[Federal Register Volume 83, Number 118 (Tuesday, June 19, 2018)]
[Notices]
[Pages 28456-28467]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2018-12506]
[[Page 28456]]
=======================================================================
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
[NRC-2018-0114]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
AGENCY: Nuclear Regulatory Commission.
ACTION: Biweekly notice.
-----------------------------------------------------------------------
SUMMARY: Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954,
as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is
publishing this regular biweekly notice. The Act requires the
Commission to publish notice of any amendments issued, or proposed to
be issued, and grants the Commission the authority to issue and make
immediately effective any amendment to an operating license or combined
license, as applicable, upon a determination by the Commission that
such amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued, from May 22, 2018, to June 4, 2018. The last
biweekly notice was published on June 5, 2018.
DATES: Comments must be filed by July 19, 2018. A request for a hearing
must be filed by August 20, 2018.
ADDRESSES: You may submit comments by any of the following methods
(unless this document describes a different method for submitting
comments on a specific subject):
Federal Rulemaking website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0114. Address
questions about NRC dockets to Jennifer Borges; telephone: 301-287-
9127; email: [email protected]. For technical questions, contact
the individual listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
Mail comments to: May Ma, Office of Administration, Mail
Stop: TWFN-7-A60M, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001.
For additional direction on obtaining information and submitting
comments, see ``Obtaining Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Janet Burkhardt, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; telephone: 301-415-1384, email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2018-0114, facility name, unit
number(s), plant docket number, application date, and subject when
contacting the NRC about the availability of information for this
action. You may obtain publicly-available information related to this
action by any of the following methods:
Federal Rulemaking website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0114.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, 301-415-4737, or by email to [email protected]. The
ADAMS accession number for each document referenced (if it is available
in ADAMS) is provided the first time that it is mentioned in this
document
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2018-0114, facility name, unit
number(s), plant docket number, application date, and subject in your
comment submission.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment into ADAMS.
II. Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses and Combined Licenses and Proposed No Significant
Hazards Consideration Determination
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in section 50.92 of title 10 of the Code
of Federal Regulations (10 CFR), this means that operation of the
facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated, or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment 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.
[[Page 28457]]
A. Opportunity To Request a Hearing and Petition for Leave To Intervene
Within 60 days after the date of publication of this notice, any
persons (petitioner) whose interest may be affected by this action may
file a request for a hearing and petition for leave to intervene
(petition) with respect to the action. Petitions shall be filed in
accordance with the Commission's ``Agency Rules of Practice and
Procedure'' in 10 CFR part 2. Interested persons should consult a
current copy of 10 CFR 2.309. The NRC's regulations are accessible
electronically from the NRC Library on the NRC's website at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of
the regulations is available at the NRC's Public Document Room, located
at One White Flint North, Room O1-F21, 11555 Rockville Pike (First
Floor), Rockville, Maryland 20852. If a petition is filed, the
Commission or a presiding officer will rule on the petition and, if
appropriate, a notice of a hearing will be issued.
As required by 10 CFR 2.309(d), the petition should specifically
explain the reasons why intervention should be permitted with
particular reference to the following general requirements for
standing: (1) The name, address, and telephone number of the
petitioner; (2) the nature of the petitioner's right under the Act to
be made a party to the proceeding; (3) the nature and extent of the
petitioner's property, financial, or other interest in the proceeding;
and (4) the possible effect of any decision or order which may be
entered in the proceeding on the petitioner's interest.
In accordance with 10 CFR 2.309(f), the petition must also set
forth the specific contentions which the petitioner seeks to have
litigated in the proceeding. Each contention must consist of a specific
statement of the issue of law or fact to be raised or controverted. In
addition, the petitioner must provide a brief explanation of the bases
for the contention and a concise statement of the alleged facts or
expert opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to the specific sources and
documents on which the petitioner intends to rely to support its
position on the issue. The petition must include sufficient information
to show that a genuine dispute exists with the applicant or licensee on
a material issue of law or fact. Contentions must be limited to matters
within the scope of the proceeding. The contention must be one which,
if proven, would entitle the petitioner to relief. A petitioner who
fails to satisfy the requirements at 10 CFR 2.309(f) with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene.
Parties have the opportunity to participate fully in the conduct of the
hearing with respect to resolution of that party's admitted
contentions, including the opportunity to present evidence, consistent
with the NRC's regulations, policies, and procedures.
Petitions must be filed no later than 60 days from the date of
publication of this notice. Petitions and motions for leave to file new
or amended contentions that are filed after the deadline will not be
entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the three factors in 10
CFR 2.309(c)(1)(i) through (iii). The petition must be filed in
accordance with the filing instructions in the ``Electronic Submissions
(E-Filing)'' section of this document.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to establish when the hearing is held. If the final determination is
that the amendment request involves no significant hazards
consideration, the Commission may issue the amendment and make it
immediately effective, notwithstanding the request for a hearing. Any
hearing would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, then any hearing held would take place before
the issuance of the amendment unless the Commission finds an imminent
danger to the health or safety of the public, in which case it will
issue an appropriate order or rule under 10 CFR part 2.
A State, local governmental body, Federally-recognized Indian
Tribe, or agency thereof, may submit a petition to the Commission to
participate as a party under 10 CFR 2.309(h)(1). The petition should
state the nature and extent of the petitioner's interest in the
proceeding. The petition should be submitted to the Commission no later
than 60 days from the date of publication of this notice. The petition
must be filed in accordance with the filing instructions in the
``Electronic Submissions (E-Filing)'' section of this document, and
should meet the requirements for petitions set forth in this section,
except that under 10 CFR 2.309(h)(2) a State, local governmental body,
or Federally-recognized Indian Tribe, or agency thereof does not need
to address the standing requirements in 10 CFR 2.309(d) if the facility
is located within its boundaries. Alternatively, a State, local
governmental body, Federally-recognized Indian Tribe, or agency thereof
may participate as a non-party under 10 CFR 2.315(c).
If a hearing is granted, any person who is not a party to the
proceeding and is not affiliated with or represented by a party may, at
the discretion of the presiding officer, be permitted to make a limited
appearance pursuant to the provisions of 10 CFR 2.315(a). A person
making a limited appearance may make an oral or written statement of
his or her position on the issues but may not otherwise participate in
the proceeding. A limited appearance may be made at any session of the
hearing or at any prehearing conference, subject to the limits and
conditions as may be imposed by the presiding officer. Details
regarding the opportunity to make a limited appearance will be provided
by the presiding officer if such sessions are scheduled.
B. Electronic Submissions (E-Filing)
All documents filed in NRC adjudicatory proceedings, including a
request for hearing and petition for leave to intervene (petition), any
motion or other document filed in the proceeding prior to the
submission of a request for hearing or petition to intervene, and
documents filed by interested governmental entities that request to
participate under 10 CFR 2.315(c), must be filed in accordance with the
NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR
46562; August 3, 2012). The E-Filing process requires participants to
submit and serve all adjudicatory documents over the internet, or in
some cases to mail copies on electronic storage media. Detailed
guidance on making electronic submissions may be found in the Guidance
for Electronic Submissions to the NRC and on the NRC website at http://www.nrc.gov/site-help/e-submittals.html. Participants may not submit
paper copies of their filings unless they seek an exemption in
accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least 10
days prior to the filing deadline, the participant should contact the
Office of the Secretary by email at
[[Page 28458]]
[email protected], or by telephone at 301-415-1677, to (1) request
a digital identification (ID) certificate, which allows the participant
(or its counsel or representative) to digitally sign submissions and
access the E-Filing system for any proceeding in which it is
participating; and (2) advise the Secretary that the participant will
be submitting a petition or other adjudicatory document (even in
instances in which the participant, or its counsel or representative,
already holds an NRC-issued digital ID certificate). Based upon this
information, the Secretary will establish an electronic docket for the
hearing in this proceeding if the Secretary has not already established
an electronic docket.
Information about applying for a digital ID certificate is
available on the NRC's public website at http://www.nrc.gov/site-help/e-submittals/getting-started.html. Once a participant has obtained a
digital ID certificate and a docket has been created, the participant
can then submit adjudicatory documents. Submissions must be in Portable
Document Format (PDF). Additional guidance on PDF submissions is
available on the NRC's public website at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the
time the document is submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an email notice confirming receipt of the document. The
E-Filing system also distributes an email notice that provides access
to the document to the NRC's Office of the General Counsel and any
others who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
document on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before adjudicatory documents are
filed so that they can obtain access to the documents via the E-Filing
system.
A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC's Electronic
Filing Help Desk through the ``Contact Us'' link located on the NRC's
public website at http://www.nrc.gov/site-help/e-submittals.html, by
email to [email protected], or by a toll-free call at 1-866-672-
7640. The NRC Electronic Filing Help Desk is available between 9 a.m.
and 6 p.m., Eastern Time, Monday through Friday, excluding government
holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
stating why there is good cause for not filing electronically and
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, 11555
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and
Adjudications Staff. Participants filing adjudicatory documents in this
manner are responsible for serving the document on all other
participants. Filing is considered complete by first-class mail as of
the time of deposit in the mail, or by courier, express mail, or
expedited delivery service upon depositing the document with the
provider of the service. A presiding officer, having granted an
exemption request from using E-Filing, may require a participant or
party to use E-Filing if the presiding officer subsequently determines
that the reason for granting the exemption from use of E-Filing no
longer exists.
Documents submitted in adjudicatory proceedings will appear in the
NRC's electronic hearing docket which is available to the public at
https://adams.nrc.gov/ehd, unless excluded pursuant to an order of the
Commission or the presiding officer. If you do not have an NRC-issued
digital ID certificate as described above, click cancel when the link
requests certificates and you will be automatically directed to the
NRC's electronic hearing dockets where you will be able to access any
publicly available documents in a particular hearing docket.
Participants are requested not to include personal privacy information,
such as social security numbers, home addresses, or personal phone
numbers in their filings, unless an NRC regulation or other law
requires submission of such information. For example, in some
instances, individuals provide home addresses in order to demonstrate
proximity to a facility or site. With respect to copyrighted works,
except for limited excerpts that serve the purpose of the adjudicatory
filings and would constitute a Fair Use application, participants are
requested not to include copyrighted materials in their submission.
For further details with respect to these license amendment
applications, see the application for amendment which is available for
public inspection in ADAMS and at the NRC's PDR. For additional
direction on accessing information related to this document, see the
``Obtaining Information and Submitting Comments'' section of this
document.
Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick
Steam Electric Plant, Unit Nos. 1 and 2, Brunswick County, North
Carolina
Date of amendment request: January 23, 2018. A publicly-available
version is in ADAMS under Accession No. ML18023A896.
Description of amendment request: The proposed amendments would
revise Technical Specification 3.6.4.1, ``Secondary Containment,''
Surveillance Requirement (SR) 3.6.4.1.2, for Brunswick Steam Electric
Plant, Units 1 and 2. The proposed changes are based on Technical
Specifications Task Force (TSTF) Traveler TSTF-551, Revision 3,
``Revise Secondary Containment Surveillance Requirements'' (ADAMS
Accession No. ML16277A226).
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 addresses conditions during which Secondary
Containment SR 3.6.4.1.2 is not met. The Secondary Containment is
not an initiator of any accident previously evaluated. As a result,
the probability of any accident previously evaluated is not
increased. The consequences of an accident previously evaluated
while utilizing the proposed change is no different than the
consequences of an accident while utilizing the existing eight hour
Completion Time for an inoperable Secondary Containment. As a
result, the consequences of an accident previously evaluated are not
significantly increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment 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 protection system design,
create new failure
[[Page 28459]]
modes, or change any modes of operation. The proposed change does
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 change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change addresses conditions during which Secondary
Containment SR 3.6.4.1.2 is not met. The allowance for both an inner
and outer Secondary Containment door to be open simultaneously for
entry and exit does not affect the safety function of the Secondary
Containment as the doors are promptly closed after entry or exit,
thereby restoring the Secondary Containment boundary.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel,
550 South Tryon Street, Mail Code DEC45A, Charlotte, NC 28202.
NRC Acting Branch Chief: Brian W. Tindell.
Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick
Steam Electric Plant, Unit Nos. 1 and 2, Brunswick County, North
Carolina
Date of amendment request: January 23, 2018. A publicly-available
version is in ADAMS under Accession No. ML18023A899.
Description of amendment request: The amendments would revise the
Technical Specifications to adopt Technical Specifications Task Force
(TSTF) Traveler TSTF-208, Revision 0, ``Extension of Time to Reach Mode
2 in LCO [Limiting Condition for Operation] 3.0.3.''
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 time frame to take response action in accordance with LCO
3.0.3 is not an initiating condition for any accident previously
evaluated. The proposed change does not authorize the addition of
any new plant equipment or systems, nor does it alter the
assumptions of any accident analyses. The small increase in the time
allowed to reach Mode 2 would not place the plant in any
significantly increased probability of an accident occurring. The
unit would already be preparing for a plant shutdown condition
because of the 1 hour requirement to initiate shutdown actions.
There is no change in the time period to reach Mode 3. The Mode 3
Condition is the point at which the plant reactor core is no longer
critical (i.e., Hot Shutdown).
Therefore, since there is no change to the time period to reach
the Hot Shutdown condition, the small change in the time to reach
Mode 2 status 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 to the allowed time to reach Mode 2 in LCO
3.0.3 does not require any modification to the plant or change
equipment operation. The proposed change will not introduce failure
modes that could result in a new accident, and the change does not
alter assumptions made in the safety analysis. The proposed change
will not alter the design configuration, or method of operation of
plant equipment beyond its normal functional capabilities. The
proposed change does not create any new credible failure mechanisms,
malfunctions, or accident initiators.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from those that have been
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change to the allowed time to reach Mode 2 in LCO
3.0.3 does not alter or exceed a design basis or safety limit. There
is no change being made to safety analysis assumptions or the safety
limits that would adversely affect plant safety as a result of the
proposed change. Margins of safety are unaffected by the proposed
change and the applicable requirements of 10 CFR 50.36(c)(2)(ii) and
10 CFR 50, Appendix A will continue to be met.
Therefore, the proposed change does not involve any reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel,
550 South Tryon Street, M/C DEC45A, Charlotte, NC 28202.
NRC Acting Branch Chief: Brian W. Tindell.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit No. 3 (Waterford 3), St. Charles Parish, Louisiana
Date of amendment request: March 8, 2018. A publicly-available
version is in ADAMS under Accession No. ML18068A705.
Description of amendment request: The amendment would update
Section 15.4.3.1 of the Updated Final Safety Analysis Report for
Waterford 3, which describes the dose consequence of the worst
undetectable single fuel assembly misload. The updated analysis would
reflect the use of Next Generation Fuel and integrated fuel burnable
absorbers.
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 the fuel assembly misload event
analysis. The analysis of the fuel assembly misload event showed
that the total number of failed fuel rods is less than other
Waterford 3 Condition 3 events that have already been demonstrated
to meet the 10 CFR 50.67 acceptance criteria. For Waterford 3, the
Excess Load with Loss of Alternating Current (LOAC) has this same
release and fuel failure that has been shown to meet the offsite
dose requirements. Since the worst undetectable misload has a fuel
failure less than the excess load with LOAC event, the fuel assembly
misload event is consistent with the Standard Review Plan 15.4.7 and
meets the 10 CFR 50.67 requirements.
This change is only analyzing the consequences of the fuel
assembly misload event and no changes are being made that would
impact the probability of the event occurring.
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 revises the fuel assembly misload event
analysis. The proposed change does not involve a physical alteration
of the plant (no new or different type of equipment will be
installed) or a change in the methods governing plant operations.
The proposed change will not introduce new failure modes or effects
and will not, in the absence of other unrelated failures, lead to an
accident whose
[[Page 28460]]
consequences exceed the consequences of accidents previously
analyzed.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change revises the fuel assembly misload event
analysis. The worst undetectable misloads have fuel failure less
than the excess load with the Excess Load with Loss of Alternating
Current (LOAC) event; the fuel assembly misload event meets the 10
CFR 50.67 criteria and is consistent with the Standard Review Plan
Section 15.4.7 guidance. The new analysis shows more adverse
consequences than were shown in previous fuel assembly misload event
analyses, but remains within the regulatory acceptance limits. Since
the event remains within the 10 CFR 50.67 requirements and is
bounded by the excess load with LOAC event, this is not a
significant reduction in margin.
Therefore, this change does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Anna Vinson Jones, Senior Counsel, Entergy
Services, Inc., 101 Constitution Avenue NW, Suite 200 East, Washington,
DC 20001.
NRC Branch Chief: Robert J. Pascarelli.
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station, Unit No. 1, DeWitt County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Unit Nos. 2 and 3, Grundy County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Unit Nos. 1 and 2, LaSalle County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Unit Nos. 1 and 2, Rock Island County,
Illinois
Date of amendment request: April 25, 2018. A publicly-available
version is in ADAMS under Accession No. ML18116A133.
Description of amendment request: The amendments would revise the
technical specification (TS) requirements for inoperable snubbers for
each facility. The amendments would also make other administrative
changes to the TSs.
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 for each site, 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 allows a delay time before declaring
supported Technical Specification (TS) systems inoperable when the
associated snubber(s) cannot perform its required safety function.
Entrance into Actions or delaying entrance into Actions is not an
initiator of any accident previously evaluated. Consequently, the
probability of an accident previously evaluated is not significantly
increased. The consequences of an accident while relying on the
delay time allowed before declaring a TS supported system inoperable
and taking its Conditions and Required Actions are no different than
the consequences of an accident under the same plant conditions
while relying on the existing TS supported system Conditions and
Required Actions. Therefore, the consequences of an accident
previously evaluated are not significantly increased by this change.
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 allows a delay time before declaring
supported TS systems inoperable when the associated snubber(s)
cannot perform its required safety function. The proposed change
does not involve a physical alteration of the plant (no new or
different type of equipment will be installed) or a change in the
methods governing normal plant operation. Therefore, the proposed
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change allows a delay time before declaring
supported TS systems inoperable when the associated snubber(s)
cannot perform its required safety function. The proposed change
restores an allowance in the pre-Improved Standard Technical
Specifications (ISTS) conversion TS that was unintentionally
eliminated by the conversion. The pre-ISTS TS were considered to
provide an adequate margin of safety for plant operation, as does
the post-ISTS conversion TS. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis for each site
and, based on this review, it appears that the three standards of 10
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to
determine that the requested amendments involve no significant hazards
consideration.
Attorney for licensee: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Branch Chief: David J. Wrona.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit No. 1, Lake County, Ohio
Date of amendment request: March 7, 2018. A publicly-available
version is in ADAMS under Accession No. ML18066A648.
Description of amendment request: The proposed amendment would
revise Technical Specification 5.5.12, ``Primary Containment Leakage
Rate Testing Program,'' to follow guidance developed by the Nuclear
Energy Institute (NEI) in topical report NEI 94-01, ``Industry
Guideline for Implementing Performance-Based Option of 10 CFR part 50,
Appendix J,'' Revision 3-A, dated July 2012, with the conditions and
limitations specified in NEI 94-01, Revision 2-A, dated October 2008.
The proposed license amendment would also revise Technical
Specification 5.5.12 by deleting two of the four listed exceptions to
program guidelines.
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 test interval extensions do not involve either a
physical change to the plant or a change in the way 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. Therefore, the proposed extensions do not involve a
significant increase in the probability of an accident previously
evaluated.
The effect resulting from changing the Type A test frequency to
1 per 15 years, measured as an increase to the total integrated
plant risk for those accident sequences influenced by Type A
testing, is
[[Page 28461]]
0.0318 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 less than or equal to
1.0 person-rem per year or less than or equal to 1 percent of the
total population dose, whichever is less restrictive for the risk
impact assessment of the extended integrated leak rate test
intervals. The results of the risk assessment calculation for the
Type A test extension meet these criteria. The risk impact for the
integrated leak rate test extension when compared to other severe
accident risks is negligible.
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 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 [American Society for Mechanical Engineers Boiler
and Pressure Vessel Code (ASME Code)], Section XI, and Technical
Specification 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
test interval extensions do not significantly increase the
consequences of an accident previously evaluated.
The proposed amendment also deletes two previously granted
exceptions to Primary Containment Leakage Rate Testing Program
guidelines. The exception regarding the performance of a Type A test
no later than a specified date would be deleted as this Type A test
has already been performed. Additionally, the exception to use the
corrections to NEI 94-01, Revision 0, would be deleted as those
corrections would no longer be in use. These changes to the
exceptions in Technical Specification 5.5.12 are administrative in
nature and do not affect the probability or consequences of an
accident previously evaluated.
Therefore, the proposed changes do 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.
Containment Type A and Type C testing requirements periodically
demonstrate the integrity of the containment and exist to ensure the
plant's ability to mitigate the consequences of an accident. These
tests do not involve any accident precursors or initiators.
The proposed change does not involve a physical modification to
the plant (that is, no new or different type of equipment will be
installed) nor does it alter the design, configuration, or change
the manner in which the plant is operated or controlled beyond the
standard functional capabilities of the equipment.
The proposed amendment also deletes two previously granted
exceptions. The exception regarding the performance of a Type A test
no later than a specified date would be deleted as this Type A test
has already been performed. Additionally, the exception to use the
corrections to NEI 94-01, Revision 0, would be deleted as those
corrections would no longer be in use. These changes to the
exceptions in Technical Specification 5.5.12 are administrative in
nature and do not create the possibility of a new or different kind
of accident from any previously evaluated.
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 license amendment does not alter the way safety
limits, limiting safety system set points, or limiting conditions
for operation are determined. The specific requirements and
conditions of the Technical Specification Primary Containment
Leakage 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 Technical Specifications 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 amendment, since they are not affected
by implementation of a performance-based containment testing
program. This ensures that the margin of safety in the plant safety
analysis is maintained.
The proposed amendment also deletes two previously granted
exceptions. The exception regarding the performance of a Type A test
no later than a specified date would be deleted as this Type A test
has already been performed. Additionally, the exception to use the
corrections to NEI 94-01, Revision 0, would be deleted as those
corrections would no longer be in use. These changes to the
exceptions in Technical Specification 5.5.12 are administrative in
nature and do not involve a significant reduction in a margin of
safety.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: David J. Wrona.
NextEra Energy, Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant (PBNP), Unit Nos. 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of amendment request: March 30, 2018. A publicly-available
version is in ADAMS under Accession No. ML18092A239.
Description of amendment request: The amendments would revise
Technical Specification (TS) 5.5.15, ``Containment Leakage Rate Testing
Program,'' to require a program 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.'' This proposed change will allow extension of the Type A
test interval up to one test in 15 years and extension of the Type C
test interval up to 75 months, based on acceptable performance history
as defined in NEI 94-01, Revision 3-A.
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 consequences of an accident previously evaluated?
Response: No.
The proposed amendment adopts the NRC-accepted guidelines of NEI
94-01, Revision 3-A, ``Industry Guideline for Implementing
Performance-Based Option of 10 CFR part 50, Appendix J,'' for
development of the PBNP performance-based containment testing
program. NEI 94-01 allows, based on risk and performance, an
extension of Type A and Type C containment leak test intervals.
Implementation of these guidelines continues to provide adequate
assurance that during design basis accidents, the primary
containment and its components will limit leakage rates to less than
the values assumed in the plant safety analyses.
The findings of the PBNP risk assessment confirm the general
findings of previous studies that the risk impact with extending the
containment leak rate is small. Per the guidance provided in
Regulatory Guide 1.174, an extension of the leak test interval in
accordance with NEI 94-01, Revision 3-A results in an estimated
change within, the very small change region.
Since the change is implementing a performance-based containment
testing program, the proposed amendment 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 requirement for containment
leakage rate acceptance will not be changed by this amendment.
[[Page 28462]]
Therefore, the containment will continue to perform its design
function as a barrier to fission product releases.
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 to implement a performance-based containment
testing program, associated with integrated leakage rate test
frequency, does not change the design or operation of structures,
systems, or components of the plant.
The proposed change would continue to ensure containment
integrity and would ensure operation within the bounds of existing
accident analyses. There are no accident initiators created or
affected by this change. Therefore, the proposed change will not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Margin of safety is related to confidence in the ability of the
fission product barriers (fuel cladding, reactor coolant system, and
primary containment) to perform their design functions during and
following postulated accidents. The proposed change to implement a
performance-based containment testing program, associated with
integrated leakage rate test and local leak rate testing frequency,
does not affect plant operations, design functions, or any analysis
that verifies the capability of a structure, system, or component of
the plant to perform a design function. In addition, this change
does not affect safety limits, limiting safety system setpoints, or
limiting conditions for operation.
The specific requirements and conditions of the TS Containment
Leakage 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. This
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
implementation of a performance-based containment testing program.
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: William Blair, Managing Attorney--Nuclear,
Florida Power & Light Company, P.O. Box 14000, 700 Universe Boulevard,
Juno Beach, FL 33408-0420.
NRC Branch Chief: David J. Wrona.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station
(HCGS), Salem County, New Jersey
Date of amendment request: April 13, 2018. A publicly-available
version is in ADAMS under Accession No. ML18103A218.
Description of amendment request: The amendment would revise
Technical Specification (TS) 3.8.3.1, ``Distribution--Operating,'' to
increase the alternating current (AC) inverters allowed outage time
(AOT) from 24 hours to 7 days. The proposed change is based on
application of the HCGS probabilistic risk assessment (PRA) in support
of a risk-informed extension, and on additional considerations and
compensatory actions.
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 TS amendment does not affect the design of the AC
inverters, the operational characteristics or function of the
inverters, the interfaces between the inverters and other plant
systems, or the reliability of the inverters. An inoperable AC
inverter is not considered an initiator of an analyzed event. In
addition, TS Actions and the associated Allowed Outage Times are not
initiators of previously evaluated accidents. Extending the Allowed
Outage Time for an inoperable AC inverter would not have a
significant impact on the frequency of occurrence of an accident
previously evaluated. The proposed amendment will not result in
modifications to plant activities associated with inverter
maintenance, but rather, provides operational flexibility by
allowing additional time to perform inverter troubleshooting,
corrective maintenance, and post-maintenance testing on-line.
The proposed extension of the Completion Time for an inoperable
AC inverter will not significantly affect the capability of the
inverters to perform their safety function, which is to ensure an
uninterruptible supply of 120-volt AC electrical power to the
associated power distribution subsystems. An evaluation, using PRA
methods, confirmed that the increase in plant risk associated with
implementation of the proposed Allowed Outage Time extension is
consistent with the NRC's Safety Goal Policy Statement, as further
described in RG [Regulatory Guide] 1.174 and RG 1.177. In addition,
a deterministic evaluation concluded that plant defense-in-depth
philosophy will be maintained with the proposed Allowed Outage Time
extension.
There will be no impact on the source term or pathways assumed
in accidents previously evaluated. No analysis assumptions will be
changed and there will be no adverse effects on onsite or offsite
doses as the result of an accident.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment does not involve physical alteration of
the HCGS. No new equipment is being introduced, and installed
equipment is not being operated in a new or different manner. There
is no change being made to the parameters with in which the HCGS is
operated. There are no setpoints at which protective or mitigating
actions are initiated that are affected by this proposed action. The
use of the alternate Class 1E power source for the AC distribution
panel is consistent with the HCGS plant design. The change does not
alter assumptions made in the safety analysis. This proposed action
will not alter the manner in which equipment operation is initiated,
nor will the functional demands on credited equipment be changed. No
alteration is proposed to the procedures that ensure the HCGS
remains with in analyzed limits, and no change is being made to
procedures relied upon to respond to an off-normal event. As such,
no new failure modes are being introduced.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Margin of safety is related to the confidence in the ability of
the fission product barriers to perform their design functions
during and following an accident. These barriers include the fuel
cladding, the reactor coolant system, and the containment system.
The proposed change, which would increase the AOT from 24 hours to 7
days for one inoperable inverter, does not exceed or alter a
setpoint, design basis or safety limit.
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.
[[Page 28463]]
Attorney for licensee: Jeffrie J. Keenan, PSEG Nuclear LLC--N21,
P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: James G. Danna.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant, Unit Nos. 3 and 4, Burke County,
Georgia
Date of amendment request: April 26, 2018. A publicly-available
version is in ADAMS under Accession No. ML18116A138.
Description of amendment request: The requested amendment proposes
changes to combined license (COL) Appendix C, with corresponding
changes to the associated plant-specific Tier 1 information, and
involves associated Tier 2 information in the Updated Final Safety
Analysis Report (UFSAR) (which includes the plant-specific Design
Control Document (DCD) Tier 2 information). Pursuant to the provisions
of 10 CFR 52.63(b)(1), also requested is an exemption from elements of
the design as certified in the 10 CFR part 52, appendix D, design
certification rule for the plant-specific DCD departures.
The requested amendment proposes changes to COL Appendix C (and
plant-specific Tier 1) to reflect a new design of containment sump
level sensors that affects the acceptance criterion for the detected
containment sump level change test and the associated minimum
detectable unidentified leakage rate in plant-specific DCD Tier 2
information.
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 is to the containment sump water level
instrumentation and its expected [reactor coolant system (RCS)]
leakage detection capability. The affected equipment is not safety-
related, but the containment sump water level sensors are
seismically qualified. The change in containment sump level
monitoring instruments has no adverse effect on the ability to
detect a 0.5 [gallons per minute (gpm)] leak in containment, and
therefore, has no adverse effect on design criteria for leak-before-
break. The change does not affect the operation of any systems or
equipment that initiate an analyzed accident or alter any
structures, systems, and components (SSC) accident initiator or
initiating sequence of events.
Because the containment sump water level monitoring channels are
still capable of detecting a 0.5 gpm leak in containment, the change
to the SSC has no effect on plant operations. 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 normal operation or postulated accident conditions.
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 does 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 change to the
containment sump water level instrumentation and its expected RCS
leakage detection capability has no adverse effect on the ability to
detect a 0.5 gpm leak in containment. The containment sump level
instrumentation functions are unchanged and leak-before-break design
criteria are not adversely affected.
Loss of coolant accidents for a spectrum of pipe sizes and
locations are already postulated in UFSAR Chapter 15, Section 15.6.
Breaks in the main steam lines inside containment are also analyzed
in UFSAR Chapter 15, Section 15.1. Unidentified leakage detection
and operator action in response to unidentified leakage are not
postulated for any of the design basis accident analyses described
in UFSAR Chapter 15.
Therefore, the requested 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 described change to the containment sump water level
instrumentation and its expected RCS leakage detection capability is
proposed to verify that the ability to detect a 0.5 gpm leak in
containment is maintained. The proposed change does not alter any
safety-related equipment, applicable design codes, code compliance,
design function, or safety analysis. By ensuring that the chosen
equipment can detect a 0.5 gpm leak in containment with the
described accuracy, guidance in Regulatory Guide 1.45, Revision 0,
as committed to in the UFSAR, and requirements in the Technical
Specifications are met which ensures that leak-before-break design
criteria are not adversely affected. 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.
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 hazard consideration
Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP,
1710 Sixth Avenue, North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer L. Dixon-Herrity.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant, Unit Nos. 3 and 4, Burke County,
Georgia
Date of amendment request: April 27, 2018. A publicly-available
version is in ADAMS under Accession No. ML18117A464.
Description of amendment request: The requested amendment proposes
to depart from Tier 2 information in the Updated Final Safety Analysis
Report (UFSAR) (which includes the plant-specific Design Control
Document Tier 2 information) and involves related changes to plant-
specific Tier 1 information, with corresponding changes to the
associated combined license (COL) Appendix C information. Specifically,
the amendment, if approved, would revise the Tier 2 information in the
UFSAR and related changes to Tier 1 and the associated COL Appendix C
to remove the fire protection system non-safety related containment
cable spray and install passive fire stops and radiant energy shields.
The changes to Tier 1 require an exemption, which is included in the
license amendment request.
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 do not affect the operation or reliability
of any system, structure or component (SSC) required to maintain a
normal power operating condition or to mitigate anticipated
transients without safety-related systems. Testing has demonstrated
that the passive fire stops prevent propagation of fires along the
length of cable trays and prevent the propagation of cable tray
fires to adjacent fire zones. The proposed changes do not affect the
operation of equipment whose failure could initiate an accident
previously analyzed. The existence or failure of passive fire stops
in fire zone 1100 AF 11300B does not affect normal equipment
operation.
[[Page 28464]]
The proposed changes do not adversely affect the reliability or
function of an SSC relied upon to mitigate an accident previously
analyzed. The existence or failure of passive fire stops in fire
zone 1100 AF 11300B will not adversely affect passive core cooling
system (PXS) performance during containment recirculation because
the passive fire stops are located outside of the zone of influence
(ZOI) of postulated high energy line breaks, and the passive fire
stops' material-of-construction complies with in-containment
refueling water storage tank (IRWST) and containment recirculation
screens design criteria for debris generation and transport.
The existing active open nozzle cable tray suppression system is
not fully automatic, is nonsafety-related, and is not credited in
the probabilistic risk assessment (PRA). Therefore, replacing the
active open nozzle cable tray suppression system with passive fire
stops does not have an impact on PRA calculations and results.
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 systems or
equipment that could 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 use of passive fire stops is
recognized by Regulatory Guide 1.189. The passive fire stops in
nonsafety-related open cable trays are more reliable than active
systems such as the current open nozzle cable tray suppression
system because they require no mechanical or human action to perform
their protective function. When protection is required, there is no
delay for operator or mechanical response. Testing has demonstrated
that the passive fire stops prevent propagation of fires along the
length of cable trays and prevent the propagation of cable tray
fires to adjacent fire zones.
The existence or failure of passive fire stops in fire zone 1100
AF 11300B will not adversely affect passive core cooling system
(PXS) performance during containment recirculation because the
passive fire stops are located outside of the zone of influence
(ZOI) of postulated high energy line breaks, and their material-of-
construction complies with in-containment refueling water storage
tank (IRWST) and containment recirculation screens design criteria
for debris generation and transport.
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 do not affect existing safety margins. The
current open nozzle cable tray suppression system is nonsafety-
related. The use of passive fire stops is recognized by Regulatory
Guide 1.189. The passive fire stops in nonsafety-related open cable
trays are more reliable than active systems such as the current open
nozzle cable tray suppression system because they require no
mechanical or human action to perform their protective function.
When protection is required, there is no delay for operator or
mechanical response. Testing has demonstrated that the passive fire
stops prevent propagation of fires along the length of cable trays
and prevent the propagation of cable tray fires to adjacent fire
zones.
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 hazard consideration.
Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP,
1710 Sixth Avenue, North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer L. Dixon-Herrity.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant, Unit Nos. 3 and 4, Burke County,
Georgia
Date of amendment request: April 27, 2018. A publicly-available
version is in ADAMS under Accession No. ML18117A464.
Description of amendment request: The requested amendment proposes
to depart from Tier 2 information in the Updated Final Safety Analysis
Report (UFSAR) (which includes the plant-specific Design Control
Document Tier 2 information) and involves related changes to plant-
specific Tier 1 information, with corresponding changes to the
associated combined license (COL) Appendix C information. Specifically,
the amendment, if approved, would revise the Tier 2 information in the
UFSAR and related changes to Tier 1 and the associated COL Appendix C
to remove the fire protection system non-safety related containment
cable spray and install passive fire stops and radiant energy shields.
The changes to Tier 1 require an exemption, which is included in the
license amendment request.
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 do not affect the operation of any systems
or equipment that initiate an analyzed accident or alter any
structures, systems, and components (SSC) accident initiator or
initiating sequence of events.
The proposed changes do not affect the physical design and
operation of the Passive Residual Heat Removal Heat Exchanger (PRHR
HX) or In-containment Refueling Water Storage Tank (IRWST) as
described in the Updated Final Safety Analysis Report (UFSAR). The
proposed changes do not affect the probability of inadvertent
operation or failure. Therefore, the probabilities of the accidents
previously evaluated in the UFSAR are not affected.
The proposed changes do not affect the ability of the PRHR HX
and IRWST to perform their design functions. The designs of the PRHR
HX and IRWST continue to meet the same regulatory acceptance
criteria, codes, and standards as required by the UFSAR. In
addition, the proposed changes maintain the capabilities of the PRHR
HX and IRWST to mitigate the consequences of an accident and to meet
the applicable regulatory acceptance criteria.
The proposed changes do not affect the prevention and mitigation
of other abnormal events (e.g. anticipated operational occurrences,
earthquakes, floods and turbine missiles), or their safety or design
analyses. Therefore, the consequences of the accidents evaluated in
the UFSAR 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 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 do not affect any other SSC design
functions or methods of operation in a manner that results in a new
failure mode, malfunction, or sequence of events that affect safety-
related or nonsafety related equipment. Therefore, this activity
does not allow for a new fission product release path, result in a
new fission product barrier failure mode, or create a new sequence
of events that result in significant fuel cladding failures.
Therefore, the requested amendment does not create the
possibility of a new or different type 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 existing safety margins. The
proposed changes verify and maintain the capabilities of the PRHR HX
and IRWST to perform their design functions. Therefore, the proposed
changes
[[Page 28465]]
satisfy the same design functions in accordance with the same codes
and standards as stated in the UFSAR. These changes do not affect
any design code, function, design analysis, safety analysis input or
result, or design/safety margin.
No safety analysis or design basis acceptance limit/criterion is
challenged or exceeded by the proposed changes, and no margin of
safety is reduced.
Therefore, the requested 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 L. Dixon-Herrity.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2 (Surry), Surry County, Virginia
Date of amendment request: March 2, 2018. A publicly-available
version is in ADAMS under Accession No. ML18075A021.
Description of amendment request: The amendments would revise the
Technical Specifications (TSs) consistent with Revision 0 to the
Technical Specifications Task Force (TSTF) Standard Technical
Specification Change Document TSTF-490, ``Deletion of E Bar Definition
and Revision to RCS Specific Activity Tech Spec.'' The proposed
amendments would adopt TSTF-490 and make the following associated
changes: (1) Adoption of a TS change to replace the current limits on
primary coolant gross specific activity with limits on primary coolant
noble gas activity, and (2) an update of the Alternative Source Term
(AST) analyses for Surry.
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:
Criterion 1. The Proposed Changes Do Not Involve a Significant Increase
in the Probability or Consequences of an Accident Previously Evaluated
Reactor coolant specific activity is not an initiator for any
accident previously evaluated, and the allowed time period when
primary coolant gross activity is not within limits is not an
initiator for any accident previously evaluated. In addition, the
current variable limit on primary coolant iodine concentration is
not an initiator to any accident previously evaluated. Updating the
Alternative Source Term analyses does not require any changes to any
plant structures, systems, or components (SSCs) and therefore does
not affect any accident initiators. As a result, the proposed
changes do not significantly increase the probability of an
accident. The proposed TS change will limit primary coolant noble
gases to concentrations consistent with the accident analyses, and
the proposed completion time when the limit may be exceeded has no
impact on the consequences of any design basis accident since the
consequences of an accident during this time period is the same as
the consequences of an accident during the existing time periods.
The revised assessments of the radiological consequences due to
design basis accidents listed in the Surry Updated Final Safety
Analysis Report, using the updated AST methodology and proposed
assumptions and inputs, conclude that the Exclusion Area Boundary
(EAB), Low Population Zone (LPZ), and Control Room doses are within
the limits of 10 CFR 50.67 and within the limits of Regulatory Guide
(RG) 1.183. As a result, the consequences of any accident previously
evaluated are not significantly increased.
Criterion 2. The Proposed Changes Do Not Create the Possibility of a
New or Different Kind of Accident From Any Accident Previously
Evaluated
The proposed TS change in specific activity limits and the
updated AST dose consequences analyses do not alter any physical
part of the plant, (i.e., no new or different type of equipment will
be installed,) nor do they affect any plant operating parameter or
create new accident precursors. Therefore, the proposed changes do
not create the potential for a new or different kind of accident
from any previously calculated.
Criterion 3. The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety.
The proposed TS change in specific activity limits is consistent
with the assumptions in -the safety analyses and will ensure the
monitored values protect the initial assumptions in the safety
analyses. The proposed changes for radiological events related to
the computer code used to calculate dose, revised X/Qs for control
room and offsite receptors (including the computer code and method
used to determine control room X/Qs for SG releases), the computer
code used to determine core inventory, the change in FHA [Fuel
Handling Accident] gap fraction methodology, and removing the LRA
[Locked Rotor Accident] from the radiological design basis have been
analyzed and result in acceptable consequences, meeting the criteria
as specified in 10 CFR 50.67 and RG 1.183. The proposed changes will
not result in plant operation in a configuration outside the
analyses or design basis and do not adversely affect systems that
are required to respond for safe shutdown of the plant and to
maintain the plant in a safe operating condition. Therefore, the
changes do not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: 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.
[[Page 28466]]
DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: August 14, 2017.
Brief description of amendment: The amendment modified Fermi 2
Technical Specification 5.5.7, ``Ventilation Filter Testing Program
(VFTP),'' by adopting the format and language of NUREG-1433, ``Standard
Technical Specifications for General Electric BWR/4 Plants,'' Revision
4.
Date of issuance: May 24, 2018.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No: 208. A publicly-available version is in ADAMS under
Accession No. ML18108A022; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-43: The amendment
revised the Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: September 26, 2017 (82
FR 44851).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 24, 2018.
No significant hazards consideration comments received: No.
Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick
Steam Electric Plant, Unit Nos. 1 and 2, Brunswick County, North
Carolina
Date of amendment request: June 29, 2017, as supplemented by
letters dated January 4, 2018, and January 23, 2018.
Brief description of amendments: The amendments adopted Technical
Specifications Task Force (TSTF) Traveler TSTF-542, Revision 2,
``Reactor Pressure Vessel Water Inventory Control,'' for Brunswick
Steam Electric Plant, Units 1 and 2. The amendments replaced existing
technical specification (TS) requirements associated with ``operations
with the potential for draining the reactor vessel,'' with revised TSs
providing alternative requirements for reactor pressure vessel water
inventory control. These alternative requirements protect Safety Limit
2.1.1.3, which states, ``Reactor vessel water level shall be greater
than the top of active irradiated fuel.''
Date of issuance: April 13, 2018.
Effective date: As of the date of issuance and shall be implemented
prior to the 2019 Unit 2 refueling outage. This Notice of Issuance
corrects the effective date of License Amendment No. 283, originally
noticed in the Federal Register on May 8, 2018 (83 FR 20865).
Amendment Nos.: 283 (Unit 1) and 311 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML18039A444; documents related
to this amendment are listed in the Safety Evaluation enclosed with the
amendments. Amendment Nos. 283 and 311 were corrected by letter dated
May 23, 2018 (ADAMS Accession No. ML18137A143).
Renewed Facility Operating License No. DPR-49: The amendments
revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: September 12, 2017 (82
FR 42846). The supplemental letters dated January 4, 2018, and January
23, 2018, provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the NRC staff's original proposed no
significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety evaluation dated April 13, 2018.
No significant hazards consideration comments received: No.
Duke Energy Progress, LLC, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: April 3, 2017, as supplemented by
letters dated April 3, 2017; May 2, 2017; September 28, 2017; and
January 8, 2018.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) to extend the required frequency of certain 18-
month Surveillance Requirements to 24 months to accommodate a 24-month
refueling cycle. In addition, the amendment revised certain programs in
TS Section 5.5, ``Programs and Manuals,'' to change 18-month
frequencies to 24 months.
Date of issuance: May 25, 2018.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the end of the next refueling outage.
Amendment No.: 258. A publicly-available version is in ADAMS under
Accession No. ML18115A150; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-23: The amendment
revised the Renewed Facility Operating License and TSs.
Date of initial notice in Federal Register: July 5, 2017 (82 FR
31092). The supplemental letters dated September 28, 2017, and January
8, 2018, provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the NRC staff's original proposed no
significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 25, 2018.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station, Unit No. 1 (Clinton), DeWitt County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station (LaSalle), Unit Nos. 1 and 2, LaSalle County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station (Limerick), Unit Nos. 1 and 2, Montgomery County,
Pennsylvania
Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit No. 2 (Nine Mile), Oswego County, New York
Date of amendment request: November 8, 2017.
Brief description of amendments: The amendments revised the
technical specification requirements for secondary containment.
Date of issuance: May 29, 2018.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: Clinton--218; LaSalle, Units 1 and 2--228 and 214;
Limerick, Units 1 and 2--229 and 192; and Nine Mile--169. A publicly-
available version is in ADAMS under Accession No. ML18113A045.
Documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Facility Operating License Nos. NPF-62, NPF-11, NPF-18, NPF-39,
NPF-85, and NPF-69: The amendments revised the Facility Operating
Licenses and Technical Specifications.
Date of initial notice in Federal Register: December 19, 2017 (82
FR 60227).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 29, 2018.
No significant hazards consideration comments received: No.
[[Page 28467]]
Southern Nuclear Operating Company, Inc.; Georgia Power Company;
Oglethorpe Power Corporation; Municipal Electric Authority of Georgia;
and City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I.
Hatch Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia
Date of amendment request: April 20, 2017, as supplemented by
letters dated September 14, 2017; February 19, 2018; and May 1, 2018.
Brief description of amendments: The amendments revised the
Technical Specifications by replacing the existing requirements related
to ``operations with a potential for draining the reactor vessel'' with
new requirements on Reactor Pressure Vessel Water Inventory Control to
protect Safety Limit 2.1.1.3, which requires reactor vessel water level
to be greater than the top of active irradiated fuel.
Date of issuance: May 31, 2018.
Effective date: As of the date of issuance and shall be implemented
prior to the commencement of the Unit No. 2 refueling outage (U2R25) in
February 2019.
Amendment Nos.: Unit 1--290, Unit 2--235. A publicly-available
version is in ADAMS under Accession No. ML18123A368; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. DPR-57 and NPF-5: The
amendments revised the Renewed Facility Operating Licenses and
Technical Specifications.
Date of initial notice in Federal Register: August 29, 2017 (82 FR
41071). The supplemental letters dated September 14, 2017; February 19,
2018; and May 1, 2018, provided additional information that clarified
the application, did not expand the scope of the application as
originally noticed, and did not change the NRC staff's original
proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 31, 2018.
No significant hazards consideration comments received: No.
Tennessee Valley Authority (TVA) Docket Nos. 50-259, 50-260, 50-296,
and 72-052, Browns Ferry Nuclear Plant, Unit Nos. 1, 2, and 3,
Limestone County, Alabama
TVA Docket Nos. 50-327, 50-328, and 72-034, Sequoyah Nuclear Plant,
Unit Nos. 1 and 2, Hamilton County, Tennessee
TVA Docket Nos. 50-390, 50-391, and 72-1048, Watts Bar Nuclear Plant,
Unit Nos. 1 and 2, Rhea County, Tennessee
Date of amendment request: January 4, 2017, as supplemented by
letters dated July 7, 2017, and July 27, 2017. (Note: This Notice of
Issuance corrects the amendments by adding the supplement dated July
27, 2017, which was inadvertently omitted from the original Federal
Register notice (January 16, 2018; 83 FR 2234).
Brief description of amendments: The amendments revised TVA
Emergency Plans for the above nuclear plants. Specifically, the
amendments adopted the NRC-endorsed Radiological Emergency Plan
Emergency Action Level schemes developed by the Nuclear Energy
Institute (NEI 99-01, Revision 6, ``Development of Emergency Action
Levels for Non-Passive Reactors'').
Date of issuance: December 22, 2017.
Effective date: As of the date of issuance and shall be implemented
within 180 days from the date of its issuance, or July 3, 2018,
whichever comes later.
Amendment Nos.: Browns Ferry Nuclear Plant--303 (Unit 1), 327 (Unit
2), and 287 (Unit 3); Sequoyah Nuclear Plant--339 (Unit 1) and 332
(Unit 2); and Watts Bar Nuclear Plant--118 (Unit 1) and 18 (Unit 2). A
publicly-available version is in ADAMS under Accession No. ML17289A032;
documents related to these amendments are listed in the Safety
Evaluations enclosed with the amendments. These amendments were
corrected by letter dated May 29, 2018 (ADAMS Accession No.
ML18138A452).
Renewed Facility Operating License Nos. DPR-33, DPR-52, DPR-68,
DPR-77, and DPR-79, and Facility Operating License Nos, NPF-90 and NPF-
96: The amendments revised the licenses.
Date of initial notice in Federal Register: June 19, 2017 (82 FR
27891). The supplemental letters dated July 7, 2017, and July 27, 2017,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the NRC staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 22, 2017.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit No. 1,
Callaway County, Missouri
Date of amendment request: April 6, 2017, as supplemented by letter
dated February 5, 2018.
Brief description of amendment: The amendment revised the Final
Safety Analysis Report to clearly describe conformance with NRC
Regulatory Guide 1.106, Revision 1, ``Thermal Overload Protection for
Electric Motors on Motor-Operated Valves.''
Date of issuance: May 30, 2018.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 218. A publicly-available version is in ADAMS under
Accession No. ML18124A026; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-30: The amendment
revised the Final Safety Analysis Report.
Date of initial notice in Federal Register: July 18, 2017 (82 FR
32885). The supplemental letter dated February 5, 2018, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the NRC staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 30, 2018.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 6th day of June 2018.
For the Nuclear Regulatory Commission.
Tara Inverso,
Acting Deputy Director, Division of Operating Reactor Licensing, Office
of Nuclear Reactor Regulation.
[FR Doc. 2018-12506 Filed 6-18-18; 8:45 am]
BILLING CODE 7590-01-P