[Federal Register Volume 82, Number 116 (Monday, June 19, 2017)]
[Notices]
[Pages 27882-27895]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2017-12732]
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NUCLEAR REGULATORY COMMISSION
[NRC-2017-0140]
Biweekly Notice: Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
AGENCY: Nuclear Regulatory Commission.
ACTION: Biweekly notice.
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SUMMARY: Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954,
as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is
publishing this regular biweekly notice. The Act requires the
Commission to publish notice of any amendments issued, or proposed to
be issued, and grants the Commission the authority to issue and make
immediately effective any amendment to an operating license or combined
license, as applicable, upon a determination by the Commission that
such amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued, from May 23, 2017, to June 2, 2017. The last
biweekly notice was published on June 6, 2017.
DATES: Comments must be filed by July 19, 2017. A request for a hearing
must be filed by August 18, 2017.
ADDRESSES: You may submit comments by any of the following methods
(unless this document describes a different method for submitting
comments on a specific subject):
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0140. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions, contact
the individual listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
Mail comments to: Cindy Bladey, Office of Administration,
Mail Stop: TWFN-8-D36M, 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].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2017-0140 facility name, unit
number(s), plant docket number, application date, and subject when
contacting the NRC about the availability of information for this
action. You may obtain publicly available information related to this
action by any of the following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0140.
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/
[[Page 27883]]
adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, 301-415-4737, or by email to [email protected]. The
ADAMS accession number for each document referenced (if it is available
in ADAMS) is provided the first time that it is mentioned in this
document.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2017-0140 facility name, unit
number(s), plant docket number, application date, and subject in your
comment submission.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment into ADAMS.
II. Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses and Combined Licenses and Proposed No Significant
Hazards Consideration Determination
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Sec. 50.92 of title 10 of the Code of
Federal Regulations (10 CFR), this means that operation of the facility
in accordance with the proposed amendment would not (1) involve a
significant increase in the probability or consequences of an accident
previously evaluated, or (2) create the possibility of a new or
different kind of accident from any accident previously evaluated; or
(3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period if circumstances change during the 30-day comment
period such that failure to act in a timely way would result, for
example in derating or shutdown of the facility. If the Commission
takes action prior to the expiration of either the comment period or
the notice period, it will publish in the Federal Register a notice of
issuance. If the Commission makes a final no significant hazards
consideration determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
A. Opportunity To Request a Hearing and Petition for Leave To Intervene
Within 60 days after the date of publication of this notice, any
persons (petitioner) whose interest may be affected by this action may
file a request for a hearing and petition for leave to intervene
(petition) with respect to the action. Petitions shall be filed in
accordance with the Commission's ``Agency Rules of Practice and
Procedure'' in 10 CFR part 2. Interested persons should consult a
current copy of 10 CFR 2.309. The NRC's regulations are accessible
electronically from the NRC Library on the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of
the regulations is available at the NRC's Public Document Room, located
at One White Flint North, Room O1-F21, 11555 Rockville Pike (first
floor), Rockville, Maryland 20852. If a petition is filed, the
Commission or a presiding officer will rule on the petition and, if
appropriate, a notice of a hearing will be issued.
As required by 10 CFR 2.309(d) the petition should specifically
explain the reasons why intervention should be permitted with
particular reference to the following general requirements for
standing: (1) The name, address, and telephone number of the
petitioner; (2) the nature of the petitioner's right under the Act to
be made a party to the proceeding; (3) the nature and extent of the
petitioner's property, financial, or other interest in the proceeding;
and (4) the possible effect of any decision or order which may be
entered in the proceeding on the petitioner's interest.
In accordance with 10 CFR 2.309(f), the petition must also set
forth the specific contentions which the petitioner seeks to have
litigated in the proceeding. Each contention must consist of a specific
statement of the issue of law or fact to be raised or controverted. In
addition, the petitioner must provide a brief explanation of the bases
for the contention and a concise statement of the alleged facts or
expert opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to the specific sources and
documents on which the petitioner intends to rely to support its
position on the issue. The petition must include sufficient information
to show that a genuine dispute exists with the applicant or licensee on
a material issue of law or fact. Contentions must be limited to matters
within the scope of the proceeding. The contention must be one which,
if proven, would entitle the petitioner to relief. A petitioner who
fails to satisfy the requirements at 10 CFR 2.309(f) with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene.
Parties have the opportunity to participate fully in the conduct of the
hearing with respect to resolution of that party's admitted
contentions, including the opportunity to present evidence, consistent
with the NRC's regulations, policies, and procedures.
Petitions must be filed no later than 60 days from the date of
publication of this notice. Petitions and motions for leave to file new
or amended contentions that are filed after the deadline will not be
entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the three factors in 10
CFR 2.309(c)(1)(i) through (iii). The petition must be filed in
accordance with the filing instructions in the ``Electronic Submissions
(E-Filing)'' section of this document.
[[Page 27884]]
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to establish when the hearing is held. If the final determination is
that the amendment request involves no significant hazards
consideration, the Commission may issue the amendment and make it
immediately effective, notwithstanding the request for a hearing. Any
hearing would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, then any hearing held would take place before
the issuance of the amendment unless the Commission finds an imminent
danger to the health or safety of the public, in which case it will
issue an appropriate order or rule under 10 CFR part 2.
A State, local governmental body, Federally-recognized Indian
Tribe, or agency thereof, may submit a petition to the Commission to
participate as a party under 10 CFR 2.309(h)(1). The petition should
state the nature and extent of the petitioner's interest in the
proceeding. The petition should be submitted to the Commission by
August 18, 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. Alternatively, a State, local governmental body, Federally-
recognized Indian Tribe, or agency thereof may participate as a non-
party under 10 CFR 2.315(c).
If a hearing is granted, any person who is not a party to the
proceeding and is not affiliated with or represented by a party may, at
the discretion of the presiding officer, be permitted to make a limited
appearance pursuant to the provisions of 10 CFR 2.315(a). A person
making a limited appearance may make an oral or written statement of
his or her position on the issues but may not otherwise participate in
the proceeding. A limited appearance may be made at any session of the
hearing or at any prehearing conference, subject to the limits and
conditions as may be imposed by the presiding officer. Details
regarding the opportunity to make a limited appearance will be provided
by the presiding officer if such sessions are scheduled.
B. Electronic Submissions (E-Filing)
All documents filed in NRC adjudicatory proceedings, including a
request for hearing and petition for leave to intervene (petition), any
motion or other document filed in the proceeding prior to the
submission of a request for hearing or petition to intervene, and
documents filed by interested governmental entities that request to
participate under 10 CFR 2.315(c), must be filed in accordance with the
NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR
46562, August 3, 2012). The E-Filing process requires participants to
submit and serve all adjudicatory documents over the internet, or in
some cases to mail copies on electronic storage media. Detailed
guidance on making electronic submissions may be found in the Guidance
for Electronic Submissions to the NRC and on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html. Participants may not submit
paper copies of their filings unless they seek an exemption in
accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least 10
days prior to the filing deadline, the participant should contact the
Office of the Secretary by email at [email protected], or by
telephone at 301-415-1677, to (1) request a digital identification (ID)
certificate, which allows the participant (or its counsel or
representative) to digitally sign submissions and access the E-Filing
system for any proceeding in which it is participating; and (2) advise
the Secretary that the participant will be submitting a petition or
other adjudicatory document (even in instances in which the
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the
Secretary will establish an electronic docket for the hearing in this
proceeding if the Secretary has not already established an electronic
docket.
Information about applying for a digital ID certificate is
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. Once a participant has obtained a
digital ID certificate and a docket has been created, the participant
can then submit adjudicatory documents. Submissions must be in Portable
Document Format (PDF). Additional guidance on PDF submissions is
available on the NRC's public Web site at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the
time the document is submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an email notice confirming receipt of the document. The
E-Filing system also distributes an email notice that provides access
to the document to the NRC's Office of the General Counsel and any
others who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
document on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before adjudicatory documents are
filed so that they can obtain access to the documents via the E-Filing
system.
A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC's Electronic
Filing Help Desk through the ``Contact Us'' link located on the NRC's
public Web site at http://www.nrc.gov/site-help/e-submittals.html, by
email to [email protected], or by a toll-free call at 1-866-672-
7640. The NRC Electronic Filing Help Desk is available between 9 a.m.
and 6 p.m., Eastern Time, Monday through Friday, excluding government
holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
stating why there is good cause for not filing electronically and
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff. Participants filing adjudicatory documents in this
manner are responsible for serving the document on all other
participants. Filing is considered complete by first-class mail as of
the time of deposit in the mail, or
[[Page 27885]]
by courier, express mail, or expedited delivery service upon depositing
the document with the provider of the service. A presiding officer,
having granted an exemption request from using E-Filing, may require a
participant or party to use E-Filing if the presiding officer
subsequently determines that the reason for granting the exemption from
use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in the
NRC's electronic hearing docket which is available to the public at
https://adams.nrc.gov/ehd, unless excluded pursuant to an order of the
Commission or the presiding officer. If you do not have an NRC-issued
digital ID certificate as described above, click cancel when the link
requests certificates and you will be automatically directed to the
NRC's electronic hearing dockets where you will be able to access any
publicly available documents in a particular hearing docket.
Participants are requested not to include personal privacy information,
such as social security numbers, home addresses, or personal phone
numbers in their filings, unless an NRC regulation or other law
requires submission of such information. For example, in some
instances, individuals provide home addresses in order to demonstrate
proximity to a facility or site. With respect to copyrighted works,
except for limited excerpts that serve the purpose of the adjudicatory
filings and would constitute a Fair Use application, participants are
requested not to include copyrighted materials in their submission.
For further details with respect to these license amendment
applications, see the application for amendment which is available for
public inspection in ADAMS and at the NRC's PDR. For additional
direction on accessing information related to this document, see the
``Obtaining Information and Submitting Comments'' section of this
document.
Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian
Point Nuclear Generating Unit Nos. 2 and 3 (IP2 and IP3), Westchester
County, New York
Date of amendment request: December 14, 2016, as supplemented by
letter dated April 19, 2017. Publicly available versions are in ADAMS
under Package Accession No. ML16355A066 and Accession No. ML17114A467,
respectively.
Description of amendment request: The amendments would revise the
Appendix C Technical Specifications (TS) Limiting Condition for
Operation (LCO) 3.1.2 for IP2 and IP3 and Appendix A TS LCO 3.7.13 for
IP2. These LCOs ensure that the fuel to be loaded into the Shielded
Transfer Canister (STC) meets the design basis for the STC and has an
acceptable rack location in the IP2 spent fuel pit before the STC is
loaded with fuel. The proposed changes to these LCOs would increase the
population of IP3 fuel eligible for transfer to the IP2 spent fuel pit
and maintain full core offload capability for IP3.
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's edits in
square brackets:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment would modify the IP2 and IP3 Technical
Specifications (TS) to incorporate the results of revised
criticality, thermal, and shielding and dose analyses and
evaluations.
[For IP2,] the proposed amendment was evaluated for impact on
the following previously evaluated events and accidents: STC
Criticality Accidents, SFP Criticality Accidents, Boron Dilution
Accidents, Fuel Handling Accidents, Loss of Spent Fuel Pool [SFP]
Cooling, and Natural Events.
[IP2] STC Criticality Accidents
The STC criticality accident considered were: Abnormal
temperature, dropped, mislocated, and misloaded fuel assemblies, and
misalignment between the active fuel region and the neutron
absorber.
The probability of an STC criticality accident will not increase
significantly due to the proposed changes because the individual
fuel assemblies will be loaded into the STC in the same manner,
using the same equipment, procedures, and other administrative
controls (i.e. fuel move sheets) that are currently used.
The consequences of an STC criticality accident are not changed
because the reactivity analysis demonstrates that the same
subcriticality criteria and requirements continue to be met for
these accidents.
[IP2] SFP Criticality Accidents
The SFP criticality accident of record considered the following
accidents (1) a dropped fuel assembly or an assembly placed
alongside a rack, (2) a misloaded fuel assembly, and (3) abnormal
heat loads. Because the IP2 and IP3 fuel assemblies are identical
[with] regards [to] those parameters that are utilized in the design
basis criticality analysis (DBA) to qualify fresh fuel these
accidents are bounding for IP3 fuel.
The probability of an SFP criticality accident will not increase
significantly due to the proposed changes because the individual
fuel assemblies will be loaded into the SFP in the same manner,
using the same equipment, procedures, and other administrative
controls (i.e. fuel move sheets) that are currently used.
The consequences of an SFP criticality accident are not changed
because the reactivity analysis demonstrates that the same
subcriticality criteria and requirements continue to be met for this
accident.
[IP2] STC Thermal Accidents
The thermal analyses demonstrate that the postulated accidents
(rupture of the HI-TRAC water jacket, 50-gallon transported fuel
tank rupture and fire, simultaneous loss of water from the water
jacket and HI-TRAC annulus, fuel misload, hypothetical tipover, and
crane malfunction) continue to meet their acceptance criteria.
The probability of an STC thermal accident will not increase
significantly because the individual fuel assemblies will be loaded
into the SFP in the same manner, using the same equipment,
procedures, and other administrative controls (i.e. fuel move
sheets) that are currently used.
The consequences of an STC thermal accident will not increase
significantly because the thermal analysis demonstrates that the
same thermal acceptance criteria and requirements continue to be met
for this accident.
[IP2] Boron Dilution Accident
The probability of a boron dilution event remains the same
because the proposed change does not alter the manner in which the
IP2 spent fuel cooling system or any other plant system is operated,
or otherwise increase the likelihood of adding significant
quantities of unborated water into the spent fuel pit.
The consequences of the boron dilution event remains the same.
The reactivity of the STC filled with the most reactive combination
of approved fuel assemblies in unborated water results in a
keff less than 0.95. Thus, even in the unlikely event of
a complete dilution of the spent fuel pit water, the STC will remain
safely subcritical.
[IP2] Fuel Handling Accident
The probability of an FHA will not increase significantly due to
the proposed changes because the individual fuel assemblies will be
moved between the STC and the spent fuel pit racks and the STC and
HI-TRAC will be moved in the same manner, using the same equipment,
procedures, and other administrative controls (i.e. fuel move
sheets) that are currently used.
The consequences of the existing fuel handling accident remain
bounding because the IP3 fuel assembly design is essentially the
same as the IP2 design and the IP3 fuel assemblies to be transferred
to IP2 will be cooled a minimum of 6 years. This compares with a
cooling time of 84 hours used in the existing FHA radiological
analysis. The 6-year cooling time results in a significant reduction
in the radioactive source term available for release from a damaged
fuel assembly compared to the source term considered in the design
basis FHA radiological analysis. The consequences of the previously
analyzed fuel assembly drop accident, therefore, continue to provide
a
[[Page 27886]]
bounding estimate of offsite dose for this accident.
[IP2] Loss of Spent Fuel Pool Cooling
The probability of a loss of spent fuel pit cooling remains the
same because the proposed change does not alter the manner in which
the IP2 spent fuel cooling loop is operated, designed or maintained.
The consequences of a loss of spent fuel pit cooling remains the
same because the thermal design basis for the spent fuel pit cooling
loop provides for all fuel pit rack locations to be filled at the
end of a full core discharge and therefore the design basis heat
load effectively includes any heat load associated with the
assemblies within the STC.
[IP2] Natural Events
The natural events considered include the following accidents
(1) a seismic event, (2) high winds, tornado and tornado missiles,
(3) flooding and (4) a lightning strike.
The probability of natural event will not increase due to the
proposed changes because there are no elements of the proposed
changes that influence the occurrence of any natural event.
The consequences of a natural event will not increase due to the
proposed changes because the structural analyses design limits
continue to be met. A lightning strike may cause ignition of the VCT
fuel but this event is addressed under STC thermal accidents.
[For IP3,] the proposed amendment was evaluated for impact on
the following previously evaluated events and accidents: STC
Criticality Accidents, SFP Criticality Accidents, Boron Dilution
Accidents, Fuel Handling Accidents, Loss of Spent Fuel Pool Cooling,
and Natural Events.
[IP3] STC Criticality Accidents
The STC criticality accident considered were: Abnormal
temperature, dropped, mislocated, and misloaded fuel assemblies, and
misalignment between the active fuel region and the neutron
absorber.
The probability of an STC criticality accident will not increase
significantly due to the proposed changes because the individual
fuel assemblies will be loaded into the STC in the same manner,
using the same equipment, procedures, and other administrative
controls (i.e. fuel move sheets) that are currently used.
The consequences of an STC criticality accident are not changed
because the reactivity analysis demonstrates that the same
subcriticality criteria and requirements continue to be met for
these accidents.
[IP3] STC Thermal Accidents
The thermal analyses demonstrate that the postulated accidents
(rupture of the HI-TRAC water jacket, 50-gallon transported fuel
tank rupture and fire, simultaneous loss of water from the water
jacket and HI-TRAC annulus, fuel mislead, hypothetical tipover, and
crane malfunction) continue to meet their acceptance criteria. The
probability of an STC thermal accident will not increase
significantly because the individual fuel assemblies will be loaded
into the SFP in the same manner, using the same equipment,
procedures, and other administrative controls (i.e. fuel move
sheets) that are currently used.
The consequences of an STC thermal accident will not increase
significantly because the thermal analysis demonstrates that the
same thermal acceptance criteria and requirements continue to be met
for this accident.
[IP3] Boron Dilution Accident
The probability of a boron dilution event remains the same
because the proposed change does not alter the manner in which the
IP3 spent fuel cooling system or any other plant system is operated,
or otherwise increase the likelihood of adding significant
quantities of unborated water into the spent fuel pit.
The consequences of the boron dilution event remains the same.
The reactivity of the STC filled with the most reactive combination
of approved fuel assemblies in unborated water results in a
keff less than 0.95. Thus, even in the unlikely event of
a complete dilution of the spent fuel pit water, the STC will remain
safely subcritical.
[IP3] Fuel Handling Accident
The probability of an FHA will not increase significantly due to
the proposed changes because the individual fuel assemblies will be
moved between the STC and the spent fuel pit racks and the STC and
HI-TRAC will be moved in the same manner, using the same equipment,
procedures, and other administrative controls (i.e. fuel move
sheets) that are currently used.
The consequences of the existing fuel handling accident remain
bounding because only IP3 fuel is moved in the IP3 spent fuel pit.
The IP3 fuel assemblies to be transferred to IP2 will be cooled a
minimum of 6 years. This compares with a cooling time of 84 hours
used in the existing FHA radiological analysis. The 6-year cooling
time results in a significant reduction in the radioactive source
term available for release from a damaged fuel assembly compared to
the source term considered in the design basis FHA radiological
analysis. The consequences of the previously analyzed fuel assembly
drop accident, therefore, continue to provide a bounding estimate of
offsite dose for this accident.
[IP3] Loss of Spent Fuel Pool Cooling
The probability of a loss of spent fuel pit cooling remains the
same because the proposed change does not alter the manner in which
the IP3 spent fuel cooling loop is operated, designed or maintained.
The consequences of a loss of spent fuel pit cooling remains the
same because the thermal design basis for the spent fuel pit cooling
loop provides for all fuel pit rack locations to be filled at the
end of a full core discharge and therefore the design basis heat
load effectively includes any heat load associated with the
assemblies within the STC.
[IP3] Natural Events
The natural events considered include the following accidents
(1) a seismic event, (2) high winds, tornado and tornado missiles,
(3) flooding and (4) a lightning strike.
The probability of natural event will not increase due to the
proposed changes because there are no elements of the proposed
changes that influence the occurrence of any natural event.
The consequences of a natural event will not increase due to the
proposed changes because the structural analyses design limits
continue to be met. A lightning strike may cause ignition of the VCT
fuel but this event is addressed under STC thermal accidents.
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 TS changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated. No new modes of operation are introduced by the proposed
changes. The proposed changes will not create any failure mode not
bounded by previously evaluated accidents.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident, from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment would modify the TS to incorporate the
results of revised criticality, thermal and shield and dose
analyses. The margin of safety required by 10 CFR 50.58(b)(4)
remains unchanged. New criticality evaluations for both the STC [and
the IP2 SFP] confirm that operation in accordance with the proposed
amendment continues to meet the required subcriticality margins. The
thermal analyses demonstrate that the postulated accidents (rupture
of the HI-TRAC water jacket, 50-gallon transported fuel tank rupture
and fire, simultaneous loss of water from the water jacket and HI-
TRAC annulus, fuel misload, hypothetical tipover, and crane
malfunction) continue to meet their acceptance criteria without a
significant loss of safety margin. The shielding and dose analyses
demonstrate that the shielding and radiation protection requirements
continue to be met without a significant loss of safety margin.
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: Jeanne Cho, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Branch Chief: James G. Danna.
[[Page 27887]]
Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert
Cliffs Nuclear Power Plant, Units 1 and 2, Calvert County, Maryland
Date of amendment request: March 28, 2017. A publicly available
version is in ADAMS under Accession No. ML17087A374.
Description of amendment request: The amendments would revise the
Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Technical
Specifications (TSs) to change the low level of the refueling water
tank (RWT) to reflect a needed increase in the required borated water
volume and change the allowable value of the RWT level-low function.
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 any accident previously
evaluated?
Response: No.
The proposed amendment increases the required volume of water in
the RWT to maintain the existing design requirements. The increase
is necessary due to an increase in the RWT Level--Low RAS
[recirculation actuation signal] setpoint, which allows more water
to stay in the tank following a LOCA [loss-of-coolant accident]. The
modification to the allowable value of the RWT level-low (function
5a) resolves a non-conservative TS per the guidance of
Administrative Letter 98-10 ``Dispositioning of Technical
Specifications That Are Insufficient to Assure Plant Safety.''
The RWT is not an accident initiator. The RWT is required to
supply adequate borated water to perform its mitigation function as
assumed in the accident analyses. With the proposed increase in the
minimum required water volume, the RWT maintains its design margin
for supplying the required amount of borated water to the reactor
core and the containment sump.
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 previously evaluated?
Response: No.
The proposed amendment increases the required volume of water in
the RWT to maintain the existing design requirements. The increase
is necessary due to an increase in the RWT Level--Low RAS setpoint,
which allows more water to stay in the tank following a LOCA. The
modification to the allowable value of the RWT level-low (function
5a) resolves a non-conservative TS per the guidance of
Administrative Letter 98-10 ``Dispositioning of Technical
Specifications That Are Insufficient to Assure Plant Safety.''
The proposed amendment does not impose any new or different
requirements. The change does not alter assumptions made in the
safety analyses. The proposed change is consistent with the safety
analyses assumptions and current plant operating practice.
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 amendment increases the required volume of water in
the RWT to maintain the existing design requirements. The increase
is necessary due to an increase in the RWT Level--Low RAS setpoint,
which allows more water to stay in the tank following a loss-of-
coolant accident. The modification to the allowable value of the RWT
level-low (function 5a) resolves a non-conservative TS per the
guidance of Administrative Letter 98-10 ``Dispositioning of
Technical Specifications That Are Insufficient to Assure Plant
Safety.''
The proposed amendment does not affect the design, operation,
and testing methods for systems, structures and components specified
in applicable codes and standards (or alternatives approved for use
by the NRC). With the proposed increase in the minimum required
water volume, the RWT maintains its design margin for supplying the
required amount of borated water to the reactor core and the
containment sump. The RWT will continue to meet all of its
requirements as described in the plant licensing basis (including
the Updated Final Safety Analysis Report and the TS Bases).
Similarly, there is no impact to Safety Analysis acceptance criteria
as described in the plant licensing basis.
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: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Branch Chief: James G. Danna.
Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York
Date of amendment request: April 5, 2017. A publicly available
version is in ADAMS under Accession No. ML17095A081.
Description of amendment request: The amendment would revise the
Nine Mile Point Nuclear Station, Unit 2, Technical Specifications to
allow greater flexibility in performing surveillance testing in Modes
1, 2, or 3 of emergency diesel generators and Class 1E batteries. The
proposed changes are based on Technical Specifications Task Force
(TSTF) Traveler TSTF-283-A, Revision 3, ``Modify Section 3.8 Mode
Restriction Notes.''
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 modify Mode restriction Notes to allow
performance of the Surveillance in whole or in part to reestablish
Emergency Diesel Generator (EDG) Operability, and to allow the
crediting of unplanned events that satisfy the Surveillances. The
EDGs and their associated emergency loads are accident mitigating
features, and are not an initiator of any accident previously
evaluated. As a result, the probability of any accident previously
evaluated is not significantly increased. To manage any increase in
risk, the proposed changes require an assessment to verify that
plant safety will be maintained or enhanced by performance of the
Surveillance in the current prohibited Modes. The radiological
consequences of an accident previously evaluated during the period
that the EDG is being tested to reestablish operability are no
different from the radiological consequences of an accident
previously evaluated while the EDG is inoperable. As a result, the
consequences of any accident previously evaluated are not increased.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes do not involve a physical alteration to the
plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The purpose of Surveillances is to verify that equipment is
capable of performing its assumed safety function. The proposed
changes will only allow the performance of the Surveillances to
reestablish Operability,
[[Page 27888]]
and the proposed changes may not be used to remove an EDG from
service. In addition, the proposed changes will potentially shorten
the time that an EDG is unavailable because testing to reestablish
Operability can be performed without a plant shutdown. The proposed
changes also require an assessment to verify that plant safety will
be maintained or enhanced by performance of the Surveillance in the
normally prohibited Modes.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Branch Chief: James G. Danna.
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: April 27, 2017. A publicly available
version is in ADAMS under Accession No. ML17121A449.
Description of amendment request: The proposed amendments would
revise Technical Specification 5.5.12, ``Primary Containment Leakage
Rate Testing Program,'' to allow for the permanent extension of the
Type A integrated leak rate testing and Type C leak rate testing
frequencies, and would also delete a one-time exception.
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 activity involves revision of the Quad Cities
Nuclear Power Station (QCNPS) Technical Specification (TS) 5.5.12,
Primary Containment Leakage Rate Testing Program, to allow the
extension of the QCNPS, Units 1 and 2, Type A containment integrated
leakage rate test interval to 15 years, and the extension of the
Type C local leakage rate test interval to 75 months. The current
Type A test interval of 120 months (10 years) would be extended on a
permanent basis to no longer than 15 years from the last Type A
test. The existing Type C test interval of 60 months for selected
components would be extended on a performance basis to no longer
than 75 months. Extensions of up to nine months (total maximum
interval of 84 months for Type C tests) are permissible only for
non-routine emergent conditions.
The proposed extension does not involve either a physical change
to the plant or a change in the manner in which the plant is
operated or controlled. The containment is designed to provide an
essentially leak tight barrier against the uncontrolled release of
radioactivity to the environment for postulated accidents. As such,
the containment and the testing requirements invoked to periodically
demonstrate the integrity of the containment exist to ensure the
plant's ability to mitigate the consequences of an accident, and do
not involve the prevention or identification of any precursors of an
accident.
The change in dose risk for changing the Type A Integrated Leak
Rate Test (ILRT) interval from three-per-ten years to once-per-
fifteen-years, measured as an increase to the total integrated dose
risk for all internal events accident sequences for QCNPS, is 1.0E-
02 person-rem/yr (0.31%) using the Electric Power Research Institute
(EPRI) guidance with the base case corrosion included. The change in
dose risk drops to 2.7E-03 person-rem/yr (0.08%) when using the EPRI
Expert Elicitation methodology. The values calculated per the EPRI
guidance are all lower than the acceptance criteria of less than or
equal to 1.0 person-rem/yr or less than 1.0% person-rem/yr defined
in Section 1.3 of Attachment 3 to this LAR. Therefore, this proposed
extension does not involve a significant increase in the probability
of an accident previously evaluated.
As documented in NUREG-1493, ``Performance-Based Containment
Leak-Test Program,'' dated January 1995, Types B and C tests have
identified a very large percentage of containment leakage paths, and
the percentage of containment leakage paths that are detected only
by Type A testing is very small. The QCNPS, Units 1 and 2 Type A
test history supports this conclusion.
The integrity of the containment is subject to two types of
failure mechanisms that can be categorized as: (1) Activity based,
and, (2) time based. Activity based failure mechanisms are defined
as degradation due to system and/or component modifications or
maintenance. Local leak rate test requirements and administrative
controls such as configuration management and procedural
requirements for system restoration ensure that containment
integrity is not degraded by plant modifications or maintenance
activities. The design and construction requirements of the
containment combined with the containment inspections performed in
accordance with American Society of Mechanical Engineers (ASME)
Section XI, and TS requirements serve to provide a high degree of
assurance that the containment would not degrade in a manner that is
detectable only by a Type A test. Based on the above, the proposed
test interval extensions do not significantly increase the
consequences of an accident previously evaluated.
The proposed amendment also deletes an exception previously
granted in amendments 220 and 214 to allow one-time extensions of
the ILRT test frequency for QCNPS, Units 1 and 2, respectively. This
exception was for an activity that has already taken place;
therefore, this deletion is solely an administrative action that
does not result in any change in how QCNPS, Units 1 and 2 are
operated.
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 amendment to TS 5.5.12, ``Primary Containment
Leakage Rate Testing Program,'' involves the extension of the QCNPS,
Units 1 and 2 Type A containment test interval to 15 years and the
extension of the Type C test interval to 75 months. The containment
and the testing requirements to periodically demonstrate the
integrity of the containment exist to ensure the plant's ability to
mitigate the consequences of an accident.
The proposed change does not involve a physical modification to
the plant (i.e., no new or different type of equipment will be
installed), nor does it alter the design, configuration, or change
the manner in which the plant is operated or controlled beyond the
standard functional capabilities of the equipment.
The proposed amendment also deletes an exception previously
granted under TS Amendments 220 and 214 to allow the one-time
extension of the ILRT test frequency for QCNPS, Units 1 and 2,
respectively. This exception was for an activity that has already
taken place; therefore, this deletion is solely an administrative
action that does not result in any change in how the QCNPS units are
operated.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed amendment to TS 5.5.12 involves the extension of
the QCNPS, Units 1 and 2 Type A containment test interval to 15
years and the extension of the Type C test interval to 75 months for
selected components. This amendment does not alter the manner in
which safety limits, limiting safety system set points, or limiting
conditions for operation are determined. The specific requirements
and conditions of the TS Containment Leak Rate Testing Program exist
to ensure that the degree of containment structural integrity and
leak-tightness that is considered in the plant safety analysis is
maintained. The overall containment leak rate limit specified by TS
is maintained.
The proposed change involves the extension of the interval
between Type A containment leak rate tests and Type C tests for
QCNPS, Units 1 and 2. The proposed surveillance interval extension
is bounded by the 15-year ILRT interval and the 75-month Type C test
interval currently authorized
[[Page 27889]]
within NEI 94-01, Revision 3-A. Industry experience supports the
conclusion that Types B and C testing detects a large percentage of
containment leakage paths and that the percentage of containment
leakage paths that are detected only by Type A testing is small. The
containment inspections performed in accordance with ASME Section Xl
and TS serve to provide a high degree of assurance that the
containment would not degrade in a manner that is detectable only by
Type A testing. The combination of these factors ensures that the
margin of safety in the plant safety analysis is maintained. The
design, operation, testing methods and acceptance criteria for Types
A, B, and C containment leakage tests specified in applicable codes
and standards would continue to be met, with the acceptance of this
proposed change, since these are not affected by changes to the Type
A and Type C test intervals.
The proposed amendment also deletes exceptions previously
granted to allow one-time extensions of the ILRT test frequency for
QCNPS, Units 1 and 2. This exception was for an activity that has
taken place; therefore, the deletion is solely an administrative
action and does not change how QCNPS is operated and maintained.
Thus, there is no reduction in any margin of 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: Tamra Domeyer, Associate General Counsel,
Exelon Nuclear Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: David J. Wrona.
Florida Power & Light Company, Docket Nos. 50-250 and 251, Turkey Point
Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida
Date of amendment request: April 9, 2017. A publicly available
version is in ADAMS under Accession No. ML17101A637.
Description of amendment request: The amendments would modify the
Technical Specifications (TSs) to remove various reporting
requirements. Specifically, the amendments would remove the
requirements to prepare various special reports, the Startup Report,
and the Annual Report. In addition, the amendments would revise the TSs
to remove the completion time for restoring spent fuel pool water level
to address inoperability of one of the two parallel flow paths in the
residual heat removal or safety injection headers for the Emergency
Core Cooling Systems and to make other administrative changes,
including updating plant staff and responsibilities and correcting a
misspelling.
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 actions, surveillance requirements, and administrative
controls associated with the proposed changes to the technical
specifications (TS) are not initiators of any accidents previously
evaluated, so the probability of accidents previously evaluated is
unaffected by the proposed changes. The proposed changes do not
alter the design, function, operation, or configuration of any plant
structure, system, or component (SSC). The capability of any
operable TS-required SSC to perform its specified safety function is
not impacted by the proposed changes. As a result, the outcomes of
accidents previously evaluated are unaffected. Therefore, the
proposed changes do not result in a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes do not challenge the integrity or
performance of any safety-related systems. No plant equipment is
installed or removed, and the changes do not alter the design,
physical configuration, or method of operation of any plant SSC. No
physical changes are made to the plant, so no new causal mechanisms
are introduced. Therefore, the proposed changes to the TS do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The ability of any operable SSC to perform its designated safety
function is unaffected by the proposed changes. The proposed changes
do not alter any safety analyses assumptions, safety limits,
limiting safety system settings, or method of operating the plant.
The changes do not adversely impact plant operating margins or the
reliability of equipment credited in the safety analyses. Therefore,
the proposed changes do not involve a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Blvd., MS LAW/JB,
Juno Beach, FL 33408-0420.
NRC Branch Chief: Undine S. Shoop.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold
Energy Center, Linn County, Iowa
Date of amendment request: April 20, 2017. A publicly available
version is in ADAMS under Accession No. ML17111A631.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs) Section 3.1.2, ``Reactivity
Anomalies,'' with a change to the method of calculating core reactivity
for the purpose of performing the reactivity anomaly surveillance. The
proposed change would allow performance of the reactivity anomaly
surveillance on a comparison of monitored to predicted core reactivity.
The reactivity anomaly verification is currently determined by a
comparison of monitored versus predicted control rod density.
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 does not affect any plant systems,
structures, or components designed for the prevention or mitigation
of previously evaluated accidents. The proposed change would only
modify how the reactivity anomaly surveillance is performed.
Verifying that the core reactivity is consistent with predicted
values ensures that accident and transient safety analyses remain
valid. This amendment changes the TS requirements such that, rather
than performing the surveillance by comparing monitored to predicted
control rod density, the surveillance is performed by a direct
comparison of core keff. Present day on-line core
monitoring systems, such as 3D MONICORE and ACUMEN, are capable of
performing the direct measurement of reactivity.
Therefore, since the reactivity anomaly surveillance will
continue to be performed by a viable method, the proposed change
does not involve a significant increase in the probability or
consequence of a previously evaluated accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
[[Page 27890]]
The proposed change does not involve any changes to the
operation, testing, or maintenance of any safety-related, or
otherwise important to safety systems. All systems important to
safety will continue to be operated and maintained within their
design bases. The proposed changes to the Reactivity Anomalies TS
will only provide a new, more efficient method of detecting an
unexpected change in core reactivity.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change is to modify the method for performing the
reactivity anomaly surveillance from a comparison of monitored to
predicted control rod density to a comparison of monitored to
predicted core keff. The direct comparison of
keff provides a technically superior method of
calculating any differences in the expected core reactivity. The
reactivity anomaly surveillance will continue to be performed at the
same frequency as is currently required by the TS, only the method
of performing the surveillance will be changed. Consequently, core
reactivity assumptions made in safety analyses will continue to be
adequately verified. The proposed change has no impact to the margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William Blair, P.O. Box 14000, Juno Beach,
FL 33408-0420.
NRC Branch Chief: David J. Wrona.
NextEra Energy Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant (Point Beach), Units 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of amendment request: March 31, 2017. A publicly available
version is in ADAMS under Accession No. ML17090A511.
Description of amendment request: The amendments would document a
risk-informed resolution strategy to resolve low risk, legacy design
code non-conformances associated with construction trusses in the
containment buildings of Point Beach, Units 1 and 2. The proposed
license amendment request (LAR) is a risk-informed licensing basis
change. The proposed change is acceptance of the final configuration of
the construction trusses, including the attached containment spray
piping and ventilation ductwork, and the containment liners/walls
adjacent to the trusses, using a risk-informed resolution. Accordingly,
the proposed change meets the criteria set forth in Regulatory Guide
(RG) 1.174, ``An Approach for Using Probabilistic Risk Assessment [PRA]
in Risk-Informed Decisions on Plant-Specific Changes to the Licensing
Basis,'' and the generic guidance in RG 1.200, ``An Approach for
Determining the Technical Adequacy of Probabilistic Risk Assessment
Results for Risk-Informed Activities.''
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 probability of an accident previously evaluated is not
changed. The containment structures and the containment spray piping
and ventilation ducts attached to the construction trusses are
accident mitigation equipment. They are not accident initiators.
The acceptance of the final configuration of Point Beach Units 1
and 2 results in a change in core damage frequency and large early
release frequency that is within acceptance guidelines and does not
involve a significant reduction in the margin of safety. Although
failures are postulated in the PRA analysis, the engineering
calculations in support of the LAR conclude that the construction
trusses and the associated structures/components remain structurally
sound in the event of a design basis seismic or thermal event and
there is no adverse impact or change to any station SSC's
[structure, system, and components] design function and there is no
change to accident mitigation response.
This change has no impact on station fire risk caused by a
seismic event.
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 install any new or different type
of equipment in the plant. The proposed change does not create any
new failure modes for existing equipment or any new limiting single
failures. Engineering calculations conclude the construction
trusses, equipment supported by the trusses, and containment liners
remain capable of withstanding design basis seismic and thermal
events and remain capable of performing their designated design
functions. Additionally, the proposed change does not involve a
change in the methods governing normal plant operation, and all
safety functions will continue to perform as previously assumed in
the accident analyses. Thus, the proposed change does not adversely
affect the design function or operation of any structures, systems
and components important to safety.
There are no new accidents identified associated with acceptance
of the final modified configuration of Unit 1 and the current
configuration of Unit 2.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The effects of the change, [Delta]CDF [core damage frequency]
and [Delta]LERF, [large early release frequency] are within the
acceptance guidelines shown in Figures 4 and 5 of Regulatory Guide
1.174. Consequently, the change does not result in a significant
reduction in the margin of safety.
The containment structures and liners, construction trusses, and
equipment supported by the trusses remain fully capable of
performing their specified design functions as concluded by
supporting engineering calculations.
Modifications associated with implementation of NFPA [National
Fire Protection Association] 805 are planned that will provide
protection of the reactor coolant system feed and bleed capability
and result in additional safety margin.
The proposed change does not affect the margin of safety
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 change does not alter any safety
analyses assumptions, safety limits, limiting safety system
settings, or methods of operating the plant. The changes do not
adversely impact the reliability of equipment credited in the safety
analyses. The proposed change does not adversely affect systems that
respond to safely shutdown the plant and to maintain the plant in a
safe shutdown condition.
The station will implement new seismic and thermal event limits
to ensure the construction trusses and associated equipment are
inspected and/or analyzed for any event exceeding elastic stress
limits to determine their capability to withstand a subsequent
design basis event prior to Unit restart.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William Blair, Managing Attorney--Nuclear,
Florida Power & Light Company, P.O. Box 14000, 700 Universe Boulevard,
Juno Beach, FL 33408-0420.
NRC Branch Chief: David J. Wrona.
[[Page 27891]]
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia
Date of amendment request: April 27, 2017. A publicly available
version is in ADAMS under Accession No. ML17118A049.
Description of amendment request: The requested amendments propose
changes to combined license (COL) Appendix C (and plant-specific Tier
1) Table 2.7.2-2 to revise the minimum chilled water flow rates to the
supply air handling units serving the Main Control Room and the Class
1E electrical rooms, and the unit coolers serving the normal residual
heat removal system and chemical and volume control system pump rooms.
The proposed COL Appendix C (and plant-specific Design Control Document
(Tier 1) changes require additional changes to corresponding Tier 2
component data information in Updated Final Safety Analysis Report
(UFSAR) Chapter 9. Because this proposed change requires a departure
from Tier 1 information in the Westinghouse Electric Company's AP1000
Design Control Document, the licensee also requested an exemption from
the requirements of the Generic Design Control Document Tier 1 in
accordance with 10 CFR 52.63(b)(1).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to COL Appendix C (and plant-specific Tier
1) Table 2.7.2-2, Updated Final Safety Analysis Report (UFSAR) Table
9.2.7-1, and associated UFSAR design information to identify the
revised equipment parameters for the nuclear island nonradioactive
ventilation system (VBS) air (VAS) unit coolers and reduced chilled
water system (VWS) cooling coil flow rates do not adversely impact
the plant response to any accidents which are previously evaluated.
The function of the cooling coils to provide chilled water to the
VBS AHUs and VAS unit coolers is not credited in the safety
analysis.
No safety-related structure, system, component (SSC) or function
is adversely affected by this change. The VWS safety-related
function of containment isolation is not affected by this change.
The change does not involve an interface with any SSC accident
initiator or initiating sequence of events, and thus, the
probabilities of the accidents evaluated in the plant-specific 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. The proposed changes do not increase the
probability or consequences of an accident previously evaluated as
the VWS, VBS and VAS do not provide safety-related functions and the
functions of each system to support required room environments are
not changed.
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 COL Appendix C (and plant-specific Tier
1) Table 2.7.2-2, UFSAR Table 9.2.7-1, and associated UFSAR design
information to identify the revised equipment parameters for VBS
AHUs and VAS unit coolers and reduced VWS cooling coil flow rates do
not affect any safety-related equipment, and do not add any new
interfaces to safety-related SSCs. The VWS function to provide
chilled water is not adversely impacted. The function of the VAS to
provide ventilation and cooling to maintain the environment of the
serviced areas within the design temperature range is not adversely
impacted by this change. No system or design function or equipment
qualification is affected by these changes as the change does not
modify the operation of any SSCs. The changes do not introduce a new
failure mode, malfunction or sequence of events that could affect
safety or safety-related equipment. Revised equipment parameters,
including the reduced cooling coil flow rates, do not adversely
impact the function of associated components.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The changes to COL Appendix C (and plant-specific Tier 1) Table
2.7.2-2, UFSAR Table 9.2.7-1, and associated UFSAR design
information do not affect any other safety-related equipment or
fission product barriers. The requested changes will not adversely
affect compliance with any design code, function, design analysis,
safety analysis input or result, or design/safety margin. No safety
analysis or design basis acceptance limit/criterion is challenged or
exceeded by the requested changes as previously evaluated accidents
are not impacted.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP,
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer Dixon-Herrity.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, 50-296, and 72-
052, Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, and
Independent Spent Fuel Storage Installation (ISFSI), Limestone County,
Alabama
Tennessee Valley Authority, Docket Nos. 50-327, 50-328, and 72-034,
Sequoyah Nuclear Plant (SQN), Units 1 and 2, and ISFSI, Hamilton
County, Tennessee
Tennessee Valley Authority (TVA), Docket Nos. 50-390, 50-391, and 72-
1048, Watts Bar Nuclear Plant (WBN), Units 1 and 2, and ISFSI, Rhea
County, Tennessee
Date of amendment request: January 4, 2017. A publicly available
version is in ADAMS under Accession No. ML17004A340.
Description of amendment request: The amendments would modify the
Emergency Plans for BFN, Units 1, 2, and 3, and its ISFSI; SQN, Units 1
and 2, and its ISFSI; and WBN, Units 1 and 2, and its ISFSI, to adopt
the Emergency Action Level (EAL) schemes based on Nuclear Energy
Institute (NEI) 99-01, Revision 6, which has been endorsed by the NRC
as documented in a letter dated March 28, 2013 (ADAMS Accession No.
ML12346A463). The proposed changes to TVA's EAL schemes to adopt the
guidance in NEI 99-01, Revision 6, do not reduce the capability to meet
the emergency planning requirements established in 10 CFR 50.47 and 10
CFR part 50, Appendix E. The proposed changes do not reduce the
functionality, performance, or capability of TVA's Emergency Response
Organization (ERO) to respond in mitigating the consequences of
accidents. The TVA ERO functions will continue to be performed as
required.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below.
1. Does the proposed amendment involve a significant increase in
the probability or consequence of an accident previously evaluated?
Response: No.
The proposed changes to TVA's EAL schemes to adopt the NRC-
endorsed guidance in NEI 99-01, Revision 6, ``Development of
Emergency Action Levels
[[Page 27892]]
for Non-Passive Reactors,'' do not reduce the capability to meet the
emergency planning requirements established in 10 CFR 50.47 and 10
CFR [Part] 50, Appendix E. The proposed changes do not reduce the
functionality, performance, or capability of TVA's ERO to respond in
mitigating the consequences of any design basis accident.
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, and
configuration of the facilities or the manner in which the plants
are operated and maintained. The proposed change does not adversely
affect the ability of structures, systems, and components (SSC) to
perform their intended safety function to mitigate the consequences
of an initiating event within the assumed acceptable limits. The
proposed changes do not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of any accident previously evaluated.
Further, the proposed changes do not increase the types and amounts
of radioactive effluent that may be released offsite, nor
significantly increase individual or cumulative occupational/public
radiation exposure.
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 to TVA's EAL schemes to adopt the NRC-
endorsed guidance in NEI 99-01, Revision 6, do not involve any
physical changes to plant systems or equipment. The proposed changes
do not involve the addition of any new plant equipment. The proposed
changes will not alter the design configuration, or method of
operation of plant equipment beyond its normal functional
capabilities. All TVA ERO functions will continue to be performed as
required. The proposed changes do not create any new credible
failure mechanisms, malfunctions, or accident initiators.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from those that have been
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes to TVA's EAL schemes to adopt the NRC-
endorsed guidance in NEI 99-01, Revision 6, do not alter or exceed a
design basis or safety limit. There is no change being made to
safety analysis assumptions, safety limits, or limiting safety
system settings that would adversely affect plant safety as a result
of the proposed changes. There are no changes to setpoints or
environmental conditions of any SSC or the manner in which any SSC
is operated. Margins of safety are unaffected by the proposed
changes to adopt the NEI 99-01, Revision 6, EAL scheme guidance. The
applicable requirements of 10 CFR 50.47 and 10 CFR [Part] 50,
Appendix E will continue to be met.
Therefore, the proposed changes do not involve any reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
NRC Branch Chief: Benjamin G. Beasley.
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 Progress Inc., Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2 (Robinson), Darlington County, South
Carolina
Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear
Power Plant, Unit 1 (Harris), Wake and Chatham Counties, North Carolina
Date of amendment request: August 19, 2015, as supplemented by
letters dated May 4, October 3, and November 17, 2016.
Brief description of amendments: The amendments revised the
Robinson Technical Specification (TS) 5.6.5.b and the Harris TS
6.9.1.6.2 to adopt the methodology reports DPC-NE-1008-P, Revision 0,
``Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse
Reactors''; DPC-NF-2010, Revision 3, ``Nuclear Physics Methodology for
Reload Design''; and DPC-NE-2011-P, Revision 2, ``Nuclear Design
Methodology Report for Core Operating Limits of Westinghouse
Reactors,'' for application specific to Robinson and Harris.
Date of issuance: May 18, 2017.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment Nos.: 253 (Robinson) and 157 (Harris). A publicly
available version is in ADAMS under Accession No. ML17102A923;
documents related to these amendments are listed in the Safety
Evaluations enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-23 and NPF-63:
Amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: February 2, 2016 (81 FR
5492). The supplemental letter dated May 4, 2016, provided additional
information that expanded the scope of the application as originally
noticed, and changed the NRC staff's original proposed no significant
hazards consideration determination as published in the Federal
Register. Accordingly, the NRC published a second proposed no
significant hazards consideration determination in the Federal Register
on August 2, 2016 (81 FR 50746). This notice superseded the original
notice in its entirety. The supplemental letters dated October 3 and
November 17, 2016, provided additional information that clarified the
application, did not expand the scope beyond the second notice, and did
not change the NRC staff's proposed no significant hazards
consideration determination as published in the Federal Register.
[[Page 27893]]
The Commission's related evaluations of the amendments are
contained in the Safety Evaluations dated May 18, 2017.
No significant hazards consideration comments received: No.
Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick
Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina
Date of amendment request: December 21, 2015, as supplemented by
letters dated June 29, July 13, August 15, November 1, November 17,
2016, and February 27, 2017.
Brief description of amendments: The amendments adopted the
approved changes to Standard Technical Specifications for General
Electric (BWR/4) [Boiling Water Reactor] Plants, NUREG-1433, Revision
4, to allow relocation of specific technical specification surveillance
frequencies to a licensee-controlled program. The changes are described
in Technical Specification Task Force (TSTF) Traveler, TSTF-425,
Revision 3, ``Relocate Surveillance Frequencies to Licensee Control--
RITSTF Initiative 5b'' (ADAMS Package Accession No. ML090850642), and
are described in the Notice of Availability published in the Federal
Register on July 6, 2009 (74 FR 31996).
Date of issuance: May 24, 2017.
Effective date: As of the date of issuance and shall be implemented
within 180 days of issuance.
Amendment Nos.: 276 (Unit 1) and 304 (Unit 2). A publicly available
version is in ADAMS under Accession No. ML17096A129; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: March 29, 2016 (81 FR
17504). The supplemental letters dated June 29, July 13, August 15,
November 1, November 17, 2016, and February 27, 2017, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the NRC staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 24, 2017.
No significant hazards consideration comments received: No.
Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear
Power Plant, Unit 1, Wake and Chatham Counties, North Carolina
Date of amendment request: May 26, 2016, as supplemented by letter
dated December 19, 2016.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) by adding a new Administrative Controls section to
establish, implement, and maintain a Diesel Fuel Oil Testing Program.
It also relocated to this program the current TS surveillance
requirements (SRs) for evaluating diesel fuel oil, along with the SRs
for draining, sediment removal, and cleaning of each main fuel oil
storage tank at least once every 10 years. In addition, the licensee
took an exception to NRC Regulatory Guide 1.137, Revision 1, ``Fuel-Oil
Systems for Standby Diesel Generators,'' to allow for the ability to
perform sampling of new fuel oil offsite prior to its addition to the
fuel oil storage tanks.
Date of issuance: May 24, 2017.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment No.: 158. A publicly available version is in ADAMS under
Accession No. ML17048A184; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-63: Amendment revised
the Renewed Facility Operating License and TSs.
Date of initial notice in Federal Register: October 11, 2016 (81 FR
70178). The supplemental letter dated December 19, 2016, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the NRC staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 24, 2017.
No significant hazards consideration comments received: No.
Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear
Power Plant, Unit 1, Wake and Chatham Counties, North Carolina
Date of amendment request: June 29, 2016, as supplemented by letter
dated November 4, 2016.
Brief description of amendment: The amendment revised the Shearon
Harris Nuclear Power Plant, Unit 1, Technical Specification (TS) 3/
4.11.1.4, ``Liquid Holdup Tanks''; TS 3/4.11.2.5, ``Explosive Gas
Mixture''; and TS 6.8.4.j, ``Gas Storage Tank Radioactivity Monitoring
Program.'' The amendment deleted TS Definition 1.16, ``GASEOUS RADWASTE
TREATMENT SYSTEM''; TS 3/4.11.1.4, ``Liquid Holdup Tanks''; and TS 3/
4.11.2.5, ``Explosive Gas Mixture.'' The amendment relocated the
deleted requirements for these TSs to licensee control under TS
6.8.4.j, ``Gas Storage Tank Radioactivity Monitoring Program.'' The
description for TS 6.8.4.j, ``Gas Storage Tank Radioactivity Monitoring
Program,'' was modified to include the controls for potentially
explosive gas mixtures contained in the Gaseous Waste Processing System
and the quantity of radioactivity contained in unprotected outdoor
liquid storage tanks. The amendment relocated requirements associated
with TS 3/4.11.1.4 and TS 3/4.11.2.5 to the licensee-controlled Plant
Programs Procedure PLP-114, ``Relocated Technical Specifications and
Design Basis Requirements.''
Date of issuance: May 25, 2017.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment No.: 159. A publicly available version is in ADAMS under
Accession No. ML17074A672; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-63: The amendment
revised the Facility Operating License and TSs.
Date of initial notice in Federal Register: October 25, 2016 (81 FR
73433). The supplemental letter dated November 4, 2016, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the NRC staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 25, 2017.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear
Plant, Van Buren County, Michigan
Date of amendment request: July 11, 2016.
Brief description of amendment: The amendment approved adoption of
NRC-approved Technical Specifications Task Force (TSTF) Standard
Technical Specifications Change Traveler TSTF-545, Revision 3, ``TS
[Technical Specification] Inservice Testing Program
[[Page 27894]]
Removal & Clarify SR [Surveillance Requirement] Usage Rule Application
to Section 5.5 Testing,'' dated October 21, 2015. Specifically, the
amendment deleted Palisades Nuclear Plant TS 5.5.7, ``Inservice Testing
Program,'' and added a new defined term, ``INSERVICE TESTING PROGRAM,''
to the TSs. All existing references to the ``Inservice Testing
Program,'' in the Palisades Nuclear Plant TS SRs are replaced with
``INSERVICE TESTING PROGRAM'' so that the SRs refer to the new
definition in lieu of the deleted program.
Date of issuance: May 30, 2017.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 262. A publicly available version is in ADAMS under
Accession No. ML17082A465; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-20: Amendment revised
the Renewed Facility Operating License and TSs.
Date of initial notice in Federal Register: August 30, 2016 (81 FR
59663).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 30, 2017.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2, Will County, Illinois
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, Units 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.
Brief description of amendments: The amendments revised the
Inservice Testing Program requirements in each plant's technical
specifications (TSs). The changes included deleting the current TS
requirements for the Inservice Testing Program, adding a new defined
term, ``INSERVICE TESTING PROGRAM,'' to the TSs, and revising other TSs
to reference this new defined term instead of the deleted program.
Date of issuance: May 26, 2017.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: 191, 192, 197, 197, 320, 298, 212, 254, 247, 223,
209, 227, 161, 313, 317, 266, 261, 124, and 290. A publicly available
version is in ADAMS under Accession No. ML17073A067. Documents related
to these amendments are listed in the Safety Evaluations enclosed with
the amendments.
Facility Operating License Nos.: NPF-72, NPF-77, NPF-37, NPF-66,
DPR-53, DPR-69, NPF-62, DPR-19, DPR-25, NPF-11, NPF-18, DPR-63, NPF-69,
DPR-44, DPR-56, DPR-29, DPR-30, DPR-18, and DPR-50. Amendments revised
the Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: November 8, 2016 (81 FR
78648).
The Commission's related evaluations of the amendments are
contained in Safety Evaluations dated May 26, 2017.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412,
Beaver Valley Power Station, Unit 2, Beaver County, Pennsylvania
Date of amendment request: June 24, 2016, as supplemented by
letters dated September 13, 2016; December 15, 2016; and March 16,
2017.
Brief description of amendment: The amendment modified the Renewed
Facility Operating License to reflect the direct transfer of Toledo
Edison Company's 18.26 percent leased interest in Beaver Valley Power
Station, Unit 2, and Ohio Edison Company's 21.66 percent leased
interest in Beaver Valley Power Station, Unit 2, from FirstEnergy
Nuclear Operating Company to FirstEnergy Nuclear Generation, LLC.
Date of issuance: May 30, 2017.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 187. A publicly available version is in ADAMS under
Accession No. ML17115A123.
Renewed Facility Operating License No. NPF-73: Amendment revised
the Renewed Facility Operating License.
Date of initial notice in Federal Register: January 23, 2017 (82 FR
7880). The supplemental letter dated March 16, 2017, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the NRC staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 14, 2017.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment request: July 21, 2016, as supplemented by letter
dated September 26, 2016.
Brief description of amendments: The amendments revised the Donald
C. Cook Nuclear Plant, Units 1 and 2, Technical Specification (TS)
Surveillance Requirements (SRs), consistent with the NRC-approved
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.'' Specifically, the change revised
the TSs to eliminate Section 5.5.6, ``Inservice Testing Program.'' A
new defined term, ``INSERVICE TESTING PROGRAM,'' was added to the TS
Definitions section. TS SRs that previously referred to the Inservice
Testing Program from Section 5.5.6 were revised to refer to the new
defined term, ``INSERVICE TESTING PROGRAM.''
Date of issuance: May 24, 2017.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment Nos.: 335 (Unit 1) and 317 (Unit 2). A publicly available
version is in ADAMS under Accession No. ML17103A106; documents related
[[Page 27895]]
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: September 27, 2016 (81
FR 66307). The supplemental letter dated September 26, 2016, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the NRC staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 24, 2017.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Units No. 1 and No. 2, Surry County, Virginia
Date of amendment request: May 18, 2016, as supplemented by letters
dated February 10, 2017; March 1, 2017; and March 10, 2017.
Brief description of amendments: The amendments revised Technical
Specification 3.14 ``Circulating and Service Water Systems,'' to extend
the Allowed Outage Time for only one operable Service Water flow path
to the Changing Pump Service Water subsystem and to the Main Control
Room/Emergency Switchgear Room air conditioning subsystem.
Date of issuance: May 31, 2017.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 289 (Unit 1) and 289 (Unit 2). A publicly available
version is in ADAMS under Accession No. ML17100A253; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised
the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: October 25, 2016 (81 FR
73443). The supplemental letters dated February 10, 2017; March 1,
2017; and March 10, 2017, provided additional information that
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the NRC staff's original
proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 31, 2017.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 6th day of June 2017.
For the Nuclear Regulatory Commission.
Eric J. Benner,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2017-12732 Filed 6-16-17; 8:45 am]
BILLING CODE 7590-01-P