[Federal Register Volume 83, Number 99 (Tuesday, May 22, 2018)]
[Notices]
[Pages 23728-23742]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2018-10565]
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NUCLEAR REGULATORY COMMISSION
[NRC-2018-0096]
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 April 24, 2018, to May 7, 2018. The last
biweekly notice was published on May 8, 2018.
DATES: Comments must be filed by June 21, 2018. A request for a hearing
must be filed by July 23, 2018.
ADDRESSES: You may submit comments by any of the following methods:
Federal Rulemaking Website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0096. 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,
[[Page 23729]]
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].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC 2018-0096, 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-2018-0096.
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-2018-0096, facility name, unit
number(s), plant docket number, application date, and subject in your
comment submission.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. The NRC will post all comment submissions at http://www.regulations.gov, as well as enter the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment into ADAMS.
II. Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses and Combined Licenses and Proposed No Significant
Hazards Consideration Determination
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Sec. 50.92 of title 10 of the Code of
Federal Regulations (10 CFR), this means that operation of the facility
in accordance with the proposed amendment would not (1) involve a
significant increase in the probability or consequences of an accident
previously evaluated, or (2) create the possibility of a new or
different kind of accident from any accident previously evaluated, or
(3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period if circumstances change during the 30-day comment
period such that failure to act in a timely way would result, for
example, in derating or shutdown of the facility. If the Commission
takes action prior to the expiration of either the comment period or
the notice period, it will publish in the Federal Register a notice of
issuance. If the Commission makes a final no significant hazards
consideration determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
A. Opportunity To Request a Hearing and Petition for Leave To Intervene
Within 60 days after the date of publication of this notice, any
persons (petitioner) whose interest may be affected by this action may
file a request for a hearing and petition for leave to intervene
(petition) with respect to the action. Petitions shall be filed in
accordance with the Commission's ``Agency Rules of Practice and
Procedure'' in 10 CFR part 2. Interested persons should consult a
current copy of 10 CFR 2.309. The NRC's regulations are accessible
electronically from the NRC Library on the NRC's website at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of
the regulations is available at the NRC's Public Document Room, located
at One White Flint North, Room O1-F21, 11555 Rockville Pike (first
floor), Rockville, Maryland 20852. If a petition is filed, the
Commission or a presiding officer will rule on the petition and, if
appropriate, a notice of a hearing will be issued.
As required by 10 CFR 2.309(d), the petition should specifically
explain the reasons why intervention should be permitted, with
particular reference to the following general requirements for
standing: (1) The name, address, and telephone number of the
petitioner; (2) the nature of the petitioner's right under the Act to
be made a party to the proceeding; (3) the nature and extent of the
petitioner's property, financial, or other interest in the proceeding;
and (4) the possible effect of any decision or order which may be
entered in the proceeding on the petitioner's interest.
In accordance with 10 CFR 2.309(f), the petition must also set
forth the specific contentions which the petitioner seeks to have
litigated in the proceeding. Each contention must consist of a specific
statement of the issue of law or fact to be raised or controverted. In
addition, the petitioner must provide a brief explanation of the bases
for the contention and a concise statement of the alleged facts or
expert opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to the specific sources and
documents on which the petitioner intends to rely to support its
position on the issue. The petition must include sufficient information
to show that a genuine dispute exists with the applicant or licensee on
a material issue
[[Page 23730]]
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 [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
[[Page 23731]]
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 10, 2018. A publicly-available
version is in ADAMS under Accession No. ML18010A344.
Description of amendment request: The amendments would modify the
licensing basis to allow for the implementation of the provisions of 10
CFR 50.69, ``Risk-informed characterization and treatment of
structures, systems, and components for nuclear reactors.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change will permit the use of a risk-informed
categorization process to modify the scope of SSCs [structures,
systems, and components] subject to NRC special treatment
requirements and to implement alternative treatments per the
regulations. The process used to evaluate SSCs for changes to NRC
special treatment requirements and the use of alternative
requirements ensures the ability of the SSCs to perform their design
function. The potential change to special treatment requirements
does not change the design and operation of the SSCs. As a result,
the proposed change does not significantly affect any initiators to
accidents previously evaluated or the ability to mitigate any
accidents previously evaluated. The consequences of the accidents
previously evaluated are not affected because the mitigation
functions performed by the SSCs assumed in the safety analysis are
not being modified. The SSCs required to safely shut down the
reactor and maintain it in a safe shutdown condition following an
accident will continue to perform their design functions.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change will permit the use of a risk-informed
categorization process to modify the scope of SSCs subject to NRC
special treatment requirements and to implement alternative
treatments per the regulations. The proposed change does not change
the functional requirements, configuration, or method of operation
of any SSC. Under the proposed change, no additional plant equipment
will be installed.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will permit the use of a risk-informed
categorization process to modify the scope of SSCs subject to NRC
special treatment requirements and to implement alternative
treatments per the regulations. The proposed change does not affect
any Safety Limits or operating parameters used to establish the
safety margin. The safety margins included in analyses of accidents
are not affected by the proposed change. The regulation requires
that there be no significant effect on plant risk due to any change
to the special treatment requirements for SSCs and that the SSCs
continue to be capable of performing their design basis functions,
as well as to perform any beyond design basis functions consistent
with the categorization process and results.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel,
550 South Tryon Street, M/C DEC45A, Charlotte, NC 28202.
NRC Acting Branch Chief: Brian W. Tindell.
Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear
Power Plant, Unit 1, Wake and Chatham Counties, North Carolina
Date of amendment request: February 1, 2018. A publicly-available
version is in ADAMS under Accession No. ML18033B768.
[[Page 23732]]
Description of amendment request: The amendment would revise the
licensing basis to allow for the implementation of the provisions of 10
CFR 50.69, ``Risk-informed categorization and treatment of structures,
systems and components for nuclear power reactors.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change will permit the use of a risk-informed
categorization process to modify the scope of [structures, systems,
and components] SSCs subject to NRC special treatment requirements
and to implement alternative treatments per the regulations. The
process used to evaluate SSCs for changes to NRC special treatment
requirements and the use of alternative requirements ensures the
ability of the SSCs to perform their design function. The potential
change to special treatment requirements does not change the design
and operation of the SSCs. As a result, the proposed change does not
significantly affect any initiators to accidents previously
evaluated or the ability to mitigate any accidents previously
evaluated. The consequences of the accidents previously evaluated
are not affected because the mitigation functions performed by the
SSCs assumed in the safety analysis are not being modified. The SSCs
required to safely shut down the reactor and maintain it in a safe
shutdown condition following an accident will continue to perform
their design functions. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change will permit the use of a risk-informed
categorization process to modify the scope of SSCs subject to NRC
special treatment requirements and to implement alternative
treatments per the regulations. The proposed change does not change
the functional requirements, configuration, or method of operation
of any SSC. Under the proposed change, no additional plant equipment
will be installed. Therefore, the proposed change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will permit the use of a risk-informed
categorization process to modify the scope of SSCs subject to NRC
special treatment requirements and to implement alternative
treatments per the regulations. The proposed change does not affect
any Safety Limits or operating parameters used to establish the
safety margin. The safety margins included in analyses of accidents
are not affected by the proposed change. The regulation requires
that there be no significant effect on plant risk due to any change
to the special treatment requirements for SSCs and that the SSCs
continue to be capable of performing their design basis functions,
as well as to perform any beyond design basis functions consistent
with the categorization process and results.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lara Nichols, Deputy General Counsel, Duke
Energy Corporation, 550 South Tryon St., M/C DEC45A, Charlotte, NC
28202.
NRC Acting Branch Chief: Brian W. Tindell.
Entergy Louisiana, LLC, and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1 (RBS), West Feliciana Parish, Louisiana
Date of amendment request: January 29, 2018. A publicly-available
version is in ADAMS under Accession No. ML18029A187.
Description of amendment request: The proposed change would modify
the RBS Updated Safety Analysis Report (USAR) and Technical
Requirements Manual to relocate the reactor core isolation cooling
(RCIC) piping injection point from the reactor vessel head spray nozzle
to the feedwater line using the residual heat removal (RHR) shutdown
cooling return line.
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.
Basis: The relocation of the RCIC injection point from the
reactor vessel head spray nozzle to the `A' Feedwater line via the
`A' RHR shutdown cooling return line does not adversely affect the
design function of an System, Structure, or Component (SSC) or a
method of performing or controlling a design function of an SSC as
described in the USAR so there is no change to the likelihood of
occurrence of a malfunction of a structure, system, or component
important to safety previously evaluated in the USAR. There is no
impact to the likelihood of occurrence of a malfunction of a
structure, system, or component because there are no structures
systems or components changed or affected by the scope of this
evaluation.
Inadvertent initiation of RCIC may be categorized as either a
Decrease in Reactor Coolant Temperature event or an Increase in
Reactor Coolant Inventory event. River Bend Transient Safety
Analysis Design Report, 6224.302-000-035A, states that three systems
were considered that could introduce a cold water perturbation
(Decrease in Reactor Coolant Temperature Event) at operating
pressures: RCIC, High Pressure Core Spray (HPCS), and the feedwater
system. This report qualifies improper startup of HPCS or RCIC as
events that would produce no significant power transients. The
proposed change relocated the injection point of the RCIC flow from
the reactor head (RPV [reactor pressure vessel]) to the feedwater
line (FWS). This change will reduce the effects of steam quenching.
However, the effect of steam quenching is not credited in any of the
safety analysis. The only portion of the RCIC system operation that
is credited is water injection at the required flow rate, and the
design function as described in the USAR of the RCIC system is to
maintain or supplement the reactor vessel water inventory. The
source of water for the Inadvertent RCIC injection remains the same.
The destination of the water for the Inadvertent RCIC injection is
still the RPV. The ability of the rerouted equipment to satisfy the
RCIC design function is not reduced from the original design
requirement to inject 600 gpm [gallons per minute] into the RPV.
This is maintained by the RCIC flow controller. The entry location
from the RPV head spray to the feedwater line has no impact to the
consequences of an inadvertent initiation of RCIC. As the
consequences of an inadvertent initiation of RCIC are unchanged, the
consequences of this event remain quantitatively bounded by the Loss
of Feedwater Heating event described in section 15.1.1 of the USAR
for the Decrease in Reactor Coolant Temperature category and bounded
by the Inadvertent HPCS Startup for the Increase in Reactor Coolant
Inventory category.
Changing the injection point of RCIC does not increase the
probability or consequences of an inadvertent RCIC injection. All
affected piping, fittings, and valve pressure boundaries are
qualified to the appropriate fluid transients and operational
conditions in accordance with the design and licensing basis. No
instrument setpoints were changed as a result of this modification.
The RCIC system's modes of operation are not changed or affected by
this modification. Therefore there is no change in the frequency of
an inadvertent initiation of RCIC event. There is no change in the
frequency of inadvertent initiation of RCIC by this modification, so
there is no impact to the probability of any previously evaluated
accident.
[[Page 23733]]
Therefore, it is concluded that this change does not
significantly increase 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.
Basis: The spurious start of RCIC accident is evaluated in the
USAR as Event 9 ``Inadvertent HPCS Pump Start (Moderator Temperature
Decrease) as shown in USAR Appendix 15A. The Inadvertent HPCS Pump
Start event bounds the inadvertent operation of RCIC event and is
quantitatively analyzed in accordance with Reg Guide 1.70 rev. 3.
This event may be classified as either a Decrease in Core Coolant
Temperature event or an Increase in Reactor Coolant Inventory Event,
however was categorized as an Increase in Reactor Coolant Inventory
Event in the RBS USAR as this is the initial effect of this event.
No new accident is created by the scope of this modification because
all aspects of the existing Decrease in Core Coolant Temperature and
Increase in Reactor Coolant Inventory events and their relationship
to the spurious start of RCIC remain applicable.
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.
Basis: The proposed change does not change any accident
analyses. The proposed change does not exceed or alter a design
basis or safety limit; therefore it does not significantly reduce
the margin of safety.
Therefore, it is concluded that 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.
Entergy Louisiana, LLC, and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: February 28, 2018. A publicly-available
version is in ADAMS under Accession No. ML18067A115.
Description of amendment request: The amendment would modify the
River Bend Station Technical Specifications (TSs) by relocating
specific surveillance frequencies to a licensee-controlled program with
the adoption of Technical Specifications Task Force (TSTF) Traveler
TSTF-425, Revision 3, ``Relocate Surveillance Frequencies to Licensee
Control--RITSTF [Risk-Informed TSTF] Initiative 5b.'' Additionally, the
change would add a new program, the Surveillance Frequency Control
Program (SFCP), to TS Chapter 5.0, ``Administrative Controls.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change relocates the specified frequencies for
periodic surveillance requirements to licensee control under a new
Surveillance Frequency Control Program. Surveillance frequencies are
not an initiator to any accident previously evaluated. As a result,
the probability of any accident previously evaluated is not
significantly increased. The systems and components required by the
technical specifications for which the surveillance frequencies are
relocated are still required to be operable, meet the acceptance
criteria for the surveillance requirements, and be capable of
performing any mitigation function assumed in the accident analysis.
As a result, the consequences of any accident previously evaluated
are not significantly increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new or different accidents result from utilizing the proposed
change. The changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements. The changes do not alter assumptions made in
the safety analysis. The proposed changes are consistent with the
safety analysis assumptions and current plant operating practice.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The design, operation, testing methods, and acceptance criteria
for systems, structures, and components (SSCs), specified in
applicable codes and standards (or alternatives approved for use by
the NRC) will continue to be met as described in the plant licensing
basis (including the final safety analysis report and bases to TS),
since these are not affected by changes to the surveillance
frequencies. Similarly, there is no impact to safety analysis
acceptance criteria as described in the plant licensing basis. To
evaluate a change in the relocated surveillance frequency, Entergy
will perform a probabilistic risk evaluation using the guidance
contained in NRC approved NEI [Nuclear Energy Institute] 04-10, Rev.
1 in accordance with the TS SFCP. NEI 04-10, Rev. 1, methodology
provides reasonable acceptance guidelines and methods for evaluating
the risk increase of proposed changes to surveillance frequencies
consistent with Regulatory Guide 1.177.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Anna Vinson Jones, Senior Counsel--Entergy
Services, Inc., 101 Constitution Avenue NW, Suite 200 East, Washington,
DC 20001.
NRC Branch Chief: Robert J. Pascarelli.
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One (ANO), Unit Nos. 1 and 2, Pope County, Arkansas
Date of amendment request: March 29, 2018. A publicly-available
version is in ADAMS under Accession No. ML18088B412.
Description of amendment request: The amendments would revise the
ANO, Units 1 and 2, currently approved Emergency Plan Emergency Action
Level (EAL) scheme, which is based on the Nuclear Energy Institute
(NEI) guidance established in NEI 99-01, Revision 5, ``Methodology for
Development of Emergency Action Levels,'' by adopting the EAL schemes
based on the guidance provided in NEI 99-01, Revision 6, ``Development
of Emergency Action Levels for Non-Passive Reactors.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or
[[Page 23734]]
consequences of an accident previously evaluated?
Response: No.
The proposed changes to the ANO EALs do not involve any physical
changes to plant equipment or systems and do not alter the
assumptions of any accident analyses. The proposed changes do not
adversely affect accident initiators or precursors and do not alter
design assumptions, plant configuration, or the manner in which the
plant is operated and maintained. The proposed changes do not
adversely affect the ability of structures, systems or components
(SSCs) to perform intended safety functions in mitigating the
consequences of an initiating event within the assumed acceptance
limits.
Therefore, the changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed changes.
The changes do not challenge the integrity or performance of any
safety-related systems. No plant equipment is installed or removed,
and the changes do not alter the design, physical configuration, or
method of operation of any plant SSC. Because EALs are not accident
initiators and no physical changes are made to the plant, no new
causal mechanisms are introduced.
Therefore, the changes do not create the possibility of a new or
different kind of accident from an accident previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
Margin of safety is associated with the ability of the fission
product barriers (i.e., fuel cladding, reactor coolant system
pressure boundary, and containment structure) to limit the level of
radiation dose to the public. The proposed changes do not impact
operation of the plant and no accident analyses are affected by the
proposed changes. The changes do not affect the Technical
Specifications or the method of operating the plant. Additionally,
the proposed changes will not relax any criteria used to establish
safety limits and will not relax any safety system settings. The
safety analysis acceptance criteria are not affected by these
changes. The proposed changes will not result in plant operation in
a configuration outside the design basis. The proposed changes do
not adversely affect systems that respond to safely shut down the
plant and to maintain the plant in a safe shutdown condition.
Therefore, the changes do not involve a significant reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Anna Vinson Jones, Senior Counsel, Entergy
Services, Inc., 101 Constitution Avenue NW, Suite 200 East, L-ENT-WDC,
Washington, DC 20001.
NRC Branch Chief: Robert J. Pascarelli.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois
Date of amendment request: February 1, 2018. A publicly-available
version is in ADAMS under Accession No. ML18036A227.
Description of amendment request: The proposed amendments would
revise the Braidwood Station licensing basis for protection from
tornado-generated missiles by identifying the TORMIS Computer Code as
the methodology used for assessing tornado-generated missile protection
of unprotected plant structures, systems, and components (SSCs).
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 NRC TORMIS Safety Evaluation Report [ADAMS Accession No.
ML080870291] states the following:
``The current Licensing criteria governing tornado missile
protection are contained in [NUREG-0800] Standard Review Plan (SRP)
Section 3.5.1.4, [Missiles Generated by Natural Phenomena] and 3.5.2
[Structures, Systems and Components [SSCs]] to be Protected from
Externally Generated Missiles]. These criteria generally specify
that safety-related systems be provided positive tornado missile
protection (barriers) from the maximum credible tornado threat.
However, SRP Section 3.5.1.4 includes acceptance criteria permitting
relaxation of the above deterministic guidance, if it can be
demonstrated that the probability of damage to unprotected essential
safety-related features is sufficiently small.''
As permitted by these SRP sections, the combined probability
will be maintained below an allowable level, i.e., an acceptance
criterion threshold, which reflects an extremely low probability of
occurrence. SRP Section 2.2.3, ``Evaluation of Potential
Accidents,'' established this threshold as approximately 1.0E-06 per
year if, ``when combined with reasonable qualitative arguments, the
realistic probability can be shown to be lower.'' The Braidwood
Station analysis approach assumes that if the sum of the individual
probabilities calculated for tornado missiles striking and damaging
portions of safety-significant SSCs is greater than or equal to
1.0E-06 per year per unit, then installation of tornado missile
protection barriers would be required for certain components to
lower the total cumulative damage probability below the acceptance
criterion of 1.0E-06 per year per unit. Conversely, if the total
cumulative damage probability remains below the acceptance criterion
of 1.0E-06 per year per unit, no additional tornado missile
protection barriers would be required for any of the unprotected
safety-significant components.
With respect to the probability of occurrence or the
consequences of an accident previously evaluated in the UFSAR
[Updated Final Safety Analysis Report], the possibility of a tornado
impacting the Braidwood Station site and causing damage to plant
SSCs is a licensing basis event currently addressed in the UFSAR.
The change being proposed (i.e., the use of the TORMIS methodology
for assessing tornado-generated missile protection of unprotected
plant SSCs), does not affect the probability of a tornado strike on
the site; however, from a licensing basis perspective, the proposed
change does affect the probability that missiles generated by a
tornado will strike and damage certain safety-significant plant
SSCs. There are a defined number of safety-significant components
that could theoretically be struck and damaged by tornado-generated
missiles. The probability of tornado-generated missile hits on these
``important'' systems and components is calculated using the TORMIS
probabilistic methodology. The combined probability of damage for
unprotected safety-significant equipment will be maintained below
the acceptance criterion of 1.0E-06 per year per unit to ensure
adequate equipment remains available to safely shutdown the
reactors, and maintain overall plant safety, should a tornado strike
occur. Consequently, the proposed change does not constitute a
significant increase in the probability of occurrence or the
consequences of an accident based on the extremely low probability
of damage caused by tornado-generated missiles and the commensurate
extremely low probability of a radiological release.
Finally, the use of the TORMIS methodology will have no impact
on accident initiators or precursors; does not alter the accident
analysis assumptions or the manner in which the plant is operated or
maintained; and does not affect the probability of operator error.
Based on the above discussion, 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 impact of a tornado strike on the Braidwood Station site is
a licensing basis event that is explicitly addressed in the
[[Page 23735]]
UFSAR. The proposed change simply involves recognition of the
acceptability of using an analysis tool (i.e., the TORMIS
methodology) to perform probabilistic tornado missile damage
calculations in accordance with approved regulatory guidance. The
proposed change does not result in the creation of any new accident
precursors; does not result in changes to any existing accident
scenarios; and does not introduce any operational changes or
mechanisms that would create the possibility of a new or different
kind of accident.
Therefore, the proposed change will not create the possibility
of a new or different kind of accident than those previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The existing Braidwood Station licensing basis regarding tornado
missile protection of safety-significant SSCs assumes that missile
protection barriers are provided for safety-significant SSCs; or the
unprotected component is assumed to be unavailable post-tornado.
The results of the Braidwood Station TORMIS analysis have
demonstrated that there is an extremely low probability, below an
established regulatory acceptance limit, that these ``important''
SSCs could be struck and subsequently damaged by tornado-generated
missiles. The change in licensing basis from protecting safety-
significant SSCs from tornado missiles, to demonstrating that there
is an extremely low probability that safety-significant SSCs will be
struck and damaged by tornado-generated missiles, does not
constitute a significant decrease in the margin of safety.
Therefore, the proposed change to use the TORMIS methodology
does not involve a significant reduction in the margin of safety.
Based on the above, EGC 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
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.
Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue County,
Minnesota
Date of amendment request: March 15, 2018. A publicly-available
version is in ADAMS under Accession No. ML18074A308.
Brief description of amendment request: The proposed amendments
would revise Prairie Island Nuclear Generating Plant Technical
Specifications (TSs) by relocating specific surveillance frequencies to
a licensee-controlled program with implementation of Nuclear Energy
Institute (NEI) 04-10, ``Risk-Informed Technical Specification
Initiative 5b, Risk-Informed Method for Control of Surveillance
Frequencies,'' Revision 1. The changes are consistent with Technical
Specifications Task Force (TSTF) Traveler TSTF-425, ``Relocate
Surveillance Frequencies to Licensee Control--Risk Informed Technical
Specifications Task Force (RITSTF) Initiative 5b,'' Revision 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 amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change relocates the specified frequencies for
periodic surveillance requirements to licensee control under a new
SFCP [surveillance frequency control program]. Surveillance
frequencies are not an initiator to any accident previously
evaluated. As a result, the probability of any accident previously
evaluated is not significantly increased. The systems and components
required by the technical specifications for which the surveillance
frequencies are relocated are still required to be operable, meet
the acceptance criteria for the surveillance requirements, and be
capable of performing any mitigation function assumed in the
accident analysis. As a result, the consequences of any accident
previously evaluated are not significantly increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any previously evaluated?
Response: No.
No new or different accidents result from utilizing the proposed
change. The changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements. The changes do not alter assumptions made in
the safety analysis. The proposed changes are consistent with the
safety analysis assumptions and current plant operating practice.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The design, operation, testing methods, and acceptance criteria
for systems, structures, and components (SSCs), specified in
applicable codes and standards (or alternatives approved for use by
the NRC) will continue to be met as described in the plant licensing
basis (including the final safety analysis report and bases to TS),
since these are not affected by changes to the surveillance
frequencies. Similarly, there is no impact to safety analysis
acceptance criteria as described in the plant licensing basis. To
evaluate a change in the relocated surveillance frequency, NSPM will
perform a probabilistic risk evaluation using the guidance contained
in NRC approved NEI 04-10, Rev. 1 in accordance with the TS SFCP.
NEI 04-10, Rev. 1, methodology provides reasonable acceptance
guidelines and methods for evaluating the risk increase of proposed
changes to surveillance frequencies consistent with Regulatory Guide
1.177.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Peter M. Glass, Assistant General Counsel,
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
NRC Branch Chief: David J. Wrona.
Northern States Power Company--Minnesota, Docket No. 50-263, Monticello
Nuclear Generating Plant, Wright County, Minnesota
Date of amendment request: March 28, 2018. A publicly-available
version is in ADAMS under Accession No. ML18087A323.
Description of amendment request: The proposed amendment would
modify the Monticello Nuclear Generating Plant licensing basis by the
addition of a license condition to allow for the implementation of the
provisions of 10 CFR 50.69, ``Risk-Informed Categorization and
Treatment of Structures, Systems and Components for Nuclear Power
Reactors.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or
[[Page 23736]]
consequences of an accident previously evaluated?
Response: No.
The proposed change will permit the use of a risk-informed
categorization process to modify the scope of Structures, Systems
and Components (SSCs) subject to NRC special treatment requirements
and to implement alternative treatments per the regulations. The
process used to evaluate SSCs for changes to NRC special treatment
requirements and the use of alternative requirements ensure the
ability of the SSCs to perform their design function. The potential
change to special treatment requirements does not change the design
and operation of the SSCs. As a result, the proposed change does not
significantly affect any initiators to accidents previously
evaluated or the ability to mitigate any accidents previously
evaluated. The consequences of the accidents previously evaluated
are not affected because the mitigation functions performed by the
SSCs assumed in the safety analysis are not being modified. The SSCs
required to safely shut down the reactor and maintain it in a safe
shutdown condition following an accident will continue to perform
their design functions.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any previously evaluated?
Response: No.
The proposed change will permit the use of a risk-informed
categorization process to modify the scope of SSCs subject to NRC
special treatment requirements and to implement alternative
treatments per the regulations. The proposed change does not change
the functional requirements, configuration, or method of operation
of any SSC. Under the proposed change, no additional plant equipment
will be installed.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change will permit the use of a risk-informed
categorization process to modify the scope of SSCs subject to NRC
special treatment requirements and to implement alternative
treatments per the regulations. The proposed change does not affect
any Safety Limits or operating parameters used to establish the
safety margin. The safety margins included in analyses of accidents
are not affected by the proposed change. The regulation requires
that there be no significant effect on plant risk due to any change
to the special treatment requirements for SSCs and that the SSCs
continue to be capable of performing their design basis functions,
as well as to perform any beyond design basis functions consistent
with the categorization process and results.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: 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, 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: April 3, 2018. A publicly-available
version is in ADAMS under Accession No. ML18094A189.
Description of amendment request: The proposed amendment would
change Functional Units 17.A and 17.B of Technical Specification (TS)
Table 4.3-1, ``Reactor Trip System Instrumentation Surveillance
Requirements.'' The Trip Actuating Device Operational Test (TADOT)
column of this table would be revised to delete the ``S/U'' frequency
and replace it with a reference to Table Notation (8), which would
state, ``Prior to entering MODE 1 whenever the unit has been in MODE
3.'' The licensee stated that the change would align the surveillance
requirements and the mode requirement for the Turbine Trip TADOT with
the TS \3/4\.3.1, Table 3.3-1, ``Reactor Trip System Instrumentation,''
channels and interlocks mode requirement.
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 changes revise the surveillance frequency for
reactor trip functions from a turbine trip event. These changes do
not alter these functions physically, or how they are maintained.
Changing the surveillance from ``prior to Startup'' to ``prior to
entering MODE 1'' will continue to ensure operability of the
function before the plant is in a condition that would benefit from
the associated actuation and prior to applicability. Since these
changes will not affect the ability of these trips to perform the
initiation of reactor trips when appropriate, the offsite dose
consequences for an accident will not be impacted. Equally, the
potential to cause an accident is not affected because no plant
system or component has been altered by the proposed changes.
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 changes only affect surveillance frequency
requirements for the turbine trip functions. This does not affect
any physical features of the plant, or the manner in which these
functions are utilized. The proposed surveillance frequency will
require the functions to be verified operable before the turbine
trip functions are applicable and able to perform their trip
functions. Changing the surveillance from ``prior to Startup'' to
``prior to entering MODE 1'' will continue to ensure operability of
the function before the plant is in a condition that would benefit
from the associated actuation. 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 changes do not alter any plant setpoints or
functions that are assumed to actuate in the event of postulated
accidents. The proposed changes do not alter any plant feature and
only alters the MODE which the surveillance tests must be performed.
The proposed changes ensure the functionality of the turbine trips
when assumed in the analysis for 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., Docket Nos. 52-025 and 52-
026, Vogtle Electric Generating Plant, Unit Nos. 3 and 4, Burke County,
Georgia
Date of amendment request: April 6, 2018. A publicly-available
version is in ADAMS under Accession No. ML18096B463.
Description of amendment request: The requested amendments require
changes to the Updated Final Safety Analysis Report (UFSAR) in the form
of
[[Page 23737]]
departures from the incorporated plant-specific Design Control Document
(DCD) Tier 2 information and related changes to the Vogtle Electric
Generating Plant, Unit Nos. 3 and 4, combined license (COL) and COL
Appendix C (and corresponding plant-specific DCD Tier 1) information.
Specifically, the requested amendments include changes to the equipment
survivability assessment requirements associated with hydrogen burns
during beyond design-basis accidents as described in the licensing
basis documents, including COL Condition 2.D(12)(g)9 and plant-specific
Tier 1 Sections 2.2.3 and 2.3.9.
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 and clarifications to the locations of
Hydrogen Igniters 27, 30, 35, 36, 37, and 38 do not adversely affect
any safety-related structure, system or component (SSC) or function.
The hydrogen ignition subsystem is designed to mitigate beyond
design basis hydrogen generation in the containment. With the
proposed changes, the hydrogen ignition subsystem continues to
maintain the designed and analyzed beyond design basis functions.
The hydrogen ignition subsystem maintains its design function to
maintain containment integrity. The proposed changes also reconcile
the as-built equipment with the list of equipment on which the
equipment survivability assessment is performed to provide
additional assurance containment penetrations and combustible gas
control components will perform their design functions after a
hydrogen burn in containment. The changes are to the equipment
assessed, not to the design functions of the equipment. The changes
do not involve an interface with any SSC accident initiator or
initiating sequence of events, and thus, the probabilities of the
accidents evaluated in the plant-specific UFSAR are not affected.
The proposed changes do not involve a change to any mitigation
sequence or the predicted radiological releases due to postulated
accident conditions, thus, the consequences of the accidents
evaluated in the UFSAR are not affected.
The maximum allowable containment vessel leakage rate specified
in the Technical Specifications is unchanged, and radiological
material release source terms are not affected; thus, the
radiological releases in the accident analyses are not affected. 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 reconcile the
as-built equipment with the list of equipment on which the equipment
survivability assessment is performed to provide additional
assurance that containment penetrations and combustible gas control
components will perform their design functions after a hydrogen burn
in containment. The equipment survivability assessment changes are
to the equipment assessed, not to the design functions of the
equipment. The VLS Hydrogen Ignition subsystem does not interface
with/affect safety-related equipment or a fission product barrier.
The subsystem is provided to address the production of hydrogen
following a beyond design basis accident in accordance with 10 CFR
50.44(c). The hydrogen ignition subsystem is a non-Class 1E
subsystem and does not interface with any safety-related system;
thus, no system or design function or equipment qualification is
affected by the proposed changes. The changes to the hydrogen
ignition subsystem do not result in a new failure mode, malfunction
or sequence of events that could affect a radioactive material
barrier or safety-related equipment. The proposed changes do not
adversely affect any system or design function or equipment
qualification as the changes do not modify any SSCs that prevent
safety functions from being performed. The changes do not introduce
a new failure mode, malfunction or sequence of events that could
adversely 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 and clarifications to the locations of
Hydrogen Igniters 27, 30, 35, 36, 37, and 38 maintain the beyond
design basis function of the hydrogen ignition subsystem. The
proposed changes also reconcile the as-built equipment with the list
of equipment on which the equipment survivability assessment is
performed to provide additional assurance containment penetrations
and combustible gas control components will perform their design
functions after a hydrogen burn in containment. The equipment
survivability assessment changes are to the equipment assessed, not
to the design functions of the equipment. The proposed changes would
not affect any safety-related design code, function, design
analysis, safety analysis input or result, or existing design/safety
margin. No safety analysis or design basis acceptance limit/
criterion is challenged or exceeded by the requested 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
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, Unit Nos. 3 and 4, Burke County,
Georgia
Date of amendment request: April 13, 2018. A publicly-available
version is in ADAMS under Accession No. ML18103A249.
Description of amendment request: The requested amendments require
changes to combined license (COL) Appendix A, Technical Specifications
and the Updated Final Safety Analysis Report (UFSAR) in the form of
departures from the incorporated plant-specific Design Control Document
Tier 2 information. Specifically, the requested amendments include
changes to the COL Appendix A, Technical Specifications related to the
statuses of the remotely operated containment isolation valves.
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.
This change clarifies that only Class 1E valves in the
nonessential containment penetration flow paths that receive the
containment isolation signal (T signal) are part of the [Post-
Accident Monitoring (PAM)] Technical Specifications and adds
additional valves to the PAM table in the UFSAR. The Normal Residual
Heat Removal System (RNS), Chemical and Volume Control System (CVS),
Component Cooling Water System (CCS), and Steam Generator System
(SGS) have containment isolation valves that do not
[[Page 23738]]
close on a T signal because they have an accident mitigation
function to be open.
The status of the valves in the essential containment flow paths
are summarized on one non-safety display screen and are separately
indicated on the safety display screens within their respective
systems. Keeping these indications separate from the ``Remotely
Operated Containment Isolation Valve Status'' which is on the
Category 1 display allows the operators to quickly verify that the
nonessential containment flow paths are isolated and then focus on
the availability of the essential flow paths for their defense-in-
depth capabilities.
The valve position indications in the essential flow paths that
penetrate containment are not Post-Accident Monitoring System (PAMS)
B1 variables. These essential flow paths support accident mitigation
functions of non-safety systems and may be intentionally opened for
extended periods of time following an accident. As a result,
excluding them from the PAMS B1 summary indication will increase the
value of the summary indication during operation of the essential
flow paths.
Furthermore, opening these essential flow paths pose low risk of
becoming an unmonitored leak path through the containment vessel.
The valves are isolated when required by separate Protection and
Safety Monitoring System (PMS) signals that are associated with each
system's post-accident functions, and the valve position indications
are designated as PAMS D2 accordingly.
No structure, system, or component (SSC) or function is changed
within this activity. Therefore, the proposed amendment does not
involve a significant increase in the probability of an accident
previously evaluated.
The proposed amendment does not affect the prevention and
mitigation of abnormal events, e.g., accidents, anticipated
operation occurrences, earthquakes, floods, turbine missiles, and
fires or their safety or design analyses. This change does not
involve containment of radioactive isotopes or any adverse effect on
a fission product barrier. There is no impact on previously
evaluated accidents.
Therefore, the proposed changes do not involve a significant
increase in the 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 involve a new failure mechanism or
malfunction, which affects an SSC accident initiator, or interface
with any SSC accident initiator or initiating sequence of events
considered in the design and licensing bases. There is no adverse
effect on radioisotope barriers or the release of radioactive
materials. The proposed amendment does not adversely affect any
accident, including the possibility of creating a new or different
kind of accident from any accident previously evaluated.
Therefore, the proposed changes do 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.
This activity clarifies that only Class 1E valves in the
nonessential containment penetration flow paths that receive the
containment isolation signal (T signal) are part of the PAM
Technical Specifications and adds additional valves to the PAM table
in the UFSAR.
The status of the valves in the essential containment flow paths
are summarized on one non-safety display screen and are separately
indicated on the safety display screens within their respective
systems. Keeping these indications separate from the ``Remotely
Operated Containment Isolation Valve Status'' which is on the
Category 1 display allows the operators to quickly verify that the
nonessential containment flow paths are isolated and then focus on
the availability of the essential flow paths for their defense-in-
depth capabilities.
The valve position indications in the essential flow paths that
penetrate containment are not PAMS B1 variables. These essential
flow paths support accident mitigation functions of non-safety
systems and may be intentionally opened for extended periods of time
following an accident. As a result, excluding them from the PAMS B1
summary indication will increase the value of the summary indication
during operation of the essential flow paths.
Furthermore, opening these essential flow paths pose low risk of
becoming an unmonitored leak path through the containment vessel.
The valves are isolated when required by separate PMS signals that
are associated with each system's post-accident functions and the
valve position indications are designated as PAMS D2 accordingly.
No SSC or function is changed within this activity. The proposed
changes would not affect any safety-related design code, function,
design analysis, safety analysis input or result, or existing
design/safety margin. No safety analysis or design basis acceptance
limit/criterion is challenged or exceeded by the requested changes.
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: M. Stanford Blanton, Balch & Bingham LLP,
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer 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 20, 2018. A publicly-available
version is in ADAMS under Accession No. ML18110A113.
Description of amendment request: The requested amendments propose
changes to the Updated Final Safety Analysis Report (UFSAR) in the form
of departures from the plant-specific Design Control Document (DCD)
Tier 2 information and involves changes to the plant-specific Tier 1
information (and associated Combined License (COL) Appendix C
information). Specifically, the amendment proposes changes to plant-
specific Tier 1 (and COL Appendix C) Table 2.5.2-3, ``PMS Automatically
Actuated Engineered Safety Features,'' to revise the nomenclature for
``Auxiliary Spray and Letdown Purification Line Isolation'' and to
include ``Component Cooling System Containment Isolation Valve
Closure.'' 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 DCD 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 nomenclature changes reflect the current plant
design. These changes provide consistency with the approved plant
design. The changes do not affect the operation of any systems or
equipment that initiate an analyzed accident or alter any
structures, systems, and components accident initiator or initiating
sequence of events. The proposed changes do not result in any
increase in probability of an analyzed accident occurring.
Therefore, the requested 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 nomenclature changes reflect the current plant
design. These changes provide consistency with the approved plant
design. The proposed changes do not affect plant electrical systems,
[[Page 23739]]
and do not 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 affected by the
proposed changes. 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 proposed nomenclature changes reflect the current plant
design. These changes provide consistency with the approved plant
design. No safety analysis or design basis acceptance limit/
criterion is involved. 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.
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.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: August 30, 2016, as supplemented by
letter dated November 20, 2017.
Brief description of amendment: The amendment revised the Columbia
Generating Station Final Safety Analysis Report to reclassify reactor
water cleanup piping, valves, pumps, and mechanical modules located
outside of the primary and secondary containment in the radwaste
building from Quality Group C to Quality Group D.
Date of issuance: April 17, 2018.
Effective date: As of its date of issuance and shall be implemented
from the date of issuance until restart after Refueling Outage 24
(spring 2019).
Amendment No.: 248. A publicly-available version is in ADAMS under
Accession No. ML18075A351; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment. This
Notice of Issuance is being reissued in its entirety to reflect a
correction to the ``Effective date'' by letter dated April 27, 2018
(ADAMS Accession No. ML18109A215).
Renewed Facility Operating License No. NPF-21: The amendment
revised the Final Safety Analysis Report.
Date of initial notice in Federal Register: December 6, 2016 (81 FR
87968). The supplemental letter dated November 20, 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 April 17, 2018.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: July 25, 2017.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3.5.1, ``ECCS [Emergency Core Cooling Systems]--
Operating,'' and deleted the Note associated with Surveillance
Requirement 3.5.1.2 to reflect the residual heat removal system design
and ensure the residual heat removal system's operation is consistent
with the TS 3.5.1 limiting condition for operation requirements.
Date of issuance: May 2, 2018.
Effective date: As of its date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 249. A publicly-available version is in ADAMS under
Accession No. ML18100A199; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-21: The amendment
revised the Renewed Facility Operating License and TSs.
Date of initial notice in Federal Register: November 7, 2017 (82 FR
51649).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 2, 2018.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: March 28, 2017, as supplemented by
letter dated February 28, 2018.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3.8.1.3, ``Diesel Fuel Oil,'' by relocating the
current required stored diesel fuel oil numerical volumes from the TSs
to the TS Bases and replacing them with comparable duration-based
requirements. In addition, the amendment revised TS 3.8.1.1 and TS
3.8.1.2, ``AC [Alternating Current] Sources Operating,'' and ``AC
Sources Shutdown,'' respectively, to relocate the specific numerical
value for feed tank fuel oil volume to the TS Bases and replace it with
the feed tank operating time requirement. The changes are consistent
with Technical Specifications Task Force (TSTF) Traveler TSTF-501,
Revision 1, ``Relocate Fuel Oil and Lube Oil Volume Values to Licensee
Control.''
[[Page 23740]]
Date of issuance: April 26, 2018.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 251. A publicly-available version is in ADAMS under
Accession No. ML18026B053; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-38: The amendment revised the
Facility Operating License and TSs.
Date of initial notice in Federal Register: July 5, 2017 (82 FR
31093). The supplemental letter dated February 28, 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 April 26, 2018.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2, Will County, Illinois, and Docket
Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle
County, Illinois
Date of amendment request: June 30, 2017.
Brief description of amendments: The amendments revised Technical
Specification (TS) 3.7.11, ``Control Room Ventilation (VC) Temperature
Control System,'' to modify the TS Actions for two inoperable VC
temperature control system trains.
Date of issuance: April 30, 2018.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 195/195; 201/201. A publicly-available version is
in ADAMS under Accession No. ML18054B436; documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Renewed Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and
NPF-66: The amendments revised the TSs and Licenses.
Date of initial notice in Federal Register: August 29, 2017 (82 FR
41068).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 30, 2018.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station, Unit No. 1, DeWitt County, Illinois
Date of amendment request: July 18, 2017.
Brief description of amendment: The amendment revised the design
value for the spent fuel storage pool in Technical Specification 4.3.2,
``Drainage,'' to an appropriate value, consistent with the original
design basis.
Date of issuance: April 30, 2018.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No: 217. A publicly-available version is in ADAMS under
Accession No. ML18072A050; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-62: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: September 12, 2017 (82
FR 42848).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 30, 2018.
No significant hazards consideration comments received: No.
Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2,
Goodhue County, Minnesota
Date of amendment request: August 4, 2017, as supplemented by
letter dated November 6, 2017.
Brief description of amendments: The amendments revised the non-
destructive examination inspection interval for special lifting devices
from annually or prior to each use, typically at each refueling outage,
to a 10-year interval.
Date of issuance: May 1, 2018.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 225 (Unit 1) and 212 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML18100A788; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. DPR-42 and DPR-60: The
amendments revised the Prairie Island Nuclear Generating Plant, Unit
Nos. 1 and 2, Updated Safety Analysis Report.
Date of initial notice in Federal Register: September 26, 2017 (82
FR 44855). The supplemental letter dated November 6, 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 May 1, 2018.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo County,
California
Date of amendment request: September 28, 2017.
Brief description of amendments: The amendments revised Technical
Specification (TS) 3.1.4, ``Rod Group Alignment Limits''; TS 3.1.5,
``Shutdown Bank Insertion Limits''; TS 3.1.6, ``Control Bank Insertion
Limits''; and TS 3.1.7, ``Rod Position Indication,'' to adopt Technical
Specifications Task Force (TSTF) Traveler TSTF-547, Revision 1,
``Clarification of Rod Position Requirements.'' The NRC approved the
TSTF and issued an associated model safety evaluation by letter dated
March 4, 2016.
Date of issuance: April 30, 2018.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: 232 (Unit 1) and 234 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML18096A054; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: November 7, 2017 (82 FR
51653).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 30, 2018.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: July 7, 2017, as supplemented by letters
dated November 1, November 27, December 14, December 19 (four letters),
and
[[Page 23741]]
December 22, 2017, and January 22, 2018.
Brief description of amendment: The amendment revised the Renewed
Facility Operating License and Technical Specifications to implement a
measurement uncertainty recapture power uprate. Specifically, the
amendment authorized an increase in the maximum licensed thermal power
level from 3,840 megawatts thermal to 3,902 megawatts thermal, which is
an increase of approximately 1.6 percent.
Date of issuance: April 24, 2018.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment No.: 212. A publicly-available version is in ADAMS under
Accession No. ML18096A542; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-57: The amendment
revised the Renewed Facility Operating License and Technical
Specifications.
Date of initial notice in Federal Register: October 3, 2017 (82 FR
46098). The supplemental letters dated November 1, November 27,
December 14, December 19 (four letters), and December 22, 2017, and
January 22, 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 April 24, 2018.
No significant hazards consideration comments received: No.
Southern California Edison Company, et al., Docket Nos. 50-206, 50-361,
and 50-362, San Onofre Nuclear Generating Station (SONGS), Unit Nos. 1,
2, and 3, San Diego County, California
Date of amendment request: December 19, 2016, as supplemented by
letters dated April 25, 2017, and November 2, 2017.
Brief description of amendments: The amendments replaced the SONGS,
Unit Nos. 1, 2, and 3, Physical Security Plan, Training and
Qualification Plan, and Safeguards Contingency Plan (the Security Plan)
with an Independent Spent Fuel Storage Installation (ISFSI) Only
Security Plan. The NRC staff determined that the proposed SONGS ISFSI-
Only Security Plan continues to meet the standards in 10 CFR 72.212,
``Conditions of general license issued under Sec. 72.210,'' paragraph
(b)(9). As such, the SONGS ISFSI-Only Security Plan provides reasonable
assurance that adequate protective measures can and will be taken in
the event of a design-basis threat of radiological sabotage related to
the spent fuel. These changes more fully reflect the status of the
facility, as well as the reduced scope of potential physical security
challenges at the site once all spent fuel has been moved to dry cask
storage within the onsite ISFSI, an activity that is currently
scheduled for completion in 2019.
Date of issuance: April 23, 2018.
Effective date: As of its date of issuance and shall be implemented
within 60 days following Southern California Edison Company's submittal
of a written certification to the NRC that all spent nuclear fuel
assemblies have been transferred out of the spent fuel pools and placed
in storage within the onsite ISFSI.
Amendment Nos.: 170 (Unit 1), 238 (Unit 2), and 231 (Unit 3). A
publicly-available version is in ADAMS under Accession No. ML17311A364;
the Safety Evaluation enclosed with the amendments includes safeguards
information and is withheld from public disclosure.
Facility Operating License Nos. DPR-13, NPF-10, and NPF-15: The
amendments revised the Facility Operating Licenses.
Date of initial notice in Federal Register: April 4, 2017 (82 FR
16422).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 5, 2018.
No significant hazards consideration comments received: No.
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: August 18, 2017, as supplemented by
letter dated December 15, 2017.
Description of amendments: The amendments authorized the Southern
Nuclear Operating Company to depart from the Vogtle Electric Generating
Plaint Updated Final Safety Analysis Report (UFSAR) Tier 2* and Tier 2
information regarding changes necessary to reflect an increase in the
design pressure of the main steam isolation valve (MSIV) compartments
from 6.0 pounds per square inch (psi) to 6.5 psi and other changes
regarding descriptions of the MSIV compartments. The Tier 2* changes
affect Wall 11 information contained in UFSAR Subsections 3H.3.3,
3H.5.1, and 3H.5.1.3. This change provides additional design margin for
the MSIV Compartments A and B at the Vogtle Electric Generating Plant.
Date of issuance: April 18, 2018.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: 122 (Unit 3) and 121 (Unit 4). Publicly-available
versions are in ADAMS Package Accession No. ML18085A932, which includes
the Safety Evaluation that references documents related to these
amendments.
Facility Combined Licenses No. NPF-91 and NPF-92: Amendments
revised the Facility Combined Licenses.
Date of initial notice in Federal Register: November 21, 2017 (82
FR 55411). The supplemental letter dated December 15, 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.
The Commission's related evaluation of the amendments is contained
in the Safety Evaluation dated April 18, 2018.
No significant hazards consideration comments received: No.
United States Maritime Administration (MARAD), Docket No. 50-238,
Nuclear Ship SAVANNAH (NSS), Baltimore, Maryland
Date of amendment request: October 31, 2017.
Brief description of amendment: The amendment permits MARAD to
begin dismantling and disposing of the NSS without prior approval of
the NRC, consistent with existing decommissioning regulations.
Date of issuance: April 23, 2018.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 15. A publicly-available version is in ADAMS under
Accession No. ML18081A134.
Facility Operating License No. NS-1: This amendment revised the
License.
Date of initial notice in Federal Register: February 13, 2018 (83
FR 6235).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 23, 2017.
No significant hazards consideration comments received: No.
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 23, 2017, as supplemented by letters
dated January 16, 2018, and March 14, 2018.
[[Page 23742]]
Brief description of amendments: The amendments revised plant
Technical Specifications Table 3.7-2 and associated Table Notations,
Table 3.7-4 and Table 4.1-1, reflecting the installation of the Class
1E 4160V negative sequence voltage (open phase) protective circuitry at
Surry Power Station, Unit Nos. 1 and 2, to address the potential for a
consequential open phase condition that could exist on one or two
phases of a primary offsite power source and that would not currently
be detected and mitigated by the existing station electrical protection
scheme.
Date of issuance: May 3, 2018.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: 292 (Unit No. 1) and 292 (Unit No. 2). A publicly-
available version is in ADAMS under Accession No. ML18106A007;
documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-32 and DPR-37: The
amendments revised the Renewed Facility Operating Licenses and
Technical Specifications.
Date of initial notice in Federal Register: October 10, 2017 (82 FR
47040). The supplemental letters dated January 16, 2018, and March 14,
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 3, 2018.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 14th day of May, 2018.
For the Nuclear Regulatory Commission.
Tara Inverso,
Acting Deputy Director, Division of Operating Reactor Licensing, Office
of Nuclear Reactor Regulation.
[FR Doc. 2018-10565 Filed 5-21-18; 8:45 am]
BILLING CODE 7590-01-P