 x
10-6 per year per unit. Therefore, the proposed change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes to the MNS UFSAR incorporate use of a NRC
approved probabilistic methodology to assess the need for additional
positive (physical) tornado missile protection for specific
features. This will not change the design function or operation of
any structure, system or component. This proposed change does not
involve any plant modifications. There are no new credible failure
mechanisms, malfunctions or accident initiators not considered in
the design and licensing bases for MNS. The proposed change involves
an already established tornado design basis event and the tornado
event is explicitly considered in the MNS UFSAR.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in the margin of safety?
Response: No.
The existing licensing basis for MNS for protecting safety-
related, safe shutdown equipment from tornado generated missiles is
to provide positive missile barriers for all safety-related
structures, systems and components. The proposed change recognizes
that there is an extremely low probability, below an established
acceptance limit, that a limited subset of the safety-related, safe
shutdown structures, systems and components could be struck and
consequently damaged. The change from requiring protection of all
safety-related, safety shutdown structures, systems and components
from tornadogenerated missiles, to only a subset of equipment, is
not considered to constitute a significant decrease in the margin of
safety due to that extremely low probability of occurrence of
tornado-generated missile strikes and consequential damage.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Kate B. Nolan, Deputy General Counsel, Duke
Energy Carolinas, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC
28202-1802.
NRC Branch Chief: Michael T. Markley.
Duke Energy Progress, LLC, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: April 5, 2018. A publicly-available
version is in ADAMS under Accession No. ML18099A130.
Description of amendment request: The proposed amendment would
revise the licensing basis, by the addition of a license condition, to
allow for the implementation of the provisions of 10 CFR 50.69, ``Risk-
informed categorization and treatment of structures, systems, and
components [SSCs] for nuclear power reactors.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change will permit the use of a risk-informed
categorization process to modify the scope of SSCs subject to NRC
special treatment requirements and to implement alternative
treatments per the regulations. The process used to evaluate SSCs
for changes to NRC special treatment requirements and the use of
alternative requirements ensures the ability of the SSCs to perform
their design function. The potential change to special treatment
requirements does not change the design and operation of the SSCs.
As a result, the proposed change does not significantly affect any
initiators to accidents previously evaluated or the ability to
mitigate any accidents previously evaluated. The consequences of the
accidents previously evaluated are not affected because the
mitigation functions performed by the SSCs assumed in the safety
analysis are not being modified. The SSCs required to safely shut
down the reactor and maintain it in a safe shutdown condition
following an accident will continue to perform their design
functions.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change will permit the use of a risk-informed
categorization process to modify the scope of SSCs subject to NRC
special treatment requirements and to implement alternative
treatments per the regulations. The proposed change does not change
the functional requirements, configuration, or method of operation
of any SSC. Under the proposed change, no additional plant equipment
will be installed.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will permit the use of a risk-informed
categorization process to
[[Page 26102]]
modify the scope of SSCs subject to NRC special treatment
requirements and to implement alternative treatments per the
regulations. The proposed change does not affect any Safety Limits
or operating parameters used to establish the safety margin. The
safety margins included in analyses of accidents are not affected by
the proposed change. The regulation requires that there be no
significant effect on plant risk due to any change to the special
treatment requirements for SSCs and that the SSCs continue to be
capable of performing their design basis functions, as well as to
perform any beyond design basis functions consistent with the
categorization process and results.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel,
Duke Energy Corporation, 550 South Tryon Street, DEC45A, Charlotte NC
28202.
NRC Acting Branch Chief: Brian W. Tindell.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: March 12, 2018, as supplemented by
letter dated April 26, 2018. Publicly-available versions are in ADAMS
under Accession Nos. ML18071A319 and ML18117A493, respectively.
Description of amendment request: The amendment would revise the
Arkansas Nuclear One, Unit No. 1 Technical Specifications (TSs) by
relocating specific surveillance frequencies to a licensee-controlled
program with the adoption of Technical Specification Task Force (TSTF)-
425, Revision 3, ``Relocate Surveillance Frequencies to Licensee
Control--RITSTF [Risk-informed TSTF] Initiative 5b.'' Additionally, the
change would add a new program, the Surveillance Frequency Control
Program, to TS Section 5.5, ``Programs and Manuals.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of any accident previously
evaluated?
Response: No.
The proposed change relocates the specified frequencies for
periodic surveillance requirements (SRs) to licensee control under a
new Surveillance Frequency Control Program [SFCP]. Surveillance
frequencies are not an initiator to any accident previously
evaluated. As a result, the probability of any accident previously
evaluated is not significantly increased. The systems and components
required by the technical specifications (TSs) for which the
surveillance frequencies are relocated are still required to be
operable, meet the acceptance criteria for the SRs, and be capable
of performing any mitigation function assumed in the accident
analysis. As a result, the consequences of any accident previously
evaluated are not significantly increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated?
Response: No.
No new or different accidents result from utilizing the proposed
change. The changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements. The changes do not alter assumptions made in
the safety analysis. The proposed changes are consistent with the
safety analysis assumptions and current plant operating practice.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in
the margin of safety?
Response: No.
The design, operation, testing methods, and acceptance criteria
for systems, structures, and components (SSCs), specified in
applicable codes and standards (or alternatives approved for use by
the NRC) will continue to be met as described in the plant licensing
basis (including the final safety analysis report and bases to TS),
since these are not affected by changes to the surveillance
frequencies. Similarly, there is no impact to safety analysis
acceptance criteria as described in the plant licensing basis. To
evaluate a change in the relocated surveillance frequency, Entergy
will perform a probabilistic risk evaluation using the guidance
contained in NRC approved NEI [Nuclear Energy Institute] 04-10, Rev.
1 in accordance with the TS SFCP. NEI 04-10, Rev. 1, methodology
provides reasonable acceptance guidelines and methods for evaluating
the risk increase of proposed changes to surveillance frequencies
consistent with Regulatory Guide 1.177.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Anna Vinson Jones, Senior Counsel, Entergy
Services, Inc., 101 Constitution Avenue NW, Suite 200 East, L-ENT-WDC,
Washington, DC 20001.
NRC Branch Chief: Robert J. Pascarelli.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: February 6, 2018, as supplemented by
letter dated March 26, 2018. Publicly-available versions are in ADAMS
under Accession Nos. ML18038B354, and ML18085A816, respectively.
Description of amendment request: The amendment would revise the
Arkansas Nuclear One, Unit No. 2 Technical Specifications (TSs) by
relocating specific surveillance frequencies to a licensee-controlled
program with the adoption of Technical Specifications Task Force
(TSTF)-425, Revision 3, ``Relocate Surveillance Frequencies to Licensee
Control--RITSTF [Risk-Informed TSTF] Initiative 5b.'' The amendment
would also add a new program, the Surveillance Frequency Control
Program, to TS Section 6.0, ``Administrative Controls.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of any accident previously
evaluated?
Response: No.
The proposed change relocates the specified frequencies for
periodic Surveillance Requirements (SRs) to licensee control under a
new Surveillance Frequency Control Program (SFCP). Surveillance
frequencies are not an initiator to any accident previously
evaluated. As a result, the probability of any accident previously
evaluated is not significantly increased. The systems and components
required by the TSs for which the surveillance frequencies are
relocated are still required to be operable, meet the acceptance
criteria for the SRs, and be capable of performing any mitigation
function assumed in the accident analysis. As a result, the
consequences of any accident previously evaluated are not
significantly increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
[[Page 26103]]
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated?
Response: No.
No new or different accidents result from utilizing the proposed
change. The changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements. The changes do not alter assumptions made in
the safety analysis. The proposed changes are consistent with the
safety analysis assumptions and current plant operating practice.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in
the margin of safety?
Response: No.
The design, operation, testing methods, and acceptance criteria
for systems, structures, and components (SSCs), specified in
applicable codes and standards (or alternatives approved for use by
the NRC) will continue to be met as described in the plant licensing
basis (including the final safety analysis report and bases to TS),
since these are not affected by changes to the surveillance
frequencies. Similarly, there is no impact to safety analysis
acceptance criteria as described in the plant licensing basis. To
evaluate a change in the relocated surveillance frequency, Entergy
will perform a probabilistic risk evaluation using the guidance
contained in NRC approved NEI [Nuclear Energy Institute] 04-10, Rev.
1, in accordance with the TS SFCP. NEI 04-10, Rev. 1, methodology
provides reasonable acceptance guidelines and methods for evaluating
the risk increase of proposed changes to surveillance frequencies
consistent with Regulatory Guide 1.177.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Anna Vinson Jones, Senior Counsel, Entergy
Services, Inc., 101 Constitution Avenue NW, Suite 200 East, L-ENT-WDC,
Washington, DC 20001.
NRC Branch Chief: Robert J. Pascarelli.
Entergy Operations, Inc.; System Energy Resources, Inc.; Cooperative
Energy, A Mississippi Electric Cooperative; and Entergy Mississippi,
Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit No. 1,
Claiborne County, Mississippi
Date of amendment request: April 10, 2018. A publicly-available
version is in ADAMS under Accession No. ML18100B304.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to adopt Technical
Specifications Task Force (TSTF) Traveler TSTF-542, Revision 2,
``Reactor Pressure Vessel Water Inventory Control.'' The proposed
change would replace existing TS requirements related to ``operations
with a potential for draining the reactor vessel'' (OPDRVs) with new
requirements on reactor pressure vessel (RPV) water inventory control
(WIC) to protect Safety Limit 2.1.1.3. Safety Limit 2.1.1.3 requires
reactor vessel water level to be greater than the top of active
irradiated fuel.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change replaces existing TS requirements related to
OPDRVs with new requirements on RPV WIC that will protect Safety
Limit 2.1.1.3. Draining of RPV water inventory in Mode 4 (i.e., cold
shutdown) and Mode 5 (i.e., refueling) is not an accident previously
evaluated and, therefore, replacing the existing TS controls to
prevent or mitigate such an event with a new set of controls has no
effect on any accident previously evaluated. RPV water inventory
control in Mode 4 or Mode 5 is not an initiator of any accident
previously evaluated. The existing OPDRV controls or the proposed
RPV WIC controls are not mitigating actions assumed in any accident
previously evaluated.
The proposed change reduces the probability of an unexpected
draining event (which is not a previously evaluated accident) by
imposing new requirements on the limiting time in which an
unexpected draining event could result in the reactor vessel water
level dropping to the top of the active fuel (TAF). These controls
require cognizance of the plant configuration and control of
configurations with unacceptably short drain times. These
requirements reduce the probability of an unexpected draining event.
The current TS requirements are only mitigating actions and impose
no requirements that reduce the probability of an unexpected
draining event.
The proposed change reduces the consequences of an unexpected
draining event (which is not a previously evaluated accident) by
requiring an Emergency Core Cooling System (ECCS) subsystem to be
operable at all times in Modes 4 and 5. The current TS requirements
do not require any water injection systems, ECCS or otherwise, to be
Operable in certain conditions in Mode 5. The change in requirement
from two ECCS subsystems to one ECCS subsystem in Modes 4 and 5 does
not significantly affect the consequences of an unexpected draining
event because the proposed Actions ensure equipment is available
within the limiting drain time that is as capable of mitigating the
event as the current requirements. The proposed controls provide
escalating compensatory measures to be established as calculated
drain times decrease, such as verification of a second method of
water injection and additional confirmations that containment and/or
filtration would be available if needed.
The proposed change reduces or eliminates some requirements that
were determined to be unnecessary to manage the consequences of an
unexpected draining event, such as automatic initiation of an ECCS
subsystem and control room ventilation. These changes do not affect
the consequences of any accident previously evaluated since a
draining event in Modes 4 and 5 is not a previously evaluated
accident and the requirements are not needed to adequately respond
to a draining event.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change replaces existing TS requirements related to
OPDRVs with new requirements on RPV WIC that will protect Safety
Limit 2.1.1.3. The proposed change will not alter the design
function of the equipment involved. Under the proposed change, some
systems that are currently required to be operable during OPDRVs
would be required to be available within the limiting drain time or
to be in service depending on the limiting drain time. Should those
systems be unable to be placed into service, the consequences are no
different than if those systems were unable to perform their
function under the current TS requirements.
The event of concern under the current requirements and the
proposed change is an unexpected draining event. The proposed change
does not create new failure mechanisms, malfunctions, or accident
initiators that would cause a draining event or a new or different
kind of accident not previously evaluated or included in the design
and licensing bases.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change replaces existing TS requirements related to
OPDRVs with new requirements on RPV WIC. The current requirements do
not have a stated safety basis and no margin of safety is
established in the licensing basis. The safety basis for the new
requirements is to protect Safety Limit 2.1.1.3. New requirements
are added to
[[Page 26104]]
determine the limiting time in which the RPV water inventory could
drain to the top of the fuel in the reactor vessel should an
unexpected draining event occur. Plant configurations that could
result in lowering the RPV water level to the TAF within one hour
are now prohibited. New escalating compensatory measures based on
the limiting drain time replace the current controls. The proposed
TS establish a safety margin by providing defense-in-depth to ensure
that the Safety Limit is protected and to protect the public health
and safety. While some less restrictive requirements are proposed
for plant configurations with long calculated drain times, the
overall effect of the change is to improve plant safety and to add
safety margin.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Anna Vinson Jones, Senior Counsel/Legal
Department, Entergy Services, Inc., 101 Constitution Avenue NW, Suite
200 East, Washington, DC 20001.
NRC Branch Chief: Robert J. Pascarelli.
Entergy Operations, Inc.; System Energy Resources, Inc.; Cooperative
Energy, A Mississippi Electric Cooperative; and Entergy Mississippi,
Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit No. 1 (GGNS),
Claiborne County, Mississippi
Date of amendment request: April 27, 2018. A publicly-available
version is in ADAMS under Accession No. ML18117A514.
Description of amendment request: The proposed amendment would
revise the Emergency Plan to adopt the Nuclear Energy Institute's
(NEI's) revised Emergency Action Level (EAL) scheme described in NEI
99-01, Revision 6, ``Development of Emergency Action Levels for Non-
Passive Reactors'' (ADAMS Accession No. ML110240324), which has been
endorsed by the NRC (ADAMS Accession No. ML12346A463).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to the GGNS EALs do not involve any
physical changes to plant equipment or systems and do not alter the
assumptions of any accident analyses. The proposed changes do not
adversely affect accident initiators or precursors and do not alter
design assumptions, plant configuration, or the manner in which the
plant is operated and maintained. The proposed changes do not
adversely affect the ability of structures, systems or components
(SSCs) to perform intended safety functions in mitigating the
consequences of an initiating event within the assumed acceptance
limits.
Therefore, the changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed changes.
The changes do not challenge the integrity or performance of any
safety-related systems. No plant equipment is installed or removed,
and the changes do not alter the design, physical configuration, or
method of operation of any plant SSC. Because EALs are not accident
initiators and no physical changes are made to the plant, no new
causal mechanisms are introduced.
Therefore, the changes do not create the possibility of a new or
different kind of accident from an accident previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
Margin of safety is associated with the ability of the fission
product barriers (i.e., fuel cladding, reactor coolant system
pressure boundary, and containment structure) to limit the level of
radiation dose to the public. The proposed changes do not impact
operation of the plant and no accident analyses are affected by the
proposed changes. The changes do not affect the Technical
Specifications or the method of operating the plant. Additionally,
the proposed changes will not relax any criteria used to establish
safety limits and will not relax any safety system settings. The
safety analysis acceptance criteria are not affected by these
changes. The proposed changes will not result in plant operation in
a configuration outside the design basis. The proposed changes do
not adversely affect systems that respond to safely shut down the
plant and to maintain the plant in a safe shutdown condition.
Therefore, the changes do not involve a significant reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Anna Vinson Jones, Senior Counsel/Legal
Department, Entergy Services, Inc., 101 Constitution Avenue NW, Suite
200 East, Washington, DC 20001.
NRC Branch Chief: Robert J. Pascarelli.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois, and Docket
Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle
County, Illinois
Date of amendment request: April 2, 2018. A publicly-available
version is in ADAMS under Accession No. ML18092B081.
Description of amendment request: The proposed amendments would
revise Technical Specification 3.2.3 to require that the axial flux
difference be maintained within the limits specified in the core
operating limits report during MODE 1 with reactor thermal power
greater or equal to 50 percent. An associated change would also be made
to the NOTE modifying surveillance 3.2.3.1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment requires that the AFD [axial flux
difference] be maintained within the limits specified in the COLR
[core operating limits report] at-all-times during MODE 1 when
reactor power is >=50% RTP [reactor thermal power]. This requirement
will ensure that all FRD [fuel rod design] performance criteria
remain satisfied during ANS [American Nuclear Society] Condition II
events (i.e., Faults of Moderate Frequency); thus, ensuring the
integrity of the fuel rod cladding. It is noted that maintaining AFD
within the COLR limits at-all-times when >=50% RTP is the normal
operating practice as specified in plant procedures.
The proposed change will have no impact on accident initiators
or precursors; does not alter accident analysis assumptions; does
not involve any physical plant modifications that would alter the
design or configuration of the facility, or the manner in which the
plant is maintained; and does not impact the probability of operator
error.
The proposed amendment will not impact the ability of
structures, systems, and components (SSCs) from performing their
intended functions to mitigate the consequences of an accident. All
accident analysis acceptance criteria will continue to be met as the
proposed change will not affect
[[Page 26105]]
the source term, containment isolation function, or radiological
release assumptions for any accident previously evaluated.
Based on the above discussion, the proposed amendment does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change formalizes the existing operating practice
of maintaining the AFD within the limits specified in the COLR at-
all-times during MODE 1 when reactor power is >= 50% RTP. This
change ensures that all FRD performance criteria remain satisfied
during ANS Condition II events. The ANS Condition II events have all
been previously evaluated in the Updated Final Safety Analysis
Report.
The proposed change does not involve a design change or other
changes that would impact safety-related SSCs from performing their
specified safety functions.
The proposed change does not result in the creation of any new
accident precursors; does not result in changes to any existing
accident scenarios; and does not introduce any operational changes
or mechanisms that would create the possibility of a new or
different kind of accident.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change to maintain the AFD within the limits
specified in the COLR at-all-times during MODE 1 when reactor power
is >= 50% RTP ensures that all FRD performance criteria remain
satisfied during ANS Condition II events; and thus, will maintain
the existing margin of safety related to FRD performance criteria
and ensure the integrity of the fuel rod cladding. The AFD limits
specified in the COLR have been established in accordance with the
analysis approach described in NRC-approved Westinghouse Topical
Reports.
In addition, this change will have no impact on the margin of
safety associated with other reactor core safety parameters such as
fuel hot channel factors, core power tilt ratios, loss of coolant
accident peak cladding temperature and peak local power density.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Branch Chief: David J. Wrona.
FirstEnergy Nuclear Operating Company, Docket No. 50-412, Beaver Valley
Power Station, Unit No. 2, Beaver County, Pennsylvania
Date of amendment request: March 28, 2018. A publicly-available
version is in ADAMS under Accession No. ML18087A293.
Description of amendment request: The amendment would revise
Technical Specification (TS) 5.5.5.2.d, ``Provisions for SG [Steam
Generator] Tube Inspection,'' and TS 5.5.5.2.f, ``Provisions for SG
Tube Repair Methods.'' More specifically, TSs 5.5.5.2.d.5 and
5.5.5.2.f.3 would be simplified and clarified, respectively, without
changing the intent of the specifications. Specification 5.5.5.2.f.3
would also be amended by changing the number of fuel cycles that
Westinghouse Electric Company, LLC leak-limiting Alloy 800 sleeves may
remain in operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Proposed amendment of Technical Specification 5.5.5.2.d.5 to
simplify the description of the required inspection region, and
Technical Specification 5.5.5.2.f.3 to clarify that this
specification is only applicable to sleeves installed in the steam
generator tubesheet and change the number of fuel cycles that an
Alloy 800 steam generator tubesheet sleeve may remain in service
from five to eight fuel cycles of operation, does not affect
structures, systems or components of the plant, plant operations,
design functions or analyses that verify the capability of
structures, systems or components to perform a design function. The
proposed amendment does not increase the likelihood of steam
generator tube sleeve leakage.
The proposed amendment of Technical Specification 5.5.5.2.d.5 to
simplify the description of the required inspection region, makes it
clear that the steam generator parent tube is to be inspected in the
areas where the joints will be established prior to installation of
the sleeve, regardless of the sleeve location. This proposed
amendment does not change the intent of the specification.
The proposed amendment of TS 5.5.5.2.f.3 includes two changes.
The first change would add the words ``installed in the hot-leg or
cold-leg tubesheet region'' after the words ``An Alloy 800 sleeve''
to make it clear that the specification only applies to Alloy 800
tube sleeves installed in the steam generator tubesheet. The design
of Alloy 800 sleeves installed in steam generator tube locations
other than the tubesheet does not include a nickel band. For these
sleeves, nondestructive examination methods have been demonstrated
to be effective and limits on sleeve operating life are not
necessary. This proposed amendment does not change the intent of the
specification.
The second change to TS 5.5.5.2.f.3, increases the number of
fuel cycles Alloy 800 tube sleeves installed in the tubesheet may
remain in service. The leak-limiting Alloy 800 sleeves are designed
using the applicable American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code and, therefore, meet the design
objectives of the original steam generator tubing. The applied
stresses and fatigue usage for the sleeves are bounded by the limits
established in the ASME Code. Mechanical testing has shown that the
structural strength of sleeves under normal, upset, emergency, and
faulted conditions provides margin to the acceptance limits. These
acceptance limits bound the most limiting (three times normal
operating pressure differential) burst margin of NRC Regulatory
Guide 1.121, ``Bases for Plugging Degraded PWR Steam Generator
Tubes.''
The leak-limiting Alloy 800 sleeve depth-based structural limit
is determined using NRC guidance and the pressure stress equation of
ASME Code, Section III with margin added to account for the
configuration of long axial cracks. Calculations show that a depth-
based limit of 45 percent through-wall degradation is acceptable.
However, Technical Specifications 5.5.5.2.c.2 and 5.5.5.2.c.3
provide additional margin by requiring an Alloy 800 sleeved tube to
be plugged on detection of any flaw in the sleeve or in the pressure
boundary portion of the original tube wall in the sleeve to tube
joint. Degradation of the original tube adjacent to the nickel band
of an Alloy 800 sleeve installed in the tubesheet, regardless of
depth, would not prevent the sleeve from satisfying design
requirements. Thus, flaw detection capabilities within the original
tube adjacent to the sleeve nickel band are a defense-in-depth
measure, and are not necessary in order to justify continued
operation of the sleeved tube.
Evaluation of repaired steam generator tube testing and analysis
indicates that there are no detrimental effects on the leak-limiting
Alloy 800 sleeve or sleeved tube assembly from reactor coolant
system flow, primary or secondary coolant chemistries, thermal
conditions or transients, or pressure conditions that may be
experienced at Beaver Valley Power Station, Unit No. 2. Westinghouse
is not aware of, and has no knowledge of any reports of parent-tube
stress corrosion cracking (SCC) in the sleeve roll joint region for
any Westinghouse sleeve design.
The proposed increase in the number of fuel cycles Alloy 800
tube sleeves installed in the tubesheet may remain in service has no
effect on sleeve operation or capability of the sleeve to perform
its design function. The mechanical and leakage tests have confirmed
[[Page 26106]]
that degradation of the parent tube adjacent to the nickel band will
not prevent the sleeve from satisfying its design function.
Consequences of a hypothetical failure of the leak-limiting
Alloy 800 sleeve and tube assembly are bounded by the current main
steam line break and steam generator tube rupture accident analyses
described in the Beaver Valley Power Station, Unit No. 2 Updated
Final Safety Analysis Report. The total number of plugged steam
generator tubes (including equivalency associated with installed
sleeves) is required to be consistent with accident analysis
assumptions. The sleeve and tube assembly leakage during plant
operation is required to be within the allowable Technical
Specification leakage limits and accident analysis assumptions.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
Proposed amendment of Technical Specification 5.5.5.2.d.5 to
simplify the description of the required inspection region, and
Technical Specification 5.5.5.2.f.3 to clarify that this
specification is only applicable to sleeves installed in the steam
generator tubesheet do not change the intent of these
specifications, and do not affect the design function or operation
of the tube sleeves. The proposed amendment of Technical
Specification 5.5.5.2.f.3 to change the number of fuel cycles that
an Alloy 800 steam generator tubesheet sleeve may remain in service
from five to eight fuel cycles of operation, does not affect the
design function or operation of the tube sleeves. Since these
changes do not create any credible new failure mechanisms,
malfunctions, or accident initiators not considered in the design or
licensing bases, the changes do not create the possibility of a new
or different kind of accident from any previously evaluated.
The leak-limiting Alloy 800 sleeves are designed using the
applicable ASME Code, and therefore meet the objectives of the
original steam generator tubing. As a result, the functions of the
steam generator will not be significantly affected by the
installation of the proposed sleeve. Therefore, the only credible
failure modes for the sleeve and tube are to leak or rupture, which
has already been evaluated. The continued integrity of the installed
sleeve and tube assembly is periodically verified as required by the
Technical Specifications, and a sleeved tube will be plugged on
detection of a flaw in the sleeve or in the pressure boundary
portion of the original tube wall in the sleeve to tube joint.
The proposed amendment to Technical Specification 5.5.5.2.f.3
increases the number of fuel cycles Alloy 800 tube sleeves installed
in the tubesheet may remain in service to eight fuel cycles of
operation. Implementation of this proposed amendment has no
significant effect on either the configuration of the plant, the
manner in which it is operated, or ability of the sleeve to perform
its design function.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Proposed amendment of Technical Specification 5.5.5.2.d.5 to
simplify the description of the required inspection region, and
Technical Specification 5.5.5.2.f.3 to clarify that this
specification is only applicable to sleeves installed in the steam
generator tubesheet, do not change the intent of these requirements
or reduce the margin of safety. The proposed amendment to Technical
Specification 5.5.5.2.f.3 to change the number of fuel cycles that
an Alloy 800 steam generator tubesheet sleeve may remain in service
from five to eight fuel cycles of operation, does not affect a
design basis or safety limit (that is, the controlling numerical
value for a parameter established in the Updated Final Safety
Analysis Report or the license) or reduce the margin of safety.
The proposed amendment to Technical Specification 5.5.5.2.f.3
increases the number of fuel cycles Alloy 800 tube sleeves installed
in the tubesheet may remain in service to eight fuel cycles of
operation. Implementation of this proposed amendment would not
affect a design basis or safety limit or reduce the margin of
safety. The repair of degraded steam generator tubes with leak-
limiting Alloy 800 sleeves restores the structural integrity of the
degraded tube under normal operating and postulated accident
conditions. Minimum reactor coolant system flow rate from the
cumulative effect of repaired (sleeved) and plugged tubes will be
greater than the flow rate limit established in the Technical
Specification limiting condition for operation 3.4.1. The design
safety factors utilized for the sleeves are consistent with the
safety factors in the American Society of Mechanical Engineers
Boiler and Pressure Vessel Code used in the original steam generator
design. Tubes with sleeves are subject to the same safety factors as
the original tubes, which are described in the performance criteria
for steam generator tube integrity in the existing Technical
Specifications. The sleeve and portions of the installed sleeve and
tube assembly that represent the reactor coolant pressure boundary
will be monitored, and a sleeved tube will be plugged if a flaw is
detected in the sleeve or in the pressure boundary portion of the
original tube wall in the leak-limiting sleeve and tube assembly.
Use of the previously-identified design criteria and design
verification testing ensures that the margin of safety is not
significantly different from the original steam generator tubes.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David W. Jenkins, FirstEnergy Nuclear
Operating Company, FirstEnergy Corporation, 76 South Main Street,
Akron, OH 44308.
NRC Branch Chief: James Danna.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: March 28, 2018. A publicly-available
version is in ADAMS under Accession No. ML18087A095.
Description of amendment request: The amendment would revise
Technical Specification (TS) 3/4.8.1, ``AC [Alternating Current]
Sources--Operating''; specifically, ACTION b concerning one inoperable
emergency diesel generator (EDG). The proposed change would remove the
Salem Nuclear Generating Station, Unit No. 3 (Salem Unit 3), gas
turbine generator and replace it with portable diesel generators.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change removes the requirement for the Salem Unit 3
gas turbine generator (GTG) and replaces it with the supplemental
power source during the existing extended allowable outage time for
the A or B EDG. The emergency diesel generators are safety related
components which provide backup electrical power supply to the
onsite Safeguards Distribution System. The emergency diesel
generators are not accident initiators; the EDGs are designed to
mitigate the consequences of previously evaluated accidents
including a loss of offsite power. (During normal operation, the
proposed portable diesel generators will not be connected to the
plant.)
The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility or the manner in which
the plant is operated and maintained. The proposed change does not
alter or prevent the ability of structures, systems, and components
(SSCs) from performing their intended function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change does not affect the source term,
containment isolation, or radiological release assumptions used in
evaluating the radiological consequences of an accident previously
evaluated. The proposed change is consistent with safety analysis
assumptions and resultant consequences.
Therefore, the proposed change does not involve a significant
increase in the
[[Page 26107]]
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change removes the requirement for the Salem Unit 3
gas turbine generator (GTG) and replaces it with the supplemental
power source during the existing extended allowable outage time for
the A or B EDG. The proposed change does not alter or involve any
design basis accident initiators. Equipment will be operated in the
same configuration and manner that is currently allowed and designed
for.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any [accident]
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not alter the permanent plant design,
including instrument set points, nor does it change the assumptions
contained in the safety analyses. The proposed change does not
impact the redundancy or availability requirements of offsite power
supplies or change the ability of the plant to cope with station
blackout [(SBO)] events.
The EDGs continue to meet their design requirements; there is no
reduction in capability or change in design configuration. The EDG
response to LOOP [loss of offsite power], LOCA [loss-of-coolant
accident], SBO, or fire is not changed by this proposed amendment;
there is no change to the EDG operating parameters. The remaining
operable emergency diesel generators are adequate to supply
electrical power to the onsite Safeguards Distribution System. The
proposed change does not alter a design basis or safety limit;
therefore it does not significantly reduce the margin of safety. The
EDGs will continue to operate per the existing design and regulatory
requirements.
Therefore, it is concluded that the proposed change does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, PSEG Nuclear LLC-N21,
P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: James G. Danna.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Unit Nos. 1 and 2 (SQN), Hamilton County, Tennessee
Date of amendment request: March 9, 2018, as supplemented by letter
dated April 11, 2018. Publicly-available versions are in ADAMS under
Accession Nos. ML18071A349 and ML18102B430, respectively.
Description of amendment request: The amendments would make changes
to the SQN Essential Raw Cooling Water (ERCW) Motor Control Centers
(MCCs) and revise the Updated Final Safety Analysis Report (UFSAR).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequence of an accident previously evaluated?
Response: No.
The proposed change does not alter the safety function of any
structure, system, or component, does not modify the manner in which
the plant is operated, and does not alter equipment out-of-service
time. In addition, this request does not degrade the ability of the
ERCW to perform its intended safety function. Therefore, the
proposed change does not involve a significant increase in the
probability or consequence of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not involve any physical changes to
plant safety related structure, system or component or alter the
modes of plant operation in a manner that is outside the bounds of
the system design analyses. The proposed change to complete the
design change for the removal of mechanical interlock device from
the feeder breakers and tie breakers for the ERCW MCCs and to revise
the ERCW System Description in Section 9.2.2.2 of the SQN UFSAR to
describe the normal and alternate power sources for the ERCW system
does not create the possibility for an accident or malfunction of a
different type than any evaluated previously in SQN's UFSAR. The
proposal does not alter the way any safety related structure, system
or component functions and does not modify the manner in which the
plant is operated. Therefore, the proposed change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change to remove the mechanical interlock device
from the feeder breakers and tie breakers for ERCW MCCs 1B-B and 2B-
B and to revise the ERCW System Description in Section 9.2.2.2 of
the SQN UFSAR to describe the normal and alternate power sources for
the ERCW system does not reduce the margin of safety because ERCW
will continue to perform its safety function. The design features
provided by the mechanical interlock device are not described in the
SQN UFSAR, are not credited in the SQN accident analysis and do not
provide any additional safety margin. The results of accident
analyses remain unchanged by this request. Therefore, the proposed
change does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
NRC Acting Branch Chief: Brian W. Tindell.
Vistra Operations Company LLC, Docket Nos. 50-445 and 50-446, Comanche
Peak Nuclear Power Plant, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: March 29, 2018. A publicly-available
version is in ADAMS under Accession No. ML18102A516.
Description of amendment request: The amendments would revise
Technical Specification 3.3.2, ``Engineered Safety Feature Actuation
System (ESFAS) Instrumentation,'' to change the applicability of when
the automatic auxiliary feedwater actuation due to the trip of all main
feedwater pumps is required to be operable at Comanche Peak Nuclear
Power Plant, Unit Nos. 1 and 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The design basis events which impose auxiliary feedwater safety
function requirements are loss of all AC [alternating current] power
to plant auxiliaries, loss of normal feedwater, steam generator
fault in either the feedwater or steam lines, and small break loss
of coolant accidents. These design basis event evaluations assume
actuation of auxiliary feedwater due to station blackout, low-low
steam generator level or a safety injection signal. The anticipatory
auxiliary feedwater automatic start signals from the main feedwater
pumps are not credited in any design basis accidents and are,
therefore, not part of the primary success path for postulated
accident mitigation as defined by 10 CFR 50.36(c)(2)(ii), Criterion
3. Modifying MODE 2 Applicability for this function will not impact
any previously evaluated design basis accidents.
[[Page 26108]]
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This technical specification change allows for an operational
allowance during MODE 2 while placing main feedwater pumps in
service. This change involves an anticipatory auxiliary feedwater
automatic start function that is not credited in the accident
analysis. Since this change only affects the conditions at which
this automatic start function needs to be operable and does not
affect the function that actuates auxiliary feedwater due to loss of
offsite power, low-low steam generator level or a safety injection
signal, it will not be an initiator to a new or different kind of
accident from any accident previously evaluated.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
This technical [s]pecification change involves the automatic
start of the auxiliary feedwater pumps due to trip of both main
feedwater pumps, which is not an assumed start signal for design
basis events. This change does not modify any values or limits
involved in a safety related function or accident analysis.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Timothy P. Matthews, Esq., Morgan, Lewis,
and Bockius, 1111 Pennsylvania Avenue NW, Washington, DC 20004.
NRC Branch Chief: Robert J. Pascarelli.
III. Notice of Issuance of Amendments to Facility Operating Licenses
and Combined Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, proposed no
significant hazards consideration determination, and opportunity for a
hearing in connection with these actions, was published in the Federal
Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment; (2) the amendment; and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment, as
indicated. All of these items can be accessed as described in the
``Obtaining Information and Submitting Comments'' section of this
document.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit No. 1, Lake County, Ohio
Date of amendment request: June 8, 2017.
Brief description of amendment: The amendment revised technical
specifications (TSs) to reflect previously approved changes made as
part of the alternative source term initiative. The amendment revised
the surveillance requirements for the control room emergency
recirculation and annulus exhaust gas treatment systems, which are
consistent with Technical Specification Task Force (TSTF) Traveler
TSTF-522, ``Revise Ventilation System Surveillance Requirement to
Operate for 10 Hours per Month.'' The amendment also deleted two TS
sections related to the fuel handling building and fuel handling
building ventilation exhaust system and increased the allowable
secondary containment leakage. Lastly, the amendment revised the TS
Table of Contents to reflect administrative changes to the titles of TS
sections.
Date of issuance: May 16, 2018.
Effective date: As of the date of issuance and shall be implemented
within 180 days of issuance.
Amendment No.: 180. A publicly-available version is in ADAMS under
Accession No. ML18110A133; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-58: The amendment revised the
Facility Operating License and TSs.
Date of initial notice in Federal Register: August 1, 2017 (82 FR
35841). The supplemental letter dated January 30, 2018, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the NRC staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 16, 2018.
No significant hazards consideration comments received: No.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold
Energy Center (DAEC), Linn County, Iowa
Date of amendment request: March 24, 2017.
Brief description of amendment: The amendment revised the DAEC
Technical Specification (TS) Table 3.3.2.1-1, ``Control Rod Block
Instrumentation,'' by relocating certain cycle-specific Minimum
Critical Power Ratio values to the DAEC Core Operating Limits Report.
The amendment also added a requirement to DAEC TS 5.6.5, ``Core
Operating Limits Report.''
Date of issuance: March 7, 2018.
Effective date: As of the date of its issuance and shall be
implemented by September 27, 2018. (Note: This Notice of Issuance
corrects the ``Effective date'' of Amendment No. 303 originally noticed
in the Federal Register on March 27, 2018 (83 FR 13153).
Amendment No.: 303. A publicly-available version is in ADAMS under
Accession No. ML18011A059; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment. Amendment
No. 303 was corrected by letter dated May 7, 2018 (ADAMS Accession No.
ML18081A074).
Renewed Facility Operating License No. DPR-49: The amendment
revised the Renewed Facility Operating License and TSs.
Date of initial notice in Federal Register: May 23, 2017 (82 FR
23627).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 7, 2018.
[[Page 26109]]
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 29th day of May, 2018.
For the Nuclear Regulatory Commission.
Gregory F. Suber,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2018-11843 Filed 6-4-18; 8:45 am]
BILLING CODE 7590-01-P