[Federal Register Volume 81, Number 216 (Tuesday, November 8, 2016)]
[Notices]
[Pages 78643-78658]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-26824]
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NUCLEAR REGULATORY COMMISSION
[NRC-2016-0226]
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
[[Page 78644]]
grants the Commission the authority to issue and make immediately
effective any amendment to an operating license or combined license, as
applicable, upon a determination by the Commission that such amendment
involves no significant hazards consideration, notwithstanding the
pendency before the Commission of a request for a hearing from any
person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued, from October 8, 2016, to October 24, 2016. The
last biweekly notice was published on October 25, 2016.
DATES: Comments must be filed by December 8, 2016. A request for a
hearing must be filed by January 9, 2017.
ADDRESSES: You may submit comments by any of the following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0226. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions, contact
the individual or individuals 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: Lynn Ronewicz, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC
20555-0001; telephone: 301-415-1927, email: [email protected].
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2016-0226, 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-2016-0226.
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-2016-0226, facility name, unit
number(s), plant docket number, application date, and subject in your
comment submission.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. The NRC will post all comment submissions at http://www.regulations.gov, as well as enter the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment into ADAMS.
I. 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 the action. Petitions shall be filed in accordance with
the Commission's ``Agency Rules of Practice and Procedure'' in 10 CFR
part 2. Interested persons should consult a current copy of 10 CFR
2.309, which is available at the NRC's PDR, located at One White Flint
North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville,
Maryland 20852. The NRC's regulations are accessible electronically
from the NRC Library on the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a petition is filed within 60 days,
the Commission or a presiding officer designated by the Commission or
by the Chief Administrative Judge of the Atomic
[[Page 78645]]
Safety and Licensing Board Panel, will rule on the petition; and the
Secretary or the Chief Administrative Judge of the Atomic Safety and
Licensing Board will issue a notice of a hearing or an appropriate
order.
As required by 10 CFR 2.309, a petition shall set forth with
particularity the interest of the petitioner in the proceeding, and how
that interest may be affected by the results of the proceeding. The
petition should specifically explain the reasons why intervention
should be permitted with particular reference to the following general
requirements: (1) The name, address, and telephone number of the
petitioner; (2) the nature of the petitioner's right under the Act to
be made a party to the proceeding; (3) the nature and extent of the
petitioner's property, financial, or other interest in the proceeding;
and (4) the possible effect of any decision or order which may be
entered in the proceeding on the petitioner's interest. The petition
must also set forth the specific contentions which the petitioner seeks
to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner shall provide a brief explanation of the bases for the
contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion to support
its position on the issue. The petition must include sufficient
information to show that a genuine dispute exists with the applicant on
a material issue of law or fact. Contentions shall be limited to
matters within the scope of the proceeding. The contention must be one
which, if proven, would entitle the petitioner to relief. A petitioner
who fails to satisfy these requirements with respect to at least one
contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing with respect to resolution of that person's admitted
contentions consistent with the NRC's regulations, policies, and
procedures.
Petitions for leave to intervene must be filed no later than 60
days from the date of publication of this notice. Requests for hearing,
petitions for leave to intervene, and motions for leave to file new or
amended contentions that are filed after the 60-day deadline will not
be entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the three factors in 10
CFR 2.309(c)(1)(i) through (iii).
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, then any hearing held would take place before
the issuance of any amendment unless the Commission finds an imminent
danger to the health or safety of the public, in which case it will
issue an appropriate order or rule under 10 CFR part 2.
A State, local governmental body, Federally-recognized Indian
Tribe, or agency thereof, may submit a petition to the Commission to
participate as a party under 10 CFR 2.309(h)(1).
The petition should state the nature and extent of the petitioner's
interest in the proceeding. The petition should be submitted to the
Commission by January 9, 2017. 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 78646]]
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 to respect 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.
Dominion Energy Kewaunee, Inc., Docket No. 50-305, Kewaunee Power
Station (KPS), Carlton, Wisconsin
Date of amendment request: September 14, 2015. A publicly available
version is in ADAMS under Accession No. ML15261A236.
Description of amendment request: The amendment would revise the
Operating License and associated Technical Specifications to reflect
removal of all KPS spent nuclear fuel from the spent fuel pool and its
transfer to dry cask storage within an Independent Spent Fuel Storage
Installation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment would modify the KPS renewed facility
operating license and Technical Specification (TS) by deleting the
portions of the license and TS that are no longer applicable to a
facility with no spent nuclear fuel stored in the spent fuel pool,
while modifying the remaining portions to correspond to all nuclear
fuel stored within an Independent Spent Fuel Storage Installation
(ISFSI). This amendment becomes effective upon removal of all spent
nuclear fuel from the KPS spent fuel pool and its transfer to dry
cask storage within an ISFSI.
The definition of safety-related structures, systems, and
components (SSCs) in 10 CFR 50.2 states that safety-related SSCs are
those relied on to remain functional during and following design
basis events to assure:
1. The integrity of the reactor coolant boundary;
2. The capability to shutdown the reactor and maintain it in a
safe shutdown condition; or
3. The capability to prevent or mitigate the consequences of
accidents which could result in potential offsite exposures
comparable to the applicable guideline exposures set forth in 10 CFR
50.43(a)(1) or 100.11.
The first two criteria (integrity of the reactor coolant
pressure boundary and safe shutdown of the reactor) are not
applicable to a plant in a permanently defueled condition. The third
criterion is related to preventing or mitigating the consequences of
accidents that could result in potential offsite exposures exceeding
limits. However, after all nuclear spent fuel assemblies have been
transferred to dry cask storage within an ISFSI, none of the SSCs at
KPS are required to be relied on for accident mitigation. Therefore,
none of the SSCs at KPS meet the definition of a safety-related SSC
stated in 10 CFR 50.2. The proposed deletion of
[[Page 78647]]
requirements in the TS does not affect systems credited in any
accident analysis at KPS.
Section 14 of the KPS Updated Safety Analysis Report (USAR)
described the design basis accidents related to the spent fuel pool.
These postulated accidents are predicated on spent fuel being stored
in the spent fuel pool. With the removal of the spent fuel from the
spent fuel pool, there are no remaining spent fuel assemblies to be
monitored and there are no credible accidents that require the
actions of a Certified Fuel Handler, Shift Manager, or a Non-
certified Operator to prevent occurrence or mitigate the
consequences of an accident.
The proposed changes do not have an adverse impact on the
remaining decommissioning activities or any of their postulated
consequences.
The proposed changes related to the relocation of certain
administrative requirements do not affect operating procedures or
administrative controls that have the function of preventing or
mitigating any accidents applicable to the safe management of
irradiated fuel or decommissioning of the facility.
Therefore, the proposed amendment does not involve a significant
increase in the consequences of a previously evaluated accident.
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 changes eliminate the operational requirements and
certain design requirements associated with the storage of the spent
fuel in the spent fuel pool, and relocate certain administrative
controls to the Quality Assurance Program Description.
After the removal of the spent fuel from the spent fuel pool and
transfer to the ISFSI, there are no spent fuel assemblies that
remain in the spent fuel pool. Coupled with a prohibition against
storage of fuel in the spent fuel pool, the potential for fuel
related accidents is removed. The proposed changes do not introduce
any new failure modes.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Do the proposed changes involve a significant reduction in
the margin of safety?
Response: No.
The removal of all spent nuclear fuel from the spent fuel pool
into storage in casks within an ISFSI, coupled with a prohibition
against future storage of fuel within the spent fuel pool, removes
the potential for fuel related accidents.
The design basis and accident assumptions within the KPS USAR
and the TS relating to safe management and safety of spent fuel in
the spent fuel pool are no longer applicable. The proposed changes
do not affect remaining plant operations, systems, or components
supporting decommissioning activities.
The requirements for systems, structures, and components (SSCs)
that have been deleted from the KPS TS are not credited in the
existing accident analysis for any applicable postulated accident;
and as such, do not contribute to the margin of safety associated
with the accident analysis.
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: Lillian M. Cuoco, Senior Counsel, Dominion
Resource Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219.
NRC Branch Chief: Bruce A. Watson.
Entergy Operations, Inc. (Entergy), Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: July 25, 2016. A publicly-available
version is in ADAMS under Accession No. ML16207A532.
Description of amendment request: The amendment would revise the
Waterford Steam Electric Station, Unit 3 (Waterford 3), Technical
Specifications (TSs) Section 6.5.8, ``Inservice Testing Program,'' to
remove requirements duplicated in the American Society of Mechanical
Engineers Code for Operation and Maintenance of Nuclear Power Plants
Case OMN-20, ``Inservice Test Frequency.'' A new defined term,
``Inservice Testing Program,'' will be added to the TS 1.0,
``Definitions,'' section. The licensee states that the proposed change
to the TS is consistent with Technical Specifications Task Force (TSTF)
Traveler TSTF-545, Revision 3, ``TS Inservice Testing Program Removal &
Clarify SR Usage Rule Application to Section 5.5 Testing'' (ADAMS
Accession No. ML15294A555). However, the Waterford 3 TSs (NUREG-0973)
are of an older standard version and have not been converted to the
Improved Standard Technical Specifications (ISTSs). Therefore, Entergy
has included in the application a table of TSs affected by the
amendment, with variations and differences between the Waterford 3 TSs
and the ISTSs listed in TSTF-545 discussed individually.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below with NRC staff edits in square
brackets:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises TS Chapter 6, ``Administrative
Controls,'' Section 6.5, ``Programs'' by eliminating the ``Inservice
Testing Program'' specification. Most requirements in the IST
Program are removed, as they are duplicative of requirements in the
ASME OM Code [American Society of Mechanical Engineers Code for
Operation and Maintenance of Nuclear Power Plants], as clarified by
Code Case OMN-20, ``Inservice Test Frequency.'' The remaining
requirements in the Section 6.5.8, IST Program are eliminated [. .
.]. 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 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
[[Page 78648]]
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 Surveillance Requirement (SR) 4.0.3
(referenced as SR 3.0.3 in the ISTS [improved standard technical
specification]) 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. [. . .] 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 B. Glew, Jr., Associate General
Counsel--Entergy Services, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Acting Branch Chief: Stephen S. Koenick.
Exelon Generation Company, LLC, Docket No. 50-244, R.E. Ginna Nuclear
Power Plant, Wayne County, New York
Date of amendment request: August 22, 2016. A publicly available
version is in ADAMS under Accession No. ML16236A300.
Description of amendment request: The amendment would (1) revise
Technical Specification (TS) 4.2.1, ``Reactor Core, Fuel Assemblies,''
to add Optimized ZIRLO\TM\ as an approved fuel rod cladding material,
(2) revise TS 5.6.5.b to add the Westinghouse topical reports for
Optimized ZIRLO\TM\ and ZIRLO[supreg], and (3) revise TS 5.6.5.b with a
non-technical change to the Reference 11 title (replace a semicolon
with a period).
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 would allow the use of Optimized
ZIRLOTM clad nuclear fuel in the reactors. The NRC
approved topical report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-
A, ``Optimized ZIRLOTM,'' prepared by Westinghouse
Electric Company LLC (Westinghouse), addresses Optimized ZIRLO and
demonstrates that Optimized ZIRLOTM has essentially the
same properties as currently licensed ZIRLO[supreg]. The fuel
cladding itself is not an accident initiator and does not affect
accident probability. With the approved exemption, use of Optimized
ZIRLOTM fuel cladding will continue to meet all 10 CFR
50.46 acceptance criteria and, therefore, will not increase the
consequences of an accident.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Use of Optimized ZIRLO\TM\ clad fuel will not result in changes
in the operation or configuration of the facility. Topical Report
WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, demonstrated that the
material properties of Optimized ZIRLOTM are similar to
those of standard ZIRLO[supreg]. Therefore, Optimized
ZIRLOTM fuel rod cladding will perform similarly to those
fabricated from standard ZIRLO[supreg], thus precluding the
possibility of the fuel cladding becoming an accident initiator and
causing a new or different type of accident. Therefore, the proposed
change does not create the possibility of a new or different kind of
accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will not involve a significant reduction in
the margin of safety. Topical Report WCAP-12610-P-A & CENPD-404-P-A,
Addendum 1-A, demonstrated that the material properties of the
Optimized ZIRLO\TM\ are not significantly different from those of
standard ZIRLO[supreg]. Optimized ZIRLO\TM\ is expected to perform
similarly to standard ZIRLO[supreg] for all normal operating and
accident scenarios, including both loss of coolant accident (LOCA)
and non-LOCA scenarios. For LOCA scenarios, where the slight
difference is Optimized ZIRLO\TM\ material properties relative to
standard ZIRLO[supreg] could have some impact on the overall
accident scenario, plant-specific LOCA analyses using Optimized
ZIRLOTM properties will demonstrate that the acceptance
criteria of 10 CFR 50.46 have been satisfied.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Branch Chief: Travis L. Tate.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2, Will County, Illinois
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert
Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station, Unit No. 1, DeWitt County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-220 and 50-410, Nine
Mile Point Nuclear Station, Units 1 and 2, Oswego County, New York
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
and Lancaster Counties, Pennsylvania
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Exelon Generation Company, LLC, Docket No. 50-244, R.E. Ginna Nuclear
Power Plant, Wayne County, New York
Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1, Dauphin County, Pennsylvania
Date of amendment request: July 26, 2016, as supplemented by letter
dated October 6, 2016. Publicly-available versions are in ADAMS under
[[Page 78649]]
Accession Nos. ML16209A218 and ML16280A402, respectively.
Description of amendment request: The amendments would revise the
Inservice Testing Program requirements in each plant's technical
specifications (TSs). For each plant, the changes include deleting the
current TS for the Inservice Testing Program, adding a new defined
term, ``INSERVICE TESTING PROGAM,'' to the TSs, and revising other TSs
to reference this new defined term instead of the deleted TS. The
licensee stated that the proposed changes are based on Technical
Specifications Task Force (TSTF) Traveler TSTF-545, Revision 3, ``TS
Inservice Testing Program Removal & Clarify SR Usage Rule Application
to Section 5.5 Testing'' (ADAMS Accession No. ML15294A555), with some
variations.
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. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's analysis is
presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises TS Chapter 5, ``Administrative
Controls,'' Section 5.5, ``Programs and Manuals,'' or equivalent, by
deleting the ``lnservice Testing Program'' specification. A new
defined term, ``INSERVICE TESTING PROGRAM,'' is added to the TS,
which references the requirements of 10 CFR 50.55a(f), ``Inservice
testing requirements.'' The regulations in 10 CFR 50.55a(f) require
that specified pumps and valves meet the inservice test requirements
in the American Society of Mechanical Engineers (ASME) Code for
Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and
addenda. Most requirements currently in the TS lnservice Testing
Program are duplicative of requirements in the ASME OM Code and
addenda, as modified by NRC-approved alternatives or reliefs. The
proposed change primarily affects the required frequency for
performing ASME OM Code required tests for pumps and valves which
are covered by the Inservice Testing Program. The proposed change
would allow a longer interval between some tests and require a
shorter interval between other tests; the effect of the change to
specific test intervals depends on the plant-specific licensing
basis.
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. Changing the required test frequency of pumps and
valves will not affect the ability of the components to mitigate any
accident previously evaluated, as the components are required to be
operable. If components required by the TSs are found to be
inoperable, the TSs specify the actions required to ensure safe
operation of the facility, and these actions are not altered by the
proposed change. Performance of inservice tests in accordance with
the ASME OM Code, as modified by NRC-approved alternatives or
reliefs, 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 significantly 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 amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not alter the 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. Changes to 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 amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change eliminates some requirements from the TSs in
lieu of requirements in the ASME OM Code, as modified by NRC-
approved alternatives or reliefs. Compliance with the ASME OM Code
is required by 10 CFR 50.55a. Changes to the required test frequency
will not affect the ability of the components to respond to an
accident, as the components are required to be operable. The
proposed change also eliminates a provision which allowed, under
certain circumstances, the licensee to delay declaring equipment
inoperable due to a missed surveillance. This change will not have a
significant effect on plant operation or safety, as the licensee
will still be required by TSs to assess component operability. If
components required by the TSs are found to be inoperable, the TSs
specify the actions required to ensure safe operation of the
facility, and these actions are not altered by the proposed change.
The proposed change also eliminates a statement that nothing in the
ASME OM Code should be construed to supersede the requirements of
any TS. Elimination of the statement will not have a significant
effect on plant operation or safety. Therefore, the proposed change
does not involve a significant reduction in a margin of safety.
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: G. Edward Miller.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station (CNS), Nemaha County, Nebraska
Date of amendment request: August 26, 2016. A publicly-available
version is in ADAMS under Accession No. ML16245A288.
Description of amendment request: The amendment would revise the
CNS Technical Specifications (TSs) to eliminate Section 5.5.6,
``Inservice Testing [IST] Program,'' to remove requirements duplicated
in the American Society of Mechanical Engineers Code for Operation and
Maintenance of Nuclear Power Plants (ASME OM Code) Case OMN-20,
``Inservice Test Frequency.'' A new defined term, ``Inservice Testing
Program,'' will be added to TS Section 1.1, ``Definitions.'' The
licensee stated that the proposed change to the TSs is consistent with
Technical Specifications Task Force (TSTF) Traveler TSTF-545, Revision
3, ``TS Inservice Testing Program Removal & Clarify SR Usage Rule
Application to Section 5.5 Testing'' (ADAMS Accession No. ML15294A555),
with no proposed technical variations or deviations. However, in some
cases, the CNS TSs use different section titles or numbering for
surveillance requirements than the Standard Technical Specifications on
which TSTF-545 was based, so the licensee changed the TSTF-545
numbering to be consistent with the CNS TS numbering.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below with NRC staff edits in [square
brackets]:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change 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.'' The remaining
requirements in the Section 5.5 IST Program
[[Page 78650]]
are eliminated [. . .]. 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 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. [. . .] 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: Mr. John C. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Acting Branch Chief: Stephen S. Koenick.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1 (FCS), Washington County, Nebraska
Date of amendment request: September 2, 2016. A publicly-available
version is in ADAMS under Accession No. ML16246A321.
Description of amendment request: The amendment would revise the
Nuclear Radiological Emergency Response Plan (RERP) for FCS for the
plant condition following permanent cessation of power operations and
defueling. The proposed FCS RERP changes would revise the shift
staffing and Emergency Response Organization (ERO) staffing.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to the FCS RERP do not impact the function
of plant structures, systems, or components (SSCs). The proposed
changes do not affect accident initiators or precursors, nor does it
alter design assumptions. The proposed changes do not prevent the
ability of the on-shift staff and ERO to perform their intended
functions to mitigate the consequences of any accident or event that
will be credible in the permanently defueled condition. The proposed
changes only remove positions that will no longer be credited in the
FCS RERP.
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 reduce the number of on-shift and ERO
positions commensurate with the hazards associated with a
permanently shut down and defueled facility. The proposed changes do
not involve installation of new equipment or modification of
existing equipment, so that no new equipment failure modes are
introduced. Also, the proposed changes do not result in a change to
the way that the equipment or facility is operated so that no new
accident initiators are created.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Margin of safety is associated with confidence in the ability of
the fission product barriers (i.e., fuel cladding, reactor coolant
system pressure boundary, and containment structure) to limit the
level of radiation dose to the public. The proposed changes are
associated with the FCS RERP staffing and do not impact operation of
the plant or its response to transients or accidents. The change
does not affect the Technical Specifications. The proposed changes
do not involve a change in the method of plant operation, and no
accident analyses will be affected by the proposed changes. Safety
analysis acceptance criteria are not affected by the proposed
changes. The revised FCS RRP will continue to provide the necessary
response staff with the proposed changes.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700
K Street NW., Washington, DC 20006-3817.
NRC Acting Branch Chief: Stephen S. Koenick.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: September 28, 2016. A publicly-
[[Page 78651]]
available version is in ADAMS under Accession No. ML16273A502.
Description of amendment request: The amendment would modify the
Technical Specifications to make administrative changes to align
staffing for decommissioning Fort Calhoun Station, Unit No. 1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes only impact administrative requirements
associated with staff qualification, staff titles, personnel
staffing levels, and clarification of systems used during
decommissioning. The proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated because: (1) The proposed amendment does not
represent a change to any system design, (2) the proposed amendment
does not alter, degrade, or prevent action described or assumed in
any accident in the USAR [updated safety analysis report] from being
performed, (3) the proposed amendment does not alter any assumptions
previously made in evaluating radiological consequences, and [(4)]
the proposed amendment does not affect the integrity of any fission
product barrier. No safety related equipment is affected by the
proposed change.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes do not alter the physical design, safety
limits, or safety analysis assumptions associated with the operation
of the plant. Hence, the proposed changes do not introduce any new
accident initiators, nor do these changes reduce or adversely affect
the capabilities of any plant structure or system in the performance
of their safety function.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes do not alter the manner in which safety
limits or limiting safety system settings are determined. The safety
analysis acceptance criteria are not affected by these proposed
changes. Further, the proposed changes do not change the design
function of any equipment assumed to operate in the event of an
accident.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700
K Street NW., Washington, DC 20006-3817.
NRC Acting Branch Chief: Stephen S. Koenick.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Generating
Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: August 30, 2016. A publicly-available
version is in ADAMS under Accession No. ML16243A233.
Description of amendment request: The amendments would revise the
Salem Generating Station, Unit Nos. 1 and 2 (Salem), Technical
Specifications (TSs), Section 6.8.4.j, ``Inservice Testing Program,''
to remove requirements duplicated in the American Society of Mechanical
Engineers (ASME) Code for Operation and Maintenance of Nuclear Power
Plants (OM Code) Case OMN-20, ``Inservice Test Frequency.'' A new
defined term, ``Inservice Testing Program,'' will be added to the TS
1.0, ``Definitions,'' section. The licensee stated that the proposed
change to the TS is consistent with Technical Specifications Task Force
(TSTF) Traveler TSTF-545, Revision 3, ``TS Inservice Testing Program
Removal & Clarify SR Usage Rule Application to Section 5.5 Testing''
(ADAMS Accession No. ML15294A555). However, the Salem TSs use different
numbering than the Standard Technical Specifications on which TSTF-545
was based, so the licensee changed the TSTF-545 numbering to be
consistent with the Salem TS numbering.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below with NRC staff edits in square
brackets:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises TS Chapter 6, ``Administrative
Controls,'' Section 6.8, ``Procedures and Programs,'' 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.'' The remaining
requirements in the Section 6.8 IST [Inservice Testing] Program are
eliminated [. . .]. 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 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
[[Page 78652]]
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 4.0.3 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. [. . .] 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: Jeffrie J. Keenan, PSEG Nuclear LLC-N21,
P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Douglas A. Broaddus.
South Carolina Electric & Gas Company and South Carolina Public Service
Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear
Station, Units 2 and 3, Fairfield, South Carolina
Date of amendment request: September 22, 2016. A publicly-available
version is in ADAMS under Accession No. ML16270A582.
Description of amendment request: The changes would amend Combined
License Nos. NPF-93 and NPF-94 for the Virgil C. Summer Nuclear
Station, Units 2 and 3, respectively. The amendments propose changes to
the Updated Final Safety Analysis Report (UFSAR) in the form of
departures from the incorporated plant-specific Design Control Document
Tier 2 information and involve related changes to the Combined
Operating License Appendix C (and corresponding plant-specific design
control document Tier 1) information. Specifically, the proposed
departures consist of changes to the design reliability assurance
program (D-RAP) to identify the covers for the in-containment refueling
water storage tank vents and overflow weirs as the risk-significant
components included in the D-RAP and to differentiate between the rod
drive motor-generator (MG) sets field control relays and the rod drive
power supply control cabinets in which the relays are located.
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 in-containment refueling water storage tank (IRWST) provides
flooding of the refueling cavity for normal refueling. The tank also
serves as a heat sink during Passive Residual Heat Removal (PRHR)
Heat Exchanger (HX) operation and in the event of a loss-of-coolant-
accident (LOCA) provides injection in support of long-term RCS
cooling. This activity adds normally closed covers to the IRWST
vents and overflow weirs to prevent debris from entering the tank,
prevent over-pressurization and accommodate volume and mass
increases in the tank. The vent and overflow weir covers open upon
differential pressures between the IRWST and containment.
The rod drive MG sets provide the power to the control rod drive
mechanisms through the reactor trip switchgear. This activity
revises the equipment description and equipment tag associated with
the risk-significant control relays which open to de-energize the
rod drive MG sets and permit rods to drop.
The proposed changes to add the IRWST vent and overflow weir
covers and to change the description of the equipment and equipment
tag related to the rod drive MG sets does not inhibit the SSCs from
performing their safety-related function. The design bases of the
IRWST vents and overflow weirs are not modified as a result of the
addition of the covers to the vents and overflow weirs and the
change to the control cabinet relay description and equipment tag.
This proposed amendment does not have an adverse impact on the
response to anticipated transients or postulated accident conditions
because the functions of the SSCs are not changed. Required IRWST
venting is not affected for any accident conditions. Required DAS
functions are not affected for any accident conditions. Safety-
related structure, system, component (SSC) or function is not
adversely affected by this change. The changes to include the IRWST
covers and to change the control cabinet relay description and tag
number do not involve an interface with any SSC accident initiator
or initiating sequence of events, and thus, the probabilities of the
accidents evaluated in the UFSAR are not affected. The proposed
changes do not involve a change to the predicted radiological
releases due to postulated accident conditions, thus, the
consequences of the accidents evaluated in the UFSAR are not
affected. Probabilistic Risk Assessment (PRA) modeling and analyses
associated with the SSCs are not impacted by this change.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes to the design of the IRWST vent and
overflow weir covers do not adversely affect any safety-related
equipment, and do not add any new interfaces to safety-related SSCs.
No system or design function or equipment qualification is affected
by these changes. The changes do not introduce a new failure mode,
malfunction or sequence of events that could affect plant safety or
safety-related equipment as the simplistic design of the cover
louvers and hinged flappers are not considered unique designs. No
new credible failure modes are introduced by the addition of the
covers.
The proposed changes to the description and equipment tag
associated with the risk-significant control relays for the rod
drive MG sets do not adversely affect any safety-related equipment,
and do not add any new interfaces to safety-related SSCs. No system
or design function or equipment qualification is affected by these
changes. The changes do not introduce a new failure mode,
malfunction or sequence of events that could affect plant safety or
safety-related equipment because the design function of the control
relays, control cabinets, or rod drive MG sets is not changed.
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 change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Kathryn M. Sutton, Morgan, Lewis & Bockius,
LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
[[Page 78653]]
NRC Branch Chief: Michael T. Markley.
Tennessee Valley Authority, Docket Nos. 50-390 and 50-391, Watts Bar
Nuclear Plant (WBN), Units 1 and 2, Rhea County, Tennessee
Date of amendment request: September 23, 2016. A publicly-available
version is in ADAMS under Accession No. ML16271A378.
Description of amendment request: The amendments would revise the
completion date for License Condition 2.C(9)b for Unit 1, and License
Condition 2.C(3) for Unit 2, regarding the date for completion of
permanent modifications to the Fort Loudoun Dam to prevent overtopping
due to the probable maximum flood. The change is needed to accommodate
the current Tennessee Department of Transportation schedule for
completion of highway construction that will facilitate access to
complete the modifications to the Fort Loudoun Dam.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to revise the completion date for License
Condition 2.C(9)b for WBN Unit 1 and License Condition 2.C(3) for
WBN Unit 2 regarding the completion of permanent modifications to
the Fort Loudoun Dam from February 1, 2017, to June 30, 2018, do not
affect the structures, systems, or components (SSCs) of the plant,
affect plant operations, or any design function or an analysis that
verifies the capability of an SSC to perform a design function. No
change is being made to any of the previously evaluated accidents in
the WBN Updated Final Safety Analysis Report (UFSAR).
The proposed changes do not (1) require physical changes to
plant SSCs; (2) prevent the safety function of any safety-related
system, structure, or component during a design basis event; (3)
alter, degrade, or prevent action described or assumed in any
accident described in the WBN UFSAR from being performed because the
safety-related SSCs are not modified; (4) alter any assumptions
previously made in evaluating radiological consequences; or (5)
affect the integrity of any fission product barrier.
Therefore, the proposed change does not involve a significant
increase in the probability or consequence 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 introduce any new accident causal
mechanisms, because no physical changes are being made to the plant,
nor do they affect any plant systems that are potential accident
initiators.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The margin of safety associated with the acceptance criteria of
any accident is unchanged. The proposed changes will have no effect
on the availability, operability, or performance of safety-related
systems and components.
The proposed change will not adversely affect the operation of
plant equipment or the function of equipment assumed in the accident
analysis.
The proposed amendment does not involve changes to any safety
analyses assumptions, safety limits, or limiting safety system
settings. The changes do not adversely affect plant-operating
margins or the reliability of equipment credited in the safety
analyses.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Sherry A. Quirk, Executive Vice President
and General Counsel, Tennessee Valley Authority, 400 West Summit Hill
Drive, 6A Tower West, Knoxville, TN 37902.
NRC Acting Branch Chief: Jeanne A. Dion.
II. Previously Published Notices of Consideration of Issuance of
Amendments to Facility Operating Licenses and Combined Licenses,
Proposed No Significant Hazards Consideration Determination, and
Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Florida Power & Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida
Date of amendment request: August 3, 2016, as supplemented by
letter dated October 4, 2016. Publicly available versions are in ADAMS
under Accession Nos. ML16230A003 and ML16291A495, respectively.
Description of amendment request: The amendments would revise the
Technical Specification (TS) requirements for the Control Room
Emergency Ventilation System (CREVS). The licensee proposed the changes
to align the CREVS TSs more closely with the applicable Standard
Technical Specifications. Consequently, the requirements to immediately
suspend irradiated fuel movement would be relocated, in most cases, to
coincide with the commencement of unit shutdown in the event the
allowable outage time (AOT) cannot be met for an inoperable CREVS
component or control room envelope (CRE) boundary. The proposed
amendments would also eliminate the TS Limiting Condition for Operation
Actions and Surveillance Requirements associated with the CREVS kitchen
and lavatory ventilation exhaust duct isolation dampers.
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.
Relocating the requirement to immediately suspend irradiated
fuel movement from the determination of inoperability to the
expiration of the AOT is consistent with the Westinghouse Standard
Technical Specifications (STS) for an inoperable CREVS train and
thereby establishes a commensurate level of safety. This change does
not impact the functioning of the fuel handling system and so does
not significantly increase the probability of a fuel handling
accident. The removal of the kitchen and lavatory area exhaust
damper requirements aligns the licensing basis with the current
design and enhances the reliability of the CRE. The CREVS is not an
initiator of an accident. Hence, neither of the proposed changes
increase the probability of an accident previously evaluated.
The proposed changes do not impair the CREVS' capability to
provide a protected environment from which operators can control the
Units for all postulated events in the presence of a single failure.
For an inoperable CRE boundary in any plant MODE, the suspension of
fuel movement for
[[Page 78654]]
the first 24 hours, during which the effectiveness of the mitigating
actions are verified, ensures no increase in the consequences of a
fuel handling accident. The proposed change aligns the licensing
bases for the kitchen and lavatory ventilation exhaust pathways with
a more reliable physical barrier design.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
Relocating the requirement to immediately suspend irradiated
fuel movement until expiration of the AOT is consistent with the
Westinghouse STS and hence does not introduce a new type of accident
than previously evaluated or change the methods governing normal
plant operation. Aligning the Control Room kitchen and lavatory
ventilation exhaust pathway licensing bases with their current
design does not introduce new failure modes for existing equipment
or result in any new limiting single failure modes. The proposed
changes do not challenge the performance or integrity of any safety-
related system.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes align the current CREVS TS ACTION(s) with
the Westinghouse STS and the licensing bases for the Control Room
kitchen and lavatory ventilation exhaust pathways with their current
design. As such, the proposed changes do not involve changes to any
safety analyses assumptions, safety limits, or limiting safety
system settings nor do they adversely impact plant operating margins
or the reliability of equipment credited in the safety analyses.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Blvd., MS LAW/JB,
Juno Beach, FL 33408-0420.
NRC Acting Branch Chief: Jeanne A. Dion.
III. Notice of Issuance of Amendments to Facility Operating Licenses
and Combined Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, proposed no
significant hazards consideration determination, and opportunity for a
hearing in connection with these actions, was published in the Federal
Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation, and/or Environmental Assessment, as
indicated. All of these items can be accessed as described in the
``Obtaining Information and Submitting Comments'' section of this
document.
Duke Energy Florida, Inc. (DEF), et al., Docket No. 50-302, Crystal
River Unit 3 Nuclear Generating Plant, Citrus County, Florida
Date of amendment request: September 22, 2015.
Brief description of amendment: The amendment approved the proposed
name change from Duke Energy Florida, Inc. to Duke Energy Florida, LLC.
Date of issuance: October 12, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 250.
Facility Operating License No. DPR-72: The amendment revised the
facility operating license.
Date of initial notice in Federal Register: August 16, 2016 (81 FR
54614).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 12, 2016.
No significant hazards consideration comments received: No.
Duke Energy Progress, LLC, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Hartsville, South Carolina
Date of amendment request: November 19, 2015, as supplemented by
letter dated August 18, 2016.
Brief description of amendment: The amendment revised the technical
specifications (TSs) to allow the extension of the Type A containment
test interval to 15 years and the extension of the Type B and Type C
test intervals for selected components to 120 months and 75 months,
respectively. The amendment also deleted from the TSs an already
implemented one-time extension of the Type A test frequency.
Date of issuance: October 11, 2016.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment No.: 247. A publicly-available version is in ADAMS under
Accession No. ML16201A195; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-23: Amendment revised
the Renewed Facility Operating License and TSs.
Date of initial notice in Federal Register: March 15, 2016 (81 FR
13841). The supplemental letter dated August 18, 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 October 11, 2016.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert
Cliffs Nuclear Power Plant, Units 1 and 2 (CCNPP 1 and 2), Calvert
County, Maryland
Date of amendment request: February 4, 2016.
[[Page 78655]]
Brief description of amendments: The amendments revised the CCNPP 1
and 2 Technical Specifications (TSs) to include Surveillance
Requirement (SR) 3.5.2.10 in the list of applicable surveillances of SR
3.5.3.1 as part of the implementation of Technical Specifications Task
Force (TSTF) Improved Standard Technical Specifications Change Traveler
TSTF-523, Revision 2, ``Generic Letter 2008-01, Managing Gas
Accumulation.''
Date of issuance: October 7, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 319 (Unit 1) and 297 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML16263A001; 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: May 24, 2016 (81 FR
32806).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 7, 2016.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
and Lancaster Counties, Pennsylvania
Date of amendment request: June 20, 2016, as supplemented by letter
dated August 11, 2016.
Brief description of amendments: The amendments revised Technical
Specification 3.8.3, ``Diesel Fuel Oil, Lube Oil, and Starting Air,''
to replace the required stored inventory of lube oil for the diesel
generators (specified in number of gallons) with inventory requirements
based on diesel generator operating time (specified in number of days).
The changes are based on Revision 1 to Technical Specifications Task
Force (TSTF) Improved Standard Technical Specifications Change Traveler
TSTF-501, ``Relocate Stored Fuel Oil and Lube Oil Volume Values to
Licensee Control.''
Date of issuance: October 14, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 310 (Unit 2) and 314 (Unit 3). A publicly-available
version is in ADAMS under Accession No. ML16235A405; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. DPR-44 and DPR-56: The
amendments revised the Renewed Facility Operating Licenses and
Technical Specifications.
Date of initial notice in Federal Register: July 19, 2016 (81 FR
46962). The supplemental letter dated August 11, 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 October 14, 2016.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station (DNPS), Unit Nos. 2 and 3, Grundy County,
Illinois
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station (QCNPS), Unit Nos. 1 and 2, Rock Island
County, Illinois
Date of amendment request: February 6, 2015, as supplemented by
letters dated September 1, 2015, and January 20, January 28, April 26,
June 22, and September 28, 2016.
Brief description of amendments: The amendments revised the
technical specifications (TSs) for both DNPS, Units Nos. 2 and 3, and
QCNPS, Unit Nos. 1 and 2, to support the use of AREVA nuclear fuel;
both facilities currently operate using a Westinghouse nuclear fuel
design. Specifically, the TSs for the core operating limits report (TS
5.6.5.b) are revised to include NRC-approved AREVA methodologies and to
delete methodologies no longer in use. The transient analyses take
credit for conservatism in the scram speed performance; therefore, a
new surveillance requirement (SR) associated with linear heat
generation rate (LHGR) is added to the TSs (SR 3.2.3.2). This
demonstrates scram speed distribution is consistent with that used in
the transient analyses. The TSs associated with the limiting condition
for operation (LCO 3.7.7) for the main turbine bypass system is revised
to include requirements to use the minimum critical power ratio limits
(LCO 3.2.2) and LHGR limits (LCO 3.2.3) during operations when at
greater than or equal to (=) 25 percent of rated thermal
power and the main turbine bypass system is inoperable.
To increase the margin to the maximum reactor pressure vessel (RPV)
acceptance criteria for certain anticipated transient without scram
(ATWS) transients, the SRs for the allowable value (AV) for the ATWS
recirculation pump trip (ATWS-RPT) on high RPV steam dome pressure are
modified (SR 3.3.4.1.4.b). The ATWS-RPT AV for DNPS, Unit Nos. 2 and 3,
is lowered to less than or equal to 1,198 pounds per square inch gauge
(psig). The ATWS-RPT AV for QCNPS, Unit Nos. 1 and 2, is lowered to
less than or equal to 1,195 psig.
Date of issuance: October 20, 2016.
Effective date: As of the date of issuance and shall be implemented
prior to entering into MODE 2 on the first plant startup following the
next refueling outage for each unit.
Amendment Nos.: 251 and 244 (DNPS, Unit Nos. 2 and 3) and 264 and
259 (GCNPS, Unit Nos. 1 and 2). A publicly-available version is in
ADAMS under Accession No. ML16221A061; documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Renewed Facility Operating License Nos. DPR-19, DPR-25, DPR-29, and
DPR-30: Amendments revised the Renewed Facility Operating Licenses and
TSs.
Date of initial notice in Federal Register: November 3, 2015 (80 FR
67800). The supplemental letters dated January 20, January 28, April
26, June 22, and September 28, 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 October 20, 2016.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of amendment request: December 14, 2015, as supplemented by
letters dated March 9, 2016, and June 1, 2016. Publicly-available
versions are in ADAMS under Accession Nos. ML15348A396, ML16069A217,
and ML16153A084, respectively.
Brief description of amendments: The amendments revised the design
bases in
[[Page 78656]]
the updated final safety analysis report to reflect the use of a new
criticality safety assessment for fuel channel bow/bulge methodology to
support the performance of criticality safety evaluation for ATRIUM-
10XM fuel design in the spent fuel pool.
Date of issuance: October 17, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 263 (Unit 1) and 258 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML16231A131; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. DPR-29 and DPR-30: The
amendments revised the Facility Operating Licenses.
Date of initial notice in Federal Register: May 3, 2016 (81 FR
26586). The March 9, 2016, supplement corrected a deficiency in the
Holtec affidavit in the original submittal and did not change the NRC
staff's initial proposed finding of no significant hazards
consideration. The June 1, 2016, supplement contained clarifying
information and did not change the NRC staff's initial proposed finding
of no significant hazards consideration.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 17, 2016.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant (CNP), Units 1 and 2, Berrien County, Michigan
Date of amendment request: November 14, 2014, as supplemented by
letters dated February 12, July 17, August 24, August 28, November 16,
and December 17, 2015, and February 19, May 6, July 12, and September
15, 2016.
Brief description of amendments: The amendments revised the CNP,
Units 1 and 2, Technical Specifications (TSs) by replacing the limit on
reactor coolant system (RCS) gross specific activity with a new limit
on RCS noble gas specific activity. The noble gas specific activity
limit is based on a new DOSE EQUIVALENT Xenon (Xe)-133 definition that
replaces the E Bar average disintegration energy definition. In
addition, the DOSE EQUIVALENT Iodine (I)-131 definition is revised to
allow the use of additional thyroid dose conversion factors. The
changes are consistent with NRC-approved industry Technical
Specifications Task Force (TSTF) Standard Technical Specification
change traveler, TSTF-490, Revision 0, ``Deletion of E-Bar Definition
and Revision to Reactor Coolant System Specific Activity Technical
Specification,'' with approved deviations. Additionally, the amendments
revised the CNP, Units 1 and 2, licensing basis and TSs to adopt the
alternative source term as allowed in 10 CFR 50.67.
Date of issuance: October 20, 2016.
Effective date: As of the date of issuance and shall be implemented
within 180 days of issuance.
Amendment Nos.: 332 for Unit 1 and 314 for Unit 2. A publicly-
available version is in ADAMS under Accession No. ML16242A111;
documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-58 and DPR-74:
Amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: March 31, 2015 (80 FR
17091). The supplemental letters dated July 17, August 24, August 28,
November 16, and December 17, 2015, and February 19, May 6, July 12,
and September 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 amendments is contained
in a Safety Evaluation dated October 20, 2016.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station (CNS), Nemaha County, Nebraska
Date of amendment request: April 21, 2016, as supplemented by
letter dated August 29, 2016.
Brief description of amendment: The amendment revised Section 2.0,
``Safety Limits (SLs),'' of the CNS Technical Specifications by
revising the two recirculation loop and single recirculation loop
safety limit minimum critical power ratio values to reflect the results
of a cycle-specific calculation.
Date of issuance: October 17, 2016.
Effective date: As of the date of issuance and shall be implemented
prior to startup from Refuel Outage 29.
Amendment No.: 257. A publicly-available version is in ADAMS under
Accession No. ML16272A137; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-46: The amendment
revised the Renewed Facility Operating License and Technical
Specifications.
Date of initial notice in Federal Register: July 5, 2016 (81 FR
43664). The supplemental letter dated August 29, 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 October 17, 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: October 14, 2015. A publicly-available
version is in ADAMS under Accession No. ML15289A233.
Brief description of amendment: The amendment revised the DAEC
Technical Specifications Section 5.5.6, ``Inservice Testing Program,''
to provide consistency with the requirements of 10 CFR 50.55a(f)(4) for
inservice testing of pumps and valves and remove requirements that are
redundant to the requirements of 10 CFR 50.55a.
Date of issuance: October 17, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 298. A publicly-available version is in ADAMS under
Accession No. ML16263A245; 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 Renewed Facility Operating License and Technical
Specifications.
Date of initial notice in Federal Register: December 22, 2015 (80
FR 79621).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 17, 2016.
No significant hazards consideration comments received: No.
[[Page 78657]]
Northern States Power Company--Minnesota, Docket No. 50-282, Prairie
Island Nuclear Generating Plant, Unit 1, Goodhue County, Minnesota
Date of amendment request: April 7, 2016.
Brief description of amendment: The amendment revised Technical
Specification (TS) Surveillance Requirement (SR) 3.8.4.3 to allow a
one-time extension of 1 month for the TS SR frequency.
Date of issuance: October 13, 2016.
Effective date: As of the date of issuance and shall be implemented
within 7 days of issuance.
Amendment No.: 218. A publicly-available version is in ADAMS under
Accession No. ML16256A514; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. DPR-42: Amendment revised the
Facility Operating License and TSs.
Date of initial notice in Federal Register: June 21, 2016 (81 FR
40360).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 13, 2016.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station
(HCGS), Salem County, New Jersey
Date of amendment request: June 8, 2016.
Brief description of amendment: The amendment revised the HCGS
Technical Specifications. Specifically, the safety limit minimum
critical power ratio for single recirculation loop operation is
revised. The change results from a cycle-specific analysis performed to
support the operation of HCGS in upcoming Cycle 21.
Date of issuance: October 13, 2016.
Effective date: As of the date of issuance and shall be implemented
prior to startup from the fall 2016 refueling outage.
Amendment No.: 200. A publicly-available version is in ADAMS under
Accession No. ML16270A038; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-57: The amendment
revised the Renewed Facility Operating License and Technical
Specifications.
Date of initial notice in Federal Register: August 2, 2016 (81 FR
50748).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 13, 2016.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Docket Nos. 50-348 and 50-364,
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama
Date of amendment request: April 25, 2016.
Brief description of amendments: The amendments updated Attachment
M, ``License Condition Changes''; Attachment S, ``Modification and
Implementation Items''; and Attachment W, ``Fire Probabilistic Risk
Analysis Insights,'' of the previously approved National Fire
Protection Association (NFPA) 805 amendment.
Date of issuance: October 17, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 205 (Unit 1) and 201 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML16232A000; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. NPF-2 and NPF-8: The
amendments revised the Renewed Facility Operating Licenses.
Date of initial notice in Federal Register: June 7, 2016 (81 FR
36623).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 17, 2016.
No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company and South Carolina Public Service
Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear
Station (VCSNS), Units 2 and 3, Fairfield County, South Carolina
Date of amendment request: January 19, 2016, as supplemented by
letter dated March 1, 2016.
Description of amendment: The amendments authorized changes to the
VCSNS, Units 2 and 3, Updated Final Safety Analysis Report in the form
of departures from the incorporated plant-specific Design Control
Document Tier 2* information. The changes are related to changes to
construction methods and construction sequence used for the composite
floors and roof of the auxiliary building.
Date of issuance: August 25, 2016.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: 51 (for Units 2 and 3). A publicly-available
version is in ADAMS under Package Accession No. ML16202A279; documents
related to these amendments are listed in the Safety Evaluation
enclosed with the amendments.
Facility Combined Licenses No. NPF-91 and NPF-92: Amendments
revised the Facility Combined Licenses.
Date of initial notice in Federal Register: March 15, 2016 (81 FR
13837). The supplemental letter dated March 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 August 25, 2016.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Unit Nos. 1 and 2, Louisa County, Virginia
Date of amendment request: December 10, 2016, as supplemented by
letter dated June 15, 2016.
Brief description of amendments: The amendments modified Technical
Specification (TS) 3.2.1, ``Heat Flux Hot Channel Factor
(FQ(Z)).'' The amendments relocate required operating space
reductions to the Core Operating Limits Report, accompanied by
verification for each reload cycle, and define TS surveillance
requirements for steady-state and transient FQ(Z) and
corresponding actions with which to apply an appropriate penalty factor
to measured results, as identified in Westinghouse Nuclear Safety
Advisory Letter (NSAL)-09-5, Revision 1, ``Relaxed Axial Offset Control
FQ Technical Specification Actions,'' and NSAL-15-1, Revision 0, ``Heat
Flux Hot Channel Factor Surveillance Requirements,'' respectively.
Date of issuance: October 17, 2016.
Effective date: As of the date of issuance and shall be implemented
before September 30, 2017.
Amendment Nos.: 278 (Unit No. 1) and 261 (Unit No. 2). A publicly
available version is in ADAMS under Accession No. ML16252A478;
documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments
revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: March 1, 2016 (81 FR
10682).
[[Page 78658]]
The supplemental letter dated June 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 amendments is contained
in a Safety Evaluation dated October 17, 2016.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 27th day of October, 2016.
For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2016-26824 Filed 11-7-16; 8:45 am]
BILLING CODE 7590-01-P