[Federal Register Volume 81, Number 187 (Tuesday, September 27, 2016)]
[Notices]
[Pages 66301-66314]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-23097]
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NUCLEAR REGULATORY COMMISSION
[NRC-2016-0202]
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.
[[Page 66302]]
This biweekly notice includes all notices of amendments issued, or
proposed to be issued, from August 30, 2016, to September 12, 2016. The
last biweekly notice was published on September 13, 2016.
DATES: Comments must be filed by October 27, 2016. A request for a
hearing must be filed by November 28, 2016.
ADDRESSES: You may submit comments by any of the following methods
(unless this document describes a different method for submitting
comments on a specific subject):
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0202. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions, contact
the individual listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
Mail comments to: Cindy Bladey, Office of Administration,
Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001.
For additional direction on obtaining information and submitting
comments, see ``Obtaining Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Janet Burkhardt, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; telephone: 301-415-1384, email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to NRC-2016-0202, facility name, unit number(s), plant
docket number, application date, and subject when contacting the NRC
about the availability of information for this action. You may obtain
publicly-available information related to this action by any of the
following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0202.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, 301-415-4737, or by email to [email protected]. The
ADAMS accession number for each document referenced (if it is available
in ADAMS) is provided the first time that it is mentioned in this
document.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2016-0202 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 submissions into ADAMS.
II. Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses and Combined Licenses and Proposed No Significant
Hazards Consideration Determination
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Sec. 50.92 of title 10 of the Code of
Federal Regulations (10 CFR), this means that operation of the facility
in accordance with the proposed amendment would not (1) involve a
significant increase in the probability or consequences of an accident
previously evaluated, or (2) create the possibility of a new or
different kind of accident from any accident previously evaluated; or
(3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period if circumstances change during the 30-day comment
period such that failure to act in a timely way would result, for
example in derating or shutdown of the facility. If the Commission
takes action prior to the expiration of either the comment period or
the notice period, it will publish in the Federal Register a notice of
issuance. If the Commission makes a final no significant hazards
consideration determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
A. Opportunity To Request a Hearing and Petition for Leave To Intervene
Within 60 days after the date of publication of this notice, any
persons (petitioner) whose interest may be affected by this action may
file a request for a hearing and a petition to intervene (petition)
with respect to issuance of the amendment to the subject facility
operating license or combined license. Petitions shall be filed in
accordance with the Commission's ``Agency Rules of Practice and
Procedure'' in 10 CFR part 2. Interested persons should consult a
current copy of 10 CFR 2.309, which is available at the NRC's PDR,
located at One White Flint North, Room O1-F21, 11555 Rockville Pike
(first floor), Rockville, Maryland 20852. The NRC's regulations are
accessible electronically from the NRC Library on the NRC's Web site at
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a petition is
filed within 60 days, the Commission or a presiding officer designated
by the Commission or by the Chief Administrative Judge of the Atomic
Safety and Licensing Board Panel, will rule on the petition; and the
Secretary or the Chief Administrative Judge of the
[[Page 66303]]
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.309, a petition shall set forth with
particularity the interest of the petitioner in the proceeding, and how
that interest may be affected by the results of the proceeding. The
petition should specifically explain the reasons why intervention
should be permitted with particular reference to the following general
requirements: (1) The name, address, and telephone number of the
petitioner; (2) the nature of the petitioner's right under the Act to
be made a party to the proceeding; (3) the nature and extent of the
petitioner's property, financial, or other interest in the proceeding;
and (4) the possible effect of any decision or order which may be
entered in the proceeding on the petitioner's interest. The petition
must also set forth the specific contentions which the petitioner seeks
to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner shall provide a brief explanation of the bases for the
contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion to support
its position on the issue. The petition must include sufficient
information to show that a genuine dispute exists with the applicant on
a material issue of law or fact. Contentions shall be limited to
matters within the scope of the amendment under consideration. The
contention must be one which, if proven, would entitle the petitioner
to relief. A petitioner who fails to satisfy these requirements with
respect to at least one contention will not be permitted to participate
as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing with respect to resolution of that person's admitted
contentions, including the opportunity to present evidence and to
submit a cross-examination plan for cross-examination of witnesses,
consistent with the NRC's regulations, policies, and procedures.
Petitions for leave to intervene must be filed no later than 60
days from the date of publication of this notice. Requests for hearing,
petitions for leave to intervene, and motions for leave to file new or
amended contentions that are filed after the 60-day deadline will not
be entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the three factors in 10
CFR 2.309(c)(1)(i)-(iii).
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, then any hearing held would take place before
the issuance of any amendment unless the Commission finds an imminent
danger to the health or safety of the public, in which case it will
issue an appropriate order or rule under 10 CFR part 2.
A State, local governmental body, Federally-recognized Indian
Tribe, or agency thereof, may submit a petition to the Commission to
participate as a party under 10 CFR 2.309(h)(1).
The petition should state the nature and extent of the petitioner's
interest in the proceeding. The petition should be submitted to the
Commission by November 28, 2016. The petition must be filed in
accordance with the filing instructions in the ``Electronic Submissions
(E-Filing)'' section of this document, and should meet the requirements
for petitions set forth in this section, except that under 10 CFR
2.309(h)(2) a State, local governmental body, or Federally-recognized
Indian Tribe, or agency thereof does not need to address the standing
requirements in 10 CFR 2.309(d) if the facility is located within its
boundaries. A State, local governmental body, Federally-recognized
Indian Tribe, or agency thereof may also have the opportunity to
participate under 10 CFR 2.315(c).
If a hearing is granted, any person who does not wish, or is not
qualified, to become a party to the proceeding may, in the discretion
of the presiding officer, be permitted to make a limited appearance
pursuant to the provisions of 10 CFR 2.315(a). A person making a
limited appearance may make an oral or written statement of position on
the issues, but may not otherwise participate in the proceeding. A
limited appearance may be made at any session of the hearing or at any
prehearing conference, subject to the limits and conditions as may be
imposed by the presiding officer. Details regarding the opportunity to
make a limited appearance will be provided by the presiding officer if
such sessions are scheduled.
B. Electronic Submissions (E-Filing)
All documents filed in NRC adjudicatory proceedings, including a
request for hearing, a petition for leave to intervene, any motion or
other document filed in the proceeding prior to the submission of a
request for hearing or petition to intervene (hereinafter
``petition''), and documents filed by interested governmental entities
participating under 10 CFR 2.315(c), must be filed in accordance with
the NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77
FR 46562; August 3, 2012). The E-Filing process requires participants
to submit and serve all adjudicatory documents over the internet, or in
some cases to mail copies on electronic storage media. Participants may
not submit paper copies of their filings unless they seek an exemption
in accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least 10
days prior to the filing deadline, the participant should contact the
Office of the Secretary by email at [email protected], or by
telephone at 301-415-1677, to request (1) a digital identification (ID)
certificate, which allows the participant (or its counsel or
representative) to digitally sign documents and access the E-Submittal
server for any proceeding in which it is participating; and (2) advise
the Secretary that the participant will be submitting a petition (even
in instances in which the participant, or its counsel or
representative, already holds an NRC-issued digital ID certificate).
Based upon this information, the Secretary will establish an electronic
docket for the hearing in this proceeding if the Secretary has not
already established an electronic docket.
Information about applying for a digital ID certificate is
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing
the E-Submittal server are available on the NRC's public Web site at
http://www.nrc.gov/site-help/e-submittals/
[[Page 66304]]
adjudicatory-sub.html. Participants may attempt to use other software
not listed on the Web site, but should note that the NRC's E-Filing
system does not support unlisted software, and the NRC Electronic
Filing Help Desk will not be able to offer assistance in using unlisted
software.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a petition.
Submissions should be in Portable Document Format (PDF). Additional
guidance on PDF submissions is available on the NRC's public Web site
at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing
is considered complete at the time the documents are submitted through
the NRC's E-Filing system. To be timely, an electronic filing must be
submitted to the E-Filing system no later than 11:59 p.m. Eastern Time
on the due date. Upon receipt of a transmission, the E-Filing system
time-stamps the document and sends the submitter an email notice
confirming receipt of the document. The E-Filing system also
distributes an email notice that provides access to the document to the
NRC's Office of the General Counsel and any others who have advised the
Office of the Secretary that they wish to participate in the
proceeding, so that the filer need not serve the documents on those
participants separately. Therefore, applicants and other participants
(or their counsel or representative) must apply for and receive a
digital ID certificate before a hearing petition to intervene is filed
so that they can obtain access to the document via the E-Filing system.
A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC Electronic
Filing Help Desk through the ``Contact Us'' link located on the NRC's
public Web site at http://www.nrc.gov/site-help/e-submittals.html, by
email to [email protected], or by a toll-free call at 1-866-672-
7640. The NRC Electronic Filing Help Desk is available between 9 a.m.
and 7 p.m., Eastern Time, Monday through Friday, excluding government
holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
stating why there is good cause for not filing electronically and
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland
20852, Attention: Rulemaking and Adjudications Staff. Participants
filing a document in this manner are responsible for serving the
document on all other participants. Filing is considered complete by
first-class mail as of the time of deposit in the mail, or by courier,
express mail, or expedited delivery service upon depositing the
document with the provider of the service. A presiding officer, having
granted an exemption request from using E-Filing, may require a
participant or party to use E-Filing if the presiding officer
subsequently determines that the reason for granting the exemption from
use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in the
NRC's electronic hearing docket which is available to the public at
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the
Commission, or the presiding officer. Participants are requested not to
include personal privacy information, such as social security numbers,
home addresses, or home phone numbers in their filings, unless an NRC
regulation or other law requires submission of such information.
However, in some instances, a petition will require including
information on local residence in order to demonstrate a proximity
assertion of interest in the proceeding. With respect to copyrighted
works, except for limited excerpts that serve the purpose of the
adjudicatory filings and would constitute a Fair Use application,
participants are requested not to include copyrighted materials in
their submission.
The Commission will issue a notice or order granting or denying a
hearing request or intervention petition, designating the issues for
any hearing that will be held and designating the Presiding Officer. A
notice granting a hearing will be published in the Federal Register and
served on the parties to the hearing.
For further details with respect to these license amendment
applications, see the application for amendment which is available for
public inspection in ADAMS and at the NRC's PDR. For additional
direction on accessing information related to this document, see the
``Obtaining Information and Submitting Comments'' section of this
document.
Energy Northwest, Docket No. 50-397, Columbia Generating Station
(Columbia), Benton County, Washington
Date of amendment request: July 14, 2016. A publicly-available
version is in ADAMS under Accession No. ML16196A419.
Description of amendment request: The amendment would change
Technical Specification (TS) 5.5.6, ``Inservice Testing [IST]
Program,'' to remove requirements duplicated in American Society of
Mechanical Engineers (ASME) Code for Operations and Maintenance of
Nuclear Power Plants (OM Code), Case OMN-20, ``Inservice Test
Frequency.'' This change, thereby, will then adopt Technical
Specification Task Force (TSTF) TSTF-545, Revision 3, ``TS Inservice
Testing Program Removal & Clarify SR [Surveillance Requirement] Usage
Rule Application to Section 5.5 Testing.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises TS Chapter 5, ``Administrative
Controls,'' Section 5.5, ``Programs and Manuals,'' by eliminating
the ``Inservice Testing Program'' specification. Most requirements
in the Inservice Testing Program are removed, as they are
duplicative of requirements in the ASME OM Code, as clarified by
Code Case OMN-20, ``Inservice Test Frequency,'' which has been
approved for use at Columbia. The remaining requirements in the
Section 5.5 IST Program are eliminated because the NRC has
determined their inclusion in the TS is contrary to regulations. A
new defined term, ``Inservice Testing Program,'' is added to the TS,
which references the requirements of 10 CFR 50.55a(f).
Performance of inservice testing is not an initiator to any
accident previously evaluated. As a result, the probability of
occurrence of an accident is not significantly affected by the
proposed change. Inservice test frequencies under Code Case OMN-20
are equivalent to the current testing period allowed by the TS with
the exception that testing frequencies greater than 2 years may be
extended by up to 6 months to facilitate test scheduling and
consideration of plant operating conditions that may not be suitable
for performance of the required testing. The testing frequency
extension will not affect the ability of the components to mitigate
any accident previously evaluated as the components are required to
be operable during the testing period extension. Performance of
inservice tests utilizing the
[[Page 66305]]
allowances in OMN-20 will not significantly affect the reliability
of the tested components. As a result, the availability of the
affected components, as well as their ability to mitigate the
consequences of accidents previously evaluated, is not affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not alter the design or configuration
of the plant. The proposed change does not involve a physical
alteration of the plant; no new or different kind of equipment will
be installed. The proposed change does not alter the types of
inservice testing performed. In most cases, the frequency of
inservice testing is unchanged. However, the frequency of testing
would not result in a new or different kind of accident from any
previously evaluated since the testing methods are not altered.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change eliminates some requirements from the TS in
lieu of requirements in the ASME Code, as modified by use of Code
Case OMN-20. Compliance with the ASME Code is required by 10 CFR
50.55a. The proposed change also allows inservice tests with
frequencies greater than 2 years to be extended by 6 months to
facilitate test scheduling and consideration of plant operating
conditions that may not be suitable for performance of the required
testing. The testing frequency extension will not affect the ability
of the components to respond to an accident as the components are
required to be operable during the testing period extension. The
proposed change will eliminate the existing TS SR 3.0.3 allowance to
defer performance of missed inservice tests up to the duration of
the specified testing frequency, and instead will require an
assessment of the missed test on equipment operability. This
assessment will consider the effect on a margin of safety (equipment
operability). Should the component be inoperable, the Technical
Specifications provide actions to ensure that the margin of safety
is protected. The proposed change also eliminates a statement that
nothing in the ASME Code should be construed to supersede the
requirements of any TS. The NRC has determined that statement to be
incorrect. However, elimination of the statement will have no effect
on plant operation or safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William A. Horin, Esq., Winston & Strawn,
1700 K Street NW., Washington, DC 20006-3817.
NRC Branch Chief: Robert J. Pascarelli.
Energy Northwest, Docket No. 50-397, Columbia Generating Station
(Columbia), Benton County, Washington
Date of amendment request: July 28, 2016. A publicly-available
version is in ADAMS under Accession No. ML16210A528.
Description of amendment request: The amendment would revise the
current Columbia Emergency Plan Emergency Action Level scheme to one
based on Nuclear Energy Institute (NEI) guidance established in NEI 99-
01, ``Development of Emergency Action Levels for Non-Passive
Reactors,'' Revision 6, 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 amendment affects the Columbia Generating Station
(Columbia) Emergency Plan (EP) and associated Emergency Action
Levels (EALs); it does not alter the Operating License or the
Technical Specifications. The proposed amendment does not change the
design function of any system, structure, or component and does not
change the way the plant is maintained or operated. The proposed
amendment does not affect any accident mitigating feature or
increase the likelihood of malfunction for plant structures,
systems, and components.
The proposed amendment will not change any of the analyses
associated with the Columbia Final Safety Analysis Report Chapter 15
accidents because plant operation, structures, systems, components,
accident initiators, and accident mitigation functions remain
unchanged.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment affects the Columbia EP and associated
EALs; it does not change the design function of any system,
structure, or component and does not change the way the plant is
operated or maintained. The proposed amendment does not create a
credible failure mechanism, malfunction, or accident initiator not
already considered in the design and licensing basis.
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 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 amendment does not impact
operation of the plant and no accident analyses are affected by the
proposed amendment. The proposed amendment does not affect the
Technical Specifications or the method of operating the plant.
Additionally, the proposed amendment will not relax any criteria
used to establish safety limits and will not relax any safety system
settings. The safety analysis acceptance criteria are not affected
by this amendment. The proposed amendment will not result in plant
operation in a configuration outside the design basis. The proposed
amendment does not adversely affect systems that respond to safely
shut down the plant and to maintain the plant in a safe shutdown
condition.
Therefore, the proposed amendment does 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 A. Horin, Esq., Winston & Strawn,
1700 K Street NW., Washington, DC 20006-3817.
NRC Branch Chief: Robert J. Pascarelli.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3 (IP3), Westchester County, New York
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant (FitzPatrick), Oswego County, New York
Date of amendment request: August 16, 2016. A publicly-available
version is in ADAMS under Accession No. ML16230A308.
Description of amendment request: The amendment would transfer the
beneficial interest in the Power Authority of the State of New York
(PASNY) Master Decommissioning
[[Page 66306]]
Trust (Master Trust), including all rights and obligations thereunder,
held by PASNY for IP3 and FitzPatrick to Entergy Nuclear Operations,
Inc. (ENO). ENO also requests the NRC's consent to amendments to the
Master Decommissioning Trust Agreement dated July 25, 1990, as amended
(Master Trust Agreement), governing the Master Trust to facilitate this
transfer. Finally, ENO seeks approval of license amendments to modify
the existing trust-related license conditions to reflect the proposed
transfer of the Master Trust to ENO and to delete other conditions so
as to apply the requirements of 10 CFR 50.75(h)(1). ENO and Exelon
Generation Company, LLC. (Exelon), jointly filed an application for a
direct license transfer of FitzPatrick to Exelon on August 18, 2016. A
separate Federal Register notice details the NRC's consideration of
approval for the FitzPatrick license transfer.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed amendments involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The requested changes delete certain license conditions
pertaining to the decommissioning trust agreements currently in
sections 2.Q to 2.X of the IP3 Operating License and sections 2.H to
2.O of the FitzPatrick Operating License. In addition, conforming
changes to 2.W and 2.X of the IP3 Operating License and 2.P and 2.Q
of the FitzPatrick Operating License are necessary [to] reflect the
transfer of the Master Trust from PASNY to ENO.
The requested changes are consistent with the types of license
amendments permitted in 10 CFR 50.75(h)(5).
The regulations of 10 CFR 50.75(h)(4) state that ``Unless
otherwise determined by the Commission with regard to a specific
application, the Commission has determined that any amendment to the
license of a utilization facility that does no more than delete
specific license conditions relating to the terms and conditions of
decommissioning trust agreements involves `no significant hazards
consideration.' ''
In addition the requested changes seek changes to the Master
Trust agreement only to the extent that they replace PASNY, a non-
licensee, with ENO, a licensee. No other changes to the Master Trust
agreement are contemplated.
This request involves changes that are administrative in nature.
No actual plant equipment or accident analyses will be affected by
the proposed changes.
Therefore, the proposed amendments do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed amendments create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This request involves administrative changes to licenses that
will be consistent with the NRC's regulations at 10 CFR 50.75(h) and
to change the name of the entity responsible under the Master Trust
for decommissioning from a non-licensee to a licensee.
No actual plant equipment or accident analyses will be affected
by the proposed changes and no failure modes not bounded by
previously evaluated accidents will be created.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Do the proposed amendments involve a significant reduction in
a margin of safety?
Response: No.
The request involves administrative changes to the licenses that
will be consistent with the NRC's regulations at 10 CFR 50.75(h) and
to change the name of the entity responsible under the Master Trust
for decommissioning from a non-licensee to a licensee.
Margin of safety is associated with confidence in the ability of
the fission product barriers to limit the level of radiation doses
to the public. No actual plant equipment or accident analyses will
be affected by the proposed change. Additionally, the proposed
changes will not relax any criteria used to establish safety limits,
will not relax any safety systems settings, or will not relax the
bases for any limiting conditions of operation.
Therefore, the proposed change does 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: Jeanne Cho, Entergy Nuclear Operations,
Inc., 440 Hamilton Avenue, White Plains, New York, 10601.
NRC Branch Chief: Travis L. Tate.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear
Plant (PNP), Van Buren County, Michigan
Date of amendment request: August 22, 2016, as supplemented by
letter dated September 8, 2016. Publicly-available versions are in
ADAMS under Accession Nos. ML16235A195 and ML16252A351, respectively.
Description of amendment request: The proposed amendment would
replace existing license condition 2.C.(4) with a new license condition
to state that technical specification (TS) surveillance requirement
(SR) 3.1.4.3 is not required for control rod drive 13 (CRD-13) during
cycle 25 until the next entry into Mode 3. In addition, the condition
would state that CRD-13 seal leakage shall be repaired prior to
entering Mode 2, following the next Mode 3 entry, and that the reactor
shall be shut down if CRD-13 seal leakage exceeds two gallons per
minute. The proposed amendment also requests replacement of the
obsolete note in TS SR 3.1.4.3 with a note to clarify that TS SR
3.1.4.3 is not required to be performed or met for CRD-13 during cycle
25 provided CRD-13 is administratively declared immovable, but
trippable, and Condition D is entered for CRD-13.
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 license amendment replaces an obsolete license
condition concerning CRD-22 testing that applied only to operating
cycle 21 with a new license condition to forgo the remaining two
required surveillance tests of CRD-13 from the PNP TS surveillance
requirement for partial movement every 92 days during cycle 25.
Since CRD-13 remains trippable, the proposed license condition does
not affect or create any accident initiators or precursors. As such,
the proposed license condition does not increase the probability of
an accident.
The proposed license amendment does not increase the
consequences of an accident. The ability to move a full-length
control rod by its drive mechanism is not an initial assumption used
in the safety analyses. The safety analyses assume full-length
control rod insertion, except the most reactive rod, upon reactor
trip. The surveillance requirement performed during the last
refueling outage verified control rod drop times are within accident
analysis assumptions. ENO [Entergy Nuclear Operations] has
determined that CRD seal leakage does not increase the likelihood of
an untrippable control rod. The assumptions of the safety analyses
will be maintained, and the consequences of an accident will not be
increased.
Therefore, operation of the facility in accordance with the
proposed license condition 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?
[[Page 66307]]
Response: No.
The proposed license amendment does not involve a physical
alteration of any structure, system or component (SSC) or change the
way any SSC is operated. The proposed license condition does not
involve operation of any required SSCs in a manner or configuration
differently from those previously recognized or evaluated. No new
failure mechanisms would be introduced by the requested SR interval
extension.
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 license amendment does not affect trippability of
the control rod. It will have the same capability to mitigate an
accident as it had prior to the proposed license condition.
Therefore, the proposed amendment would 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: Jeanne Cho, Senior Counsel, Entergy Nuclear
Operations, Inc., 440 Hamilton Ave., White Plains, NY 10601.
NRC Branch Chief: David J. Wrona.
Exelon Generation Company, LLC, Docket No. 50-219, Oyster Creek Nuclear
Generating Station (OCNGS), Ocean County, New Jersey; and Docket No.
50-220, Nine Mile Point Nuclear Station, Unit 1 (NMP1), Oswego County,
New York
Date of amendment request: August 1, 2016. A publicly-available
version is in ADAMS under Accession No. ML16215A128.
Description of amendment request: The amendments would revise
OCNGS's Technical Specification (TS) Section 2.1, ``Safety Limit--Fuel
Cladding Integrity,'' and NMP1's TS Section 2.1.1, ``Fuel Cladding
Integrity,'' to reduce the steam dome pressure.
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 edits in [brackets]:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to the OCNGS TS for the reactor steam dome
pressure in Reactor Core Safety Limits 2.1.A and 2.1.B does not
alter the use of the analytical methods used to determine the safety
limits that have been previously reviewed and approved by the NRC.
Additionally, the proposed change to NMP1 for the reactor steam dome
pressure in Reactor Core Safety Limits 2.1.1.a and 2.1.1.b does not
alter the use of the analytical methods used to determine the safety
limits that have been previously reviewed and approved by the NRC.
The proposed change is in accordance with an NRC approved critical
power correlation methodology, and as such, maintains required
safety margins. The proposed change does not adversely affect
accident initiators or precursors, nor does it alter the design
assumptions, conditions, or configuration of the facility or the
manner in which the plant is operated and maintained.
The proposed change does not alter or prevent the ability of
structures, systems, and components (SSCs) from performing their
intended function to mitigate the consequences of an initiating
event within the assumed acceptance limits. The proposed change does
not require any physical change to any plant SSCs nor does it
require any change in systems or plant operations. The proposed
change is consistent with the safety analysis assumptions and
resultant consequences.
Lowering the value of reactor steam dome pressure in the TS has
no physical effect on plant equipment and therefore, no impact on
the course of plant transients. The change is an analytical exercise
to demonstrate the applicability of correlations and methodologies.
There are no known operational or safety benefits.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed reduction in the reactor dome pressure safety limit
from 800 psia [pounds per square inch absolute] to 700 psia is a
change based upon previously approved documents and does not involve
changes to the plant hardware or its operating characteristics. As a
result, no new failure modes are being introduced. There are no
hardware changes nor are there any changes in the method by which
any plant systems perform a safety function. No new accident
scenarios, failure mechanisms, or limiting single failures are
introduced as a result of the proposed change.
The proposed change does not introduce any new accident
precursors, nor does it involve any physical plant alterations or
changes in the methods governing normal plant operation. Also, the
change does not impose any new or different requirements or
eliminate any existing requirements. The change does not alter
assumptions made in the safety analysis.
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 margin of safety is established through the design of the
plant structures, systems, and components, and through the
parameters for safe operation and setpoints for the actuation of
equipment relied upon to respond to transients and design basis
accidents. Evaluation of the 10 CFR part 21 condition by GE [General
Electric] determined that since the MCPR [minimum critical power
ratio] improves during the PRFO [pressure regulator failure-maximum
demand (open)] transient, there is no decrease in the safety margin
and therefore there is not a threat to fuel cladding integrity. The
proposed change in reactor dome pressure supports the current safety
margin, which protects the fuel cladding integrity during a
depressurization transient, but does not change the requirements
governing operation or availability of safety equipment assumed to
operate to preserve the margin of safety. The change does not alter
the behavior of plant equipment, which remains unchanged.
The proposed change to Reactor Core Safety Limits 2.1.A and
2.1.B is consistent with and within the capabilities of the
applicable NRC approved critical power correlation for the fuel
designs in use at OCNGS. Additionally, the proposed change to
Reactor Core Safety Limits 2.1.1.a and 2.1.1.b is consistent with
and within the capabilities of the NRC approved critical power
correlation for the fuel designs in use at NMP1. No setpoints at
which protective actions are initiated are altered by the proposed
change. The proposed change does not alter the manner in which the
safety limits are determined. This change is consistent with plant
design and does not change the TS operability requirements; thus,
previously evaluated accidents are not affected by this proposed
change.
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: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Acting Branch Chief: Shaun M. Anderson.
Indiana Michigan Power Company (I&M), Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment request: July 21, 2016. A publicly-available
version is in ADAMS under Accession No. ML16208A076.
Description of amendment request: The proposed changes are
consistent
[[Page 66308]]
with the NRC-approved Technical Specifications Task Force (TSTF)
Traveler, TSTF-545, Revision 3, ``TS [Technical Specification]
Inservice Testing [IST] Program Removal & Clarify SR [Surveillance
Requirement] Usage Rule Application to Section 5.5 Testing.'' The
proposed change would revise the TSs to eliminate the Section 5.5.6,
``Inservice Testing Program.'' A new defined term, ``INSERVICE TESTING
PROGRAM,'' would be added to the TS Definitions section. TS SRs that
currently refer to the Inservice Testing Program from Section 5.5.6
would be revised to refer to the new defined term, ``INSERVICE TESTING
PROGRAM.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises TS Chapter 5, ``Administrative
Controls,'' Section 5.5, ``Programs and Manuals,'' by eliminating
the ``Inservice Testing Program'' specification. Most requirements
in the IST Program are removed, as they are duplicative of
requirements in the ASME [American Society of Mechanical Engineers]
OM [Operation and Maintenance] Code, as clarified by Code Case OMN-
20, ``Inservice Test Frequency.'' The remaining requirements in the
Section 5.5.6 IST Program are eliminated because the NRC has
determined their inclusion in the TS is contrary to regulations. A
new defined term, ``Inservice Testing Program,'' is added to the TS,
which references the requirements of 10 CFR 50.55a(f).
Performance of IST is not an initiator to any accident
previously evaluated. As a result, the probability of occurrence of
an accident is not significantly affected by the proposed change.
Inservice test frequencies under Code Case OMN-20 are equivalent to
the current testing period allowed by the TS with the exception that
testing frequencies greater than 2 years may be extended by up to 6
months to facilitate test scheduling and consideration of plant
operating conditions that may not be suitable for performance of the
required testing. The testing frequency extension will not affect
the ability of the components to mitigate any accident previously
evaluated as the components are required to be operable during the
testing period extension. Performance of inservice tests utilizing
the allowances in OMN-20 will not significantly affect the
reliability of the tested components. As a result, the availability
of the affected components, as well as their ability to mitigate the
consequences of accidents previously evaluated, is not affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not alter the design or configuration
of the plant. The proposed change does not involve a physical
alteration of the plant; no new or different kind of equipment will
be installed. The proposed change does not alter the types of
inservice testing performed. In most cases, the frequency of IST is
unchanged. However, the frequency of testing would not result in a
new or different kind of accident from any previously evaluated
since the testing methods are not altered.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change eliminates some requirements from the TS in
lieu of requirements in the ASME Code, as modified by use of Code
Case OMN-20. Compliance with the ASME Code is required by 10 CFR
50.55a. The proposed change also allows inservice tests with
frequencies greater than 2 years to be extended by 6 months to
facilitate test scheduling and consideration of plant operating
conditions that may not be suitable for performance of the required
testing. The testing frequency extension will not affect the ability
of the components to respond to an accident as the components are
required to be operable during the testing period extension. The
proposed change will eliminate the existing TS SR 3.0.3 allowance to
defer performance of missed inservice tests up to the duration of
the specified testing frequency, and instead will require an
assessment of the missed test on equipment operability. This
assessment will consider the effect on a margin of safety (equipment
operability). Should the component be inoperable, the TS provide
actions to ensure that the margin of safety is protected. The
proposed change also eliminates a statement that nothing in the ASME
Code should be construed to supersede the requirements of any TS.
The NRC has determined that statement to be incorrect. However,
elimination of the statement will have no effect on plant operation
or safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Robert B. Haemer, Senior Nuclear Counsel,
One Cook Place, Bridgman, MI 49106.
NRC Branch Chief: David J. Wrona.
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 11, 2016. A publicly-available
version is in ADAMS under Accession No. ML16224B122.
Description of amendment request: The amendment request proposes
changes to plant-specific Tier 2 information incorporated into the
Updated Final Safety Analysis Report (UFSAR), and involves changes to
combined license Appendix C (and corresponding plant-specific Tier 1
information). The proposed changes are to information identifying the
frontal face area and screen surface area for the In-Containment
Refueling Water Storage Tank (IRWST) screens, the location and
dimensions of the protective plate located above the containment
recirculation (CR) screens, and increasing the maximum Normal Residual
Heat Removal System flowrate through the IRWST and CR screens. Pursuant
to the provisions of 10 CFR 52.63(b)(1), an exemption from elements of
the design as certified in the 10 CFR part 52, appendix D, design
certification rule is also requested for the plant-specific Design
Control Document Tier 1 material departures.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below with the 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 to the location and dimensions of the
protective plate continues to provide sufficient space surrounding
the containment recirculation screens for debris to settle before
reaching the screens as confirmed by an evaluation demonstrating
that the protective plate continues to fulfill its design function
of preventing debris from reaching the screens. In addition, the
increase to the minimum IRWST screen size reinforces the ability of
the screens to perform their design function with the increased
[Residual Heat Removal System (RNS)] maximum flowrate proposed. The
proposed changes do not adversely affect any accident initiating
component, and thus the probabilities of the accidents previously
evaluated are not affected. The affected equipment does not
adversely affect the ability of equipment to contain radioactive
material. Because the proposed change does not affect a release path
or increase the
[[Page 66309]]
expected dose rates, the potential radiological releases in the
UFSAR accident analyses are unaffected.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed activity to change the location and dimensions of
the protective plate above the containment recirculation screens, to
change the minimum IRWST screen size, and to increase the maximum
RNS flowrate through the IRWST and CR screens does not alter the
method in which safety functions are accomplished. The analyses
demonstrate that the screens are able to perform their functions in
a similar manner and perform adequately in response to an accident,
and no new failure modes are introduced by the proposed change.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change to the design does not change any of the
codes or standards to which the IRWST screens, containment
recirculation screens, and containment recirculation screen
protective plate are designed as documented in the UFSAR. The
containment recirculation screen protective plate continues to
prevent debris from reaching the CR screens, and the IRWST and CR
screens maintain their ability to block debris while at the proposed
increase in RNS maximum flowrate.
No safety analysis or design basis acceptance limit/criterion is
challenged or exceeded by the proposed changes.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
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: August 23, 2016. A publicly-available
version is in ADAMS under Accession No. ML16236A265.
Description of amendment request: The amendment request proposes
changes to the Fire Pump Head and Diesel Fuel Day Tank. Because, this
proposed change requires a departure from Tier 1 information in the
Westinghouse Electric Company's AP1000 Design Control Document (DCD),
the licensee also requested an exemption from the requirements of the
Generic DCD Tier 1 in accordance with 10 CFR 52.63(b)(1).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The increase in head pressure by the proposed change to the fire
protection system (FPS) motor-driven and diesel-driven fire pumps
maintains compliance with National Fire Protection Association
(NFPA) Standard NFPA-14, Standard for the Installation of Standpipe,
Private Hydrants, and Hose Systems, 2000 Edition, requirements by
providing adequate pressure in the standpipe and automatic sprinkler
system to maintain the ability to fight and/or contain a postulated
fire. The proposed change to the diesel-driven fire pump fuel day
tank volume maintains the availability of the diesel-driven fire
pump for service upon failure of the electric motor-driven fire pump
or a loss of offsite power by providing a fuel day tank that is
reserved exclusively for the diesel-driven pump and meets the
minimum capacity requirements of NFPA 20, Standard for the
Installation of Stationary Pumps for Fire Protection, 1999 Edition.
These changes do not affect the operation of any systems or
equipment that initiate an analyzed accident or alter any
structures, systems, and [components (SSCs)] accident initiator or
initiating sequence of events.
These changes have no adverse impact on the support, design, or
operation of mechanical and fluid systems. The response of systems
to postulated accident conditions is not adversely affected by the
proposed changes. There is no change to the predicted radioactive
releases due to normal operation or postulated accident conditions.
Consequently, the plant response to previously evaluated accidents
is not impacted, nor does the proposed change create any new
accident precursors.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes do not affect the operation of any systems
or equipment that may initiate a new or different kind of accident,
or alter any SSC such that a new accident initiator or initiating
sequence of events is created. The proposed changes to the fire pump
performance specifications and fire pump fuel day tank volume do not
affect any safety-related equipment, nor do they add any new
interface to safety-related SSCs. No system or design function or
equipment qualification is affected by this change. The changes do
not introduce a new failure mode, malfunction, or sequence of events
that could affect safety or safety-related equipment.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes maintain compliance with the applicable
Codes and Standards, thereby maintaining the margin of safety
associated with these SSCs. The proposed changes do not alter any
applicable design codes, code compliance, design function, or safety
analysis. Consequently, no safety analysis or design basis
acceptance limit/criterion is challenged or exceeded by the proposed
change, thus the margin of safety is not reduced.
Because no safety analysis or design basis acceptance limit/
criterion is challenged or exceeded by these changes, no margin of
safety is reduced.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. M. Stanford Blanton, Balch & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-321 and 50-
366, Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2, Appling County,
Georgia
Date of amendment request: July 28, 2016. A publicly-available
version is in ADAMS under Accession No. ML16214A252.
Description of amendment request: The amendments would revise the
technical specifications (TSs) at the Edwin I. Hatch Nuclear Plant,
Units 1 and 2, to eliminate the ``lnservice Testing Program'' from TS
5.5, ``Programs and Manuals,'' and add a new defined term, ``INSERVICE
TESTING PROGRAM,'' to TS 1.1, ``Definitions.'' This request is
submitted in accordance with Technical
[[Page 66310]]
Specifications Task Force (TSTF) Traveler TSTF-545, Revision 3, ``TS
lnservice Testing Program Removal & Clarify SR [Surveillance
Requirement] Usage Rule Application to Section 5.5 Testing.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises TS Chapter 5, ``Administrative
Controls,'' Section 5.5, ``Programs and Manuals,'' by eliminating
the ``lnservice Testing Program'' specification. Most requirements
in the lnservice Testing Program are removed, as they are
duplicative of requirements in the ASME OM [American Society of
Mechanical Engineers Operation and Maintenance] Code, as clarified
by Code Case OMN-20, ``lnservice Test Frequency.'' The remaining
requirements in the Section 5.5 IST [Inservice Testing] Program are
eliminated because the NRC has determined their inclusion in the TS
is contrary to regulations. A new defined term, ``INSERVICE TESTING
PROGRAM,'' is added to the TS, which references the requirements of
10 CFR 50.55a(f).
Performance of inservice testing is not an initiator to any
accident previously evaluated. As a result, the probability of
occurrence of an accident is not significantly affected by the
proposed change. lnservice test frequencies under Code Case OMN-20
are equivalent to the current testing period allowed by the TS with
the exception that testing frequencies greater than 2 years may be
extended by up to 6 months to facilitate test scheduling and
consideration of plant operating conditions that may not be suitable
for performance of the required testing. The testing frequency
extension will not affect the ability of the components to mitigate
any accident previously evaluated as the components are required to
be operable during the testing period extension. Performance of
inservice tests utilizing the allowances in OMN-20 will not
significantly affect the reliability of the tested components. As a
result, the availability of the affected components, as well as
their ability to mitigate the consequences of accidents previously
evaluated, is not affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not alter the design or configuration
of the plant. The proposed change does not involve a physical
alteration of the plant; no new or different kind of equipment will
be installed. The proposed change does not alter the types of
inservice testing performed. In most cases, the frequency of
inservice testing is unchanged. However, the frequency of testing
would not result in a new or different kind of accident from any
previously evaluated since the testing methods are not altered.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change eliminates some requirements from the TS in
lieu of requirements in the ASME Code, as modified by use of Code
Case OMN-20. Compliance with the ASME Code is required by 10 CFR
50.55a. The proposed change also allows inservice tests with
frequencies greater than 2 years to be extended by 6 months to
facilitate test scheduling and consideration of plant operating
conditions that may not be suitable for performance of the required
testing. The testing frequency extension will not affect the ability
of the components to respond to an accident as the components are
required to be operable during the testing period extension.
The proposed change will eliminate the existing TS SR 3.0.3
allowance to defer performance of missed inservice tests up to the
duration of the specified testing frequency, and instead will
require an assessment of the missed test on equipment operability.
This assessment will consider the effect on a margin of safety
(equipment operability). Should the component be inoperable, the
Technical Specifications provide actions to ensure that the margin
of safety is protected. The proposed change also eliminates a
statement that nothing in the ASME Code should be construed to
supersede the requirements of any TS. The NRC has determined that
statement to be incorrect. However, elimination of the statement
will have no effect on plant operation or safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jennifer M. Buettner, Associate General
Counsel, Southern Nuclear Operating Company, Inc., 40 Inverness Center
Parkway, Birmingham, AL 35242.
NRC Branch Chief: Michael T. Markley.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendment request: July 28, 2016. A publicly-available
version is in ADAMS under Accession No. ML16214A252.
Description of amendment request: The amendments would revise the
technical specifications (TSs) at the Joseph M. Farley Nuclear Plant,
Units 1 and 2, to eliminate the ``lnservice Testing Program'' from TS
5.5, ``Programs and Manuals,'' and add a new defined term, ``INSERVICE
TESTING PROGRAM,'' to TS 1.1, ``Definitions.'' This request is
submitted in accordance with Technical Specifications Task Force (TSTF)
Traveler TSTF-545, Revision 3, ``TS lnservice Testing Program Removal &
Clarify SR [Surveillance Requirement] Usage Rule Application to Section
5.5 Testing.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises TS Chapter 5, ``Administrative
Controls,'' Section 5.5, ``Programs and Manuals,'' by eliminating
the ``lnservice Testing Program'' specification. Most requirements
in the lnservice Testing Program are removed, as they are
duplicative of requirements in the ASME OM Code [American Society of
Mechanical Engineers Operation and Maintenance Code], as clarified
by Code Case OMN-20, ``lnservice Test Frequency.'' The remaining
requirements in the Section 5.5 IST [Inservice Testing] Program are
eliminated because the NRC has determined their inclusion in the TS
is contrary to regulations. A new defined term, ``INSERVICE TESTING
PROGRAM,'' is added to the TS, which references the requirements of
10 CFR 50.55a(f).
Performance of inservice testing is not an initiator to any
accident previously evaluated. As a result, the probability of
occurrence of an accident is not significantly affected by the
proposed change. lnservice test frequencies under Code Case OMN-20
are equivalent to the current testing period allowed by the TS with
the exception that testing frequencies greater than 2 years may be
extended by up to 6 months to facilitate test scheduling and
consideration of plant operating conditions that may not be suitable
for performance of the required testing. The testing frequency
extension will not affect the ability of the components to mitigate
any accident previously evaluated as the
[[Page 66311]]
components are required to be operable during the testing period
extension. Performance of inservice tests utilizing the allowances
in OMN-20 will not significantly affect the reliability of the
tested components. As a result, the availability of the affected
components, as well as their ability to mitigate the consequences of
accidents previously evaluated, is not affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not alter the design or configuration
of the plant. The proposed change does not involve a physical
alteration of the plant; no new or different kind of equipment will
be installed. The proposed change does not alter the types of
inservice testing performed. In most cases, the frequency of
inservice testing is unchanged. However, the frequency of testing
would not result in a new or different kind of accident from any
previously evaluated since the testing methods are not altered.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change eliminates some requirements from the TS in
lieu of requirements in the ASME Code, as modified by use of Code
Case OMN-20. Compliance with the ASME Code is required by 10 CFR
50.55a. The proposed change also allows inservice tests with
frequencies greater than 2 years to be extended by 6 months to
facilitate test scheduling and consideration of plant operating
conditions that may not be suitable for performance of the required
testing. The testing frequency extension will not affect the ability
of the components to respond to an accident as the components are
required to be operable during the testing period extension.
The proposed change will eliminate the existing TS SR 3.0.3
allowance to defer performance of missed in service tests up to the
duration of the specified testing frequency, and instead will
require an assessment of the missed test on equipment operability.
This assessment will consider the effect on a margin of safety
(equipment operability). Should the component be inoperable, the
Technical Specifications provide actions to ensure that the margin
of safety is protected. The proposed change also eliminates a
statement that nothing in the ASME Code should be construed to
supersede the requirements of any TS. The NRC has determined that
statement to be incorrect. However, elimination of the statement
will have no effect on plant operation or safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jennifer M. Buettner, Associate General
Counsel, Southern Nuclear Operating Company, Inc., 40 Iverness Center
Parkway, Birmingham, AL 35242.
NRC Branch Chief: Michael T. Markley.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of amendment request: July 28, 2016. A publicly-available
version is in ADAMS under Accession No. ML16214A252.
Description of amendment request: The amendments would revise the
technical specifications (TSs) at the Vogtle Electric Generating Plant,
Units 1 and 2, to eliminate the ``lnservice Testing Program'' from the
TS 5.5, ``Programs and Manuals,'' section and to add a new defined
term, ``INSERVICE TESTING PROGRAM,'' to the TS 1.1, ``Definitions,''
section. This request is submitted in accordance with Technical
Specifications Task Force (TSTF) Traveler TSTF-545, Revision 3, ``TS
lnservice Testing Program Removal & Clarify SR Usage Rule Application
to Section 5.5 Testing.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises TS Chapter 5, ``Administrative
Controls,'' Section 5.5, ``Programs and Manuals,'' by eliminating
the ``lnservice Testing Program'' specification. Most requirements
in the lnservice Testing Program are removed, as they are
duplicative of requirements in the ASME OM [American Society of
Mechanical Engineers Operation and Maintenance] Code, as clarified
by Code Case OMN-20, ``lnservice Test Frequency.'' The remaining
requirements in the Section 5.5 IST [Inservice Testing] Program are
eliminated because the NRC has determined their inclusion in the TS
is contrary to regulations. A new defined term, ``INSERVICE TESTING
PROGRAM,'' is added to the TS, which references the requirements of
10 CFR 50.55a(f).
Performance of inservice testing is not an initiator to any
accident previously evaluated. As a result, the probability of
occurrence of an accident is not significantly affected by the
proposed change. lnservice test frequencies under Code Case OMN-20
are equivalent to the current testing period allowed by the TS with
the exception that testing frequencies greater than 2 years may be
extended by up to 6 months to facilitate test scheduling and
consideration of plant operating conditions that may not be suitable
for performance of the required testing. The testing frequency
extension will not affect the ability of the components to mitigate
any accident previously evaluated as the components are required to
be operable during the testing period extension. Performance of
inservice tests utilizing the allowances in OMN-20 will not
significantly affect the reliability of the tested components. As a
result, the availability of the affected components, as well as
their ability to mitigate the consequences of accidents previously
evaluated, is not affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not alter the design or configuration
of the plant. The proposed change does not involve a physical
alteration of the plant; no new or different kind of equipment will
be installed. The proposed change does not alter the types of
inservice testing performed. In most cases, the frequency of
inservice testing is unchanged. However, the frequency of testing
would not result in a new or different kind of accident from any
previously evaluated since the testing methods are not altered.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change eliminates some requirements from the TS in
lieu of requirements in the ASME Code, as modified by use of Code
Case OMN-20. Compliance with the ASME Code is required by 10 CFR
50.55a. The proposed change also allows inservice tests with
frequencies greater than 2 years to be extended by 6 months to
facilitate test scheduling and consideration of plant operating
conditions that may not be suitable for performance of the required
testing. The testing frequency extension will not affect the ability
of the components to respond to an accident as the components are
required to be operable during the testing period extension.
The proposed change will eliminate the existing TS SR 3.0.3
allowance to defer performance of missed in service tests up to the
duration of the specified testing frequency, and instead will
require an
[[Page 66312]]
assessment of the missed test on equipment operability. This
assessment will consider the effect on a margin of safety (equipment
operability). Should the component be inoperable, the Technical
Specifications provide actions to ensure that the margin of safety
is protected. The proposed change also eliminates a statement that
nothing in the ASME Code should be construed to supersede the
requirements of any TS. The NRC has determined that statement to be
incorrect. However, elimination of the statement will have no effect
on plant operation or safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jennifer M. Buettner, Associate General
Counsel, Southern Nuclear Operating Company, Inc., 40 Inverness Center
Parkway, Birmingham, AL 35242.
NRC Branch Chief: Michael T. Markley.
III. Notice of Issuance of Amendments to Facility Operating Licenses
and Combined Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, proposed no
significant hazards consideration determination, and opportunity for a
hearing in connection with these actions, was published in the Federal
Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items can be accessed as described in the
``Obtaining Information and Submitting Comments'' section of this
document.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos.
1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: October 9, 2015, as
supplemented by letter dated May 12, 2016.
Brief description of amendments: The amendments approve a revision
to the emergency action levels from a scheme based on Nuclear Energy
Institute (NEI) 99-01, Revision 5, ``Methodology for Development of
Emergency Action Levels,'' to a scheme provided in the subsequent
Revision 6 of NEI 99-01.
Date of issuance: September 8, 2016.
Effective date: As of the date of issuance and shall be implemented
within 365 days from the date of issuance.
Amendment Nos.: Unit 1--198; Unit 2--198; Unit 3--198. A publicly-
available version is in ADAMS under Accession No. ML16180A109;
documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. NPF-41, NPF-51, and NPF-74:
The amendments revised the Operating Licenses.
Date of initial notice in Federal Register: December 8, 2015 (80 FR
76318). The supplemental letter dated May 12, 2016, provided additional
information that clarified the application, incorporated recent
emergency preparedness frequently asked questions, did not expand the
scope of the application as originally noticed, and did not change the
staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 8, 2016.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert
Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland
Date of amendment request: September 24, 2013, as supplemented by
letters dated February 9, March 11, April 13, July 6, and August 13,
2015; and February 24 and April 22, 2016.
Brief description of amendments: These amendments modify the
operating licenses and technical specifications (TSs) to incorporate a
new fire protection licensing basis in accordance with 10 CFR 50.48(c).
The amendments authorize the transition of the licensee's fire
protection program to a risk-informed, performance-based program based
on the 2001 Edition of National Fire Protection Association Standard
805, ``Performance-Based Standard for Fire Protection for Light Water
Reactor Electric Generating Plants.''
Date of issuance: August 30, 2016.
Effective date: As of the date of issuance and shall be implemented
in accordance with the schedule contained in the revised paragraph 2.E.
and page 12 of Appendix C, Additional Conditions to the Renewed
Facility Operating Licenses.
Amendment Nos.: 318 and 296. A publicly-available version is in
ADAMS under Accession No. ML16175A359; documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: August 5, 2014 (79 FR
45488). The supplemental letters dated February 9, March 11, April 13,
July 6, and August 13, 2015; and February 24 and April 22, 2016,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the 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 August 30, 2016.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-220 and 50-410, Nine
Mile Point Nuclear Station, Units 1 and 2, Oswego County, New York
Date of amendment request: October 8, 2015, as supplemented by
letter dated April 7, 2016.
[[Page 66313]]
Brief description of amendment: The amendments modified the
technical specifications (TSs) to allow for brief, inadvertent
simultaneous opening of redundant secondary containment personnel
access doors during brief entry and exit conditions.
Date of issuance: August 31, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 223 (Unit 1) and 157 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML16197A486; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. DPR-63 and NPF-69:
Amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: January 5, 2016 (81 FR
262). The supplemental letter dated April 7, 2016, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 31, 2016.
No significant hazards consideration comments received: No.
Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: August 31, 2015, as supplemented by
letters dated April 20 and July 15, 2016.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) consistent with Technical Specification
Task Force Traveler 422, Revision 2, ``Change in Technical
Specifications End States (CE NPSD-1186).''
Date of issuance: August 30, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 234 and 184. A publicly-available version is in
ADAMS under Accession No. ML16210A374; documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
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: November 24, 2015 (80
FR 73237). The supplemental letters dated April 20 and July 15, 2016,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 30, 2016.
No significant hazards consideration comments received: No.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold
Energy Center (DAEC), Linn County, Iowa
Date of amendment request: May 18, 2016.
Brief description of amendment: The amendment revised the DAEC
technical specifications (TSs) Section 2.1.1, ``Reactor Core [Safety
Limits],'' to change the Safety Limit Minimum Critical Power Ratio
(SLMCPR) for two recirculation loop operation and for single
recirculation loop operation. The changes reflected the cycle-specific
analysis. The amendment also removed an outdated historical footnote
from TS Table 3.3.5.1-1.
Date of issuance: September 12, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 297. A publicly-available version is in ADAMS under
Accession No. ML16211A514; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-49: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: July 5, 2016 (81 FR
43665).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 12, 2016.
No significant hazards consideration comments received: No.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold
Energy Center, Linn County, Iowa
Date of amendment request: August 18, 2015, as supplemented by
letters dated January 29, April 14, and May 31, 2016.
Brief description of amendment: The amendment revised Technical
Specification (TS) 5.5.12, ``Primary Containment Leakage Rate Testing
Program,'' to state that the program shall be in accordance with
Nuclear Energy Institute (NEI) 94-01, Revision 3-A, ``Industry
Guideline for Implementing Performance-Based Option of 10 CFR part 50,
appendix J.''
Date of issuance: August 30, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 296. A publicly-available version is in ADAMS under
Accession No. ML16210A008; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendments.
Renewed Facility Operating License No. DPR-49: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: October 27, 2015 (80 FR
65814). The supplemental letters dated January 29, April 14, and May
31, 2016, provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the staff's original proposed no
significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 30, 2016.
No significant hazards consideration comments received: No.
NextEra Energy, Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: June 26, 2013, as supplemented by
letters dated September 16, 2013, July 29, August 28, September 25,
November 14, December 19, 2014; January 16, May 12, August 26, 2015;
and February 22, April 7, and May 3, 2016.
Brief description of amendments: The amendments authorized the
transition of the Point Beach fire protection program to a risk-
informed, performance-based program based on National Fire Protection
Association Standard 805 (NFPA 805), ``Performance-Based Standard for
Fire Protection for Light Water Reactor Electric Generating Plants,''
2001 Edition, in accordance with 10 CFR 50.48(c).
Date of issuance: September 8, 2016.
Effective date: As of the date of issuance and shall be implemented
as described in the Transition License Conditions.
Amendment Nos.: 256 and 260. A publicly-available version is in
ADAMS under Accession No. ML16196A093;
[[Page 66314]]
documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-24 and DPR-27:
Amendments revised the Facility Operating License and Technical
Specifications.
Date of initial notice in Federal Register: July 8, 2014 (79 FR
28580). The supplemental letters dated September 16, 2013, July 29,
August 28, September 25, November 14, December 19, 2014; January 16,
May 12, August 26, 2015; and February 22, April 7, and May 3, 2016,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 8, 2016.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-391, Watts Bar Nuclear Plant,
Unit 2, Rhea County, Tennessee
Date of amendment request: December 15, 2015, as supplemented by
letters dated May 4, 2016, and June 1, 2016.
Brief description of amendment: The amendment revised the Technical
Specifications to allow implementation of the F* (F-star) alternate
repair criterion for steam generator tubes.
Date of issuance: September 6, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 2. A publicly-available version is in ADAMS under
Accession No. ML16203A365; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-96: Amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: February 16, 2016 (81
FR 7844). The supplemental letters dated May 4, 2016, and June 1, 2016,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 6, 2016.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 19th day of September 2016.
For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2016-23097 Filed 9-26-16; 8:45 am]
BILLING CODE 7590-01-P