[Federal Register Volume 81, Number 196 (Tuesday, October 11, 2016)]
[Notices]
[Pages 70175-70190]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-24321]
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NUCLEAR REGULATORY COMMISSION
[NRC-2016-0207]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
AGENCY: Nuclear Regulatory Commission.
ACTION: Biweekly notice.
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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 September 13, 2016 to September 26, 2016.
The last biweekly notice was published on September 27, 2016.
DATES: Comments must be filed by November 10, 2016. A request for a
hearing must be filed by December 12, 2016.
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 Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0207. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions, contact
the individual listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
Mail comments to: Cindy Bladey, Office of Administration,
Mail Stop: OWFN-12-H08, 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: Kay Goldstein, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC
20555-0001; telephone: 301-415-1506, email: [email protected].
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2016-0207, 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 Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0207.
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-2016-0207, facility name, unit
number(s), 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 posts all comment submissions at http://www.regulations.gov as well as entering 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 submissions 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
[[Page 70176]]
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 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 a petition 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, which is available at the NRC's PDR, located at One White Flint
North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville,
Maryland 20852. The NRC's regulations are accessible electronically
from the NRC Library on the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a petition is filed within 60 days,
the Commission or a presiding officer designated by the Commission or
by the Chief Administrative Judge of the Atomic Safety and Licensing
Board Panel, will rule on the petition; and the Secretary or the Chief
Administrative Judge of the Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition shall set forth with
particularity the interest of the petitioner in the proceeding, and how
that interest may be affected by the results of the proceeding. The
petition should specifically explain the reasons why intervention
should be permitted with particular reference to the following general
requirements: (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. The petition
must also set forth the specific contentions which the petitioner seeks
to have litigated at 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 shall 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 those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion to support
its position on the issue. The petition must include sufficient
information to show that a genuine dispute exists with the applicant on
a material issue of law or fact. Contentions shall 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 these requirements 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,
and have the opportunity to participate fully in the conduct of the
hearing with respect to resolution of that person's admitted
contentions, including the opportunity to present evidence and to
submit a cross-examination plan for cross-examination of witnesses,
consistent with the NRC's regulations, policies, and procedures.
Petitions for leave to intervene must be filed no later than 60
days from the date of publication of this notice. Requests for hearing,
petitions for leave to intervene, and motions for leave to file new or
amended contentions that are filed after the 60-day 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).
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 decide 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 held
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 any 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 by December 12, 2016. 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. A State, local governmental body, Federally-recognized
Indian Tribe, or agency thereof may also have the opportunity to
participate under 10 CFR 2.315(c).
If a hearing is granted, any person who does not wish, or is not
qualified, to become a party to the proceeding may, in 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
[[Page 70177]]
written statement of 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, a petition for leave to intervene, any motion or
other document filed in the proceeding prior to the submission of a
request for hearing or petition to intervene (hereinafter
``petition''), and documents filed by interested governmental entities
participating 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 documents over the internet, or in
some cases to mail copies on electronic storage media. 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 request (1) a digital identification (ID)
certificate, which allows the participant (or its counsel or
representative) to digitally sign documents and access the E-Submittal
server for any proceeding in which it is participating; and (2) advise
the Secretary that the participant will be submitting a petition (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 Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing
the E-Submittal server are available on the NRC's public Web site at
http://www.nrc.gov/site-help/e-submittals/adjudicatory-sub.html.
Participants may attempt to use other software not listed on the Web
site, but should note that the NRC's E-Filing system does not support
unlisted software, and the NRC Electronic Filing Help Desk will not be
able to offer assistance in using unlisted software.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a petition.
Submissions should be in Portable Document Format (PDF). Additional
guidance on PDF submissions is available on the NRC's public Web site
at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing
is considered complete at the time the documents are 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 documents on those
participants separately. Therefore, applicants and other participants
(or their counsel or representative) must apply for and receive a
digital ID certificate before a hearing petition to intervene is filed
so that they can obtain access to the document via the E-Filing system.
A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC Electronic
Filing Help Desk through the ``Contact Us'' link located on the NRC's
public Web site 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 7 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, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville,
Maryland, 20852, Attention: Rulemaking and Adjudications Staff.
Participants filing a document 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
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the
Commission, or the presiding officer. Participants are requested not to
include personal privacy information, such as social security numbers,
home addresses, or home phone numbers in their filings, unless an NRC
regulation or other law requires submission of such information.
However, in some instances, a petition will require including
information on local residence in order to demonstrate a proximity
assertion of interest in the proceeding. 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.
The Commission will issue a notice or order granting or denying a
hearing request or intervention petition, designating the issues for
any hearing that will be held and designating the Presiding Officer. A
notice granting a hearing will be published in the Federal Register and
served on the parties to the hearing.
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.
[[Page 70178]]
Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear
Power Plant, Unit 1, New Hill, North Carolina
Date of amendment request: May 26, 2016. A publicly-available
version is in ADAMS under Accession No. ML16151A001.
Description of amendment request: The amendment would revise the
Shearon Harris Nuclear Power Plant, Unit 1, technical specifications
(TSs) to institute a new administrative program TS for the
establishment, implementation, and maintenance of a Diesel Fuel Oil
Testing Program, the specifics of which will be contained in a
licensee-controlled document. It also relocates to this program the
current TS surveillance requirements (SRs) for evaluating diesel fuel
oil, along with the SRs for the draining, sediment removal, and
cleaning of each main fuel oil storage tank at least once every 10
years. In addition, an exception is proposed to Regulatory Guide (RG)
1.137, Revision 1, ``Fuel Oil Systems for Standby Diesel Generators,''
for the allowance of performing sampling of new fuel oil offsite prior
to its addition to the fuel oil storage tanks.
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 amendment institutes a new administrative program
TS for the establishment, implementation, and maintenance of a
Diesel Fuel Oil Testing Program. The specifics of this program will
be contained in a licensee-controlled document. The current TS SR
for evaluating new and stored diesel fuel oil and the cleaning of
the fuel oil storage tanks will be relocated to this program. The
American Society for Testing and Materials (ASTM) standard
references pertaining to new and stored fuel oil will be relocated
to the aforementioned program; however, requirements to perform
testing in accordance with applicable ASTM standards are retained in
the TS. Requirements to perform surveillances of both new and stored
diesel fuel oil are also retained in the TS. Evaluations of future
changes to the licensee-controlled document will be conducted
pursuant to the requirements of 10 CFR 50.59. A more rigorous
testing of water and sediment content is added to the ``clear and
bright'' test used to establish the acceptability of new fuel oil
for use prior to its addition to the fuel oil storage tanks.
Additionally, an exception to RG 1.137 is proposed to allow for the
performance of new fuel oil sampling offsite. These changes will not
affect nor degrade the ability of the emergency diesel generators
(DGs) to perform their specified safety functions as the diesel fuel
oil continues to be properly evaluated.
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, and
configuration of the facility or the manner in which the plant is
operated and maintained. The proposed changes do not alter or
prevent the ability of structures, systems or components from
performing their intended function to mitigate the consequences on
an initiating event with the assumed acceptance limits. The proposed
changes do not affect the source term, containment isolation, or
radiological release assumptions used in evaluating the radiological
consequences of an accident previously evaluated. Further, the
proposed changes do not increase the types and amounts of
radioactive effluent that may be released offsite, nor significantly
increase individual or cumulative occupational or public radiation
exposure.
Therefore, the proposed changes do 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 amendment institutes a new administrative program
TS for the establishment, implementation, and maintenance of a
Diesel Fuel Oil Testing Program, of which the current TS SR for
evaluating new and stored diesel fuel oil and the cleaning of the
fuel oil storage tanks are relocated, including pertinent ASTM
standard references. A more rigorous testing of water and sediment
content is added to the ``clear and bright'' test used to establish
the acceptability of new fuel oil for use prior to its addition to
the fuel oil storage tanks. Additionally, an exception to RG 1.137
is proposed to allow for the performance of new fuel oil sampling
offsite. These changes do not alter the way any structure, system,
or component functions and does not modify the manner in which the
plant is operated. The requirements retained in the TS continue to
require testing of the diesel fuel oil to ensure the proper
functioning of the DGs.
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 amendment involve a significant reduction
in the margin of safety?
Response: No.
The proposed amendment institutes a new administrative program
TS for the establishment, implementation, and maintenance of a
Diesel Fuel Oil Testing Program, the specifics of which will be
contained in a licensee-controlled document. The current TS SR for
evaluating new and stored diesel fuel oil and the cleaning of the
fuel oil storage tanks will be relocated to this program, along with
the pertinent ASTM standard references. Changes to the licensee-
controlled document are performed in accordance with the provisions
of 10 CFR 50.59, thereby providing an effective level of regulatory
control and ensures that diesel fuel oil testing is conducted such
that there is no significant reduction in a margin of safety.
A more rigorous testing of water and sediment content is added
to the ``clear and bright'' test used to establish the acceptability
of new fuel oil for use prior to its addition to the fuel oil
storage tanks. Additionally, an exception to RG 1.137 is proposed to
allow for the performance of new fuel oil sampling offsite. The
margin of safety provided by the DGs is unaffected by the proposed
changes since there continue to be TS requirements to ensure fuel
oil is of the appropriate quality and reliability for emergency DG
use. The proposed changes provide the flexibility needed to improve
fuel oil sampling and analysis methodologies, while maintaining
sufficient controls to preserve the current margins of safety.
Based on the above, Duke Energy concludes that the proposed
amendment does not involve a significant hazards consideration under
the standards set forth in 10 CFR 50.92, and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
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 Business Services, 550 South Tryon Street, Mail Code
DEC45A, Charlotte, NC 28202.
NRC Acting Branch Chief: Jeanne A. Dion.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: August 29, 2016. A publicly-available
version is in ADAMS under Accession No. ML16242A332.
Description of amendment request: The amendment would revise
technical specification (TS) 5.5.6, Primary Containment Leak Rate
Testing Program. These revisions would extend the Type A Primary
Containment Integrated Leak Rate Test interval to 15 years and extend
the Type C Local Leak Rate Test testing interval up to 75 months.
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
[[Page 70179]]
consequences of an accident previously evaluated?
Response: No.
The proposed amendment to the TS involves the extension of the
JAF [James A. FitzPatrick Nuclear Power Plant] Type A containment
test interval to 15 years and the extension of the Type C test
interval to 75 months. The current Type A test interval of 120
months (10 years) would be extended on a permanent basis to no
longer than 15 years from the last Type A test. The current Type C
test interval of 60 months for selected components would be extended
on a performance basis to no longer than 75 months. Extensions of up
to nine months (total maximum interval of 84 months for Type C
tests) are permissible only for non-routine emergent conditions. The
proposed extension does not involve either a physical change to the
plant or a change in the manner in which the plant is operated or
controlled. The containment is designed to provide an essentially
leak tight barrier against the uncontrolled release of radioactivity
to the environment for postulated accidents. As such, the
containment and the testing requirements invoked to periodically
demonstrate the integrity of the containment exist to ensure the
plant's ability to mitigate the consequences of an accident, and do
not involve the prevention or identification of any precursors of an
accident. The change in dose risk for changing the Type A test
frequency from three-per-ten years to once-per-fifteen-years,
measured as an increase to the total integrated plant risk for those
accident sequences influenced by Type A testing, is 0.0087 person
rem/year. EPRI [Electric Power Research Institute] Report No.
1009325, Revision 2-A states that a very small population dose is
defined as an increase of <= 1.0 person-rem per year, or <= 1% of
the total population dose, whichever is less restrictive for the
risk impact assessment of the extended ILRT intervals. The results
of the risk assessment for this amendment meet these criteria.
Moreover, the risk impact for the ILRT extension when compared to
other severe accident risks is negligible. Therefore, this proposed
extension does not involve a significant increase in the probability
of an accident previously evaluated.
As documented in NUREG-1493, Type B and C tests have identified
a very large percentage of containment leakage paths, and the
percentage of containment leakage paths that are detected only by
Type A testing is very small. The JAF Type A test history supports
this conclusion.
The integrity of the containment is subject to two types of
failure mechanisms that can be categorized as: (1) Activity based,
and; (2) time based. Activity based failure mechanisms are defined
as degradation due to system and/or component modifications or
maintenance. Local leak rate test requirements and administrative
controls such as configuration management and procedural
requirements for system restoration ensure that containment
integrity is not degraded by plant modifications or maintenance
activities. The design and construction requirements of the
containment combined with the containment inspections performed in
accordance with ASME [American Society of Mechanical Engineers]
Section Xl, the Maintenance Rule, and TS requirements serve to
provide a high degree of assurance that the containment would not
degrade in a manner that is detectable only by a Type A test. Based
on the above, the proposed extensions do not significantly increase
the consequences of an accident previously evaluated.
The proposed amendment also deletes exceptions previously
granted to allow one time extensions of the ILRT test frequency for
JAF. These exceptions were for activities that would have already
taken place by the time this amendment is approved; therefore, their
deletion is solely an administrative action that has no effect on
any component and no impact on how the unit is operated.
Therefore, the proposed change does not result in 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 amendment to the TS involves the extension of the
JAF Type A containment test interval to 15 years and the extension
of the Type C test interval to 75 months. The containment and the
testing requirements to periodically demonstrate the integrity of
the containment exist to ensure the plant's ability to mitigate the
consequences of an accident do not involve any accident precursors
or initiators. The proposed change does not involve a physical
change to the plant (i.e., no new or different type of equipment
will be installed) or a change to the manner in which the plant is
operated or controlled.
The proposed amendment also deletes exceptions previously
granted to allow one time extensions of the ILRT test frequency for
JAF. These exceptions were for activities that would have already
taken place by the time this amendment is approved; therefore, their
deletion is solely an administrative action that does not result in
any change in how the unit is operated.
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 change involve a significant reduction in a
margin of safety?
Response: No.
The proposed amendment to TS 5.5.6 involves the extension of the
JAF Type A containment test interval to 15 years and the extension
of the Type C test interval to 75 months for selected components.
This amendment does not alter the manner in which safety limits,
limiting safety system set points, or limiting conditions for
operation are determined. The specific requirements and conditions
of the TS Containment Leak Rate Testing Program exist to ensure that
the degree of containment structural integrity and leak-tightness
that is considered in the plant safety analysis is maintained. The
overall containment leak rate limit specified by TS is maintained.
The proposed change involves only the extension of the interval
between Type A containment leak rate tests and Type C tests for JAF.
The proposed surveillance interval extension is bounded by the 15-
year ILRT Interval and the 75-month Type C test interval currently
authorized within NEI 94-01, Revision 3-A. Industry experience
supports the conclusion that Type B and C testing detects a large
percentage of containment leakage paths and that the percentage of
containment leakage paths that are detected only by Type A testing
is small. The containment inspections performed in accordance with
ASME Section Xl, TS and the Maintenance Rule serve to provide a high
degree of assurance that the containment would not degrade in a
manner that is detectable only by Type A testing. The combination of
these factors ensures that the margin of safety in the plant safety
analysis is maintained. The design, operation, testing methods and
acceptance criteria for Type A, B, and C containment leakage tests
specified in applicable codes and standards would continue to be
met, with the acceptance of this proposed change, since these are
not affected by changes to the Type A and Type C test intervals.
The proposed amendment also deletes exceptions previously
granted to allow one time extensions of the ILRT test frequency for
JAF. These exceptions were for activities that would have already
taken place by the time this amendment is approved; therefore, their
deletion is solely an administrative action and does not change how
the unit is operated and maintained. Thus, there is no reduction in
any margin of safety.
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. 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: Ms. Jeanne Cho, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Branch Chief: Travis L. Tate.
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station (CPS), Unit No.1, DeWitt County, Illinois
Date of amendment request: July 28, 2016. A publicly-available
version is in ADAMS under Accession No. ML16210A300.
Description of amendment request: The proposed changes supports
changes to the organization, staffing, and training requirements
contained in
[[Page 70180]]
Section 5.0 of the technical specifications (TSs) after the license no
longer authorizes operation of the reactor or placement or retention of
fuel in the reactor pressure vessel.
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 would not take effect until CPS has
permanently ceased operation and entered a permanently defueled
condition. The proposed changes would revise the CPS TS by deleting
or modifying certain portions of the TS administrative controls
described in Section 5.0 of the TS that are no longer applicable to
a permanently shutdown and defueled facility.
The proposed changes do not involve any physical changes to
plant structures, systems, and components (SSCs) or the manner in
which SSCs are operated, maintained, modified, tested, or inspected.
The proposed changes do not involve a change to any safety limits,
limiting safety system settings, limiting control settings, limiting
conditions for operation, surveillance requirements, or design
features.
The deletion and modification of provisions of the facility
administrative controls do not affect the design of SSCs necessary
for safe storage of spent irradiated fuel or the methods used for
handling and storage of such fuel in the Spent Fuel Pool (SFP). The
proposed changes are administrative in nature and do not affect any
accidents applicable to the safe management of spent irradiated fuel
or the permanently shutdown and defueled condition of the reactor.
In a permanently defueled condition, the only credible accidents
are the Fuel Handling Accident (FHA), Postulated Radioactive
Releases Due to Liquid Radwaste Tank Failures, and Cask Drop
Accident. Other accidents such as Loss of Coolant Accident, Loss of
Feedwater, and Reactivity and Power Distribution Anomalies will no
longer be applicable to a permanently defueled reactor plant.
The probability of occurrence of previously evaluated accidents
is not increased, since extended operation in a permanently defueled
condition will be the only operation allowed, and therefore, bounded
by the existing analyses. Additionally, the occurrence of postulated
accidents associated with reactor operation is no longer credible in
a permanently defueled reactor. This significantly reduces the scope
of applicable accidents.
The proposed changes in the administrative controls do not
affect the ability to successfully respond to previously evaluated
accidents and do not affect radiological assumptions used in the
evaluations. The proposed changes narrow the focus of nuclear safety
concerns to those associated with safely maintaining spent nuclear
fuel. These changes remove the implication that CPS can return to
operation once the final certification required by 10 CFR
50.82(a)(1)(ii) is submitted to the NRC. Any event involving safe
storage of spent irradiated fuel or the methods used for handling
and storage of such fuel in the SFP would evolve slowly enough that
no immediate response would be required to protect the health and
safety of the public or station personnel. Adequate communications
capability is provided to allow facility personnel to safely manage
storage and handling of irradiated fuel. As a result, no changes to
radiological release parameters are involved. There is no effect on
the type or amount of radiation released, and there is no effect on
predicted offsite doses in the event of an accident.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequence 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 changes to delete and/or modify certain TS
administrative controls have no impact on facility SSCs affecting
the safe storage of spent irradiated fuel, or on the methods of
operation of such SSCs, or on the handling and storage of spent
irradiated fuel itself. The proposed changes do not result in
different or more adverse failure modes or accidents than previously
evaluated because the reactor will be permanently shut down and
defueled and CPS will no longer be authorized to operate the
reactor.
The proposed changes will continue to require proper control and
monitoring of safety significant parameters and activities. The
proposed changes do not result in any new mechanisms that could
initiate damage to the remaining relevant safety barriers in support
of maintaining the plant in a permanently shutdown and defueled
condition (e.g., fuel cladding and SFP cooling). Since extended
operation in a defueled condition will be the only operation
allowed, and therefore bounded by the existing analyses, such a
condition does not create the possibility of a new or different kind
of accident.
The proposed changes do not alter the protection system design
or create new failure modes. The proposed changes do not involve a
physical alteration of the plant, and no new or different kind of
equipment will be installed. Consequently, there are no new
initiators that could result in a new or different kind of accident.
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 a
margin of safety?
Response: No.
The proposed changes involve deleting and/or modifying certain
TS administrative controls once the CPS facility has been
permanently shutdown and defueled. As specified in 10 CFR
50.82(a)(2), the 10 CFR 50 license for CPS will no longer authorize
operation of the reactor or emplacement or retention of fuel into
the reactor vessel following submittal of the certifications
required by 10 CFR 50.82(a)(1). As a result, the occurrence of
certain design basis postulated accidents are no longer considered
credible when the reactor is permanently defueled. The only
remaining credible accidents are the FHA, the Postulated Radioactive
Releases Due to Liquid Radwaste Tank Failures, and the Cask Drop
Accident. The FHA is the limiting Chapter 15 dose event for CPS in
its decommissioned state.
The proposed changes do not adversely affect the inputs or
assumptions of any of the design basis analyses that impact the FHA.
The proposed changes are limited to those portions of the TS
administrative controls that are not related to the safe storage and
maintenance of spent irradiated fuel.
These proposed changes do not directly involve any physical
equipment limits or parameters. The requirements that are proposed
to be revised and/or deleted from the CPS TS are not credited in the
existing accident analysis for the remaining applicable postulated
accidents; therefore, they do not contribute to the margin of safety
associated with the accident analysis. Certain postulated DBAs
[design-basis accidents] involving the reactor are no longer
possible because the reactor will be permanently shut down and
defueled and CPS will no longer be authorized to operate the
reactor.
Therefore, the proposed changes do not involve a significant
reduction in the 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: Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear,. 4300 Winfield Road, Warrenville, IL 60555.
Acting NRC Branch Chief: G. Edward Miller.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold
Energy Center, Linn County, Iowa
NextEra Energy Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit
No. 1, Rockingham County, New Hampshire
Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
[[Page 70181]]
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida
Date of amendment request: July 28, 2016. A publicly-available
version is in ADAMS under Accession No. ML16214A276.
Description of amendment request: The amendments would revise the
Technical Specifications (TS) consistent with Technical Specifications
Task Force Traveler 545, Revision 3, ``TS Inservice Testing [IST]
Program Removal & Clarify SR [Surveillance Requirement] Usage Rule
Application to Section 5.5 Testing'' (ADAMS Accession No. ML15294A555).
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 eliminates Technical Specifications (TS)
Section 5.5.6 and 5.5.7, ``Inservice Testing Program,'' for Duane
Arnold and Point Beach, respectively, and eliminates TS Section
6.8.4.i, ``Inservice Testing Program'' for St. Lucie Units 1 and 2.
The proposed change eliminates the requirements regarding [IST] from
TS 4.0.5 in the Seabrook and Turkey Point TS. Most requirements in
the [IST] Program are removed, as they are duplicative of
requirements in the ASME OM [American Society of Mechanical
Engineers Operation and Maintenance] Code, as clarified by Code Case
OMN-20, ``Inservice Test Frequency.'' The remaining requirements
related to the IST Program are eliminated because the NRC has
determined their inclusion in the TS is contrary to regulations. A
new defined term, ``Inservice Testing Program,'' is added to the TS,
which references the requirements of 10 CFR 50.55a(f).
Performance of [IST] is not an initiator to any accident
previously evaluated. As a result, the probability of occurrence of
an accident is not significantly affected by the proposed change.
Inservice test frequencies under Code Case OMN-20 are equivalent to
the current testing period allowed by the TS with the exception that
testing frequencies greater than 2 years may be extended by up to 6
months to facilitate test scheduling and consideration of plant
operating conditions that may not be suitable for performance of the
required testing. The testing frequency extension will not affect
the ability of the components to mitigate any accident previously
evaluated as the components are required to be operable during the
testing period extension. Performance of inservice tests utilizing
the allowances in OMN-20 will not significantly affect the
reliability of the tested components. As a result, the availability
of the affected components, as well as their ability to mitigate the
consequences of accidents previously evaluated, is not affected.
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.
The proposed change does not alter the design or configuration
of the plant. The proposed change does not involve a physical
alteration of the plant; no new or different kind of equipment will
be installed. The proposed change does not alter the types of [IST]
performed. In most cases, the frequency of [IST] is unchanged.
However, the frequency of testing would not result in a new or
different kind of accident from any previously evaluated since the
testing methods are not altered.
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 change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change eliminates some requirements from the TS in
lieu of requirements in the ASME Code, as modified by use of Code
Case OMN-20. Compliance with the ASME Code is required by 10 CFR
50.55a. The proposed change also allows inservice tests with
frequencies greater than 2 years to be extended by 6 months to
facilitate test scheduling and consideration of plant operating
conditions that may not be suitable for performance of the required
testing. The testing frequency extension will not affect the ability
of the components to respond to an accident as the components are
required to be operable during the testing period extension. The
proposed change will eliminate the existing TS allowance to defer
performance of missed inservice tests up to the duration of the
specified testing frequency, and instead will require an assessment
of the missed test on equipment operability. This assessment will
consider the effect on margin of safety (equipment operability).
Should the component be inoperable, the TS provide actions to ensure
that the margin of safety is protected. The proposed change also
eliminates a statement that nothing in the ASME Code should be
construed to supersede the requirements of any TS. The NRC has
determined that statement to be incorrect. However, elimination of
the statement will have no effect on plant operation or safety.
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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William Blair, Managing Attorney--Nuclear,
Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-
0420.
Acting NRC Branch Chief: Jeanne A. Dion.
Northern States Power Company--Minnesota (NSPM), Docket No. 50-263,
Monticello Nuclear Generating Plant, Wright County, Minnesota
Date of amendment request: July 28, 2016. A publicly-available
version is in ADAMS under Accession No. ML16210A030.
Description of amendment request: The proposed amendment would
eliminate technical specification (TS), Section 5.5.5, ``Inservice
Testing [IST] Program,'' to remove requirements duplicated in American
Society of Mechanical Engineers (ASME) Code for Operation and
Maintenance of Nuclear Power Plants (OM Code), Case OMN-20, ``Inservice
Test Frequency.'' A new defined term, ``Inservice Testing Program,'' is
added to TS Section 1.1, ``Definitions.'' The proposed change to the TS
is consistent with TSTF-545, Revision 3, ``TS Inservice Testing Program
Removal & Clarify SR [surveillance requirement] Usage Rule Application
to Section 5.5 Testing.'' TS SRs that currently refer to the IST
Program from Section 5.5.6 would be revised to refer to the new defined
term, ``Inservice Testing Program.''
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 revises TS Chapter 5, ``Administrative
Controls,'' Section 5.5, ``Programs and Manuals,'' by eliminating
the ``Inservice Testing Program'' specification. Most requirements
in the IST Program are removed as they are duplicative of
requirements in the ASME OM Code, as clarified by Code Case OMN-20,
``Inservice Test Frequency.'' The remaining requirements in the
Section 5.5 IST Program are eliminated because the NRC has
determined their inclusion in the TS is contrary to the regulations.
A new defined term, ``Inservice Testing Program,'' is added to the
TS, which references the requirements of 10 CFR 50.55a(f).
Performance of inservice testing is not an initiator to any
accident previously evaluated. As a result, the probability of
occurrence of an accident is not significantly affected by the
proposed change. Inservice test frequencies under Code Case OMN-20
[[Page 70182]]
are equivalent to the current testing period allowed by the TS with
the exception that testing frequencies greater than 2 years may be
extended by up to 6 months to facilitate test scheduling and
consideration of plant operating conditions that may not be suitable
for performance of the required testing. The testing frequency
extension will not affect the ability of the components to mitigate
any accident previously evaluated as the components are required to
be operable during the testing period extension. Performance of
inservice tests utilizing the allowances in OMN-20 will not
significantly affect the reliability of the tested components. As a
result, the availability of the affected components, as well as
their ability to mitigate the consequences of accidents previously
evaluated, is not affected.
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 does not alter the design or configuration
of the plant. The proposed change does not involve a physical
alteration of the plant; no new or different kind of equipment will
be installed. The proposed change does not alter the types of
inservice testing performed. In most cases, the frequency of
inservice testing is unchanged. However, the frequency of testing
would not result in a new or different kind of accident from any
previously evaluated since the testing methods are not altered.
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 a
margin of safety?
Response: No.
The proposed change eliminates some requirements from the TS in
lieu of requirements in the ASME Code, as modified by use of Code
Case OMN-20. Compliance with the ASME Code is required by 10 CFR
50.55a. The proposed change also allows inservice tests with
frequencies greater than 2 years to be extended by 6 months to
facilitate test scheduling and consideration of plant operating
conditions that may not be suitable for performance of the required
testing. The testing frequency extension will not affect the ability
of the components to respond to an accident as the components are
required to be operable during the testing period extension. The
proposed change will eliminate the existing TS SR 3.0.3 allowance to
defer performance of missed inservice tests up to the duration of
the specified frequency, and will instead require an assessment of
the missed test on equipment operability. This assessment will
consider the effect on a margin of safety (equipment operability).
Should the component be inoperable, the TS provide actions to ensure
that the margin of safety is protected. The proposed change also
eliminates a statement that nothing in the ASME Code should be
construed to supersede the requirements of any TS. The NRC has
determined that statement to be incorrect. However, elimination of
the statement will have no effect on plant operation or safety.
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: Peter M. Glass, Assistant General Counsel,
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
NRC Branch Chief: David J. Wrona.
South Carolina Electric & Gas Company and South Carolina Public Service
Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear
Station, Units 2 and 3, Fairfield County, South Carolina
Date of amendment request: August 12, 2016. A publicly-available
version is in ADAMS under Accession No. ML16225A437.
Description of amendment request: The amendment request proposes
changes to plant-specific Tier 2 information incorporated into the
Updated Final Safety Analysis Report (UFSAR), and involves changes to
combined license Appendix C (and corresponding plant-specific Tier 1
information). The proposed changes are to information identifying the
frontal face area and screen surface area for the In-Containment
Refueling Water Storage Tank (IRWST) screens, the location and
dimensions of the protective plate located above the containment
recirculation (CR) screens, and increasing the maximum Normal Residual
Heat Removal System (RNS) flowrate through the IRWST and CR screens.
Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from
elements of the design as certified in the 10 CFR part 52, appendix D,
design certification rule is also requested for the plant-specific
Design Control Document Tier 1 material departures.
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 location and dimensions of the
protective plate continues to provide sufficient space surrounding
the containment recirculation screens for debris to settle before
reaching the screens as confirmed by an evaluation demonstrating
that the protective plate continues to fulfill its design function
of preventing debris from reaching the screens. In addition, the
increase to the minimum IRWST screen size reinforces the ability of
the screens to perform their design function with the increased RNS
maximum flowrate proposed. The proposed changes do not adversely
affect any accident initiating component, and thus the probabilities
of the accidents previously evaluated are not affected. The affected
equipment does not adversely affect the ability of equipment to
contain radioactive material. Because the proposed change does not
affect a release path or increase the expected dose rates, the
potential radiological releases in the UFSAR accident analyses are
unaffected.
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.
The proposed activity to change the location and dimensions of
the protective plate above the containment recirculation screens, to
change the minimum IRWST screen size, and to increase the maximum
RNS flowrate through the IRWST and CR screens does not alter the
method in which safety functions are accomplished. The analyses
demonstrate that the screens are able to perform accident, and no
new failure modes are introduced by the proposed change.
Therefore, the proposed amendment 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 the design does not change any of the
codes or standards to which the IRWST screens, containment
recirculation screens, and containment recirculation screen
protective plate are designed as documented in the UFSAR. The
containment recirculation screen protective plate continues to
prevent debris from reaching the CR screens, and the IRWST and CR
screens maintain their ability to block debris while at the proposed
increase in RNS maximum flowrate.
No safety analysis or design basis acceptance limit/criterion is
challenged or exceeded by the proposed changes.
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
[[Page 70183]]
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis &
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC, 20004-2514.
NRC Branch Chief: Jennifer Dixon-Herrity.
South Carolina Electric & Gas Company and South Carolina Public Service
Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear
Station, Units 2 and 3, Fairfield, South Carolina
Date of amendment request: September 8, 2016. A publicly-available
version is in ADAMS under Accession No. ML16252A200.
Description of amendment request: The amendment request proposes
changes to the Fire Pump Head and Diesel Fuel Day Tank. Because, this
proposed change requires a departure from Tier 1 information in the
Westinghouse Electric Company's AP1000 Design Control Document (DCD),
the licensee also requested an exemption from the requirements of the
Generic DCD Tier 1 in accordance with 10 CFR 52.63(b)(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 amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The increase in head pressure by the proposed change to the fire
protection system (FPS) motor-driven and diesel-driven fire pumps
maintains compliance with National Fire Protection Association
(NFPA) Standard NFPA-14, Standard for the Installation of Standpipe,
Private Hydrants, and Hose Systems, 2000 Edition, requirements by
providing adequate pressure in the standpipe and automatic sprinkler
system to maintain the ability to fight and/or contain a postulated
fire. The proposed change to the diesel-driven fire pump fuel day
tank volume maintains the availability of the diesel-driven fire
pump for service upon failure of the electric motor-driven fire pump
or a loss of offsite power by providing a fuel day tank that is
reserved exclusively for the diesel-driven pump and meets the
minimum capacity requirements of NFPA 20, Standard for the
Installation of Stationary Pumps for Fire Protection, 1999 Edition.
These changes do not affect the operation of any systems or
equipment that initiate an analyzed accident or alter any
structures, systems, and component's (SSC's) accident initiator or
initiating sequence of events.
These changes have no adverse impact on the support, design, or
operation of mechanical and fluid systems. The response of systems
to postulated accident conditions is not adversely affected by the
proposed changes. There is no change to the predicted radioactive
releases due to normal operation or postulated accident conditions.
Consequently, the plant response to previously evaluated accidents
is not impacted, nor does the proposed change create any new
accident precursors.
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.
The proposed changes do not affect the operation of any systems
or equipment that may initiate a new or different kind of accident,
or alter any SSC such that a new accident initiator or initiating
sequence of events is created. The proposed changes to the fire pump
performance specifications and fire pump fuel day tank volume do not
affect any safety-related equipment, nor do they add any new
interface to safety-related SSCs. No system or design function or
equipment qualification is affected by this change. The changes do
not introduce a new failure mode, malfunction, or sequence of events
that could affect safety or safety-related equipment.
Therefore, the proposed amendment 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 changes maintain compliance with the applicable
Codes and Standards, thereby maintaining the margin of safety
associated with these SSCs. The proposed changes do not alter any
applicable design codes, code compliance, design function, or safety
analysis. Consequently, no safety analysis or design basis
acceptance limit/criterion is challenged or exceeded by the proposed
change, thus the margin of safety is not reduced.
Because no safety analysis or design basis acceptance limit/
criterion is challenged or exceeded by these changes, no margin of
safety is reduced.
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: Ms. Kathryn M. Sutton, Morgan, Lewis &
Bockius, LLC, 1111 Pennsylvania NW., Washington, DC 20004-2514.
NRC Branch Chief: Jennifer Dixon-Herrity.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit
No. 1, Fairfield County, South Carolina
Date of amendment request: August 29, 2016. A publicly-available
version is in ADAMS under Accession No. ML16243A463.
Description of amendment request: The amendment would remove the
administrative controls associated with the Limiting Condition for
Operation (LCO) of Technical Specification (TS) 3.5.4, ``Refueling
Water Storage Tank.''
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, with NRC staff edits in square
brackets:
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 an administrative note added by
Amendment No. 192. The administrative control applied by Amendment
No. 192 was issued to prevent or reduce the risk for drainage of the
Reactor Water Storage Tank (RWST) when aligned to the non-safety,
non-seismic purification system. The station has implemented a
modification that qualifies the interconnection of the RWST to the
purification system. The installed design prevents the RWST being
drained below the current Technical Specifications minimum volume
requirement due to a failure in the non-safety purification system.
The RWST will continue to perform its safety function and the
overall system performance has not been affected [by] this proposed
amendment. Assumptions previously made in evaluating the
consequences of the accident are not altered, and the consequences
of the accident are not increased. Therefore, the proposed change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated. The Purification
Loop supports the Spent Fuel System and is not credited for safe
shutdown of the plant or accident mitigation. Therefore, the
proposed change has insignificant impact on the probability and
consequences of an accident previously evaluated. A combination of
design and administrative controls ensure that the Purification Loop
maintains RWST boron concentration and water volume requirements
whenever the contents of the RWST are processed through the system.
The RWST is operated under System Operating Procedure for the Spent
Fuel Cooling System and is protected by maintaining the isolation
valve for the lower return line locked closed in modes 1 through 4.
2. Does the proposed change create the possibility of a new or
different kind of
[[Page 70184]]
accident from any accident previously evaluated?
Response: No.
The proposed change does not introduce a new or different
accident previously evaluated. The station implemented a qualified
design that prevents the RWST from being drained below the current
TS 3.5.4.a minimum volume requirement. The proposed change does not
alter the design requirements of the RWST or any Structure, System
or Component or its function during accident conditions. The changes
do not alter assumptions made in the safety analysis and the current
TS LCO are maintained. The Purification Loop supports the Spent Fuel
System and is not credited for safe shutdown of the plant or
accident mitigation. The proposed change removes a note added by
Amendment No. 192 that applied an administrative control to manage
the risk of a postulated RWST drainage scenario by the purification
system.
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 change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change removes a note added by Amendment No. 192.
The proposed change does not alter the safety limits, limiting
safety system settings or limiting conditions for operation of the
RWST. The modification preserved the current licensing and design
bases of the RWST, therefore the margin of safety for the RWST are
not affected. The proposed changes do not adversely affect systems
that respond to safely shutdown the plant and to maintain the plant
in a safe shutdown condition. The Purification Loop supports the
Spent Fuel System and is not credited for safe shutdown of the plant
or accident mitigation.
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 M. Sutton, Morgan, Lewis & Bockius
LLP, 1111 Pennsylvania Avenue NW., Washington, DC 20004.
NRC Branch Chief: Michael T. Markley.
Southern Nuclear Operating Company, Inc. (SNC); Georgia Power Company;
Oglethorpe Power Corporation; Municipal Electric Authority of Georgia;
City of Dalton, Georgia, Docket No. 50-366, Edwin I. Hatch Nuclear
Plant (HNP), Unit No. 2, Appling County, Georgia
Date of amendment request: August 29, 2016. A publicly-available
version is in ADAMS under Accession No. ML16245A257.
Description of amendment request: The amendment would revise the
values for the reactor core Safety Limit 2.1.1.2 for Minimum Critical
Power Ratios for both single and dual recirculation loop 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, with NRC staff edits in
brackets:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The Safety Limit Minimum Critical Power Ratio (SLMCPR) ensures
that 99.9% of the fuel rods in the core will not be susceptible to
boiling transition during normal operation or the most limiting
postulated design-basis transient event. The new SLMCPR values
preserve the existing margin to the onset of transition boiling;
therefore, the probability of fuel damage is not increased as a
result of this proposed change. The determination of the revised HNP
Unit 2 SLMCPRs has been performed using NRC-approved methods of
evaluation. These plant-specific calculations are performed each
operating cycle and may require changes for future cycles. The
revised SLMCPR values do not change the method of operating the
plant; therefore, they have no effect on the probability of an
accident, initiating event, or transient:
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 changes result only from a specific analysis for
the HNP Unit 2 core reload design. These changes do not involve any
new or different methods for operating the facility. No new
initiating events or transients result from these changes.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The new SLMCPRs have been calculated using NRC-approved methods
of evaluation with plant and cycle-specific input values for the
fuel and core design for the upcoming cycle of operation. The SLMCPR
values ensure that 99.9% of the fuel rods in the core will not be
susceptible to boiling transition during normal operation or the
most limiting postulated design-basis transient event. The operating
MCPR limit is set appropriately above the safety limit value to
ensure adequate margin when the cycle-specific transients are
evaluated. Accordingly, the margin of safety is maintained with the
revised values.
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: Jennifer M. Buettner, Associate General
Counsel, Southern Nuclear Operating Company, 40 Iverness Center
Parkway, Birmingham, AL 35242.
NRC Branch Chief: Michael T. Markley.
Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 52-
026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke
County, Georgia
Date of amendment request: July 29, 2016. A publicly-available
version is in ADAMS under Accession No. ML16211A436.
Description of amendment request: The amendment request proposes to
add to License Condition 2.D.(1) of the VEGP Units 3 and 4 combined
licenses an Interim Amendment Request process for changes during
construction when emergent conditions are present.
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 with NRC staff's edits in
square brackets:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment would add an Interim Amendment Request
process to Condition 2.0.(1) of the Vogtle 3 and 4 COLs [combined
licenses] to allow construction to continue, at SNC's [Southern
Nuclear Operating Company] own risk, in emergent conditions, where a
non-conforming condition that has little or no safety significance
is discovered and the work activity cannot be adjusted. The Interim
Amendment Request process would require SNC to submit a Nuclear
Construction Safety Assessment which (1) identifies the proposed
change; (2) evaluates whether emergent conditions are present; (3)
evaluates whether the change would result in any material decrease
in safety; and (4) evaluates whether continued construction would
make the non-conforming condition irreversible. Only if the
continued construction would have no material decrease in safety
would the NRC issue a determination that construction could continue
pending SNC's initiation of the COL-ISG-025 PAR [preliminary
amendment request]/LAR [license amendment request] process. The
requirement to include a Nuclear Construction Safety Assessment
ensures that the proposed amendment would not involve a significant
increase in the probability or consequences of an accident
previously evaluated. If the continued construction would result a
material decrease in safety, then continued construction would not
be authorized.
[[Page 70185]]
The proposed amendment does not modify the design, construction,
or operation of any plant structures, systems, or components (SSCs),
nor does it change any procedures or method of control for any SSCs.
Because the proposed amendment does not change the design,
construction, or operation of any SSCs, it does not adversely affect
any design function as described in the Updated Final Safety
Analysis Report.
The proposed amendment does not affect the probability of an
accident previously evaluated. Similarly, because the proposed
amendment does not alter the design or operation of the nuclear
plant or any plant SSCs, the proposed amendment does not represent a
change to the radiological effects of an accident, and therefore,
does not involve an increase in the consequences of an accident
previously evaluated.
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.
The proposed amendment would add an Interim Amendment Request
process to Condition 2.0.(1) of the Vogtle 3 and 4 COLs to allow
construction to continue, at SNC's own risk, in emergent conditions,
where a non-conforming condition that has little or no safety
significance is discovered and the work activity cannot be adjusted.
The Interim Amendment Request process would require SNC to submit a
Nuclear Construction Safety Assessment which (1) identifies the
proposed change; (2) evaluates whether emergent conditions are
present; (3) evaluates whether the change would result in any
material decrease in safety; and (4) evaluates whether continued
construction would make the non-conforming condition irreversible.
Only if the continued construction would have no material decrease
in safety would NRC issue a determination that construction could
continue pending SNC's initiation of the COL-ISG-025 PAR/LAR
process.
The proposed amendment is not a modification, addition to, or
removal of any plant SSCs. Furthermore, the proposed amendment is
not a change to procedures or method of control of the nuclear plant
or any plant SSCs. The proposed amendment only adds a new screening
process and does not change the design, construction, or operation
of the nuclear plant or any plant operations.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment would add an Interim Amendment Request
process to Condition 2.0.(1) of the Vogtle 3 and 4 COLs to allow
construction to continue, at SNC's own risk, in emergent conditions,
where a non-conforming condition that has little or no safety
significance is discovered and the work activity cannot be adjusted.
The Interim Amendment Request process would require SNC to submit a
Nuclear Construction Safety Assessment which (1) identifies the
proposed change; (2) evaluates whether emergent conditions are
present; (3) evaluates whether the change would result in any
material decrease in safety; and (4) evaluates whether continued
construction would make the non-conforming condition irreversible.
Only if the continued construction would have no material decrease
in safety would the NRC issue determination that construction could
continue pending SNC's initiation of the COL-ISG-025 PAR/LAR
process.
The proposed amendment is not a modification, addition to, or
removal of any plant SSCs. Furthermore, the proposed amendment is
not a change to procedures or method of control of the nuclear plant
or any plant SSCs. The proposed amendment does not alter any design
function or safety analysis. Consequently, no safety analysis or
design basis acceptance limit/criterion is challenged or exceeded by
the proposed amendment, thus the margin of safety is not reduced.
The only impact of this activity is the addition of an Interim
Amendment Request process.
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: M. Stanford Blanton, Balch & Bingham LLP,
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Inc. Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant Units 3 and 4, Burke County, Georgia
Date of amendment request: September 9, 2016. A publicly-available
version is in ADAMS under Accession No. ML16253A412.
Description of amendment request: The amendment request proposes
changes to update the Protection and Safety Monitoring System (PMS)
design, specifically the description of the roles of the Qualified Data
Processing System (QDPS) and the safety displays. The proposed changes
add Main Control Room (MCR) safety-related display divisions A and D to
plant-specific Tier 1 (and associated COL Appendix C) and the Updated
Final Safety Analysis Report (UFSAR), and correct the name of the QDPS
in the UFSAR by referring to the QDPS as a system, rather than a
subsystem. Because, this proposed change requires a departure from Tier
1 information in Westinghouse Electric Company's AP1000 Design Control
Document (DCD), the licensee also requested an exemption from the
requirements of the Generic DCD Tier 1 in accordance with 10 CFR
52.63(b)(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 amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to the roles of the qualified data
processing system (QDPS) and safety-related displays, as well as the
change to add Division A and Division D of the main control room
(MCR) safety-related displays to the listing of PMS equipment, as
identified in Combined License (COL) Appendix C (and plant-specific
Tier 1) Table 2.5.2-1 and Updated Final Safety Analysis Report
(UFSAR) Table 3.11-1 and 3l.6-2 do not alter any accident initiating
component/system failure or event, thus the probabilities of the
accidents previously evaluated are not affected.
The proposed changes do not adversely affect safety-related
equipment or a radioactive material barrier, and this activity dos
not involve the containment of radioactive material.
The radioactive material source terms and release paths used in
the safety analysis are unchanged, thus the radiological releases in
the UFSAR accident analysis are not affected.
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.
The proposed change to the roles of the QDPS and safety-related
displays, as well as the change to add Division A and Division D of
the MCR safety-related displays to the listing of PMS equipment, as
identified in COL Appendix C (and plant-specific Tier 1) Table
2.5.2-1 and UFSAR Table 3.11-1 and 3l.6-2 does not create the
possibility of a new or different kind of accident from any accident
previously evaluated. The proposed changes do not alter the design
or capability of any sensors which provide input to the QDPS. The
functionality of the QDPS to process the input obtained from sensors
into data to be sent to the safety displays is not affected by the
proposed changes. The proposed changes do not affect any functions
performed by the safety displays, nor do the proposed changes affect
the capability of the safety displays to display the data received
from the QDPS.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
[[Page 70186]]
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
There is no safety-related structure, system or component (SSC)
or function adversely affected by the proposed change to the roles
of the QDPS and safety-related displays, nor by the change to add
Division A and Division D of the MCR safety-related displays to the
listing of Protection and Safety Monitoring System (PMS) equipment.
The proposed changes do not alter the mechanisms by which system
components are actuated or controlled. Because no safety analysis or
design basis acceptance limit/criterion is challenged or exceeded by
the proposed changes, no margin of safety is reduced.
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: Mr. M. Stanford Blanton, Balch & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 52-
026, Vogtle Electric Generating Plant Units 3 and 4, Burke County,
Georgia
Date of amendment request: September 9, 2016. A publicly-available
version is in ADAMS under Accession No. ML16253A204.
Description of amendment request: The amendment request proposes
changes to revise plant-specific Tier 1, plant-specific Tier 2, and
combined license (COL) Appendix C information concerning the details of
the Class 1E direct current and uninterruptible power supply system
(IDS), specifically adding seven Class 1E fuse panels to the IDS
design. These proposed changes provide electrical isolation between the
non-Class 1E IDS battery monitors and their respective Class 1E battery
banks. Because, this proposed change requires a departure from Tier 1
information in the Westinghouse Electric Company's AP1000 Design
Control Document (DCD), the licensee also requested an exemption from
the requirements of the Generic DCD Tier 1 in accordance with 10 CFR
52.63(b)(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 with NRC staff edits in square
brackets:
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 revise plant-specific Tier 1, COL
Appendix C, and [Updated Final Safety Analysis Report (UFSAR)]
information concerning details of the IDS, specifically the addition
of seven Class 1E fuse isolation panels at the interconnection of
the non-Class 1E IDS battery monitors and Class 1E IDS circuits, are
necessary to conform to Regulatory Guide 1.75 Rev. 2 (consistent
with UFSAR Appendix 1A exceptions) and IEEE 384-1981 to prevent a
fault on non-Class 1E circuits or equipment from degrading the
operation of Class 1E IDS circuits and equipment below an acceptable
level. The proposed changes do not adversely affect the design
functions of the IDS, including the Class 1E battery banks and the
battery monitors.
These proposed changes to revise plant-specific Tier 1, COL
Appendix C, and UFSAR information concerning details of the IDS,
specifically the addition of seven Class 1E fuse isolation panels at
the interconnection of the non-Class 1E IDS battery monitors and
Class 1E IDS circuits as described in the current licensing basis do
not have an adverse effect on any of the design functions of any
plant systems. The proposed changes do not adversely affect any
plant electrical system and do not affect the support, design, or
operation of mechanical and fluid systems required to mitigate the
consequences of an accident. There is no change to plant systems or
the response of systems to postulated accident conditions. There is
no change to the predicted radioactive releases due to postulated
accident conditions. The plant response to previously evaluated
accidents or external events is not adversely affected, nor do the
proposed changes create any new accident precursors.
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.
The proposed changes to revise plant-specific Tier 1, COL
Appendix C, and UFSAR information concerning details of the IDS,
specifically the addition of seven Class 1E fuse isolation panels at
the interconnection of the non-Class 1E IDS battery monitors and
Class 1E IDS circuits, are necessary to conform to Regulatory Guide
1.75 Rev. 2 (consistent with UFSAR Appendix 1A exceptions) and IEEE
384-1981 to prevent a fault on non-Class 1E circuits or equipment
from degrading the operation of Class 1E IDS circuits and equipment
below an acceptable level. The proposed changes do not adversely
affect any plant electrical system and do not adversely affect the
design function, support, design, or operation of mechanical and
fluid systems. The proposed changes do not result in a new failure
mechanism or introduce any new accident precursors. No design
function described in the UFSAR is adversely affected by the
proposed changes.
Therefore, the proposed amendment 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.
There is no safety-related [structure, system, and component
(SSC)] or function adversely affected by the proposed change to add
IDS fuse isolation panels to non-Class 1E IDS battery monitors and
Class 1E IDS circuits. No safety analysis or design basis acceptance
limit/criterion is challenged or exceeded by the proposed changes
and no margin or safety is reduced.
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: Mr. M. Stanford Blanton, Balch & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 52-
026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County,
Georgia
Date of amendment request: September 13, 2016. A publicly-available
version is in ADAMS under Accession No. ML16257A711.
Description of amendment request: The amendment request proposes
changes to the Updated Final Safety Analysis Report (UFSAR) in the form
of departures from the incorporated plant-specific Design Control
Document Tier 2* information. The proposed departure consists of
changes to Tier 2* information in the UFSAR to change the provided
minimum reinforcement area in the column line 7.3 wall from elevation
82'-6'' to elevation 100'-0''.
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.
As indicated in the UFSAR Subsection 3H.5.1.2, the wall at
column line 7.3 is a shear wall that connects the shield building
and the nuclear island exterior wall at column line I. Deviations
were identified in
[[Page 70187]]
the constructed wall from the design requirements. The wall was
repaired in accordance with American Concrete Institute (ACI) 349-
01. This change impacts UFSAR Table 3H.5-5. For the south face of
the Vogtle Unit 3 column line 7.3 wall, the provided minimum steel
for wall section 11 for the vertical reinforcement from the wall
segment of elevation 82'-6'' to 100'-0'' is decreased from 3.12
in\2\/ft to 3.08 in\2\/ft. The change of the provided versus
required vertical reinforcing steel does not change the performance
of the affected portion of the auxiliary building for postulated
loads. The criteria and requirements of ACI 349-01 provide a margin
of safety to structural failure. The design of the auxiliary
building structure conforms to criteria and requirements in ACI 349-
01 and therefore maintains the margin of safety. This change does
not involve any accident initiating components or events, thus
leaving the probabilities of an accident unaltered. The reduced
margin does not adversely affect any safety-related structures or
equipment nor does the reduced margin reduce the effectiveness of a
radioactive material barrier. Thus, the proposed change would not
affect any safety-related accident mitigating function served by the
containment internal structures.
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.
The reduction of the provided versus required vertical
reinforcing steel does not change the performance of the affected
portion of the auxiliary building. As demonstrated by the continued
conformance to the applicable codes and standards governing the
design of the structures, the wall withstands the same effects as
previously evaluated. There is no change to the design function of
the wall, and no new failure mechanisms are identified as the same
types of accidents are presented to the wall before and after the
change.
Therefore, the proposed amendment 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 of the provided versus required vertical
reinforcing steel, identified in UFSAR Table 3H.5-5, is not a
significant reduction in the margin of safety. For the south face of
the Vogtle Unit 3 column line 7.3 wall, the provided minimum steel
for wall section 11 for the vertical reinforcement from the wall
segment of elevation 82'-6'' to 100'-0'' is decreased from 3.12
in\2\/ft to 3.08 in\2\/ft. The change of the provided versus
required vertical reinforcing steel does not change the performance
of the affected portion of the auxiliary building for postulated
loads. The criteria and requirements of ACI 349-01 provide a margin
of safety to structural failure. The design of the auxiliary
building structure conforms to criteria and requirements in ACI 349-
01 and therefore maintains the margin of safety. The reduction in
margin does not alter any design function, design analysis, or
safety analysis input or result, and sufficient margin exists to
justify departure from the Tier 2 * requirements for the wall. As
such, because the system continues to respond to design basis
accidents in the same manner as before without any changes to the
expected response of the structure, no safety analysis or design
basis acceptance limit/criterion is challenged or exceeded by the
proposed changes. Accordingly, no significant safety margin is
reduced by the change.
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: M. Stanford Blanton, Balch & Bingham LLP,
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer Dixon-Herrity.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama
Date of amendment request: August 12, 2016. A publicly-available
version is in ADAMS under Accession No. ML16225A663.
Description of amendment request: The amendments would modify the
Technical Specifications (TSs) for Units 1, 2, and 3 by revising TS
4.3.1.2, ``Fuel Storage Criticality,'' to preclude the placement of
fuel in the new fuel storage vaults. This TS change would remove the
existing TS 4.3.1.2 criticality criteria wording in its entirety, and
replaces it with language that specifically restricts the placement of
fuel in the new fuel storage vaults.
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 amendment does not change the fuel handling
processes, the fuel handling equipment, or require alteration of the
plant fuel storage systems. The amendment places a restriction on
use of the new fuel storage vaults, requiring that new fuel be
placed only in the spent fuel pool racks. Because no changes to fuel
handling equipment, fuel storage systems, or fuel handling processes
are involved, the proposed amendment does not increase the
probability or consequences of a fuel handling accident.
Therefore, the proposed change does not increase the probability
or consequences of a previously evaluated accident.
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 modification to the Technical Specifications does
not require changes to the plant hardware or alter the operating
characteristics of any plant system. As a result, no new failure
modes are being introduced. Therefore, the change does not introduce
a new or different kind of accident from those previously evaluated.
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 amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change to TS 4.3.1.2 ensures that the criticality
margins of safety for fuel storage are maintained, by excluding the
new fuel storage vault as an approved fuel storage location. The
change restricts the storage of new fuel to the spent fuel pool
racks, which are fully analyzed from a criticality standpoint. The
change does not physically alter the fuel storage systems, or modify
fuel storage requirements in such a way as to degrade the margins of
criticality safety.
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: Sherry A. Quirk, General Counsel, Tennessee
Valley Authority, 400 West Summit Hill Dr., WT 6A, Knoxville, TN 37902.
NRC Acting Branch Chief: Jeanne A. Dion.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: May 10, 2016. A publicly-available
version is in ADAMS under Accession No. ML16134A069.
Description of amendment request: The amendments would extend the
Surry Power Station, Unit Nos. 1 and 2, Technical Specification 3.2,
``Chemical and Volume Control System,'' paragraph E requirements for
primary grade water
[[Page 70188]]
(PG) lockout from being applicable in Refueling Shutdown and Cold
Shutdown to being applicable in Refueling Shutdown, Cold Shutdown,
Intermediate Shutdown, and Hot Shutdown (except during the approach to
critical and within 1 hour following reactor shutdown from reactor
critical or power 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 change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change conservatively imposes additional
operational controls on the highest capacity flow path of PG to the
Reactor Coolant System (RCS). These controls are currently credited
in the boron dilution analysis in Refueling Shutdown and Cold
Shutdown modes. The proposed change extends these controls into
Intermediate and Hot Shutdown modes. As such, the change will
provide defense against rapid reactivity insertions due to boron
dilution events and reduce the probability of boron dilution events.
The proposed change will have no impact on normal operating plant
releases and will not increase the predicted radiological
consequences of accidents postulated in the UFSAR [Updated Final
Safety Analysis Report]. The proposed change makes no physical
modifications and does not change plant design.
Therefore, neither the probability of occurrence nor the
consequences of any accident previously evaluated is significantly
increased.
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 is an extension of existing operational
controls on PG flow to the RCS to include additional operating
modes. The change precludes high flow rate boron dilutions in
Intermediate and Hot Shutdown modes similar to the current TS
requirement in Refueling and Cold Shutdown modes. It does not affect
the operation of the emergency boration function of the Chemical and
Volume Control System (CVCS).
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously analyzed.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change provides defense against rapid reactivity
insertions to potential boron dilution events in shutdown operating
modes and reduces the probability of boron dilution events. As such,
it increases the margin of safety for the boron dilution event.
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: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219.
NRC Branch Chief: Michael T. Markley.
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.
Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: January 18, 2016, as supplemented by
letter dated June 20, 2016.
Brief description of amendments: The amendments revised Technical
Specification (TS) 5.5.2, ``Containment Leakage Rate Testing Program,''
to allow (1) an increase in the existing Type A Integrated Leakage Rate
Testing Program test interval from 10 years to 15 years, in accordance
with Nuclear Energy Institute (NEI) Topical Report NEI 94-01, Revision
3-A, ``Industry Guideline for Implementing Performance-Based Option of
10 CFR part 50, appendix J,'' and the conditions and limitations
specified in NEI 94-01, Revision 2-A; (2) adoption of an extension of
the containment isolation valve leakage testing (Type C) frequency from
the 60 months currently permitted by 10 CFR part 50, appendix J, Option
B, to a 75-month frequency for Type C leakage rate testing of selected
components, in accordance with NEI 94-01, Revision 3-A; (3) adoption of
the use of American National Standards Institute/American Nuclear
Society (ANSI/ANS)-56.8-2002, ``Containment System Leakage Testing
Requirements''; and (4) adoption of a more conservative grace interval
of 9 months for Type A, Type B, and Type C leakage tests, in accordance
with NEI 94-01, Revision 3-A.
The amendments also made the following administrative changes: (1)
Deletion of the information regarding the performance of containment
visual inspections as required by Regulatory Position C.3, as the
containment inspections are addressed in TS Surveillance Requirement
3.6.1.1, and (2) deletion of the information regarding the performance
of the next Catawba Nuclear Station, Unit 1, Type A test no later than
November 13, 2015, and the next Catawba Nuclear Station, Unit 2, Type A
test no later than February 6, 2008, as both Type A tests have already
occurred.
Date of issuance: September 12, 2016.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment Nos.: 286 (Unit 1) and 282 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML16229A113; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments
[[Page 70189]]
revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: March 15, 2016 (81 FR
13839). The supplemental letter dated June 20, 2016, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 12, 2016.
No significant hazards consideration comments received: No.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: February 18, 2016, as supplemented by
letter dated June 30, 2016.
Brief description of amendment: The amendments modified Technical
Specification (TS) 5.5.2, ``Containment Leakage Rate Testing Program,''
for a one-time extension to the 10-year frequency of the integrated
leakage rate test (ILRT) or Type A test. This revision extends the
period from 10 years to 10.5 years between successive tests, changing
the performance of the next ILRT from fall 2017 to spring 2019 for Unit
1 and from spring 2017 to fall 2018 for Unit 2.
Date of issuance: September 26, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 290 (Unit 1) and 269 (Unit 2). A publicly available
version is in ADAMS under Accession No. ML16236A053; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: May 10, 2016 (81 FR
28894). The supplemental letter dated June 30, 2016, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 26, 2016.
No significant hazards consideration comments received: No.
Duke Energy Progress, Inc., Docket Nos. 50-325 and 50-324, Brunswick
Steam Electric Plant, Units 1 and 2 (BSEP), Brunswick County, North
Carolina
Duke Energy Progress, Inc., Docket No. 50-261; H. B. Robinson Steam
Electric Plant Unit No. 2 (RNP), Darlington County, South Carolina
Duke Energy Progress, Inc., Docket No. 50-400, Shearon Harris Nuclear
Power Plant, Unit 1, (HNP), Wake and Chatham Counties, North Carolina
Date of amendment request: February 1, 2016.
Description of amendment request: The amendments revised the
licensee's name from Duke Energy Progress, Inc. to Duke Energy
Progress, LLC.
Date of issuance: September 13, 2016.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: 271 and 299 (BSEP); 152 (HNP); 246 (RNP). A
publicly-available version is in ADAMS under Accession No. ML16217A118;
documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-71, DPR-62 (BSEP), NPF-
63 (HNP), and NFP-23 (RNP): Amendments revised the Renewed Facility
Operating Licenses.
Date of initial notice in Federal Register: April 12, 2016 (81 FR
21596).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 13, 2016.
No significant hazards consideration comments received: No.
Duke Energy Progress, Inc., Docket No. 50-400, Shearon Harris Nuclear
Power Plant (HNP), Unit 1, Wake and Chatham Counties, North Carolina
Date of amendment request: October 29, 2015, as supplemented by
letters dated, February 16, 2016, August 8 and 26, 2016, and September
8 and 16, 2016.
Brief description of amendment: The amendment revised Technical
Specifications to allow the `A' Emergency Service Water (ESW) pump to
be inoperable for 14 days to allow for the replacement of the `A' Train
ESW pump. The amendment is applicable on a one-time basis.
Date of issuance: September 16, 2016.
Effective date: As of the date of issuance and shall be implemented
by October 29, 2016.
Amendment No.: 153. A publicly-available version is in ADAMS under
Accession No. ML16253A059; documents related to this amendment are
listed in the Safety Evaluation (SE) enclosed with the amendment.
Renewed Facility Operating License No. NPF-63: Amendment revised
the Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: January 5, 2016 (81 FR
260). The supplemental letters dated February 16, 2016, August 8 and
26, 2016, and September 8 and 16, 2016, provided additional information
that clarified the application, did not expand the scope of the
application as originally noticed, and did not change the 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 an SE dated September 16, 2016.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-220 and 50-410, Nine
Mile Point Nuclear Station, Units 1 and 2, Oswego County, New York
Date of application for amendment: March 18, 2016.
Brief description of amendment: The amendments revised the
technical specifications (TSs) on a change to the method of calculating
core reactivity for the purpose of performing the Reactivity Anomalies
surveillance.
Date of issuance: September 15, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: Unit 1-224 and Unit 2-158. A publicly-available
version is in ADAMS under Accession No. ML16188A029; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License No. DPR-63 and NPF-69: The
amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: May 10, 2016 (81 FR
28897).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 15, 2016.
No significant hazards consideration comments received: No.
[[Page 70190]]
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant (PNPP), Unit No. 1, Lake County, Ohio
Date of amendment request: October 29, 2015, as supplemented by
letter dated April 22, 2016.
Brief description of amendment: The amendment revised the PNPP
emergency action level (EAL) scheme to one based on the Nuclear Energy
Institute (NEI) guidance in NEI 99-01, Revision 6, ``Development of
Emergency Action Levels for Non-Passive Reactors.''
Date of issuance: September 14, 2016.
Effective date: As of the date of issuance and shall be implemented
within 180 days of issuance.
Amendment No.: 173. A publicly-available version is in ADAMS under
Accession No. ML16158A331; 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 to authorize revision to the PNPP emergency
plan.
Date of initial notice in Federal Register: December 22, 2015 (80
FR 79620). The supplemental letter dated April 22, 2016, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the 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 September 14, 2016.
No significant hazards consideration comments received: No.
Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: January 19, 2016, as supplemented by a
letter dated May 6, 2016.
Brief description of amendments: The amendments revised the
Operating Licenses' licensing basis to allow elimination of the end-of-
cycle moderator temperature coefficient (MTC) surveillance test as
supported by NRC-Approved Topical Report CE NPSD-91 1-A and Amendment
1-A, ``Analysis of Moderator Temperature Coefficients in Support of a
Change in the Technical Specification End of Cycle Negative MTC
Limit,'' and St. Lucie specific supporting information. The amendments
also add NRC-approved Westinghouse PARAGON Topical Report WCAP-16045-P-
A, Revision 0, ``Qualification of the Two-Dimensional Transport Code
PARAGON,'' to the Technical Specification list of Core Operating Limits
Report methodologies.
Date of issuance: September 19, 2016.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: 235 and 185. A publicly-available version is in
ADAMS under Accession No. ML16183A138; documents related to these
amendments are listed in the Safety Evaluation (SE) enclosed with the
amendments.
Renewed Facility Operating License Nos. DPR-67 and NPF-16:
Amendments revised the Renewed Facility Operating Licenses and
Technical Specifications.
Date of initial notice in Federal Register: March 29, 2016 (81 FR
17506). The supplemental letter dated May 6, 2016, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the 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 an SE dated September 19, 2016.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 28th day of September 2016.
For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2016-24321 Filed 10-7-16; 8:45 am]
BILLING CODE 7590-01-P