[Federal Register Volume 77, Number 152 (Tuesday, August 7, 2012)]
[Notices]
[Pages 47123-47131]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2012-19004]


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NUCLEAR REGULATORY COMMISSION

[NRC-2012-0181]


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses and Combined Licenses Involving No Significant 
Hazards Considerations

Background

    Pursuant to Section 189a. (2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC) is publishing this regular biweekly notice. The Act 
requires the Commission 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 July 12, 2012 to July 25, 2012. The last 
biweekly notice was published on July 24, 2012 (77 FR 43374).

ADDRESSES: You may access information and comment submissions related 
to this document, which the NRC possesses and are publicly available, 
by searching on http://www.regulations.gov under Docket ID NRC-2012-
0181. You may submit comments by any of the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0181. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; email: Carol.Gallagher@nrc.gov.
     Mail comments to: Cindy Bladey, Chief, Rules, 
Announcements, and Directives Branch (RADB), Office of Administration, 
Mail Stop: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
     Fax comments to: RADB at 301-492-3446.
    For additional direction on accessing information and submitting 
comments, see ``Accessing Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

SUPPLEMENTARY INFORMATION:

I. Accessing Information and Submitting Comments

A. Accessing Information

    Please refer to Docket ID NRC-2012-0181 when contacting the NRC 
about the availability of information regarding this document. You may 
access information related to this document, which the NRC possesses 
and are publicly available, by any of the following methods:
     Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0181.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may access publicly available documents online in the NRC 
Library at http://www.nrc.gov/reading-rm/adams.html. To begin the 
search, select ``ADAMS Public Documents'' and then select ``Begin Web-
based ADAMS Search.'' For problems with ADAMS, please contact the NRC's 
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to pdr.resource@nrc.gov. Documents may be viewed in 
ADAMS

[[Page 47124]]

by performing a search on the document date and docket number.
     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-2012-0181 in the subject line of your 
comment submission, in order to ensure that the NRC is able to make 
your comment submission available to the public in this docket.
    The NRC cautions you not to include identifying or contact 
information in comment submissions that you do not want to be publicly 
disclosed. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into 
ADAMS, and the NRC does not 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 in their comment submissions 
that they do not want to be publicly disclosed. Your request should 
state that the NRC will not edit comment submissions to remove such 
information before making the comment submissions available to the 
public or entering the comment submissions into ADAMS.

Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses and Combined Licenses, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in Title 10 of the Code of Federal 
Regulations (10 CFR) 50.92, 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 should 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. Should the 
Commission take 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. Should the Commission make 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.
    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license or 
combined license. Requests for a hearing and a petition for leave to 
intervene shall be filed in accordance with the Commission's ``Rules of 
Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2. 
Interested person(s) should consult a current copy of 10 CFR 2.309, 
which is available at the NRC's PDR, located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. The NRC regulations are accessible electronically from the NRC 
Library on the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to 
intervene is filed by the above date, the Commission or a presiding 
officer designated by the Commission or by the Chief Administrative 
Judge of the Atomic Safety and Licensing Board Panel, will rule on the 
request and/or petition; and the Secretary or the Chief Administrative 
Judge of the Atomic Safety and Licensing Board will issue a notice of a 
hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/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 requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held. If 
the final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing

[[Page 47125]]

held would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, then any hearing held would take place before 
the issuance of any amendment.
    All documents filed in the NRC adjudicatory proceedings, including 
a request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC E-Filing rule (72 FR 49139; 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at hearing.docket@nrc.gov, or by 
telephone at 301-415-1677, to request (1) a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a request or 
petition for hearing (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/apply-certificates.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the agency's public Web 
site at http://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with the NRC guidance 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC's Web site 
at http://www.nrc.gov/site-help/e-submittals.html, by email at 
MSHD.Resource@nrc.gov, or by a toll-free call at 1-866 672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 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 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, 
Maryland, 20852, Attention: Rulemaking and Adjudications Staff. 
Participants filing a document in this manner are responsible for 
serving the document on all other participants. Filing is considered 
complete by first-class mail as of the time of deposit in the mail, or 
by courier, express mail, or expedited delivery service upon depositing 
the document with the provider of the service. A presiding officer, 
having granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 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.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Non-timely filings 
will not be entertained absent a determination by the presiding officer 
that the petition or request should be granted or the contentions 
should be admitted, based on a balancing of the factors specified in 10 
CFR 2.309(c)(1)(i)-(viii).
    For further details with respect to this license amendment 
application, see the

[[Page 47126]]

application for amendment which is available for public inspection at 
the NRC's PDR, located at One White Flint North, Room O1-F21, 11555 
Rockville Pike (first floor), Rockville, Maryland 20852. Publicly 
available documents created or received at the NRC are accessible 
electronically through ADAMS in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to ADAMS or who 
encounter problems in accessing the documents located in ADAMS, should 
contact the NRC's PDR Reference staff at 1-800-397-4209, 301-415-4737, 
or by email to pdr.resource@nrc.gov.

Carolina Power and Light Company, Docket No. 50-261, H. B. Robinson 
Steam Electric Plant, Unit 2, (HBRSEP) Darlington County, South 
Carolina

    Date of amendment request: June 8, 2012.
    Description of amendment request: The proposed change would revise 
the Technical Specifications (TSs) 3.1.4, ``Rod Group Alignment 
Limits,'' and TS 3.1.7, ``Rod Position Indication,'' to allow up to 1 
hour of soak time following substantial rod movement during which 
individual rod position indicators may not be within its limits.
    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.
    This license amendment request proposes to allow up to one hour 
of soak time following substantial rod movement during which time 
the rod position indication may be outside its limits. This would 
allow an additional hour for rod position indication to be 
inoperable or a control rod to be misaligned prior to entry into a 
TS LCO [Limiting Condition for Operation] Condition and Required 
Actions. RPI [Rod Position Indicators] instrumentation is not an 
assumed accident initiator; however, the HBRSEP, Unit No. 2 safety 
analyses consider two types of rod misalignment events, static 
misalignment and a dropped rod.
    The safety analyses show that for the static misalignment event, 
without any operator intervention, a single fully withdrawn rod 
event does not result in any fuel pin failure; therefore, the static 
rod misalignment event is not time dependent and an additional hour, 
with the misalignment undetected and unmitigated does not increase 
the consequences of the event. Multiple rod misalignment events are 
bounded by the single rod misalignment analyses and therefore an 
additional hour would not have any impact on this event.
    The safety analyses also show that a single dropped rod event, 
without any operator intervention, does not result in any fuel pin 
failure; therefore, the rod drop event is not time dependent and an 
additional hour with the misalignment undetected and unmitigated 
does not increase the consequences of the event. Multiple rod drop 
events cause the reactor to trip and therefore an additional hour 
would not have any impact on that event.
    Although this license amendment request may allow a misaligned 
rod to be undetected for an additional hour, the additional time for 
discovery does not change the probability of a misaligned control 
rod event because the one hour time extension does not affect the 
control rod drive system features that would result in either type 
of misalignment.
    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 previously evaluated?
    Response: No.
    This proposed change does not alter the design, function, or 
operation of any plant component and does not install any new or 
different equipment. No new accident scenarios, transient 
precursors, failure mechanisms, or limiting single failures are 
introduced as a result of these changes. No new equipment 
performance burdens are imposed.
    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 
the margin of safety?
    Response: No.
    The RPI system is an instrumentation system that provides 
indication to the operators that a control rod may be misaligned. 
Inoperable individual RPI instrumentation does not, by itself in any 
way, harm or impact reactor operation. Inoperable rod position 
indication may impair the ability of the operators to detect a 
misaligned rod. However, the impact of inoperable RPI 
instrumentation may be offset by availability of other indications 
that a rod is misaligned such as nuclear instrumentation indication 
that reactor power has shifted to one side of the core or 
thermocouple indication that the core temperatures increased in one 
region of the core and/or decreased in another region of the core. 
Based on plant experience, the likelihood of a misaligned rod at 
HBRSEP, Unit No. 2 is considered to be small and the likelihood of a 
misaligned rod coincident with inoperable rod position indication 
during the allowed one hour extension is even smaller. In addition, 
these proposed changes may enhance plant safety and reliability 
because the one hour soak time will allow the operators and 
engineers to focus on monitoring the reactor performance without 
unnecessary entry into TS LCO Conditions and Required Actions.
    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: David T. Conley, Manager--Senior Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Acting Branch Chief: Jessie F. Quichocho.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Miami-Dade County, Florida

    Date of amendment request: April 30, 2012.
    Description of amendment request: This amendment request proposes 
to permanently revise technical specification (TS) 6.8.4.j, Steam 
Generator (SG) Surveillance Program, to exclude portions of the SG tube 
below the top of the SG tubesheet from periodic tube inspections.
    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 previously analyzed accidents are initiated by the failure 
of plant structures, systems, or components. The proposed change 
that alters the SG inspection and reporting criteria does not have a 
detrimental impact on the integrity of any plant structure, system, 
or component that initiates an analyzed event. The proposed change 
will not alter the operation of, or otherwise increase the failure 
probability of any plant equipment that initiates an analyzed 
accident.
    Of the applicable accidents previously evaluated, the limiting 
transients with consideration to the proposed change to the SG tube 
inspection and repair criteria are the SG tube rupture (SGTR) event 
and the steam line break (SLB) postulated accident.
    Addressing the SGTR event, the required structural integrity 
margins of the SG tubes and the tube-to-tubesheet joint over the H* 
distance will be maintained. Tube rupture in tubes with cracks 
within the tubesheet is precluded by the constraint provided by the 
presence of the tubesheet and the tube-to-tubesheet joint. Tube 
burst cannot occur within the thickness of the tubesheet. The tube-
to-tubesheet joint constraint results from the hydraulic expansion 
process, thermal expansion mismatch between the tube and

[[Page 47127]]

tubesheet, and from the differential pressure between the primary 
and secondary side, and tubesheet rotation. The structural margins 
against burst, as discussed in Regulatory Guide (RG) 1.121, ``Bases 
for Plugging Degraded PWR [Pressurized-Water Reactors] Steam 
Generator Tubes'' [Reference 7] and NEI [Nuclear Energy Institute] 
97-06, ``Steam Generator Program Guidelines'', [Reference 3] are 
maintained for both normal and postulated accident conditions.
    For the portion of the tube outside of the tubesheet, the 
proposed change also has no impact on the structural or leakage 
integrity. Therefore, the proposed change does not result in a 
significant increase in the probability of the occurrence of a SGTR 
accident.
    At normal operating pressures, leakage from primary water stress 
corrosion cracking below the proposed limited inspection depth is 
limited by the tube-to-tubesheet crevice. Consequently, negligible 
normal operating leakage is expected from degradation below the 
inspected depth within the tubesheet region. The consequences of an 
SGTR event are not affected by the primary to secondary leakage flow 
during the event as primary to secondary leakage flow through a 
postulated tube that has been pulled out of the tubesheet is 
essentially equivalent to a tube rupture. Therefore, the proposed 
change does not result in a significant increase in the consequences 
of an SGTR. In addition, the selected H* value envelopes the depth 
within the tubesheet required to prevent a tube pullout.
    The probability of a SLB is unaffected by the potential failure 
of a SG tube as the failure of a tube is not an initiator for a SLB 
event.
    The leak rate factor of 1.82 for Turkey Point Units 3 and 4, for 
a postulated SLB, has been calculated as shown in References 2, 9 
and 19. Turkey Point Units 3 and 4 will apply the factor of 1.82 to 
the normal operating leakage associated with the tubesheet expansion 
region in the condition monitoring (CM) and operational assessment 
(OA). Through application of the limited tubesheet inspection scope, 
the existing operating leakage limit provides assurance that 
excessive leakage (i.e., greater than accident analysis assumptions) 
will not occur. Multiplying the TS operational leak rate limit of 
150 gpd (at room temperature) through any one SG by a factor of 1.82 
shows that the maximum primary to secondary accident induced leak 
rate is limited to 273 gpd. This leakage rate is bounded by the 
current licensing basis assumed primary to secondary accident leak 
rate of 0.20 gpm (288 gpd) through any one SG for SLB. Since the 
existing limit on operational leakage continues to ensure that the 
SLB assumed accident induced leakage will not be exceeded, the 
consequences of a SLB accident are not increased.
    For the CM assessment, the component of leakage from the prior 
cycle from below the H* distance will be multiplied by a factor of 
1.82 and added to the total leakage from any other source and 
compared to the allowable accident induced leak rate. For the OA, 
the difference in the leakage between the allowable leakage and the 
calculated accident induced leakage from sources other than the 
tubesheet expansion region will be divided by 1.82 and compared to 
the observed operational leakage.
    Based on the above, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No
    The proposed change that alters the SG inspection and reporting 
criteria does not introduce any new equipment, create new failure 
modes for existing equipment, or create any new limiting single 
failures. Plant operation will not be altered, and all safety 
functions will continue to perform as previously assumed in accident 
analyses. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Response: No
    The proposed change defines the safety significant portion of 
the tube that must be inspected and repaired. WCAP-17345, Rev. 2 
[Reference 9] identifies the specific inspection depth below which 
any type of tube degradation is shown to have no impact on the 
performance criteria in NEI 97-06 Rev. 3, ``Steam Generator Program 
Guidelines'' [Reference 3] and TS 6.8.4.j, ``Steam Generator (SG) 
Program.''
    The proposed change that alters the SG inspection and reporting 
criteria maintains the required structural margins of the SG tubes 
for both normal and accident conditions. Nuclear Energy Institute 
97-06, ``Steam Generator Program Guidelines'' [Reference 3], and NRC 
Regulatory Guide (RG) 1.121, ``Bases for Plugging Degraded PWR Steam 
Generator Tubes'' [Reference 7], are used as the bases in the 
development of the limited tubesheet inspection depth methodology 
for determining that SG tube integrity considerations are maintained 
within acceptable limits. RG 1.121 describes a method acceptable to 
the NRC for meeting General Design Criteria (GDC) 14, ``Reactor 
Coolant Pressure Boundary,'' GDC 15, ``Reactor Coolant System 
Design,'' GDC 31, ``Fracture Prevention of Reactor Coolant Pressure 
Boundary,'' and GDC 32, ``Inspection of Reactor Coolant Pressure 
Boundary,'' by reducing the probability and consequences of a SGTR. 
RG 1.121 concludes that by determining the limiting safe conditions 
for tube wall degradation, the probability and consequences of a 
SGTR are reduced. This RG uses safety factors on loads for tube 
burst that are consistent with the requirements of Section III of 
the American Society of Mechanical Engineers (ASME) Code.
    For axially oriented cracking located within the tubesheet, tube 
burst is precluded due to the presence of the tubesheet. For 
circumferentially oriented cracking, Westinghouse WCAP-17091-P, Rev. 
0 [Reference 2] and WCAP-17345, Rev. 2 [Reference 9] define a length 
of degradation-free expanded tubing that provides the necessary 
resistance to tube pullout due to the pressure induced forces, with 
applicable safety factors applied. Application of the limited hot 
and cold leg tubesheet inspection criteria will preclude 
unacceptable primary to secondary leakage during all plant 
conditions. The SLB leak rate factor for Turkey Point Units 3 and 4 
is 1.82 (Table 9-7 in WCAP-17091-P). Multiplying the TS operational 
leak rate limit of 150 gpd through any one SG by the leak rate 
factor of 1.82 shows that the maximum primary to secondary accident 
induced leak rate is limited to 273 gpd. This leakage rate is 
bounded by the current licensing basis assumed primary to secondary 
accident leak rate of 0.20 gpm (288 gpd) through any one SG for SLB.
    Therefore, the proposed change does not involve a significant 
reduction in any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Acting Branch Chief: Jessie F. Quichocho.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: May 30, 2012.
    Description of amendment request: The proposed amendment would 
revise Technical Specification Section 2.0, ``Safety Limits.'' 
Specifically, the proposed amendment would revise two recirculation 
loop and single recirculation loop Safety Limit Minimum Critical Power 
Ratio (SLMCPR) values to reflect results of a cycle-specific 
calculation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    Four accidents have been evaluated previously as reflected in 
the CNS [Cooper Nuclear Station] Updated Safety Analysis Report 
(USAR). These four accidents are (1) loss-of-coolant, (2) control 
rod drop, (3) main steam line break, and (4) fuel handling. The 
probability of an evaluated accident is derived from the 
probabilities of the

[[Page 47128]]

individual precursors to that accident. Changing the SLMCPR values 
does not increase the probability of an evaluated accident. The 
change does not require any physical modifications to the plant or 
any components, nor does it require a change in plant operation. 
Therefore, no individual precursors of an accident are affected.
    The consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences. This proposed change makes no modification to the 
design or operation of the systems that are used in mitigation of 
accidents. Limits have been established, consistent with Nuclear 
Regulatory Commission (NRC) approved methods, to ensure that fuel 
performance during normal, transient, and accident conditions is 
acceptable. The proposed change to the values of the SLMCPR 
continues to conservatively establish this safety limit such that 
the fuel is protected during normal operation and during any plant 
transients or anticipated operational occurrences.
    Based on the above, NPPD [Nebraska Public Power District] 
concludes that the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Creation of the possibility of a new or different kind of 
accident from an accident previously evaluated would require 
creation of precursors of that accident. New accident precursors may 
be created by modification of the plant configuration or changes in 
how the plant is operated. The proposed change does not involve a 
modification of the plant configuration or in how the plant is 
operated. The proposed change to the SLMCPR values assures that 
safety criteria are maintained.
    Based on the above, NPPD concludes that the proposed change does 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The values of the proposed SLMCPR provides a margin of safety by 
ensuring that no more than 0.1% of fuel rods are expected to be in 
boiling transition if the Minimum Critical Power Ratio limit is not 
violated. The proposed change will ensure the appropriate level of 
fuel protection is maintained. Additionally, operational limits are 
established based on the proposed SLMCPR to ensure that the SLMCPR 
is not violated during all modes of operation. This will ensure that 
the fuel design safety criteria are met (i.e., that at least 99.9% 
of the fuel rods do not experience transition boiling during normal 
operation as well as anticipated operational occurrences).
    Based on the above, NPPD concludes that 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: Mr. John C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Branch Chief: Michael T. Markley.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit 1, Washington County, Nebraska

    Date of amendment request: February 10, 2012.
    Description of amendment request: The proposed amendment would 
establish the limiting condition for operation (LCO) requirements for 
the reactor protective system (RPS) actuation circuits in Technical 
Specification (TS) 2.15, ``Instrumentation and Control Systems.'' 
Specifically, the proposed change: renumbers LCO 2.15(1) through 
2.15(4) to 2.15.1(1) through 2.15.1(4), renumbers LCO 2.15(5) to LCO 
2.15.3 with an associated Table 2-6, and implements a new LCO 2.15.2 
for the RPS logic and trip initiation channels. The Table of Contents 
will also be revised to reflect the renumbering and addition of the LCO 
for the RPS logic and trip initiation channels and the new Table 2-6.
    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 reactor protective system logic and trip initiation channels 
meets Criterion 3 of 10 CFR 50.36 for inclusion into Technical 
Specification (TS) as a component that is part of the primary 
success path and which functions or actuates to mitigate a design 
basis accident or transient. The TSs currently does not have 
limiting conditions for operations (LCO) specific for this 
circuitry, but does contain surveillance requirements. The addition 
of LCOs provides additional restrictions on the operation of the 
plant and provides required actions and time limits if these 
components are incapable of performing their function. As such, the 
proposed change does not increase the probability of an accident. 
The proposed changes do not alter the physical design of the RPS, or 
any other plant structure, system or component (SSC) at Fort Calhoun 
Station (FCS).
    The proposed changes conform to the Nuclear Regulatory 
Commission's (NRC's) regulatory guidance regarding the content of 
plant TS as identified in 10 CFR 50.36 and NRC publication NUREG 
1432.
    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 alter the physical design, safety 
limits, or safety analysis assumptions associated with the operation 
of the plant. Hence, the proposed changes do not introduce any new 
accident initiators, nor do they reduce or adversely affect the 
capabilities of any plant structure or system in the performance of 
their safety function.
    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 TS operability requirements for the RPS logic and trip 
initiation channels ensure there is adequate components operable to 
assure safe reactor operation and are necessary to ensure safety 
systems accomplish their safety function for design basis accident 
events. The proposed TS would revise the applicability for when the 
RPS logic and trip initiation channels are required to be operable 
to include whenever control element assemblies (CEAs) are capable of 
being withdrawn and the reactor coolant system (RCS) is not at 
refueling boron concentration. When the RCS boron concentration is 
at refueling boron concentration, or when no more than one trippable 
control rod is capable of being withdrawn, the RPS function is 
already fulfilled. These proposed TS changes for the RPS are aligned 
with the applicability and operability requirements provided in 
NUREG 1432.
    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: David A. Repka, Esq., Winston & Strawn, 1700 
K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Michael T. Markley.

ZionSolutions LLC, Docket Nos. 50-295 and 50-304, Zion Nuclear Power 
Station (Zion), Units 1 and 2, Lake County, Illinois

    Date of amendment request: May 31, 2012.

[[Page 47129]]

    Description of amendment request: The proposed amendments would 
approve methods of analysis, use of the upgraded fuel handling building 
crane system as a single-failure proof crane, and a NUREG 0612 
compliant heavy loads handling program.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The existing DSAR [Defueled Safety Analysis Report] analysis 
assumes that a spent fuel cask drop occurs. In this analysis, the 
physics of the drop, coupled with concrete bumpers on the cask 
loading pit and pool edge were used to demonstrate that a postulated 
drop of the spent fuel cask near the Spent Fuel Pool neither 
impacted the spent fuel directly nor damaged the pool structure in a 
manner that adversely affected the spent fuel, when a cask was to be 
handled in the cask loading pit. The proposed License Amendment 
Request to operate a single-failure proof Fuel Building Crane 
demonstrates that no analysis is required for the cask drop event 
based on the design and the associated programmatic controls. A drop 
of the spent fuel cask handled with a single-failure proof crane 
(designed to ASME NOG-1 [``Rules for Construction of Overhead and 
Gantry Cranes (Top Running Bridge, Multiple Girder)''] and compliant 
with NUREG-0554 [``Single-Failure-Proof Cranes for Nuclear Power 
Plants'', ML110450636]), operated in accordance with the 
administrative controls of NUREG-0612 [``Control of Heavy Loads at 
Nuclear Power Plants,'' ML070250180] has an acceptably low 
probability so as to effectively preclude consideration of the 
event. The risk of such a drop event using the new single-failure 
proof crane operated in accordance with the Heavy Loads Program 
procedures, qualitatively, is lower than the event previously 
analyzed which postulate the event without evaluation of its 
likelihood.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The location and design functions of the Fuel Building crane are 
not changed from those currently described in the DSAR. Because the 
new crane has a single-failure proof design the uncontrolled 
lowering, or drop, of a heavy load will not be considered credible. 
Evaluations show that individual malfunctions or component failures 
of the crane will not result in load drop. The new single-failure 
proof crane['s] primary use[s] will be to move a loaded or unloaded 
MAGNASTOR transfer cask between the cask loading pit [and] the 
decontamination pit, and transfer [the cask] to the low profile cart 
rail transport in the Fuel Handling Building. No components that are 
classified as Important to the Defueled Condition, other than the 
Fuel Building crane, will be affected by these movements. Based on 
the design and programmatic controls on the crane, no load will 
lower uncontrollably or drop in or around the spent fuel pool or 
near an open cask containing spent fuel nor will a cask containing 
spent fuel drop or be lowered uncontrollably during operation of the 
crane. Hence no new accidents will be initiated.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    This proposed License Amendment Request involves the replacement 
of the existing non-single-failure proof Fuel Building Crane with a 
new single-failure proof crane. The new crane has been designed to 
meet the specifications found in ASME NOG-1-2004, which has been 
endorsed by the NRC in RIS 2005-25, as supplemented, as an 
acceptable means of meeting the criteria in NUREG-0554, ``Single-
failure Proof Cranes for Nuclear Power Plants.'' to provide adequate 
protection and safety margin against the uncontrolled lowering of 
the lifted load. The occurrence of a cask load drop accident is 
considered not credible when the load is lifted with a single-
failure proof lifting system meeting the guidance in NUREG-0612, 
``Control of Heavy Loads at Nuclear Power Plants'' Section 5.1.6, 
``Single-Failure-Proof Handling Systems.'' As a result, the proposed 
change, replacing the existing non-single-failure proof crane, has 
no adverse impact on stored spent fuel, or structural integrity of 
the pool.
    The configuration of the crane and the primary load, a spent 
fuel cask containing spent fuel, is changed from that of the DSAR. 
The specific analysis dealing with a drop of the cask will no longer 
be applicable and [will be] removed from the DSAR, since the new 
single-proof crane makes that event of low enough probability to not 
be considered credible. The maximum critical lift capacity of the 
crane has not been changed, though the load to be lifted is larger. 
The structural analyses of the crane and its support structure, 
however, show acceptable margin under the acceptance criteria of 
NOG-I for operation of the crane.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Russ Workman, Deputy General Counsel, 
EnergySolutions, 423 West 300 South, Suite 200, Salt Lake City, UT 
84101.
    NRC Branch Chief: Bruce Watson.

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 are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland 20852. Publicly available documents created or received at the 
NRC are accessible electronically through the Agencywide Documents 
Access and Management System (ADAMS) in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR's Reference staff at 1-800-397-4209, 301-415-4737 or by 
email to pdr.resource@nrc.gov.

[[Page 47130]]

Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414, 
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of application for amendments: July 21, 2011.
    Brief description of amendments: The amendments revised Technical 
Specifications 3.3.2, ``Engineered Safety Feature Actuation System 
(ESFAS) Instrumentation,'' 3.5.4, ``Refueling Water Storage Tank 
(RWST),'' and 3.6.6, ``Containment Spray System.''
    Date of issuance: July 25, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1-269 and Unit 2-265.
    Renewed Facility Operating License Nos. NPF-35 and NPF-52: 
Amendments revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: March 20, 2012 (77 FR 
16274).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 25, 2012.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian 
Point Nuclear Generating Units 2 and 3 (IP2 and IP3), Westchester 
County, New York

    Date of application for amendment: July 8, 2009, as supplemented by 
letters dated September 28, 2009, October 26, 2009, October 5, 2010, 
October 28, 2010, July 28, 2011, August 23, 2011, October 28, 2011, 
December 15, 2011, January 11, 2012, March 2, 2012, April 23, 2012, and 
May 7, 2012.
    Brief description of amendment: The amendment authorizes the 
transfer of spent fuel from the IP3 spent fuel pool to the IP2 spent 
fuel pool, using a newly-designed shielded transfer canister, for 
further transfer to the on-site Independent Spent Fuel Storage 
Installation.
    Date of issuance: July 13, 2012.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 268 and 246.
    Facility Operating License Nos. DPR-26 and DPR-64: The amendment 
revised the License and the Technical Specifications.
    Date of initial notice in Federal Register: January 21, 2010 (75 FR 
3497).
    The supplements provided additional information that clarified the 
application but did not expand the scope of the application as 
originally noticed.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 13, 2012.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

    Date of application for amendment: September 8, 2010, as 
supplemented by letters dated November 18, 2010, November 23, 2010, 
February 23, 2011 (four letters), March 9, 2011 (two letters), March 
22, 2011, March 30, 2011, March 31, 2011, April 14, 2011, April 21, 
2011, May 3, 2011, May 5, 2011, May 11, 2011, June 8, 2011, June 15, 
2011, June 21, 2011, June 23, 2011, July 6, 2011, July 28, 2011, August 
25, 2011, August 29, 2011, August 30, 2011, September 2, 2011, 
September 9, 2011, September 12, 2011, September 15, 2011, September 
26, 2011, October 10, 2011, October 24, 2011, November 14, 2011, 
November 25, 2011, November 28, 2011, December 19, 2011, February 6, 
2012, February 15, 2012, February 20, 2012, March 13, 2012, March 21, 
2012, April 5, 2012, April 18, 2012 (two letters), April 26, 2012, May 
9, 2012, and June 12, 2012.
    Brief description of amendment: The amendment increased the maximum 
steady-state reactor core power level from 3,898 megawatts thermal 
(MWt) to 4,408 MWt, which is an increase of approximately 15 percent 
from the original licensed thermal power level of 3,833 MWt. The 
proposed increase in power level is considered an extended power 
uprate.
    Date of issuance: July 18, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment No: 191.
    Facility Operating License No. NPF-29: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: January 10, 2011 (76 FR 
1464). The supplemental letters dated November 18, 2010, November 23, 
2010, February 23, 2011 (four letters), March 9, 2011 (two letters), 
March 22, 2011, March 30, 2011, March 31, 2011, April 14, 2011, April 
21, 2011, May 3, 2011, May 5, 2011, May 11, 2011, June 8, 2011, June 
15, 2011, June 21, 2011, June 23, 2011, July 6, 2011, July 28, 2011, 
August 25, 2011, August 29, 2011, August 30, 2011, September 2, 2011, 
September 9, 2011, September 12, 2011, September 15, 2011, September 
26, 2011, October 10, 2011, October 24, 2011, November 14, 2011, 
November 25, 2011, November 28, 2011, December 19, 2011, February 6, 
2012, February 15, 2012, February 20, 2012, March 13, 2012, March 21, 
2012, April 5, 2012, April 18, 2012 (two letters), April 26, 2012, May 
9, 2012, and June 12, 2012, provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the staff's original proposed 
no significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 18, 2012.
    No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket Nos. 50-220, and 50-410, 
Nine Mile Point Nuclear Station, Units 1 and 2, Oswego County, New York

    Date of application for amendments: July 20, 2011, as supplemented 
on November 3, 2011, and January 12, 2012.
    Brief description of amendments: The amendments revised the NMP1 
Technical Specification (TS) Section 5.1, ``Site,'' and associated TS 
Figure 5.1-1, ``Site Boundaries, Nine Mile Point-Unit 1,'' and the NMP2 
TS Figure 4.1-1, ``Site Area and Land Portion of Exclusion Area 
Boundaries,'' to reflect the transfer of a portion of the Nine Mile 
Point Nuclear Station, LLC (NMPNS) site real property located outside 
of the NMPNS Protected Area but within the current NMPNS Owner 
Controlled Area, as well as specified easements over the remainder of 
the NMPNS site, to Nine Mile Point 3 Nuclear Project, LLC (NMP3), a 
subsidiary of UniStar Nuclear Energy, LLC.
    Date of issuance: July 12, 2012.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 212 for Unit 1 and 142 for Unit 2.
    Renewed Facility Operating License Nos. DPR-63 and NPF-69: 
Amendments revised the License and Technical Specifications.
    Date of initial notice in Federal Register: December 27, 2011 (76 
FR 80977).
    The supplements dated November 3, 2011, and January 12, 2012, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the Nuclear Regulatory Commission (NRC) staff's initial proposed 
no significant hazards consideration determination noticed in

[[Page 47131]]

the Federal Register on December 27, 2011 (76 FR 80977).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 12, 2012.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 26th day of July 2012.

    For the Nuclear Regulatory Commission.
Louise Lund,
Deputy Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2012-19004 Filed 8-6-12; 8:45 am]
BILLING CODE 7590-01-P