[Federal Register Volume 82, Number 223 (Tuesday, November 21, 2017)]
[Notices]
[Pages 55401-55421]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2017-25063]
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NUCLEAR REGULATORY COMMISSION
[NRC-2017-0220]
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 October 24, 2017 to November 6, 2017. The
last biweekly notice was published on November 7, 2017.
DATES: Comments must be filed by December 21, 2017. A request for a
hearing must be filed by January 22, 2018.
ADDRESSES: You may submit comments by any of the following methods
(unless this document describes a different method for submitting
comments on a specific subject):
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0220. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions, contact
the individual listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
Mail comments to: May Ma, Office of Administration, Mail
Stop: OWFN-2-A13, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001.
For additional direction on obtaining information and submitting
comments, see ``Obtaining Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Shirley Rohrer, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; telephone: 301-415-5411, email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2017-0220, facility name, unit
number(s), plant docket number, application date, and subject when
contacting the NRC about the availability of information for this
action. You may obtain publicly-available information related to this
action by any of the following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0220.
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-2017-0220, 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
[[Page 55402]]
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 Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of
the regulations is available at the NRC's Public Document Room, located
at One White Flint North, Room O1-F21, 11555 Rockville Pike (First
Floor), Rockville, Maryland 20852. If a petition is filed, the
Commission or a presiding officer will rule on the petition and, if
appropriate, a notice of a hearing will be issued.
As required by 10 CFR 2.309(d) the petition should specifically
explain the reasons why intervention should be permitted with
particular reference to the following general requirements for
standing: (1) The name, address, and telephone number of the
petitioner; (2) the nature of the petitioner's right under the Act to
be made a party to the proceeding; (3) the nature and extent of the
petitioner's property, financial, or other interest in the proceeding;
and (4) the possible effect of any decision or order which may be
entered in the proceeding on the petitioner's interest.
In accordance with 10 CFR 2.309(f), the petition must also set
forth the specific contentions which the petitioner seeks to have
litigated in the proceeding. Each contention must consist of a specific
statement of the issue of law or fact to be raised or controverted. In
addition, the petitioner must provide a brief explanation of the bases
for the contention and a concise statement of the alleged facts or
expert opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to the specific sources and
documents on which the petitioner intends to rely to support its
position on the issue. The petition must include sufficient information
to show that a genuine dispute exists with the applicant or licensee on
a material issue of law or fact. Contentions must be limited to matters
within the scope of the proceeding. The contention must be one which,
if proven, would entitle the petitioner to relief. A petitioner who
fails to satisfy the requirements at 10 CFR 2.309(f) with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene.
Parties have the opportunity to participate fully in the conduct of the
hearing with respect to resolution of that party's admitted
contentions, including the opportunity to present evidence, consistent
with the NRC's regulations, policies, and procedures.
Petitions must be filed no later than 60 days from the date of
publication of this notice. Petitions and motions for leave to file new
or amended contentions that are filed after the deadline will not be
entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the three factors in 10
CFR 2.309(c)(1)(i) through (iii). The petition must be filed in
accordance with the filing instructions in the ``Electronic Submissions
(E-Filing)'' section of this document.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to establish when the hearing is held. If the final determination is
that the amendment request involves no significant hazards
consideration, the Commission may issue the amendment and make it
immediately effective, notwithstanding the request for a hearing. Any
hearing would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, then any hearing held would take place before
the issuance of the amendment unless the Commission finds an imminent
danger to the health or safety of the public, in which case it will
issue an appropriate order or rule under 10 CFR part 2.
A State, local governmental body, Federally-recognized Indian
Tribe, or agency thereof, may submit a petition to the Commission to
participate as a party under 10 CFR 2.309(h)(1). The petition should
state the nature and extent of the petitioner's interest in the
proceeding. The petition should be submitted to the Commission no later
than 60 days from the date of publication of this notice. The petition
must be filed in accordance with the filing instructions in the
``Electronic Submissions (E-Filing)'' section of this document, and
should meet the requirements for petitions set forth in this section,
except that under 10 CFR 2.309(h)(2) a State, local governmental body,
or federally recognized Indian Tribe, or agency thereof does not need
to address the standing requirements in 10 CFR 2.309(d) if the facility
is located within its boundaries. Alternatively, a State, local
governmental body, Federally-recognized Indian Tribe, or agency thereof
may participate as a non-party under 10 CFR 2.315(c). If a hearing is
granted, any person who is not a party to the proceeding and is not
affiliated with or represented by a party may, at the discretion of the
presiding officer, be permitted to make a limited appearance pursuant
to the provisions of 10 CFR 2.315(a). A person making a limited
appearance may make an oral or written statement of his or her position
on the issues but may not otherwise participate in the proceeding. A
limited appearance may be made at any session of the hearing or at any
prehearing conference, subject to the limits and conditions as may be
imposed by the presiding officer. Details regarding the opportunity to
make a limited appearance will be provided by the presiding officer if
such sessions are scheduled.
B. Electronic Submissions (E-Filing)
All documents filed in NRC adjudicatory proceedings, including a
request for hearing and petition for leave to intervene (petition), any
motion or other document filed in the proceeding prior to the
submission of a request for hearing or petition to intervene, and
documents filed by interested governmental entities that request to
participate under 10 CFR 2.315(c), must be filed in accordance with the
NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR
46562, August 3, 2012). The E-Filing process requires participants to
submit and serve all adjudicatory documents over the internet, or in
some cases to mail copies on electronic storage media. Detailed
guidance on making electronic submissions may be found in the Guidance
for Electronic Submissions to the NRC and on the NRC Web site 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
[[Page 55403]]
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 Web site 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 Web site at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the
time the document is submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an email notice confirming receipt of the document. The
E-Filing system also distributes an email notice that provides access
to the document to the NRC's Office of the General Counsel and any
others who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
document on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before adjudicatory documents are
filed so that they can obtain access to the documents via the E-Filing
system.
A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC's Electronic
Filing Help Desk through the ``Contact Us'' link located on the NRC's
public Web site at http://www.nrc.gov/site-help/e-submittals.html, by
email to [email protected], or by a toll-free call at 1-866-672-
7640. The NRC Electronic Filing Help Desk is available between 9 a.m.
and 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.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear
Plant (PNP), Van Buren County, Michigan
Date of amendment request: August 31, 2017. A publicly-available
version is in ADAMS under Accession No. ML17248A389.
Description of amendment request: The proposed amendment would
revise the PNP Site Emergency Plan (SEP) for the permanently shut down
and defueled condition. The proposed PNP SEP changes would revise the
shift staffing and Emergency Response Organization (ERO) staffing.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to the PNP SEP do not impact the function
of plant structures, systems, or components (SSCs). The proposed
changes do not affect accident initiators or precursors, nor does it
alter design assumptions. The proposed changes do not prevent the
ability of the on-shift staff and augmented ERO to perform their
intended functions to mitigate the consequences of any accident or
event that will be credible in the permanently shut down and
defueled condition. The proposed changes only remove positions that
will no longer be credited in the PNP SEP.
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?
[[Page 55404]]
Response: No.
The proposed changes reduce the number of on-shift and augmented
ERO positions commensurate with the hazards associated with a
permanently shut down and defueled facility. The proposed changes do
not involve installation of new equipment or modification of
existing equipment, so that no new equipment failure modes are
introduced. Also, the proposed changes do not result in a change to
the way that the equipment or facility is operated so that no new
accident initiators are created.
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.
Margin of safety is associated with confidence in 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 are
associated with the PNP SEP and do not impact operation of the plant
or its response to transients or accidents. The change does not
affect the Technical Specifications. The proposed changes do not
involve a change in the method of plant operation, and no accident
analyses will be affected by the proposed changes. Safety analysis
acceptance criteria are not affected by the proposed changes. The
revised PNP SEP will continue to provide the necessary response
staff with the proposed changes.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William Dennis, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White Plains, NY
10601.
NRC Branch Chief: Douglas A. Broaddus.
Exelon Generation Company (EGC), 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: September 1, 2017. A publicly-available
version is in ADAMS under Accession No. ML17244A093.
Description of amendment request: The amendments would modify the
licensing basis by the addition of a license condition to allow for the
implementation of the provisions of 10 CFR, Section 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 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 [Nuclear Regulatory
Commission] 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 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: Tamra Domeyer, Associate General Counsel,
Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: David J. Wrona.
Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket Nos. 50-277
and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and
Lancaster Counties, Pennsylvania
Date of amendment request: August 30, 2017, as supplemented by
letter dated October 24, 2017. Publicly-available versions are in ADAMS
under Accession Nos. ML17243A014 and ML17297B521, respectively.
Description of amendment request: The amendments would modify the
licensing basis by the addition of a license condition to allow for the
implementation of the provisions of 10 CFR 50.69, ``Risk-informed
categorization and treatment of structures, systems and components 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, with NRC staff edits shown in
square brackets:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change 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
[[Page 55405]]
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: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Rd., Warrenville, IL
60555.
NRC Branch Chief: James G. Danna.
Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket Nos. 50-277
and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and
Lancaster Counties, Pennsylvania
Date of amendment request: September 29, 2017. A publicly-available
version is in ADAMS under Accession No. ML17275A069.
Description of amendment request: The amendments would revise
Technical Specification (TS) requirements related to the direct current
(DC) electrical power system. The proposed changes are based on
Technical Specifications Task Force (TSTF) Traveler TSTF-500, Revision
2, ``DC Electrical Rewrite--Update to TSTF-360.''
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 restructures the TS for the direct current
(DC) electrical power system. The proposed changes add actions to
specifically address battery charger inoperability. The DC
electrical power system, including associated battery chargers, is
not an initiator of any accident sequence analyzed in the Updated
Final Safety Analysis Report (UFSAR). Operation in accordance with
the proposed TS ensures that the DC electrical power system is
capable of performing its function as described in the UFSAR.
Therefore, the mitigative functions supported by the DC electrical
power system will continue to provide the protection assumed by the
analysis, and the probability of previously analyzed accidents will
not increase by implementing these changes.
The relocation of preventive maintenance surveillances, and
certain operating limits and actions, to a newly created licensee-
controlled Battery Monitoring and Maintenance Program will not
challenge the ability of the DC electrical power system to perform
its design function. Appropriate monitoring and maintenance,
consistent with industry standards, will continue to be performed.
In addition, the DC electrical power system is within the scope of
10 CFR 50.65, ``Requirements for monitoring the effectiveness of
maintenance at nuclear power plants,'' which will ensure the control
of maintenance activities associated with the DC electrical power
system.
The integrity of fission product barriers, plant configuration,
and operating procedures as described in the UFSAR will not be
affected by the proposed changes. Therefore, the consequences of
previously analyzed accidents will not increase by implementing
these 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 change involves restructuring the TS for the DC
electrical power system. The DC electrical power system, including
associated battery chargers, is not an initiator to any accident
sequence analyzed in the UFSAR. Rather, the DC electrical power
system is used to supply equipment used to mitigate an accident.
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 margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. The proposed changes will not adversely affect
operation of plant equipment. These changes will not result in a
change to the setpoints at which protective actions are initiated.
Sufficient DC capacity to support operation of mitigation equipment
is ensured. The changes associated with the new battery maintenance
and monitoring program will ensure that the station batteries are
maintained in a highly reliable manner. The equipment fed by the DC
electrical sources will continue to provide adequate power to safety
related loads in accordance with analysis assumptions.
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: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Rd., Warrenville, IL
60555.
NRC Branch Chief: James G. Danna.
Exelon Generation Company, LLC, Docket No. 50-333, James A. FitzPatrick
Nuclear Power Plant, Oswego County, New York
Date of amendment request: October 2, 2017. A publicly-available
version is in ADAMS under Accession No. ML17275A520.
[[Page 55406]]
Description of amendment request: The amendment would revise the
James A. FitzPatrick Nuclear Power Plant Technical Specifications (TSs)
to adopt Technical Specifications Task Force (TSTF) Traveler TSTF-542,
Revision 2, ``Reactor Pressure Vessel Water Inventory Control'' (ADAMS
Accession No. ML16074A448). Specifically, the licensee proposed changes
to replace TS requirements related to operations with a potential for
draining the reactor vessel (OPDRVs) with new requirements on reactor
pressure vessel (RPV) water inventory control (WIC) to protect Safety
Limit 2.1.1.3.
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 replace existing TS requirements related to
OPDRVs with new requirements on RPV WIC that will protect Safety
Limit 2.1.1.3. Draining of RPV water inventory in Mode 4 (i.e., cold
shutdown) and Mode 5 (i.e., refueling) is not an accident previously
evaluated, and therefore replacing the existing TS controls to
prevent or mitigate such an event with a new set of controls has no
effect on any accident previously evaluated. RPV water inventory
control in Mode 4 or Mode 5 is not an initiator of any accident
previously evaluated. The existing OPDRV controls or the proposed
RPV WIC controls are not mitigating actions assumed in any accident
previously evaluated.
The proposed changes reduce the probability of an unexpected
draining event (which is not a previously evaluated accident) by
imposing new requirements on the limiting time in which an
unexpected draining event could result in the reactor vessel water
level dropping to the top of the active fuel (TAF). These controls
require cognizance of the plant configuration and control of
configurations with unacceptably short drain times. These
requirements reduce the probability of an unexpected draining event.
The current TS requirements are only mitigating actions and impose
no requirements that reduce the probability of an unexpected
draining event.
The proposed changes reduce the consequences of an unexpected
draining event (which is not a previously evaluated accident) by
requiring an Emergency Core Cooling System (ECCS) subsystem to be
operable at all times in Modes 4 and 5. The current TS requirements
do not require any water injection systems, ECCS or otherwise, to be
Operable in certain conditions in Mode 5. The change in requirement
from two ECCS subsystems to one ECCS subsystem in Modes 4 and 5 does
not significantly affect the consequences of an unexpected draining
event because the proposed Actions ensure equipment is available
within the limiting drain time that is as capable of mitigating the
event as the current requirements. The proposed controls provide
escalating compensatory measures to be established as calculated
drain times decrease, such as verification of a second method of
water injection and additional confirmations that containment and/or
filtration would be available if needed.
The proposed changes reduce or eliminate some requirements that
were determined to be unnecessary to manage the consequences of an
unexpected draining event, such as automatic initiation of an ECCS
subsystem and control room ventilation. These changes do not affect
the consequences of any accident previously evaluated since a
draining event in Modes 4 and 5 is not a previously evaluated
accident and the requirements are not needed to adequately respond
to a draining event.
Therefore, the proposed changes do 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 replace existing TS requirements related to
OPDRVs with new requirements on RPV WIC that will protect Safety
Limit 2.1.1.3. The proposed changes will not alter the design
function of the equipment involved. Under the proposed changes, some
systems that are currently required to be operable during OPDRVs
would be required to be available within the limiting drain time or
to be in service depending on the limiting drain time. Should those
systems be unable to be placed into service, the consequences are no
different than if those systems were unable to perform their
function under the current TS requirements.
The event of concern under the current requirements and the
proposed changes are an unexpected draining event. The proposed
changes do not create new failure mechanisms, malfunctions, or
accident initiators that would cause a draining event or a new or
different kind of accident not previously evaluated or included in
the design and licensing bases.
Therefore, the proposed changes do 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 replace existing TS requirements related to
OPDRVs with new requirements on RPV WIC. The current requirements do
not have a stated safety basis and no margin of safety is
established in the licensing basis. The safety basis for the new
requirements is to protect Safety Limit 2.1.1.3. New requirements
are added to determine the limiting time in which the RPV water
inventory could drain to the top of the fuel in the reactor vessel
should an unexpected draining event occur. Plant configurations that
could result in lowering the RPV water level to the TAF within one
hour are now prohibited. New escalating compensatory measures based
on the limiting drain time replace the current controls. The
proposed TS establish a safety margin by providing defense-in-depth
to ensure that the Safety Limit is protected and to protect the
public health and safety. While some less restrictive requirements
are proposed for plant configurations with long calculated drain
times, the overall effect of the change is to improve plant safety
and to add safety margin.
Therefore, the proposed 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: Donald P. Ferraro, Assistant General
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Suite 305,
Kennett Square, PA 19348.
NRC Branch Chief: James G. Danna.
Florida Power & Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Nuclear Generating Unit Nos. 3, and 4, Miami-Dade County, Florida
Date of amendment request: August 23, 2017, as supplemented by
letter dated October 19, 2017. Publicly-available versions are in ADAMS
under Accession Nos. ML17235B008 and ML17292A789, respectively.
Description of amendment request: The amendments would modify the
Technical Specifications (TSs) to relocate the Explosive Gas Monitoring
Instrumentation, Explosive Gas Mixture, and Gas Decay Tanks System
requirements to licensee-controlled documents and establish a Gas Decay
Tank Explosive Gas and Radioactivity Monitoring Program. The proposed
amendments also relocate the Standby Feedwater System requirements to
licensee-controlled documents and modify related Auxiliary Feedwater
(AFW) System requirements.
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.
[[Page 55407]]
The proposed license amendments modify the Turkey Point TS by
relocating the Explosive Gas Monitoring Instrumentation, Explosive
Gas Mixture, Gas Decay Tanks and Standby Feedwater System
requirements to licensee controlled documents, by relatedly
modifying the AFW System requirements and by establishing a Gas
Decay Tank Explosive Gas and Radioactivity Monitoring Program. The
proposed changes are administrative in nature and do not alter any
plant equipment or the manner in which plant equipment is operated
and maintained. All equipment limitations, applicable methodologies
and surveillances are maintained by the proposed changes. In
addition, the proposed changes to the AFW System requirements
enhance plant safety. As such, the proposed changes cannot affect
the initiators, the likelihood or the expected outcomes of any
analyzed accidents.
Therefore, facility operation in accordance with the proposed
changes would 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 license amendments modify the Turkey Point TS by
relocating the Explosive Gas Monitoring Instrumentation, Explosive
Gas Mixture, Gas Decay Tanks and Standby Feedwater System
requirements to licensee controlled documents, by relatedly
modifying the AFW System requirements and by establishing a Gas
Decay Tank Explosive Gas and Radioactivity Monitoring Program. The
proposed changes neither install or remove plant equipment nor alter
any plant equipment design, configuration, or method of operation.
Hence, no new failure mechanisms are introduced as a result of the
proposed changes.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed license amendments modify the Turkey Point TS by
relocating the Explosive Gas Monitoring Instrumentation, Explosive
Gas Mixture, Gas Decay Tanks and Standby Feedwater System
requirements to licensee controlled documents, by relatedly
modifying the AFW System requirements and by establishing a Gas
Decay Tank Explosive Gas and Radioactivity Monitoring Program. The
proposed changes neither involve changes to safety analyses
assumptions, safety limits, or limiting safety system settings nor
adversely impact plant operating margins or the reliability of
equipment credited in safety analyses.
Therefore, operation of the facility in accordance with the
proposed changes will not involve a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Blvd., MS LAW/JB,
Juno Beach, FL 33408-0420.
NRC Branch Chief: Undine Shoop.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold
Energy Center (DAEC), Linn County, Iowa
Date of amendment request: September 5, 2017. A publicly-available
version is in ADAMS under Accession No. ML17248A284.
Description of amendment request: The proposed amendment would
revise DAEC Technical Specifications 3.5.1, ``ECCS [emergency core
cooling system]-Operating.'' The proposed change would decrease the
nitrogen supply requirement for the Automatic Depressurization System
(ADS) in Surveillance Requirement (SR) 3.5.1.3 from 100 days to 30
days.
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 modifies a SR for verification of the
nitrogen supply for the ADS accumulators. Accidents are initiated by
the malfunction of plant equipment, or the catastrophic failure of
plant structures, systems or components. The performance of this
surveillance is not a precursor to any accident previously evaluated
and does not change the manner in which the ADS operates. Technical
evaluation of the change concluded that a 30-day nitrogen supply is
more than adequate to ensure that the reactor is depressurized, so
the consequences of an accident remain unchanged.
Therefore, the proposed change does not involve a significant
increase in the probability or consequence of a previously evaluated
accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve physical alterations to the
plant. No new or different type of equipment will be installed, and
there are no physical modifications required to existing installed
equipment associated with the proposed change. The proposed change
does not create any failure mechanism, malfunction or accident
initiator not already considered in the design and licensing basis.
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.
Although the proposed change will decrease the required supply
of nitrogen for the ADS accumulators from 100 days to 30 days, the
assessment above has shown that the reactor would be depressurized
within 3 days following any postulated accident or event that would
create a hostile environment in the drywell. Once initial
depressurization is completed, long term core cooling can be assured
without ADS.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William Blair, P.O. Box 14000, Juno Beach,
FL 33408-0420.
NRC Branch Chief: David J. Wrona.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold
Energy Center, Linn County, Iowa
Date of amendment request: August 31, 2017. A publicly-available
version is in ADAMS under Accession No. ML17243A469.
Description of amendment request: The proposed amendment would
modify the licensing basis by the addition of a license condition to
allow for the implementation of the provisions of 10 CFR, part 50.69,
``Risk-Informed Categorization and Treatment of Structures, Systems,
and Components (SSCs) for Nuclear Power Reactors.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change will permit the use of a risk-informed
categorization process to modify the scope of SSCs subject to NRC
special treatment requirements and to implement alternative
treatments per the regulations. The process used to evaluate SSCs
for changes to NRC special treatment
[[Page 55408]]
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: William Blair, P.O. Box 14000, Juno Beach,
FL 33408-0420.
NRC Branch Chief: David J. Wrona.
NextEra Energy, Point Beach Nuclear Plant (PBNP), LLC, Docket Nos. 50-
266 and 50-301, Point Beach Nuclear Plant, Units 1 and 2, Town of Two
Creeks, Manitowoc County, Wisconsin
Date of amendment request: June 23, 2017, as supplemented by letter
dated August 21, 2017. Publicly-available versions are in ADAMS under
Accession Nos. ML17174A458, and ML17233A283, respectively.
Description of amendment request: The amendments would revise the
Emergency Plan for PBNP to adopt the Nuclear Energy lnstitute's (NEl's)
revised Emergency Action Level (EAL) scheme described in NEI 99-01,
Revision 6, ``Development of Emergency Action Levels for Non-Passive
Reactors,'' which has been endorsed by the NRC.
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 does not impact the physical configuration
or function of plant structures, systems, or components (SSCs) or
the manner in which SSCs are operated, maintained, modified, tested,
or inspected. No actual facility equipment or accident analyses are
affected by the proposed changes.
The change revises the NextEra Emergency Action Levels to be
consistent with the NRC endorsed EAL scheme contained in NEI 99-01,
Revision 6, ``Methodology for Development of Emergency Action
Levels,'' but does not alter any of the requirements of the
Operating License or the Technical Specifications.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
The proposed change does not create any new failure modes for
existing equipment or any new limiting single failures.
Additionally, the proposed change does not involve a change in the
methods governing normal plant operation, and all safety functions
will continue to perform as previously assumed in the accident
analyses. Thus, the proposed change does not adversely affect the
design function or operation of any structures, systems, and
components important to safety. No new accident scenarios, failure
mechanisms, or limiting single failures are introduced as a result
of the proposed change. The proposed change does not challenge the
performance or integrity of any safety-related system.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The margin of safety associated with the acceptance criteria of
any accident is unchanged. The proposed change will have no affect
on the availability, operability, or performance of safety-related
systems and components. The proposed change will not adversely
affect the operation of plant equipment or the function of equipment
assumed in the accident analysis. The proposed amendment does not
involve changes to any safety analyses assumptions, safety limits,
or limiting safety system settings. The changes do not adversely
impact plant operating margins or the reliability of equipment
credited in the safety analyses.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William Blair, Managing Attorney--Nuclear,
Florida Power & Light Company, P.O. Box 14000, 700 Universe Boulevard,
Juno Beach, FL 33408-0420.
NRC Branch Chief: David J. Wrona.
PSEG Nuclear LLC and Exelon Generation Company, LLC, Docket Nos. 50-272
and 50-311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of amendment request: September 27, 2017. A publicly-available
version is in ADAMS under Accession No. ML17270A076.
Description of amendment request: The amendments would relocate the
reactor coolant system pressure isolation valve (RCS PIV) table from
the technical specifications (TSs) to the technical requirements manual
(TRM). The request would also remove references to the table and move
all notes and leakage acceptance criteria from the table to the TS
surveillance requirements.
[[Page 55409]]
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 to the TS will not alter the way any
structure, system, or component (SSC) functions, and will not alter
the manner in which the plant is operated. The proposed changes do
not alter the design of any SSC. The relocation of the RCS PIV valve
lists from the TS to the TRM is an administrative change. Future
revisions to the TRM are subject to 10 CFR 50.59. Therefore the
probability of an accident previously evaluated is not significantly
increased.
The proposed changes do not alter the RCS PIV leakage limits
contained in the TS nor do they alter the frequency for testing of
the RCS PIV. Therefore, the consequences of an accident previously
evaluated are not increased.
Therefore, these proposed changes do not represent 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 do not involve a modification to the
physical configuration of the plant or changes in the methods
governing normal plant operation. The proposed changes will not
impose any new or different requirement or introduce a new accident
initiator, accident precursor, or malfunction mechanism. The
proposed changes are administrative in nature.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes to the RCS PIV TS are administrative in
nature. The proposed changes do not alter the RCS PIV leakage limits
contained in the TS nor do they alter the frequency for testing of
the RCS PIV. The proposed changes will not result in changes to
system design or setpoints that are intended to ensure timely
identification of plant conditions that could be precursors to
accidents or potential degradation of accident mitigation systems.
The proposed amendment will not result in a design basis or
safety limit being exceeded or altered. Therefore, since the
proposed changes do not impact the response of the plant to a design
basis accident, 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: Jeffrie J. Keenan, PSEG Nuclear LLC--N21,
P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: James G. Danna.
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: October 6, 2017. A publicly-available
version is in ADAMS under Accession No. ML17279A715.
Description of amendment request: The proposed amendment would
increase the Integrated Leak Rate Test (ILRT) Peak Calculated
Containment Internal Pressure, Pa, listed in Technical
Specification (TS) 6.8.4.g, ``Containment Leakage Rate Testing
Program,'' to remove the reference to Regulatory Guide (RG) 1.163,
``Performance-Based Containment Leak Test Program,'' dated September
1995 and ANSI/ANS (American National Standards Institute/American
Nuclear Society)--56.8-2002, ``Containment System Leakage Testing
Requirements,'' and to replace the reference of Nuclear Energy
Institute (NEI) 94-01, Revision 3-A, ``Industry Guideline for
Implementing Performance-Based option of 10 CFR part 50, Appendix J,''
with NEI 94-01, Revision 2-A.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below, with NRC staff edits in square
brackets:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes involve removal of RG 1.163 and ANSl/ANS-
56.8-2002 references, replacement of NEI 94-01, Revision 3-A with
NEI 94-01, Revision 2-A, and an increase in the Pa [Peak
Calculated Containment Internal Pressure] value for containment
leakage testing. The activity does not involve a physical change to
the plant or a change in the manner in which the plant is operated
or controlled. The containment is designed to provide an essentially
leak tight barrier against the uncontrolled release of radioactivity
to the environment for postulated accidents. As such, the reactor
containment itself and the testing requirements invoked to
periodically demonstrate the integrity of the reactor containment
exist to ensure the plant's ability to mitigate the consequences of
an accident, and do not involve the prevention or identification of
any precursors of an accident.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The integrity of the reactor containment is subject to two types
of failure mechanisms which can be categorized as (1) activity based
and (2) time based. Activity based failure mechanisms are defined as
degradation due to system and/or component modifications or
maintenance. Local leak rate test requirements and administrative
controls such as configuration management and procedural
requirements for system restoration ensure that containment
integrity is not degraded by plant modifications or maintenance
activities. The updated Pa value reflects the updated
mass and energy release and containment response calculations,
ensuring a sound technical basis for the local and integrated
leakage tests.
To mitigate time-based mechanisms, the design and construction
requirements of the containment itself combined with the containment
inspections performed in accordance with ASME [American Society of
Mechanical Engineers], Section XI and the Maintenance Rule serve to
provide a high degree of assurance that the containment will not
degrade in a manner that is detectable only by a Type A test. The
change to the Pa value is less than 1 psid [per square
inch differential]. Radiological consequences will continue to be
evaluated at the Technical Specification allowed leakage,
La [allowed leakage] of 0.20 percent by weight of air,
which will not be increased despite the increase in Pa.
As described in Section 3.5, past leakage testing yielded values
well under La. Based on the above, neither the reference
changes nor the Pa change involves a significant increase
in the 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 involve removal of RG 1.163 and ANSl/ANS-
56.8-2002 references, replacement of NEI 94-01, Revision 3-A with
NEI 94-01, Revision 2-A, and an increase in the Pa value
for containment leakage testing. The reactor containment and the
testing requirements invoked to periodically demonstrate the
integrity of the reactor containment exist to ensure the plant's
ability to mitigate the consequences of an accident. There are not
any accident initiators or precursors affected by the revision. The
proposed TS change does not involve a physical change to the plant
or the manner in which the plant is operated or controlled.
Therefore, the proposed TS 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?
[[Page 55410]]
Response: No.
The proposed changes involve removal of RG 1.163 and ANSl/ANS-
56.8-2002 references, replacement of NEI 94-01, Revision 3-A with
NEI 94-01, Revision 2-A, and an increase in the Pa value
for containment leakage testing. The proposed TS change does not
involve a physical change to the plant or a change in the manner in
which the plant is operated or controlled. Using the same analysis
methodology as described in WCAP-10325-P-A [Westinghouse LOCA [loss-
of-accident coolant] Mass and Energy Release Model for Containment
Design], the updated mass and energy release and containment
response analyses corrected input errors identified in the NSALs
[Westinghouse Nuclear Safety Advisory Letters] described previously.
As shown in Figure 1 [October 6, 2017, submittal], the correction of
these errors resulted in a slightly higher predicted peak pressure
than that of the current licensing basis but does not pose a
significant challenge to the design limit.
The specific requirements and conditions of the Primary
Containment Leak Rate Testing Program, as defined in the Technical
Specifications, exist to ensure that the degree of reactor
containment structural integrity and leak-tightness that is
considered in the plant safety analysis is maintained. The overall
containment leak rate limit specified by the Technical Specification
is maintained. The containment inspections performed in accordance
with ASME, Section XI and the Maintenance Rule serve to provide a
high degree of assurance that the containment will not degrade in a
manner that is detectable only by Type A testing. The combination of
these factors ensures that the margin of safety that is in plant
safety analysis is maintained. The design, operation, testing
methods and acceptance criteria for Type A, B, and C containment
leakage tests specified in applicable codes and standards will
continue to be met.
Therefore, the proposed TS 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, Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia
Date of amendment request: July 28, 2017. A publicly-available
version is in ADAMS under Accession No. ML17209A759.
Description of amendment request: The amendment request proposes to
revise Technical Specification Section 1.1 (TS), Definition of
Actuation Logic Test, by adding a new TS Section 1.1 Definition of
Actuation Logic Output Test (ALOT), revising existing Surveillance
Requirements 3.3.15.1 and 3.3.16.1 and adding new Surveillance
Requirements 3.3.15.2 and 3.3.16.2 to implement the new ALOT. This
submittal requests approval of the license amendment that is necessary
to implement these changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(A), licensee has provided
its analysis of the issue on no significant hazards consideration
determination, 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.
There are no design changes associated with the proposed
amendment. All design, material, and construction standards that
were applicable prior to this amendment request will continue to be
applicable.
The [Processor Module Self-Diagnostic (PMS)] will continue to
function in a manner consistent with the plant design basis. There
will be no changes to the PMS operating limits. The existing
ACTUATION LOGIC TEST Surveillance Requirements are revised such that
different portions of the PMS logic circuitry are tested on
appropriate surveillance test frequencies.
The proposed change will not adversely affect accident
initiators or precursors or adversely alter the design assumptions,
conditions, and configuration of the facility, or the manner in
which the plant is operated and maintained, with respect to such
initiators or precursors.
The proposed changes will not alter the ability of structures,
systems, and components (SSCs) to perform their specified safety
functions to mitigate the consequences of an initiating event within
the assumed acceptance limits.
Accident analysis acceptance criteria will continue to be met
with the proposed changes. The proposed changes will not affect the
source term, containment isolation, or radiological release
assumptions used in evaluating the radiological consequences of any
accident previously evaluated. The proposed changes will not alter
any assumptions or change any mitigation actions in the radiological
consequence evaluations in the Updated Final Safety Analysis Report
(UFSAR).
The applicable radiological dose acceptance criteria will
continue to be met.
The proposed change revises the frequency of testing certain
portions of the PMS logic circuitry, but does not physically alter
any safety-related systems.
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
different kind of accident from any accident previously evaluated?
Response: No.
With respect to any new or different kind of accident, there are
no proposed design changes nor are there any changes in the method
by which any safety-related plant SSC performs its specified safety
function. The proposed change will not affect the normal method of
plant operation or change any operating parameters. No equipment
performance requirements will be affected. The proposed change will
not alter any assumptions made in the safety analyses.
The proposed change revises the frequency of testing certain
portions of the PMS logic circuitry. The proposed change does not
involve a physical modification of the plant.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures will be introduced as a
result of this amendment. There will be no adverse effect or
challenges imposed on any safety-related system as a result of this
amendment.
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 existing ACTUATION LOGIC TEST Surveillance Requirements are
revised such that different portions of the PMS logic circuitry are
tested on appropriate surveillance test frequencies. The reliability
of the PMS is such that not testing the Component Interface Module
(CIM) logic and driver output circuits when the reactor is at power
will have a net positive impact on Engineered Safety Feature
Actuation System (ESFAS) availability. There will be a reduction in
the potential for challenges to the safety systems, coupled with
less time that the safety systems are unavailable.
There will be no effect on those plant systems necessary to
effect the accomplishment of protection functions.
No instrument setpoints or system response times are affected.
None of the acceptance criteria for any accident analysis will be
changed.
The proposed change will have no impact on the radiological
consequences of a design basis accident.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
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.
[[Page 55411]]
NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia
Date of amendment request: August 18, 2017. A publicly-available
version is in ADAMS under Accession No. ML17230A365.
Description of amendment request: The requested amendment proposes
to depart from approved AP1000 Design Control Document (DCD) Tier 2
information (text) and involved Tier 2* information (as incorporated
into the Updated Final Safety Analysis Report (UFSAR) as plant-specific
DCD information).
This amendment request proposes increasing the design pressure of
the main steam (MS) isolation valve (MSIV) compartments from 6.0 to 6.5
psi and proposes other changes to the licensing basis regarding
descriptions of the MSIV compartments.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below with Nuclear Regulatory
Commission (NRC) staff's edits in square brackets:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes do not adversely affect the operation of
any structures, systems, and components inside or outside the
auxiliary building that could initiate or mitigate abnormal events,
e.g., accidents, anticipated operational occurrences, earthquakes,
floods, tornado missiles, and turbine missiles, or their safety or
design analyses, evaluated in the UFSAR. The changes do not
adversely affect any design function of the auxiliary building or
the structures, systems, and components contained therein. The
ability of the affected auxiliary building main steam isolation
valve compartments and adjacent rooms, including the main control
room, to withstand the pressurization effects from the postulated
pipe ruptures is not adversely affected by the increase in design
pressure, since the structures, systems, and components therein
remain qualified for this service.
Therefore, the proposed activity 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 might initiate a new or different kind of
accident, or alter any [structure, system, and component (SSC)] such
that a new accident initiator or initiating sequence of events is
created. The proposed changes do not adversely affect the physical
design and operation of the [in-containment refueling water storage
tank (IRWST)] injection, drain, containment recirculation, and
fourth-stage [automatic depressurization system (ADS)] valves,
including as-installed inspections, and maintenance requirements, as
described in the UFSAR. Therefore, the operation of the IRWST
injection, drain, containment recirculation, and fourth-stage ADS
valves is not adversely affected. These proposed changes do not
adversely affect any other SSC design functions or methods of
operation in a manner that results in a new failure mode,
malfunction, or sequence of events that affect safety-related or
nonsafety-related equipment. Therefore, this activity does not allow
for a new fission product release path, result in a new fission
product barrier failure mode, or create a new sequence of events
that result in significant fuel cladding failures.
Therefore, the 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 margin of safety for the design of the auxiliary building is
maintained through continued use of approved codes and standards as
stated in the UFSAR, and adherence to the assumptions used in the
analyses of this structure and the events associated with this
structure. The auxiliary building continues to be a seismic Category
I building with all current structural safety margins maintained.
The 3-hour fire rating requirements for the impacted auxiliary
building walls are maintained. The equipment housed in the main
steam isolation valve compartments continue to be environmentally
qualified for their intended service in accordance with the approved
codes and standards stated within the UFSAR. Thus, the requested
changes will not adversely affect any safety-related equipment,
design code, function, design analysis, safety analysis input or
result, or design/safety margin. No safety analysis or design basis
acceptance limit/criterion is challenged or exceeded by the
requested change, thus, no margin of safety is reduced. 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: 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, Units 3 and 4, Burke County, Georgia
Date of amendment request: October 6, 2017. A publicly-available
version is in ADAMS under Accession No. ML17279A084.
Description of amendment request: The amendment request proposes to
depart from Tier 2 information in the Updated Final Safety Analysis
Report (UFSAR) (which includes the plant-specific Design Control
Document (DCD) Tier 2 information) and involves related changes to
plant-specific Tier 1 information, with corresponding changes to the
associated combined license (COL) Appendix C information. 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. Specifically, the requested amendment proposes to
depart from Tier 2 information in UFSAR Subsection 8.3.2.4 describing
raceway and cable routing criteria and hazard protection, and involves
related changes to plant-specific Tier 1 Table 3.3-6, inspections,
tests, analyses, and acceptance criteria information, with
corresponding changes to the associated COL Appendix C information.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below with NRC staff edits in square
brackets:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Changes 1, 3 and 4 are clarifications only and do not represent
a change to the minimum required separation distance between
raceways. Change 2 reduces the required separation distances between
raceways from those documented in [Institute of Electrical and
Electronics Engineers (IEEE)] 384-1981. These reduced separation
distances are based on specific tests performed on the specified
raceway configurations, and the recommendations from those tests
contained in the associated report. The NRC staff previously
reviewed the descriptions of the ten tests documented in this
report, including the ones applicable to the existing UFSAR
exceptions, and concluded that they were acceptable, as documented
in NUREG-1793, ``Final Safety Evaluation Report Related to
Certification of
[[Page 55412]]
the AP1000 Standard Design,'' (Initial Report) Subsection 8.3.2.2.
The reduced separation does not adversely impact the ability to
safely shutdown the plant, and maintain it shutdown. The referenced
test report has shown a failure of a faulted cable will not
propagate to a nearby target cable in way that adversely impacts its
function.
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.
Changes 1, 3 and 4 are clarifications only and do not represent
a change to the minimum required separation distance between
circuits. Change 2 reduces the required separation distances between
circuits from those documented in IEEE 384-1981. This change does
not result in a new accident initiator or impact a current accident
initiator.
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.
Changes 1, 3 and 4 are clarifications only and do not represent
a change to the minimum required separation distance between
circuits. Change 2 reduces the required separation distances between
circuits from those documented in IEEE 384-1981. These reduced
separation distances are based on specific tests performed on the
specified raceway configurations, and the recommendations from those
tests contained in the associated report. The NRC staff previously
reviewed the descriptions of the ten tests documented in this
report, including the ones applicable to the existing UFSAR
exceptions, and concluded that they were acceptable, as documented
in NUREG-1793, ``Final Safety Evaluation Report Related to
Certification of the AP1000 Standard Design,'' (Initial Report)
Subsection 8.3.2.2.
The reduced separation does not adversely impact the ability to
safely shutdown the plant, and maintain it shutdown. The referenced
test report has shown a failure of a faulted cable will not
propagate to a nearby target cable in a way that adversely impacts
its function.
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., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia
Date of amendment request: July 10, 2017. A publicly-available
version is in ADAMS under Accession No. ML17191B163.
Description of amendment request: The amendments would revise the
technical specifications (TSs) by: (1) Adding a Note to the
surveillance requirements (SRs) of TS 3.7.7, ``Main Turbine Bypass
System,'' to clarify that the SRs are not required to be met when the
limiting condition for operation (LCO) does not require the Main
Turbine Bypass System to be operable, (2) clarifying that LCO 3.2.3,
``LINEAR HEAT GENERATION RATE (LHGR),'' also has limits for an
inoperable Main Turbine Bypass System that are made applicable as
specified in the Core Operating Limits Report, and (3) deleting an
outdated footnote for LCO 3.2.3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change (1) adds a Note to the Surveillance
Requirements (SRs) of the Hatch Nuclear Plant (HNP) Unit 1 and Unit
2 Technical Specifications (TS) 3.7.7 clarifying that the SRs are
not required to be met when the LCO does not require the Main
Turbine Bypass System to be Operable, (2) clarifies that LCO 3.2.3,
``LINEAR HEAT GENERATION RATE'' also has limits for an inoperable
Main Turbine Bypass System that are made applicable as specified in
the Core Operating Limits Report, and (3) deletes an outdated
footnote for LCO 3.2.3. The proposed change does not affect the
requirement to meet the LCO, nor does it affect the requirements to
perform the SRs when the Main Turbine Bypass System is being used to
meet the LCO. This change simply clarifies the existing allowance to
apply the Main Turbine Bypass System inoperable limits to minimum
critical power ratio (MCPR) and linear heat generation rate (LHGR)
in lieu of the requirement for the Main Turbine Bypass System to be
Operable. The current safety analysis evaluation is unaffected by
this proposed change. The change regarding the outdated footnote has
no effect on the actual TS requirements.
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 (1) adds a Note to the Surveillance
Requirements (SRs) of the Hatch Nuclear Plant (HNP) Unit 1 and Unit
2 Technical Specifications (TS) 3.7.7 clarifying that the SRs are
not required to be met when the LCO does not require the Main
Turbine Bypass System to be Operable, (2) clarifies that LCO 3.2.3,
``LINEAR HEAT GENERATION RATE'' also has limits for an inoperable
Main Turbine Bypass System that are made applicable as specified in
the Core Operating Limits Report, and (3) deletes an outdated
footnote for LCO 3.2.3. This change simply clarifies the existing
allowance to apply the Main Turbine Bypass System inoperable limits
to minimum critical power ratio (MCPR) and linear heat generation
rate (LHGR) in lieu of the requirement for the Main Turbine Bypass
System to be Operable. The change regarding the outdated footnote
has no effect on the actual TS requirements. The current safety
analysis evaluation is unaffected by these proposed changes.
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 (1) adds a Note to the Surveillance
Requirements (SRs) of the Hatch Nuclear Plant (HNP) Unit 1 and Unit
2 Technical Specifications (TS) 3.7.7 clarifying that the SRs are
not required to be met when the LCO does not require the Main
Turbine Bypass System to be Operable, (2) clarifies that LCO 3.2.3,
``LINEAR HEAT GENERATION RATE'' also has limits for an inoperable
Main Turbine Bypass System that are made applicable as specified in
the Core Operating Limits Report, and (3) deletes an outdated
footnote for LCO 3.2.3. This change simply clarifies the existing
allowance to apply the Main Turbine Bypass System inoperable limits
to minimum critical power ratio (MCPR) and linear heat generation
rate (LHGR) in lieu of the requirement for the Main Turbine Bypass
System to be Operable. The applicable safety analyses for TS 3.7.7
is unaffected by this clarification. The change regarding the
outdated footnote has no effect on the actual TS requirements.
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
[[Page 55413]]
amendment request involves no significant hazards consideration.
Attorney for licensee: Jennifer M. Buettner, Associate General
Counsel, Southern Nuclear Operating Company, 40 Inverness Center
Parkway, Birmingham, AL 35242.
NRC Branch Chief: Michael T. Markley.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia
Date of amendment request: September 13, 2017. A publicly-available
version is in ADAMS under Accession No. ML17256A626.
Description of amendment request: The requested amendment proposes
to depart from approved AP1000 Design Control Document (DCD) Tier 2
information as incorporated into the Updated Final Safety Analysis
Report (UFSAR) as plant-specific DCD information, and from Technical
Specifications as incorporated in Appendix A of the Combined License
(COL). Specifically, the proposed changes revise COL Appendix A
Technical Specification 3.6.8 to identify the trisodium phosphate (TSP)
mass value required in the pH adjustment baskets. The TSP mass value
adjusts the pH of the containment water to >7.0 following a postulated
accident.
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 activity revises the mass of trisodium phosphate
(TSP), which raises the pH of post-accident containment water to 7.0
or greater following a postulated accident. The change to the TSP
mass value does not adversely impact the ability to support
radionuclide retention with high radioactivity in containment and
helps prevent corrosion of containment equipment during long-term
floodup conditions. The proposed changes do not adversely impact
previously evaluated accidents, because pH control capability is
provided to mitigate already postulated accidents. As described in
Updated Final Safety Analysis Report (UFSAR) Subsection
15.6.5.3.1.3, the passive core cooling system (PXS) is assumed to
provide sufficient TSP to the post-loss-of-coolant accident (LOCA)
cooling solution to maintain the pH at greater than or equal to 7.0
following a LOCA. The pH adjustment baskets provide for long-term pH
control. Long-term pH control is not adversely impacted as the pH
adjustment baskets contain the required amount of TSP to support pH
control requirements following a design basis accident (DBA).
No safety-related structure, system, component (SSC) or function
is adversely affected by this change. The change does not involve an
interface with any SSC accident initiator or initiating sequence of
events, and thus, the probabilities of the accidents evaluated in
the UFSAR are not affected. The proposed changes do not involve a
change to the predicted radiological releases due to postulated
accident conditions, thus, 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 activity revises the mass of TSP, which raises the
pH of containment to 7.0 or greater following a postulated accident.
The proposed activity does not create the possibility of a new or
different kind of accident as pH adjustment is used to support
proper containment chemistry requirements following an accident. The
proposed activity does not adversely affect any safety related
equipment, and does not add any new interfaces to safety-related
SSCs that adversely affect safety functions. No system or design
function or equipment qualification is adversely affected by these
changes as the changes do not modify any SSCs that prevent safety
functions from being performed. The capability to maintain a maximum
containment pH below 9.5 is not adversely impacted by these changes.
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 activity revises the mass of TSP, which raises the
pH of containment to 7.0 or greater following a postulated accident.
The proposed activity does not affect any other safety-related
equipment or fission product barriers. Containment water pH
adjustment is not adversely impacted. The requested changes will not
adversely affect compliance with any design code, function, design
analysis, safety analysis input or result, or design/safety margin.
No safety analysis or design basis acceptance limit/criterion is
challenged or exceeded by the requested changes as previously
evaluated accidents are not impacted.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and based on
this review it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazard consideration.
Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP,
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County,
Georgia
Date of amendment request: September 29, 2017. A publicly-available
version is in ADAMS under Accession No. ML17272A957.
Description of amendment request: The requested amendment proposes
to depart from Tier 2* and associated Tier 2 information in the Updated
Final Safety Analysis Report (UFSAR) (which includes the plant-specific
DCD Tier 2 information). The requested amendment proposes to depart
from UFSAR Tier 2* information regarding resolution of human
engineering deficiencies (HEDs) contained in Westinghouse Electric
Company's report APP-OCS-GEH-320, ``AP1000 Human Factors Engineering
Integrated Systems Validation Plan,'' which is incorporated by
reference into the VEGP Units 3 and 4 UFSAR.
The proposed changes would revise the licensing basis of the
combined licenses regarding the process for addressing and re-testing
of HEDs identified during the integrated system validation (ISV) as
described in Tier 2* document, APPOCS-GEH-320 ``AP1000 Human Factors
Engineering Integrated System Validation Plan.'' APPOCS-GEH-320
references APP-OCS-GEH-420, ``Human Factors Engineering Discrepancy
Resolution Process,'' which defines the process for tracking,
resolution, and closure of HEDs. The proposed changes to APP-OCS-GEH-
320 do not impact APP-OCS-GEH-420.
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 Integrated System Validation (ISV) provides a comprehensive
human performance-based assessment of the design of the AP1000
Human-System Interface (HSI) resources, based on their realistic
operation
[[Page 55414]]
within a simulator driven Main Control Room (MCR). The ISV is part
of the overall AP1000 Human Factors Engineering (HFE) program. The
changes to APP-OCS-GEH-320, which is incorporated by reference into
the UFSAR, clarify the resources and methodology used during re-
testing performed to verify the effectiveness of Human Engineering
Deficiency (HED) resolution. The ISV Plan does not affect the plant
itself. Changing APP-OCS-GEH-320 and the UFSAR does not affect
prevention and mitigation of abnormal events, e.g., accidents,
anticipated operational occurrences, earthquakes, floods and turbine
missiles, or their safety or design analyses. No safety-related
structure, system, component (SSC) or function is adversely
affected. The changes neither involve nor interface with any SSC
accident initiator or initiating sequence of events, and thus, the
probabilities of the accidents evaluated in the UFSAR are not
affected. Because the changes do not involve any safety-related SSC
or function used to mitigate an accident, 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 changes to APP-OCS-GEH-320 and the VEGP 3 and 4 UFSAR affect
only the testing and validation of the MCR design and HSI using a
plant simulator. Therefore, the changes do not affect the safety-
related equipment itself, nor do they affect equipment which, if it
failed, could initiate an accident or a failure of a fission product
barrier. No analysis is adversely affected. No system or design
function or equipment qualification is adversely affected by the
changes. This activity does not allow for a new fission product
release path, result in a new fission product barrier failure mode,
or create a new sequence of events that would result in significant
fuel cladding failures. In addition, the changes do not result in a
new failure mode, malfunction or sequence of events that could
affect safety or safety related equipment.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The changes to APP-OCS-GEH-320 and the UFSAR affect the testing
and validation of the MCR design and HSI using a plant simulator.
Therefore, the changes do not affect the assessments or the plant
itself. These changes do not affect safety-related equipment or
equipment whose failure could initiate an accident, nor does it
adversely interface with safety-related equipment or fission product
barriers. No safety analysis or design basis acceptance limit/
criterion is challenged or exceeded by the requested change.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and based on
this review it appears that the three standards of 10 CFR 50.92 (c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazard consideration.
Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP,
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer Dixon-Herrity.
Susquehanna Nuclear, LLC, Docket Nos. 50-387 and 50-388, Susquehanna
Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of amendment request: September 20, 2017. A publicly-available
version is in ADAMS under Package Accession No. ML17265A434.
Description of amendment request: The amendments would revise
technical specification (TS) requirements related to ``operations with
a potential for draining the reactor vessel'' (OPDRVs) with new
requirements on reactor pressure vessel (RPV) water inventory control
(WIC) to protect Safety Limit 2.1.1.3. Safety Limit 2.1.1.3 requires
RPV water level to be greater than the top of active irradiated fuel.
The proposed changes are based on Technical Specifications Task Force
(TSTF) Traveler TSTF-542, Revision 2, ``Reactor Pressure Vessel Water
Inventory Control,'' dated December 20, 2016.
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 replace existing TS requirements related to
OPDRVs with new requirements on RPV WIC that will protect Safety
Limit 2.1.1.3. Draining of RPV water inventory in Mode 4 (i.e., cold
shutdown) and Mode 5 (i.e., refueling) is not an accident previously
evaluated and, therefore, replacing the existing TS controls to
prevent or mitigate such an event with a new set of controls has no
effect on any accident previously evaluated. RPV water inventory
control in Mode 4 or Mode 5 is not an initiator of any accident
previously evaluated. The existing OPDRV controls or the proposed
RPV WIC controls are not mitigating actions assumed in any accident
previously evaluated.
The proposed changes reduce the probability of an unexpected
draining event (which is not a previously evaluated accident) by
imposing new requirements on the limiting time in which an
unexpected draining event could result in the reactor vessel water
level dropping to the top of the active fuel (TAF). These controls
require cognizance of the plant configuration and control of
configurations with unacceptably short drain times. These
requirements reduce the probability of an unexpected draining event.
The current TS requirements are only mitigating actions and impose
no requirements that reduce the probability of an unexpected
draining event.
The proposed changes reduce the consequences of an unexpected
draining event (which is not a previously evaluated accident) by
requiring an Emergency Core Cooling System (ECCS) subsystem to be
operable at all times in Modes 4 and 5. The current TS requirements
do not require any water injection systems, ECCS or otherwise, to be
Operable in certain conditions in Mode 5. The change in requirement
from two ECCS subsystems to one ECCS subsystem in Modes 4 and 5 does
not significantly affect the consequences of an unexpected draining
event because the proposed Actions ensure equipment is available
within the limiting drain time that is as capable of mitigating the
event as the current requirements. The proposed controls provide
escalating compensatory measures to be established as calculated
drain times decrease, such as verification of a second method of
water injection and additional confirmations that containment and/or
filtration would be available if needed.
The proposed changes reduce or eliminate some requirements that
were determined to be unnecessary to manage the consequences of an
unexpected draining event, such as automatic initiation of an ECCS
subsystem and the Control Room Emergency Outside Air Supply (CREOAS)
system. These changes do not affect the consequences of any accident
previously evaluated since a draining event in Modes 4 and 5 is not
a previously evaluated accident and the requirements are not needed
to adequately respond to a draining event.
The administrative update to delete expired completion time
notes is purely administrative in nature.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes replace existing TS requirements related to
OPDRVs with new requirements on RPV WIC that will protect Safety
Limit 2.1.1.3. The proposed changes will not alter the design
function of the equipment involved. Under the proposed changes, some
systems that are currently required to be operable during OPDRVs
would be required to be available within the
[[Page 55415]]
limiting drain time or to be in service depending on the limiting
drain time. Should those systems be unable to be placed into
service, the consequences are no different than if those systems
were unable to perform their function under the current TS
requirements. The event of concern under the current requirements
and the proposed changes are an unexpected draining event. The
proposed changes do not create new failure mechanisms, malfunctions,
or accident initiators that would cause a draining event or a new or
different kind of accident not previously evaluated or included in
the design and licensing bases.
The administrative update to delete expired completion time
notes is purely administrative in nature.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes replace existing TS requirements related to
OPDRVs with new requirements on RPV WIC. The current requirements do
not have a stated safety basis and no margin of safety is
established in the licensing basis. The safety basis for the new
requirements is to protect Safety Limit 2.1.1.3. New requirements
are added to determine the limiting time in which the RPV water
inventory could drain to the top of the fuel in the reactor vessel
should an unexpected draining event occur. Plant configurations that
could result in lowering the RPV water level to the TAF within one
hour are now prohibited. New escalating compensatory measures based
on the limiting drain time replace the current controls. The
proposed TS establish a safety margin by providing defense-in-depth
to ensure that the Safety Limit is protected and to protect the
public health and safety. While some less restrictive requirements
are proposed for plant configurations with long calculated drain
times, the overall effect of the change is to improve plant safety
and to add safety margin.
The administrative update to delete expired completion time
notes is purely administrative in nature.
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: Damon D. Obie, Associate General Counsel,
Talen Energy Supply, LLC, 835 Hamilton St., Suite 150, Allentown, PA
18101.
NRC Branch Chief: James G. Danna.
Tennessee Valley Authority (TVA), Docket Nos. 50-259, 50-260, and 50-
296, Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, Limestone
County, Alabama
Date of amendment request: August 15, 2017. A publicly-available
version is in ADAMS under Accession No. ML17228A490.
Description of amendment request: The amendments would revise the
BFN, Units 1, 2, and 3 Technical Specification (TS) 5.5.12, ``Primary
Containment Leakage Rate Testing Program,'' by adopting Nuclear Energy
Institute (NEI) 94-01, Revision 3-A, ``Industry Guideline for
Implementing Performance-Based Option of 10 CFR part 50, Appendix J,''
as the implementation document for the performance-based Option B of 10
CFR part 50, Appendix J. The proposed changes permanently extend the
Type A containment integrated leak rate testing (ILRT) interval from 10
years to 15 years and the Type C local leakage rate testing (LLRT)
intervals from 60 months to 75 months.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequence of an accident previously evaluated?
Response: No.
The proposed revision to TS 5.5.12 changes the testing period to
a permanent 15-year interval for Type A testing (10 CFR part 50,
Appendix J, Option B, ILRT) and a 75-month interval for Type C
testing (10 CFR part 50, Appendix J, Option B, LLRT). The current
Type A test interval of 10 years would be extended to 15 years from
the last Type A test. The proposed extension to Type A testing does
not involve a significant increase in the consequences of an
accident because research documented in NUREG-1493, ``Performance-
Based Containment System Leakage Testing Requirements''
[``Performance-Based Containment Leak-Test Program''], September
1995, has found that, generically, very few potential containment
leakage paths are not identified by Type B and C tests. NUREG-1493
concluded that reducing the Type A testing frequency to one per 20
years was found to lead to an imperceptible increase in risk. A high
degree of assurance is provided through testing and inspection that
the containment will not degrade in a manner detectable only by Type
A testing. The last Type A test (performed November 19, 2010 for
BFN, Unit 1, June 3, 2009 for BFN, Unit 2 and May 12, 2012 for BFN,
Unit 3) shows leakage to be below acceptance criteria, indicating a
very leak tight containment. Inspections required by the ASME Code
[American Society of Mechanical Engineers Boiler and Press Vessel
Code] Section Xl (Subsection IWE) and Maintenance Rule monitoring
(10 CFR 50.65, ``Requirements for Monitoring the Effectiveness of
Maintenance at Nuclear Power Plants'') are performed in order to
identify indications of containment degradation that could affect
that leak tightness. Types B and C testing required by TSs will
identify any containment opening such as valves that would otherwise
be detected by the Type A tests. These factors show that a Type A
test interval extension will not represent a significant increase in
the consequences of an accident.
The proposed amendment involves changes to the BFN, Units 1, 2,
and 3, 10 CFR 50 Appendix J Testing Program Plan. The proposed
amendment does not involve a physical change to the plant or a
change in the manner in which the units are operated or controlled.
The primary containment function is to provide an essentially leak
tight barrier against the uncontrolled release of radioactivity to
the environment for postulated accidents. As such, the containment
itself and the testing requirements to periodically demonstrate the
integrity of the containment exist to ensure the plant's ability to
mitigate the consequences of an accident, and do not involve any
accident precursors or initiators. Therefore, the probability of
occurrence of an accident previously evaluated is not significantly
increased by the proposed amendment.
The proposed amendment adopts the NRC-accepted guidelines of NEI
94-01, Revision 3-A, for development of the BFN, Units 1, 2, and 3,
performance-based leakage testing program. Implementation of these
guidelines continues to provide adequate assurance that during
design basis accidents, the primary containment and its components
will limit leakage rates to less than the values assumed in the
plant safety analyses. The potential consequences of extending the
ILRT interval from 10 years to 15 years have been evaluated by
analyzing the resulting changes in risk. The increase in risk in
terms of person-rem [roentgen equivalent man] per year resulting
from design basis accidents was estimated to be very small, and the
increase in the LERF [large early release frequency] resulting from
the proposed change was determined to be within the guidelines
published in NRC RG [Regulatory Guide] 1.174. Additionally, the
proposed change maintains defense-in-depth by preserving a
reasonable balance among prevention of core damage, prevention of
containment failure, and consequence mitigation. TVA has determined
that the increase in CCFP [conditional containment failure
probability] due to the proposed change would be very small.
Based on the above discussions, the proposed changes do not
involve an 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.
[[Page 55416]]
The proposed revision to TS 5.5.12 changes the testing period to
a permanent 15-year interval for Type A testing (10 CFR part 50,
Appendix J, Option B, ILRT) and a 75-month interval for Type C
testing (10 CFR part 50, Appendix J, Option B, LLRT). The current
test interval of 10 years, based on past performance, would be
extended to 15 years from the last Type A test (performed November
19, 2010 for BFN, Unit 1, June 3, 2009 for BFN, Unit 2 and May 12,
2012 for BFN, Unit 3). The proposed extension to Type A and Type C
test intervals does not create the possibility of a new or different
type of accident because there are no physical changes being made to
the plant and there are no changes to the operation of the plant
that could introduce a new failure mode creating an accident or
affecting the mitigation of an accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed revision to TS 5.5.12 changes the testing period to
a permanent 15-year interval for Type A testing (10 CFR part 50,
Appendix J, Option B, ILRT) and a 75-month interval for Type C
testing (10 CFR part 50, Appendix J, Option B, LLRT). The current
test interval of 10 years, based on past performance, would be
extended to 15 years from the last Type A test (performed November
19, 2010 for BFN, Unit 1, June 3, 2009 for BFN, Unit 2 and May 12,
2012 for BFN, Unit 3). The proposed extension to Type A testing will
not significantly reduce the margin of safety. NUREG-1493,
``Performance-Based Containment System Leakage Testing
Requirements'' [``Performance-Based Containment Leak-Test
Program''], September 1995, generic study of the effects of
extending containment leakage testing, found that a 20 year
extension to Type A leakage testing resulted in an imperceptible
increase in risk to the public. NUREG-1493 found that, generically,
the design containment leakage rate contributes about 0.1% to the
individual risk and that the decrease in Type A testing frequency
would have a minimal effect on this risk since 95% of the potential
leakage paths are detected by Type C testing. Regular inspections
required by the ASME Code Section Xl (Subsection IWE) and
maintenance rule monitoring (10 CFR 50.65, ``Requirements for
Monitoring the Effectiveness of Maintenance at Nuclear Power
Plants'') will further reduce the risk of a containment leakage path
going undetected.
The proposed amendment adopts the NRC-accepted guidelines of NEI
94-01, Revision 3-A, for development of the BFN, Units 1, 2, and 3,
performance-based leakage testing program, and establishes a 15-year
interval for the performance of the primary containment ILRT and a
75-month interval for Type C testing. The amendment does not alter
the manner in which safety limits, limiting safety system setpoints,
or limiting conditions for operation are determined. The specific
requirements and conditions of the 10 CFR part 50, Appendix J
Testing Program Plan, as defined in the TS, ensure that the degree
of primary containment structural integrity and leak-tightness that
is considered in the plant safety analyses is maintained. The
overall containment leakage rate limit specified by the TS is
maintained, and the Type A, B, and C containment leakage tests will
continue to be performed at the frequencies established in
accordance with the NRC-accepted guidelines of NEI 94-01, Revision
3-A.
Containment inspections performed in accordance with other plant
programs serve to provide a high degree of assurance that the
containment will not degrade in a manner that is detectable only by
an ILRT. This ensures that evidence of containment structural
degradation is identified in a timely manner. Furthermore, a risk
assessment using the current BFN, Units 1, 2, and 3, PRA
[probabilistic risk assessment] model concluded that extending the
ILRT test interval from 10 years to 15 years results in a very small
change to the BFN, Units 1, 2, and 3, risk profile.
Accordingly, 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Dr., WT 6A, Knoxville, TN 37902.
NRC Branch Chief: Undine Shoop.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2 (SQN), Hamilton County, Tennessee
Tennessee Valley Authority, Docket Nos. 50-390 and 50-391, Watts Bar
Nuclear Plant, Units 1 and 2 (WBN), Rhea County, Tennessee
Date of amendment request: August 7, 2017. A publicly-available
version is in ADAMS under Accession No. ML17219A505.
Description of amendment request: The amendment would revise
Technical Specification (TS) 3.2.4, ``Quadrant Power Tilt Ratio
(QPTR),'' and TS 3.3.1, ``Reactor Trip System (RTS) Instrumentation,''
to avoid confusion as to when an incore power distribution measurement
for QPTR is required. The amendment would also revise the WBN TSs for
consistency with the existing SQN TSs and Westinghouse Standard TSs in
NUREG-1431, Revision 4.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, or
configuration of the facility or the manner in which the plant is
operated and maintained. The proposed changes do not alter or
prevent the ability of structures, systems, and components (SSCs)
from performing their intended function to mitigate the consequences
of an initiating event within the assumed acceptance limits. The
proposed changes do not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of an accident previously evaluated.
Further, the proposed changes do not increase the types or amounts
of radioactive effluent that may be released offsite, nor
significantly increase individual or cumulative occupational/public
radiation exposures. The proposed changes do not significantly
increase the probability of an accident and are consistent with
safety analysis assumptions and resultant consequences.
Therefore, the changes do not increase 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 result in a change in the manner in
which the reactor trip system (RTS) and engineered safety feature
actuation system (ESFAS) provide plant protection. The RTS and ESFAS
will continue to have the same setpoints after the proposed changes
are implemented. There are no design changes associated with the
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 changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes do not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The safety analysis acceptance criteria
are not impacted by these changes. Redundant RTS and ESFAS trains
are maintained, and diversity with regard to the signals that
provide reactor trip and engineered safety features actuation is
also
[[Page 55417]]
maintained. All signals credited as providing primary or secondary
protection, and all operator actions credited in the accident
analyses will remain the same. The proposed changes will not result
in plant operation in a configuration outside the design basis.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
NRC Branch Chief: Undine Shoop.
III. Notice of Issuance of Amendments to Facility Operating Licenses
and Combined Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, proposed no
significant hazards consideration determination, and opportunity for a
hearing in connection with these actions, was published in the Federal
Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items can be accessed as described in the
``Obtaining Information and Submitting Comments'' section of this
document.
Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment requests: December 15, 2016.
Brief description of amendments: The amendments modified Technical
Specification (TS) 3.4.10, ``Pressurizer Safety Valves,'' TS 3.7.4,
``Steam Generator Power Operated Relief Valves (SG PORVs),'' and TS
3.7.6, ``Condensate Storage System,'' to revise the Completion Times
for Limiting Condition for Operation (LCO) of TS LCO 3.4.10 Required
Action B.2, TS LCO 3.7.4 Required Action C.2, and TS LCO 3.7.6 Required
Action B.2 from 12 to 24 hours. The proposed changes are consistent
with Technical Specifications Task Force (TSTF) Traveler TSTF-352-A,
Revision 1, ``Provide Consistent Completion Time to Reach MODE 4.''
Date of issuance: October 23, 2017.
Effective date: These license amendments are effective as of its
date of issuance and shall be implemented within 120 days of issuance.
Amendment Nos.: 294 (Unit 1) and 290 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML17254A144; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the renewed licenses and technical specifications.
Date of initial notice in Federal Register: April 25, 2017 (82 FR
19099).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 23, 2017.
No significant hazards consideration comments received: No.
Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment requests: December 15, 2016.
Brief description of amendments: The amendments modified technical
specification (TS) limiting condition for operation (LCO) 3.7.5,
``Auxiliary Feedwater (AFW) System,'' Condition A and Required Action
A.1. Condition A was revised to include the situation when one turbine-
driven AFW pump is inoperable in MODE 3, immediately following a
refueling outage, only applicable if MODE 2 has not been entered
following the refueling outage. Required Action A.1 was revised to
include the turbine-driven AFW addition to Condition A. The amendments
are consistent with Technical Specifications Task Force (TSTF) Traveler
TSTF-340-A, Revision 3, ``Allow 7 day Completion Time for a turbine-
driven AFW pump inoperable.''
Date of issuance: October 23, 2017.
Effective date: These license amendments are effective as of its
date of issuance and shall be implemented within 120 days of issuance.
Amendment Nos.: 295 (Unit 1) and 291 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML17257A297; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the renewed licenses and TSs.
Date of initial notice in Federal Register: April 25, 2017 (82 FR
19100).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 23, 2017.
No significant hazards consideration comments received: No.
Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment requests: December 15, 2016.
Brief description of amendments: The amendments revised Technical
Specification 3.1.2, ``Core Reactivity,'' to revise the Completion
Times of Required Actions A.1 and A.2 from 72 hours to 7 days. This
proposed change is consistent with Technical Specifications Task Force
(TSTF) Traveler TSTF-142-A, Revision 0, ``Increase the Completion Time
when the Core Reactivity Balance is Not Within Limit.''
Date of issuance: October 23, 2017.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment Nos.: 296 (Unit 1) and 292 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML17261B290; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the Renewed Licenses and Technical Specifications.
[[Page 55418]]
Date of initial notice in Federal Register: April 11, 2017 (82 FR
17457).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 23, 2017.
No significant hazards consideration comments received: No.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment requests: January 11, 2017.
Brief description of amendments: The amendments modified Technical
Specification (TS) 3.8.1, ``AC Sources--Operating,'' to allow greater
flexibility in performing Surveillance Requirements (SRs) by modifying
Mode restriction notes in TS SRs 3.8.1.8, 3.8.1.11, 3.8.1.16, 3.8.1.17,
and 3.8.1.19. This proposed change was consistent with Technical
Specifications Task Force (TSTF) Traveler TSTF-283-A, Revision 3,
``Modify Section 3.8 Mode Restriction Notes.''
Date of issuance: October 25, 2017.
Effective date: These license amendments are effective as of its
date of issuance and shall be implemented within 120 days of issuance.
Amendment Nos.: 300 (Unit 1) and 279 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML17269A055; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the renewed facility operating licenses and
technical specifications.
Date of initial notice in Federal Register: May 23, 2017 (82 FR
23620).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 25, 2017.
No significant hazards consideration comments received: Yes. One
comment from a member of the public was received, however it was not
related to the no significant hazards consideration determination nor
the license amendment request.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment requests: January 11, 2017.
Brief description of amendments: The amendments modified Technical
Specification (TS) 3.1.8, ``PHYSICS TESTS Exceptions,'' to allow the
numbers of channels required by the Limiting Condition for Operation
(LCO) section of TS 3.3.1, ``Reactor Trip System (RTS)
Instrumentation,'' to be reduced from ``4'' to ``3'' to allow one
nuclear instrumentation channel to be used as an input to the
reactivity computer for physics testing without placing the nuclear
instrumentation channel in a tripped condition. This proposed change is
consistent with Technical Specifications Task Force (TSTF) Traveler
TSTF-315-A, Revision 0, ``Reduce plant trips due to spurious signals to
the NIS [Nuclear Instrumentation System] during physics testing.''
Date of issuance: October 25, 2017.
Effective date: These license amendments are effective as of their
date of issuance and shall be implemented within 120 days of issuance.
Amendment Nos.: 301 (Unit 1) and 280 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML17261B218; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the renewed facility operating licenses and
technical specifications.
Date of initial notice in Federal Register: May 23, 2017 (82 FR
23621).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 25, 2017.
No significant hazards consideration comments received: Yes. One
comment from a member of the public was received, however it was not
related to the proposed no significant hazards consideration
determination or to the license amendment request.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment requests: January 11, 2017.
Brief description of amendments: The amendments modify the limiting
condition for operation (LCO) Required Action B.2 for Technical
Specification (TS) 3.4.10, ``Pressurizer Safety Valves,'' LCO Required
Action C.2 for TS 3.7.4, ``Steam Generator Power Operated Relief Valves
(SG PORVs),'' and LCO Required Action G.1 for TS 3.4.12, ``Low
Temperature Overpressure Protection (LTOP) System.'' Specifically, the
Completion Times are revised from 12 hours to 24 hours for TS LCO
3.4.10, Required Action B.2, and TS LCO 3.7.4, Required Action C.2; and
from 8 hours to 12 hours for TS LCO 3.4.12, Required Action G.1. The
changes are consistent with Technical Specifications Task Force (TSTF)
Traveler TSTF-352-A, Revision 1, ``Provide Consistent Completion Time
to Reach MODE 4.''
Date of issuance: October 31, 2017.
Effective date: These license amendments are effective as of their
date of issuance and shall be implemented within 120 days of issuance.
Amendment Nos.: 302 (Unit 1) and 281 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML17269A198; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the Renewed Licenses and Technical Specifications.
Date of initial notice in Federal Register: May 23, 2017 (82 FR
23622).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 31, 2017.
No significant hazards consideration comments received: Yes. One
comment from a member of the public was received, however it was not
related to the proposed no significant hazards consideration
determination or to the license amendment request.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment requests: January 11, 2017.
Brief description of amendments: The amendments modify Technical
Specification (TS) 3.7.5, ``Auxiliary Feedwater (AFW) System,''
Limiting Condition for Operation (LCO) Condition A and Required Action
A.1. The proposed changes modify Condition A to expand the condition to
include when one turbine driven AFW pump is inoperable in MODE 3. This
expanded condition is applicable immediately following a refueling
outage and only if MODE 2 has not been entered. Required Action A.1 is
revised to state ``affected equipment'' as opposed to ``steam supply''
as a result of the addition of the turbine driven AFW pump to Condition
A. The changes are consistent with Technical Specifications Task Force
(TSTF) Traveler TSTF-340-A, Revision 3, ``Allow 7 day Completion Time
for a turbine-driven AFW pump inoperable.''
Date of issuance: October 31, 2017.
Effective date: These license amendments are effective as of their
date of issuance and shall be
[[Page 55419]]
implemented within 120 days of issuance.
Amendment Nos.: 304 (Unit 1) and 283 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML17277A313; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the renewed facility operating licenses and
technical specifications.
Date of initial notice in Federal Register: May 23, 2017 (82 FR
23621).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 31, 2017.
No significant hazards consideration comments received: Yes. One
comment from a member of the public was received, however it was not
related to the proposed no significant hazards consideration
determination or to the license amendment request.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: January 11, 2017.
Brief description of amendments: The amendments modify Technical
Specification (TS) Limiting Condition for Operation (LCO) 3.9.6,
``Residual Heat Removal (RHR) and Coolant Circulation--Low Water
Level,'' to add a note which allows all RHR pumps to be secured for
less than or equal to 15 minutes to support the switching of the
shutdown cooling loops from one train to another. The changes are
consistent with Technical Specifications Task Force (TSTF) Travelers
TSTF-349-A, Revision 1, ``Add Note to LCO 3.9.5 Allowing Shutdown
Cooling Loops Removal from Operation,'' TSTF-361-A, Revision 2, ``Allow
standby [Shutdown Cooling] SDC/RHR/[Decay Heat Removal] DHR loop to
[be] inoperable to support testing,'' and TSTF-438-A, Revision 0,
``Clarify Exception Notes to be Consistent with the Requirement Being
Excepted.''
Date of issuance: October 31, 2017.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment Nos.: Unit 1--303; Unit 2--282. A publicly-available
version is in ADAMS under Accession No. ML17271A034; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the Renewed Licenses and Technical Specifications.
Date of initial notice in Federal Register: May 23, 2017 (82 FR
23623).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 31, 2017.
No significant hazards consideration comments received: Yes. One
comment from a member of the public was received, however it was not
related to the proposed no significant hazards consideration
determination or the license amendment request.
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: December 2, 2016, as supplemented by
letters dated April 25, May 22, and October 2, 2017.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) to (1) relocate cycle-specific parameters to the
Core Operating Limits Report (COLR) consistent with Technical
Specification Task Force (TSTF)-339, ``Relocate TS Parameters to
COLR;'' (2) delete duplicate reporting requirements in the
Administrative Section of TSs consistent with TSTF-5, ``Delete Safety
Limit Violation Notification Requirements,'' Revision 1; and (3) delete
reference to plant procedure PLP-6, ``Technical Specification Equipment
List Program and Core Operating Limits Report,'' in TSs as it pertains
to the COLR.
Date of issuance: November 6, 2017.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment No.: 161. A publicly-available version is in ADAMS under
Accession No. ML17250A202; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-63: Amendment revised
the Facility Operating License and TSs.
Date of initial notice in Federal Register: February 14, 2017 (82
FR 10595). The supplemental letters dated April 25, May 22, and October
2, 2017, provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the staff's original proposed no
significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 6, 2017.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: November 8, 2016, as supplemented by
letter dated July 11, 2017.
Brief description of amendment: The amendment would, on a one-time
basis, extend the completion time from 7 days to 14 days for the
Residual Heat Removal Train A subsystem to operable status associated
with Technical Specification (TS) 3.5.1, ``ECCS [Emergency Core Cooling
System]--Operating''; TS 3.6.1.5, ``Residual Heat Removal (RHR) Drywell
Spray''; and TS 3.6.2.3, ``Residual Heat Removal (RHR) Suppression Pool
Cooling.'' This amendment will be used to support preventive
maintenance, which replaces the RHR Train A subsystem's pump and motor.
Date of issuance: October 30, 2017.
Effective date: As of its date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 245. A publicly-available version is in ADAMS under
Accession No. ML17290A127; 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 Technical
Specifications.
Date of initial notice in Federal Register: February 14, 2017 (82
FR 10596). The supplemental letter dated July 11, 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 October 30, 2017.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket No. 50-277,
Peach Bottom Atomic Power Station, Unit 2, York and Lancaster Counties,
Pennsylvania
Date of amendment request: May 19, 2017, as supplemented by letter
dated August 29, 2017.
[[Page 55420]]
Brief description of amendment: The amendment revised the Technical
Specifications to decrease the number of safety relief valves and
safety valves required to be operable when operating at a power level
less than or equal to 3,358 megawatts thermal. This change is
applicable only to the current Cycle 22 that is scheduled to end in
October 2018.
Date of issuance: October 25, 2017.
Effective date: As of the date of issuance and shall be implemented
within 5 days.
Amendment No.: 315. A publicly-available version is in ADAMS under
Accession No. ML17249A151; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-44: The amendment
revised the Renewed Facility Operating License and Technical
Specifications.
Date of initial notice in Federal Register: July 5, 2017 (82 FR
31094). The supplemental letter dated August 29, 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 October 25, 2017.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station (LSCS), Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: October 27, 2016, as
supplemented by the letters dated July 28, 2017, August 30, 2017, and
October 19, 2017.
Brief description of amendments: The amendments revised the
suppression pool swell design analysis. The new analysis utilizes a
different computer code and incorporates different analysis assumptions
than the current analysis. The changes are necessary because the
current design analysis determining the suppression pool swell response
to a loss-of-coolant accident was determined to be non-conservative.
These changes to the suppression pool swell design analysis do not
require any changes to the LSCS Technical Specifications. Changes to
the LSCS updated final safety analysis report related to changes to the
suppression pool swell design analysis shall be made in accordance with
10 CFR 50.71(e) based on the NRC approval of these changes.
Date of issuance: October 30, 2017.
Effective date: These license amendments are effective as of the
date of its issuance and shall be implemented within 60 days from the
date of issuance.
Amendment Nos.: 225 for NPF-11 and 211 for NPF-18. A publicly-
available version is in ADAMS under Accession No. ML17257A304;
documents related to this amendment are listed in the Safety Evaluation
enclosed with the amendment.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
approved to revise the LSCS updated final safety analysis report
related to changes to the suppression pool swell design analysis and
the Licenses.
Date of initial notice in Federal Register: March 8, 2017 (82 FR
13022). The supplements dated July 28, 2017, August 30, 2017, and
October 19, 2017, contained clarifying information and did not change
the NRC staff's initial proposed finding of no significant hazards
consideration.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 30, 2017.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point
Nuclear Station (Nine Mile Point), Unit 2, Oswego County, New York
Date of amendment request: December 13, 2016, as supplemented by
letter dated February 17, 2017.
Brief description of amendment: The amendment revised the Nine Mile
Point, Unit 2, Technical Specification (TS) safety limit (SL) to
increase the low pressure isolation setpoint allowable value, which
will result in earlier main steam line isolation. The revised main
steam line low pressure isolation capability and the revised SL are
intended to ensure that Nine Mile Point, Unit 2, remains within the TS
SLs in the event of a pressure regulator failure maximum demand
transient.
Date of issuance: October 31, 2017.
Effective date: As of the date of issuance and shall be implemented
within 180 days of issuance.
Amendment No.: 164. A publicly-available version is in ADAMS under
Accession No. ML17268A263; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-69: Amendment revised
the Renewed Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: March 28, 2017 (82 FR
15381). The supplemental letter dated February 17, 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 October 31, 2017.
No significant hazards consideration comments received: No.
Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: January 23, 2017, as supplemented by
letter dated July 3, 2017.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) by limiting the MODE of applicability
for the Reactor Protection System, Startup, and Operating Rate of
Change of Power--High, functional unit trip. Additionally, the
amendments added new Limiting Condition for Operation (LCO) 3.0.5 and
relatedly modified LCO 3.0.1 and LCO 3.0.2, to provide for placing
inoperable equipment under administrative control for the purpose of
conducting testing required to demonstrate OPERABILITY.
Date of issuance: November 2, 2017.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: 243 and 194. A publicly-available version is in
ADAMS under Accession No. ML17257A015; documents related to this
amendment are listed in the Safety Evaluation enclosed with the
amendment.
Renewed Facility Operating License Nos. DPR-67 and NPF-16:
Amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: March 28, 2017 (82 FR
15383). The supplemental letter dated July 3, 2017, provided additional
information that expanded the scope of the application as originally
noticed and changed the NRC staff's original proposed no significant
hazards consideration (NSHC) determination as published in the Federal
Register. Accordingly, the NRC published a second proposed no
significant hazards consideration determination in the Federal Register
on September 12, 2017
[[Page 55421]]
(82 FR 42849). This notice superseded the original notice in its
entirety. It also provided an opportunity to request a hearing by
November 13, 2017, but indicated that if the Commission makes a final
NSHC determination, any such hearing would take place after issuance of
the amendments.
The Commission's related evaluation of the amendments and final
NSHC are contained in a Safety Evaluation dated November 2, 2017.
No significant hazards consideration comments received: No.
Florida Power & Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida
Date of amendment request: December 21, 2016.
Brief description of amendments: The amendments modify the
Technical Specifications by deleting high-range noble gas effluent
monitors' requirements and relocating the requirements to the Turkey
Point Offsite Dose Calculation Manual.
Date of issuance: October 26, 2017.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos: 277 and 272. A publicly-available version is in
ADAMS under Accession No. ML17228A563. Documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Renewed Facility Operating License Nos. DPR-31 and DPR-41:
Amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: March 14, 2017 (82 FR
13666).
The Commission's related evaluation of the amendments is contained
in a safety evaluation dated October 26, 2017.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Docket Nos. 50-348 and 50-364,
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama
Date of amendment request: August 11, 2017.
Brief description of amendments: The amendments request an
extension to the time to achieve full compliance with 10 CFR 50.48(c),
National Fire Protection Association (NPFA) 805, from November 6, 2017,
to the conclusion of the FNP, Unit 1, Spring 2018 Refueling Outage
(1R28). The amendments update Attachment S, ``Modification and
Implementation Items''; of the previously approved NFPA-805 amendment.
Date of issuance: November 1, 2017.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: 215 (Unit 1) and 212 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML17269A166; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. NPF-2 and NPF-8: The
amendments revised the Renewed Facility Operating Licenses.
Date of initial notice in Federal Register: August 29, 2017 (82 FR
41059).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 1, 2017.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 14th day of November 2017.
For the Nuclear Regulatory Commission.
Kathryn M. Brock,
Acting Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2017-25063 Filed 11-20-17; 8:45 am]
BILLING CODE 7590-01-P