[Federal Register Volume 78, Number 53 (Tuesday, March 19, 2013)]
[Notices]
[Pages 16876-16889]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2013-06164]


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NUCLEAR REGULATORY COMMISSION

[NRC-2013-0049]


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 February 21, 2013, to March 6, 2013. The 
last biweekly notice was published on March 4, 2013 (78 FR 14126).

ADDRESSES: You may access information and comment submissions related 
to this document, which the NRC possesses and is publically available, 
by searching on http://www.regulations.gov under Docket ID NRC-2013-
0049. 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-2013-0049. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; email: [email protected].
     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-2013-0049 when contacting the NRC 
about the availability of information regarding this document. You may 
access information related to this document by any of the following 
methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0049.
     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 [email protected]. Documents may be viewed in 
ADAMS 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-2013-0049 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 that 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

[[Page 16877]]

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 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 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 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 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 [email protected], or by 
telephone at 301-415-1677, to request (1) a digital

[[Page 16878]]

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 
[email protected], or by a toll-free call at 1-866ndash;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. 
However, a request to intervene will require including information on 
local residence in order to demonstrate a proximity assertion of 
interest in the proceeding. With respect to copyrighted works, except 
for limited excerpts that serve the purpose of the adjudicatory filings 
and would constitute a Fair Use application, participants are requested 
not to include copyrighted materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day deadline will not 
be entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the following three 
factors in 10 CFR 2.309(c)(1): (i) The information upon which the 
filing is based was not previously available; (ii) the information upon 
which the filing is based is materially different from information 
previously available; and (iii) the filing has been submitted in a 
timely fashion based on the availability of the subsequent information.
    For further details with respect to this license amendment 
application, see the 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 [email protected].

[[Page 16879]]

Dairyland Power Cooperative, Docket Nos.: 50-409 and 72-046, La Crosse 
Boiling Water Reactor (LACBWR), La Crosse County, Wisconsin

    Date of amendment request: December 10, 2012.
    Description of amendment request: The proposed amendment would 
revise certain license conditions and to remove TS definitions, 
operational requirements, and specific design requirements that are no 
longer applicable with all spent fuel in dry cask storage at the 
Independent Spent Fuel Storage Installation (ISFSI). The proposed 
changes to the TS also remove administrative control requirements that 
have been relocated to the LACBWR Quality Assurance Program Description 
(QAPD) or are superseded by regulation or other guidance.
    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 proposed changes reflect the complete transfer of all spent 
nuclear fuel from the Fuel Element Storage Well (FESW) to the 
Independent Spent Fuel Storage Installation (ISFSI). Design basis 
SAFSTOR accidents related to the FESW were discussed in the LACBWR 
Decommissioning Plan. These postulated accidents were predicated on 
spent nuclear fuel being stored in the FESW. With the removal of the 
spent fuel from the FESW, there are no remaining important to safety 
systems required to be monitored and there are no remaining credible 
accidents that require that actions of a Certified Fuel Handler to 
prevent occurrence or mitigate the consequences.
    The LACBWR Decommissioning Plan provided a discussion of 
radiological events postulated to occur during SAFSTOR with the 
bounding consequence resulting from a materials handling event. The 
proposed changes do not have an adverse impact on decommissioning 
activities or any postulated consequences.
    The proposed change to the Design Features section of the 
Technical Specifications clarifies that the spent fuel is being 
stored in dry casks within an ISFSI. The probability or consequences 
of accidents at the ISFSI are evaluated in the dry cask vendor's 
FSAR and are independent of the SAFSTOR accidents that were 
evaluated in the LACBWR Decommissioning Plan.
    Therefore, the proposed amendment 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 changes reflect the reduced operational risks as a 
result of the spent nuclear fuel being transferred to dry casks 
within an ISFSI. The proposed changes do not modify any physical 
systems, or components. The plant conditions for which the LACBWR 
Decommissioning Plan design basis accidents relating to spent fuel 
were evaluated are no longer applicable. The proposed changes do not 
affect any of the parameters or conditions that could contribute to 
the initiation of an accident. Design basis accidents associated 
with the dry cask storage of spent fuel are already considered in 
the dry cask system's Final Safety Analysis Report. No new accident 
scenarios are created as a result of deleting non-applicable 
operational and administrative requirements.
    Therefore, the proposed amendment 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.
    As described above, the proposed changes reflect the reduced 
operational risks as a result of the spent nuclear fuel being 
transferred to dry casks within an ISFSI. The design basis and 
accident assumptions within the LACBWR Decommissioning Plan and the 
Technical Specifications relating to spent fuel are no longer 
applicable. The proposed changes do not affect remaining plant 
operations, systems, or components supporting decommissioning 
activities. In addition, the proposed changes do not result in a 
change in initial conditions, system response time, or in any other 
parameter affecting the SAFSTOR accident analysis.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Thomas Zaremba, Wheeler, Van Sickle and 
Anderson, Suite 801, 25 West Main Street, Madison, WI 53703-3398.
    NRC Branch Chief: Bruce Watson.

Detroit Edison, Docket No. 50-016, Fermi 1, Monroe County, Michigan

    Date of amendment request: December 21, 2012.
    Description of amendment request: The proposed amendment 
(ML13002A037) would revise the Fermi 1 operating license to change its 
name on the license to ``DTE Electric Company.'' This name change is 
purely administrative in nature. Detroit Edison is a wholly owned 
subsidiary of DTE Energy Company, and this name change is part of a set 
of name changes of DTE Energy subsidiaries to conform their names to 
the ``DTE'' brand name. No other changes are contained within this 
request. This request does not involve a transfer of control over or of 
an interest in the license for Fermi 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. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed amendment changes the name of the owner licensee. 
The proposed amendment is purely administrative in nature. The 
functions, powers, resources and management of the owner licensee 
will not change. Detroit Edison, which will be renamed DTE Electric 
Company, will remain the licensee of the facility. The proposed 
changes do not adversely affect accident initiators or precursors, 
and do not alter the design assumptions, conditions, or 
configuration of the plant or the manner in which the plant is 
operated and maintained. The ability of structures, systems, and 
components to perform their intended safety functions is not altered 
or prevented by the proposed changes, and the assumptions used in 
determining the radiological consequences of previously evaluated 
accidents are not affected.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed amendment is purely administrative in nature. The 
functions of the owner licensee will not change. These changes do 
not involve any physical alteration of the plant (i.e., no new or 
different type of equipment will be installed), and installed 
equipment is not being operated in a new or different manner. Thus, 
no new failure modes are introduced.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The proposed amendment is a name change to reflect the new name 
of the owner licensee. The proposed amendment is purely 
administrative in nature. The functions of the owner licensee will 
not change. Detroit Edison, which will be renamed DTE Electric 
Company, will remain the licensee of the facility, and its functions 
will not change. The proposed changes do not alter the manner in 
which safety limits, limiting safety system settings, or limiting 
conditions for operation are determined. There are no changes to 
setpoints at which protective

[[Page 16880]]

actions are initiated, and the operability requirements for 
equipment assumed to operate for accident mitigation are not 
affected.
    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: Bruce R. Masters, DTE Energy, General 
Council--Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226-1279.
    NRC Branch Chief: Bruce Watson.

Detroit Edison, Docket No. 50-341, Fermi 2, Monroe County, Michigan

    Date of amendment request: January 11, 2013.
    Description of amendment request: The proposed amendment would 
update the Fermi 2 Updated Final Safety Analysis Report (UFSAR) to 
describe methodology and results of the analysis performed to evaluate 
the protection of the plant's structures, systems and components (SSCs) 
from tornado generated missiles. The analysis is consistent with the 
guidance provided in Regulatory Issue Summary 2008-14, ``Use of TORMIS 
Computer Code for Assessment of Tornado Missile Protection.''
    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.
    Proposed for NRC review and approval are changes to the Fermi 2 
Updated Final Safety Analysis Report (UFSAR) which in essence 
constitute a license amendment to incorporate use of an NRC approved 
methodology to assess the need for additional positive (physical) 
tornado missile protection of specific features at the Fermi 2 site. 
The UFSAR changes will reflect use of the Electric Power Research 
Institute (EPRI) Topical Report ``Tornado Missile Risk Evaluation 
Methodology'' (EPRI NP-2005), Volumes I and II. As noted in the NRC 
Safety Evaluation Report on this topic dated October 26, 1983, the 
current licensing criteria governing tornado missile protection are 
contained in Standard Review Plan (SRP) Sections 3.5.1.4 and 3.5.2. 
These criteria generally specify that safety-related systems be 
provided positive tornado missile protection (barriers) from the 
maximum credible tornado threat. However, SRP Section 3.5.1.4 
includes acceptance criteria permitting relaxation of the above 
deterministic guidance, if it can be demonstrated that the 
probability of damage to unprotected essential safety-related 
features is sufficiently small.
    As permitted in NRC Standard Review Plan (NUREG-0800) sections, 
the combined probability will be maintained below an allowable 
level, i.e., an acceptance criterion threshold, which reflects an 
extremely low probability of occurrence. The Fermi 2 approach 
assumes that if the sum of the individual probabilities calculated 
for tornado missiles striking and damaging portions of important 
systems or components is greater than or equal to 10\-6\ per year 
per unit, then installation of unique missile barriers would be 
needed to lower the total cumulative probability below the 
acceptance criterion of 10-6 per year per unit.
    With respect to the probability of occurrence or the 
consequences of an accident previously evaluated in the UFSAR, the 
possibility of a tornado reaching the Fermi 2 site and causing 
damage to plant structures, systems and components is a design basis 
event considered in the Updated Final Safety Analysis Report. The 
changes being proposed do not affect the probability that the 
natural phenomenon (a tornado) will reach the plant, but from a 
licensing basis perspective they do affect the probability that 
missiles generated by the winds of the tornado might strike and 
damage certain plant systems or components. There are a limited 
number of safety-related components that could theoretically be 
struck and consequently damaged by tornado-generated missiles. The 
probability of tornado-generated missile strikes on ``important'' 
systems and components (as discussed in Regulatory Guide 1.117, 
``Tornado Design Classification'') is what is to be analyzed using 
the probability methods discussed above. The combined probability of 
damage will be maintained below an extremely low acceptance 
criterion to ensure overall plant safety. The proposed change is not 
considered to constitute a significant increase in the probability 
of occurrence or the consequences of an accident, due to the 
extremely low probability of damage due to tornado-generated 
missiles and thus an extremely low probability of a radiological 
release.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of previously evaluated 
accidents.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The possibility of a tornado reaching the Fermi 2 site is a 
design basis event that is explicitly considered in the UFSAR. This 
change involves recognition of the acceptability of performing 
tornado missile probability calculations in accordance with 
established regulatory guidance. The change therefore deals with an 
established design basis event (the tornado). Therefore, the 
proposed change would not contribute to the possibility of a new or 
different kind of accident from those previously analyzed. The 
probability and consequences of such a design basis event are 
addressed in Question 1 above.
    Based on the above discussions, the proposed change will not 
create the possibility of a new or different kind of accident than 
those previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The existing Fermi 2 licensing basis for protection of safety-
related equipment required for safe shutdown from design basis 
tornado generated missiles is to provide positive missile barriers 
for all safety-related systems and components. With the change, it 
will be recognized that there is an extremely low probability, below 
an established acceptance limit, that a limited subset of the 
``important'' systems and components could be struck and 
consequently damaged. The change from protecting all safety-related 
systems and components to ensuring an extremely low probability of 
occurrence of tornado-generated missile strikes and consequential 
damage on portions of important systems and components is not 
considered to constitute a significant decrease in the margin of 
safety due to that extremely low probability.
    Therefore, the changes associated with this license amendment 
request 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: Bruce R. Masters, DTE Energy, General 
Council--Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226-1279.
    NRC Branch Chief: Robert D. Carlson.

Entergy Nuclear Vermont Yankee (VY), LLC and Entergy Nuclear 
Operations, Inc., Docket No. 50-271, Vermont Yankee Nuclear Power 
Station (VYNPS), Vernon, Vermont

    Date of amendment request: December 17, 2012.
    Description of amendment request: The proposed amendment would 
revise VYNPS Technical Specification (TS) 3.3.B to provide an action 
statement for inoperable control rods consistent with the Standard 
Technical Specification (STS) provision (NUREG-1433, Revision 4).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:


[[Page 16881]]


    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment does not significantly increase the 
probability or consequences of an accident. The adding of an 
additional, restrictive action statement for inoperable equipment, 
consistent with the STS does not alter any accident analysis.
    Therefore, the proposed amendment 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 does not involve any new modes of 
operation. The change establishes additional restrictive controls 
for equipment that is considered inoperable. The proposed amendment 
does not change how the control rod system is operated or change the 
design configuration of the control rods. No new accident precursors 
are introduced. No new or different types of equipment will be 
installed. The methods governing plant operation remain bounded by 
current safety analysis assumptions.
    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 does not involve any new methods of 
operation. The change establishes additional restrictive controls 
for equipment that is considered inoperable. The proposed amendment 
does not change how the control rod system is operated or change the 
design configuration of the control rods. No new or different types 
of equipment will be installed. The methods governing plant 
operation remain bounded by current safety analysis assumptions.
    Therefore, the proposed amendment will not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: George Wilson.

Entergy Nuclear Vermont Yankee (VY), LLC and Entergy Nuclear 
Operations, Inc., Docket No. 50-271, Vermont Yankee Nuclear Power 
Station (VYNPS), Vernon, Vermont

    Date of amendment request: December 21, 2012.
    Description of amendment request: The proposed amendment would 
revise the licensing basis relative to how the station satisfies the 
requirements in 10 CFR 50.63, ``Loss of all alternating current 
power.'' The VYNPS currently relies on the Vernon Hydroelectric Station 
(VHS) as the alternate alternating current (AAC) power source providing 
acceptable capability to withstand station blackout under 10 CFR 
50.63(c)(2). The VYNPS proposes to replace the VHS with an onsite 
diesel generator as the AAC power source providing this capability 
which would involve changes to the facility and procedures described in 
the VYNPS Updated Final Safety Analysis Report.
    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 amendment does not significantly increase the 
probability or consequences of an accident. The proposed amendment 
replaces one AAC power source (the VHS) with an additional onsite 
AAC power source (diesel generator). This equipment can not initiate 
a design basis accident and is not used to mitigate the consequences 
of design basis accidents. The equipment is used to mitigate the 
consequences of a station blackout as required by 10 CFR 50.63. 
Station blackout events are not considered design basis accidents 
and do not result in radiological consequences.
    Therefore, the proposed amendment 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 does not involve any new modes of 
operation. The change provides an alternate means to provide AAC 
power to the station. The location of the SBO DG does not create the 
possibility of a different kind of accident. No new accident 
precursors are introduced. Station procedures will be revised to 
align the AAC source to provide the required power within 
established coping times. The methods governing plant operation 
remain bounded by current safety analysis assumptions.
    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 design of the new AAC source will accommodate the loading 
associated with the proceduralized station blackout response and 
safety margins will be maintained. The design of the system will 
meet regulatory guidance and be within station design analysis. The 
station safety analysis results are unchanged and margin to 
regulatory limits is not affected.
    Therefore, the proposed amendment will not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: George Wilson.

Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1, Dauphin County, Pennsylvania

    Date of amendment request: December 14, 2012, as supplemented by 
letter dated January 31, 2013.
    Description of amendment request: The proposed amendment would 
modify the pressure-temperature limit curves and low temperature 
overpressure protection limits in the Three Mile Island Nuclear 
Station, Unit 1 Technical Specification (TS) Section 3.1.2, 
``Pressurization Heatup and Cooldown Limitations,'' TS Section 3.1.12, 
``Pressurizer Power Operated Relief Valve, Block Valve, and Low-
Temperature Overpressure Protection,'' and TS Section 4.5.2, 
``Emergency Core Cooling System.'' The proposed changes reflect revised 
fluence projections out to 50.2 effective full-power years (EFPY) as 
compared to the current projections which go to 29 EFPY. The submittal, 
dated December 14, 2012, also includes a corresponding exemption 
request to use an alternate initial reference temperature for nil-
ductility transition (RTNDT) for Linde 80 weld materials per 
NRC-approved Topical Report BAW-2308, ``Initial RTNDT of 
Linde 80 Weld Materials,'' Revisions 1-A and 2-A.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the

[[Page 16882]]

licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment will revise the reactor coolant system 
heatup, cooldown, and inservice leak hydrostatic test limitations 
(Technical Specification (TS) Section 3.1.2 (``Pressurization Heatup 
and Cooldown Limitations'')) for the Reactor Coolant System (RCS) to 
a maximum of 50.2 Effective Full Power Years (EFPY) in accordance 
with 10 CFR Part 50, Appendix G. Further, the proposed amendment 
revises TMI, Unit 1 Technical Specification Sections 3.1.12 
(``Pressurizer Power Operated Relief Valve (PORV), Block Valve, and 
Low Temperature Overpressure Protection (LTOP)''), and 4.5.2 
(``Emergency Core Cooling System'') for Low Temperature Overpressure 
Protection (LTOP) requirements to reflect the revised P-T limits of 
the reactor vessel. P-T limits for the TMI, Unit 1 reactor vessel 
were developed in accordance with the requirements of 10 CFR Part 
50, Appendix G (``Fracture Toughness Requirements''), utilizing the 
analytical methods and flaw acceptance criteria of Topical Report 
BAW-10046A (AREVA NP Document BAW-10046A, Rev. 2, ``Methods of 
Compliance with Fracture Toughness and Operational Requirements of 
10 CFR Part 50, Appendix G,'' by H. W. Behnke et al., June 1986) and 
ASME Code Section XI, Appendix G (``Fracture Toughness Criteria for 
Protection Against Failure,'' 1995 Edition with Addenda through 
1996) which are previously approved NRC standards for the 
preparation of P-T limit curves. Updating the P-T limit curves for 
additional EFPY maintains the level of assurance that Reactor 
Coolant Pressure Boundary integrity will be maintained, as specified 
in 10 CFR Part 50, Appendix G. Additionally, this proposed amendment 
deletes administrative requirements contained in TS 3.1.2.4 and 
3.1.2.5 which provide reporting requirements related to the 
preparation and submittal of P-T curves that are outdated or 
contained in regulation.
    The proposed changes do not adversely affect accident initiators 
or precursors, and do not alter the design assumptions, conditions, 
or configuration of the plant or the manner in which the plant is 
operated and maintained. The ability of structures, systems, and 
components to perform their intended safety functions is not altered 
or prevented by the proposed changes, and the assumptions used in 
determining the radiological consequences of previously evaluated 
accidents are not affected.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes incorporate methodologies that either have 
been approved or accepted for use by the NRC (provided that any 
conditions/limitations are satisfied). The P-T limit curves and LTOP 
limits will provide the same level of protection to the Reactor 
Coolant Pressure Boundary as was previously evaluated. Reactor 
Coolant Pressure Boundary integrity will continue to be maintained 
in accordance with 10 CFR Part 50, Appendix G, and the assumed 
accident performance of plant structures, systems and components 
will not be affected. Additionally, this proposed amendment deletes 
administrative requirements contained in TS 3.1.2.4 and 3.1.2.5. 
These changes do not involve any physical alteration of the plant 
(i.e., no new or different type of equipment will be installed), and 
installed equipment is not being operated in a new or different 
manner. Thus, no new failure modes are introduced.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not affect the function of the Reactor 
Coolant Pressure Boundary or its response during plant transients. 
By calculating the P-T limits and associated LTOP limits using NRC-
approved methodology, adequate margins of safety relating to Reactor 
Coolant Pressure Boundary integrity are maintained. Additionally, 
this proposed amendment deletes administrative requirements 
contained in TS 3.1.2.4 and 3.1.2.5. The proposed changes do not 
alter the manner in which safety limits, limiting safety system 
settings, or limiting conditions for operation are determined. These 
changes will ensure that protective actions are initiated and the 
operability requirements for equipment assumed to operate for 
accident mitigation are not affected.
    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: J. Bradley Fewell, Esquire, Associate 
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Meena Khanna.

Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1, Dauphin County, Pennsylvania

    Date of amendment request: February 4, 2013.
    Description of amendment request: The proposed amendment would 
delete various reporting requirements contained in the Technical 
Specifications (TSs). Specifically, the proposed amendment will delete 
the Sealed Source Contamination Special Report and the Startup Report, 
as well as the plant-specific annual reports regarding periodic Leak 
Reduction Program tests, Pressurizer Power Operated Relief Valve and 
Pressurizer Safety Valve challenges, specific activity analysis in 
which the primary coolant exceeds the limits of TS 3.1.4.1, and major 
changes to radioactive waste treatment systems.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not involve the modification of any 
plant equipment or affect plant operation. The proposed changes will 
have no impact on any safety related structures, systems, or 
components. The reporting requirements proposed for deletion are not 
required because the requirements are adequately addressed by other 
regulatory requirements, or are no longer warranted.
    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 have no impact on the design, function or 
operation of any plant structure, system or component. The proposed 
changes do not affect plant equipment or accident analyses. The 
reporting requirements proposed for deletion are not required 
because the requirements are adequately addressed by other 
regulatory requirements, or are no longer warranted.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes do not adversely affect existing plant 
safety margins or the reliability of the equipment assumed to 
operate in the safety analyses. 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.
    Margins of safety are unaffected by deletion of the reporting 
requirements.

    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.

[[Page 16883]]

    Attorney for licensee: J. Bradley Fewell, Esquire, Associate 
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Meena Khanna.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, 
Davis-Besse Nuclear Power Station, Unit 1 (DBNPS), Ottawa County, Ohio

    Date of amendment request: January 18, 2013.
    Description of amendment request: The amendment would revise DBNPS 
Technical Specification (TS) 3.4.17, ``Steam Generator (SG) Tube 
Integrity''; TS 3.7.18, ``Steam Generator Level''; TS 5.5.8, ``Steam 
Generator (SG) Program''; and TS 5.6.6, ``Steam Generator Tube 
Inspection Report.'' The proposed revision to these TSs is to support 
plant operations following the replacement of the original SGs which is 
scheduled to be completed in April 2014. The proposed changes to TS 
3.4.17, TS 5.5.8, and TS 5.6.6 would impose requirements that reflect 
the analysis and tube materials of the replacement SGs. These changes 
are consistent with Technical Specifications Task Force (TSTF) traveler 
TSTF-510, Revision 2, ``Revision to Steam Generator Program Inspection 
Frequencies and Tube Sample Selection,'' which was approved by the U.S. 
Nuclear Regulatory Commission on October 27, 2011. The proposed 
revision to TS 5.5.8 also includes minor editorial changes and 
eliminates the requirements for special visual inspections of the 
internal auxiliary feedwater header, since this component will not be 
part of the replacement SGs.
    The proposed changes to TS 3.7.18 would impose inventory limits on 
the secondary-side that reflect the design characteristics and 
dimensions of the replacement SGs. The revised limits will ensure that 
plant operations with the replacement SGs is bounded by the values used 
in the existing main steam line break analysis presented in the DBNPS 
updated safety analysis report.
    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.
    For TS 3.4.17, ``Steam Generator (SG) Tube Integrity,'' a steam 
generator tube rupture (SGTR) event is the relevant design basis 
accident analyzed in the licensing basis for DBNPS. TS 3.4.17 and TS 
5.5.8, ``Steam Generator (SG) Program,'' impose monitoring and 
inspection requirements that ensure tube integrity is maintained. 
The proposed changes to these TSs would implement monitoring and 
inspection requirements appropriate for the design and materials of 
the replacement SGs. The proposed SG tube inspection frequency and 
sample selection criteria will continue to ensure that the SG tubes 
are inspected such that that the integrity of the SG tubes is 
verified to be maintained at a level that prevents an increase in 
the probability of a SGTR.
    Therefore the proposed changes to these TSs will not increase 
the probability of a SGTR.
    The radiological consequences of a SGTR are bounded by using 
conservative assumptions in the design basis accident analysis, and 
are dependent upon the pre-existing primary-to-secondary leak rate, 
the flow rate through the ruptured tube, the radiological isotopic 
content of the RCS [reactor coolant system] and the release paths. 
The monitoring and inspection requirements imposed by TS 3.4.17 and 
TS 5.5.8 are intended to ensure that SG tube integrity is 
maintained. The proposed changes to these TSs would implement 
monitoring and inspection requirements appropriate for the design 
and materials of the replacement SGs and would not affect 
radiological releases in the event of an SGTR. The radiological 
isotopic content of the RCS and the release paths are not affected 
by any of the requirements in the current TS 3.4.17 or TS 5.5.8 or 
proposed revisions thereto. Therefore, the proposed changes to these 
TSs will not increase the consequences of a SGTR.
    TS 5.6.6, ``Steam Generator Tube Inspection Report,'' specifies 
information that is to be reported to the NRC following SG 
inspections performed in accordance with the Steam Generator Program 
requirements contained in TS 5.5.8. The requirement to provide this 
report is administrative in nature and the content of this report 
can have no effect on the probability or the consequences of an 
accident previously evaluated.
    LCO [limiting condition for operation] 3.7.18, ``Steam Generator 
Level'' ensures that the plant is operated within the SG inventory 
limits that were used as initial conditions in the current accident 
analysis for a Main Steam Line Break (MSLB). The SG inventory is not 
an accident initiator and does not affect any accident initiator. 
Therefore, the proposed changes in SG inventory limits will not 
increase the probability of a MSLB accident.
    The radiological consequences of a MSLB are dependent upon the 
total SG inventory released, the SG primary-to-secondary leakage 
rate, the radiological isotopic content of the RCS, and the release 
paths. The revision to LCO 3.7.18 will ensure that the total 
inventory released remains bounded by the existing analysis. None of 
the other factors listed above are affected by the revised operating 
limits on SG inventory that are proposed in the revisions to LCO 
3.7.18.
    Therefore, the proposed changes in SG inventory limits will not 
increase the consequences of a MSLB.
    Based on the above, 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 support replacement of the SGs at the 
DBNPS. Replacement of the SGs is being performed as a design 
modification in accordance with the provisions of 10 CFR 50.59, 
``Changes, tests and experiments.'' The proposed changes to TS 
3.4.17, TS 5.5.8 and TS 5.6.6 would implement monitoring and 
inspection requirements appropriate for the design and materials of 
the replacement SGs, and establish appropriate reporting 
requirements. These changes would not affect the method of operation 
of the SGs. The proposed changes to TS 3.7.18 would ensure that the 
replacement SGs will be operated in accordance with existing 
analyses. None of the proposed changes would introduce any changes 
to the plant design. In addition, the proposed changes would not 
impact any other plant system or component.
    The proposed changes would continue to prevent loss of SG tube 
integrity, and would ensure operation within the bounds of existing 
accident analyses. There are no accident initiators created or 
affected by these changes. Therefore, the proposed changes will 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 SG tubes in pressurized water reactors are an integral part 
of the reactor coolant system (RCS) pressure boundary and, as such, 
are relied upon to maintain the primary system's pressure and 
inventory. As part of the RCS pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
also isolate the radioactive fission products in the primary coolant 
from the secondary system. In summary, the safety function of a SG 
is maintained by ensuring the integrity of its tubes and the ability 
to remove residual heat from the primary system.
    The proposed changes will ensure that the existing margins of 
safety are maintained following the replacement of SGs. The changes 
to LCO 3.4.17 and TSs 5.5.8 and 5.6.6 impose requirements for SG 
tube integrity monitoring, inspection, and reporting that will 
ensure that there is no reduction in the ability of the tubes to 
perform their RCS pressure boundary and heat transfer functions. The 
changes to LCO 3.7.18 ensure the MSLB accident analyses remain 
bounding.
    Therefore, the proposed changes do not involve a significant 
reduction[middot] in a margin of safety.


[[Page 16884]]


    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 W. Jenkins, FirstEnergy Corporation, 
76 South Main Street, Akron, Ohio 44308.
    NRC Branch Chief: Jeremy S. Bowen.

Florida Power and Light Company, Docket Nos. 50-335 and 50-389, St. 
Lucie Plant, Units 1 and 2, St. Lucie County, Florida

    Date of amendment request: December 27, 2012.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to align St. Lucie TSs with 
Combustion Engineering Owners Group TSs language describing required 
licensed Senior Reactor Operator (SRO) duties during fuel handling 
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 proposed changes will not result in any significant increase 
in the probability or consequences of an accident previously 
evaluated, as the proposed TS changes are consistent with Standard 
Technical Specifications. Further, not requiring licensed SRO 
oversight of fuel handling operations other than core alterations 
does not introduce additional risk or a greater potential for 
consequences of an accident that has not previously been evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes are administrative in nature and do not 
involve a physical modification of the plant. No new or different 
type of equipment will be installed. The methods for conducting core 
alterations and other fuel handling operations will remain the same. 
The proposed changes will not introduce new failure modes/effects 
that could lead to an accident for which consequences exceed that of 
accidents previously analyzed. 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 change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes will not involve a significant reduction in 
a margin of safety in that the changes are administrative in nature. 
No plant equipment or accident analyses will be affected. 
Additionally, the proposed changes will not relax any criteria used 
to establish safety limits, safety system settings, or the bases for 
any limiting conditions for operation. Safety analysis acceptance 
criteria are not affected. Plant operation will continue within the 
design basis. The proposed changes do not adversely affect systems 
that respond to safely shutdown the plant and maintain the plant in 
a safe shutdown condition. Consequently, 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: James Petro, Managing Attorney--Nuclear, 
Florida Power & Light, P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Branch Chief: Jessie F. Quichocho.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of amendment request: June 6, 2012.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to eliminate the requirements 
that the average power range monitoring (APRM) system ``Upscale'' and 
``Inoperative'' scram and control rod withdrawal block functions be 
operable in Operational Condition (OPCON) 5, refueling operations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below with the NRC staff's edits in 
square brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The APRM system is not an initiator of or a precursor to any 
accident or transient. The APRM system monitors the neutron flux 
level in the power operating range from approximately one percent to 
greater than rated thermal power and initiates automatic protective 
actions for postulated at-power reactivity insertion events. Thus, 
the proposed changes to the TS operability requirements for the APRM 
system will not impact the probability of any previously evaluated 
accident.
    The design of plant equipment is not being modified by the 
proposed amendment. The TSs will continue to require operability of 
the APRM system ``Upscale'' and ``Inoperative'' scram and control 
rod withdrawal block functions when the reactor is in the Startup 
and Run modes (OPCON 2 and OPCON 1) to provide core protection for 
postulated reactivity insertion events occurring during power 
operating conditions. Thus, the consequences of previously evaluated 
at-power reactivity insertion events are not affected by the 
proposed amendment.
    The proposed elimination of the TS requirements that the APRM 
system ``Upscale'' and ``Inoperative'' scram and control rod 
withdrawal block functions be operable when the reactor is in the 
Refueling mode (OPCON 5) also does not increase the consequences of 
an accident previously evaluated. The possibility of inadvertent 
criticality due to a control rod withdrawal error during refueling 
is minimized by design features and procedural controls that are not 
affected by the proposed amendment. Since the core is designed to 
meet shutdown requirements with the highest worth rod withdrawn, the 
core remains subcritical even with one rod withdrawn. Any attempt to 
withdraw a second rod results in a rod block by the Refueling 
Interlocks (RI). In addition, since reactor neutron flux levels 
during refueling are below the APRM indicating range, the APRM 
system does not provide any meaningful core monitoring or protection 
in the refueling operating condition (OPCON 5). The source range 
(SRM) and intermediate range (IRM) neutron monitoring systems 
provide adequate neutron flux monitoring during refueling and 
automatically initiate protective actions (scram or control rod 
withdrawal block) when required during refueling.
    Additionally, if the infrequently performed TS 3/4.10.3, 
``Shutdown Margin Demonstrations,'' is performed in OPCON 5, the 
additional controls and restrictions in place during this test are 
sufficiently robust even without the RIs when the mode switch is 
temporarily placed in Startup. In addition to the OPCON 5 SRM and 
IRM protective actions, the SRM RPS [reactor protection system] trip 
is made operable, the RWM [rod worth minimizer] is operable and 
programmed for the shutdown margin demonstration, use of the ``rod-
out-notch-override'' control is prohibited, and no other core 
alterations are allowed. Therefore, during this infrequent 
operation, operability of the APRMs is not required as they would 
not provide any meaningful core monitoring or protection.
    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 changes to the TS operability requirements for the 
APRM system do not introduce any new accident precursors and do not 
involve any physical plant alterations or changes in the methods 
governing normal

[[Page 16885]]

plant operation that could initiate a new or different kind of 
accident. The proposed amendment does not alter the intended 
function of the APRM system and does not affect the ability of the 
system to provide core protection for at-power reactivity insertion 
events. The other existing TS-required neutron monitoring systems 
(SRM and IRM) provide for core monitoring and protection in the 
refueling mode (OPCON 5). Additionally, if the infrequently 
performed TS 3/4.10.3, ``Shutdown Margin Demonstrations'' is 
performed in OPCON 5, the additional controls and restrictions in 
place during this test are sufficiently robust even without the RIs 
when the mode switch is temporarily placed in ``Startup.''
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the amendment involve a significant reduction in a 
margin of safety?
    Response: No.
    Margin of safety is related to the ability of the fission 
product barriers (fuel cladding, reactor coolant system, and primary 
containment) to perform their design functions during and following 
postulated accidents. The proposed amendment does not alter 
setpoints or limits established or assumed by the accident analyses. 
The proposed TS changes to eliminate the requirements that the APRM 
system ``Upscale'' and ``Inoperative'' scram and control rod 
withdrawal block functions be operable when in OPCON 5 have no 
impact on the performance of the fission product barriers. These 
APRM functions do not provide any meaningful core monitoring or 
protection in the Refueling operating condition, including the 
infrequently performed special test TS 3/4.10.3. The other existing 
TS required neutron monitoring systems (SRM and IRM) provide for 
core monitoring and protection in the refueling mode (OPCON 5). In 
the Startup and Run modes the TSs will continue to require 
operability of these APRM functions to provide core protection for 
postulated reactivity insertion events occurring during power 
operating conditions, consistent with the plant safety analyses.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, and with the changes noted above in square brackets, it 
appears that the three standards of 10 CFR 50.92(c) are satisfied. 
Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, PSEG Nuclear LLC--N21, 
P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Meena K. Khanna.

Southern Nuclear Operating Company Docket Nos.: 52-025 and 52-026, 
Vogtle Electric Generating Plant (VEGP) Units 3 and 4, Burke County, 
Georgia

    Date of amendment request: February 15, 2013.
    Description of amendment request: The proposed change would amend 
Combined Licenses Nos.: NPF-91 and NPF-92 for Vogtle Electric 
Generating Plant (VEGP) Units 3 and 4 by departing from the plant-
specific design control document Tier 2* material by revising reference 
document APP-OCS-GEH-320, ``AP1000 Human Factors Engineering Integrated 
System Validation Plan'' from Revision D to Revision 2. APP-OCS-GEH-320 
is incorporated by reference in the updated final safety analysis 
report (UFSAR) as a means to implement the activities associated with 
the human factors engineering verification and validation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The Integrated System Validation (ISV) provides a comprehensive 
human performance-based assessment of the design of the AP1000 
Human-System Interface (HSI) resources, based on their realistic 
operation within a simulator-driven Main Control Room (MCR). The ISV 
is part of the overall AP1000 Human Factors Engineering (HFE) 
program. The changes are to the ISV Plan to clarify the scope and 
amend the details of the methodology. The ISV Plan is needed to 
perform, in the simulator, the scenarios described in the document. 
The functions and tasks allocated to plant personnel can still be 
accomplished after the proposed changes. The performance of the 
tests governed by the ISV Plan provides additional assurances that 
the operators can appropriately respond to plant transients. The ISV 
Plan does not affect the plant itself. Changing the ISV Plan does 
not affect prevention and mitigation of abnormal events, e.g., 
accidents, anticipated operational occurrences, earthquakes, floods 
and turbine missiles, or their safety or design analyses. No safety-
related structure, system, component (SSC) or function is adversely 
affected. The changes do not involve nor interface with any SSC 
accident initiator or initiating sequence of events, and thus, the 
probabilities of the accidents evaluated in the UFSAR are not 
affected. Because the changes do not involve any safety-related SSC 
or function used to mitigate an accident, the consequences of the 
accidents evaluated in the UFSAR are not affected.
    Therefore, there is no significant increase in the probability 
or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The changes to the ISV Plan affect the testing and validation of 
the Main Control Room and Human System Interface using a plant 
simulator.
    Therefore, the changes do not affect the safety-related 
equipment itself, nor do they affect equipment which, if it failed, 
could initiate an accident or a failure of a fission product 
barrier. No analysis is adversely affected. No system or design 
function or equipment qualification will be adversely affected by 
the changes. This activity will not allow for a new fission product 
release path, nor will it result in a new fission product barrier 
failure mode, nor create a new sequence of events that would result 
in significant fuel cladding failures. In addition, the changes do 
not result in a new failure mode, malfunction or sequence of events 
that could affect safety or safety-related equipment.
    Therefore, this activity does not create the possibility of a 
new or different kind of accident than any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The changes to the ISV Plan affect the testing and validation of 
the Main Control Room and Human System Interface using a plant 
simulator. Therefore, the changes do not affect the assessments or 
the plant itself. These changes do not affect safety-related 
equipment or equipment whose failure could initiate an accident, nor 
does it adversely interface with safety-related equipment or fission 
product barriers. No safety analysis or design basis acceptance 
limit/criterion is challenged or exceeded by the requested change.

    Therefore, there is no significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. M. Stanford Blanton, Blach & Bingham 
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Acting Branch Chief: Lawrence Burkhart.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant 
(WBN), Unit 1, Rhea County, Tennessee

    Date of amendment request: November 19, 2012.
    Description of amendment request: The proposed amendment would

[[Page 16886]]

change the Technical Specification (TS) 3.7.10 to require a unit 
shutdown within the TS 3.7.10 Actions instead of entering Limiting 
Condition for Operation (LCO) 3.0.3 when both Control Room Emergency 
Ventilation System (CREVS) trains are inoperable in MODE 1, 2, 3, or 4 
due to actions taken as a result of a tornado warning and the 
Completion Time of 8 hours for restoration of at least one CREVS train 
to OPERABLE status is not met.
    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 modify WBN Unit 1 TS 3.7.10 to resolve a 
potential conflict in applying the appropriate actions for not 
meeting the Required Action and associated Completion Time of 
Condition E. These proposed changes are acceptable in the event that 
both CREVS trains are inoperable in MODE 1, 2, 3, or 4 due to 
actions taken as a result of a tornado warning and the Completion 
Time of 8 hours for restoration of at least one CREVS train to 
OPERABLE status is not met because the requirements to shutdown the 
unit to Mode 3 and Mode 5 are similar to the current requirements, 
the required Completion Times are 1 hour less than the existing LCO 
3.0.3 Completion Times that currently apply, and do not impact the 
design and operation of the CREVS, or the ultimate Actions required 
to be taken by TS 3.7.10 upon inoperability of the CREVS in MODE 1, 
2, 3, or 4 due to actions taken as a result of a tornado warning. 
The proposed changes do not (1) require physical changes to plant 
systems, structures, or components; (2) prevent the safety function 
of any safety-related system, structure, or component during a 
design basis event; (3) alter, degrade, or prevent action described 
or assumed in any accident described in the WBN Unit 1 UFSAR from 
being performed since the safety-related systems, structures, or 
components are not modified; (4) alter any assumptions previously 
made in evaluating radiological consequences; or (5) affect the 
integrity of any fission product barrier.
    Therefore, the proposed change does not involve a significant 
increase in the probability or 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 modify WBN Unit 1 TS 3.7.10 to resolve a 
potential conflict in applying the appropriate, actions for not 
meeting the Required Action and associated Completion Time of 
Condition E. These proposed changes are acceptable in the event that 
both CREVS trains are inoperable in MODE 1, 2, 3, or 4 due to 
actions taken as a result of a tornado warning and the Completion 
Time of 8 hours for restoration of at least one CREVS train to 
OPERABLE status is not met because the requirements to shutdown the 
unit to Mode 3 and Mode 5 are similar to the current requirements, 
the required Completion Times are 1 hour less than the existing LCO 
3.0.3 Completion Times that currently apply, and do not impact the 
design and operation of the CREVS, or the ultimate Actions required 
to be taken by TS 3.7.10 upon inoperability of the CREVS in MODE 1, 
2, 3, or 4 due to actions taken as a result of a tornado warning. 
The proposed changes do not introduce any new accident causal 
mechanisms, since no physical changes are being made to the plant, 
nor do they impact any plant systems that are potential accident 
initiators.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes modify WBN Unit 1 TS 3.7.10 to resolve a 
potential conflict in applying the appropriate actions for not 
meeting the Required Action and associated Completion Time of 
Condition E. These proposed changes are acceptable in the event that 
both CREVS trains are inoperable in MODE 1, 2, 3, or 4 due to 
actions taken as a result of a tornado warning and the Completion 
Time of 8 hours for restoration of at least one CREVS train to 
OPERABLE status is not met because the requirements to shutdown the 
unit to Mode 3 and Mode 5 are similar to the current requirements, 
the required Completion Times are 1 hour less than the existing LCO 
3.0.3 Completion Times that currently apply, and do not impact the 
design and operation of the CREVS, or the ultimate Actions required 
to be taken by TS 3.7.10 upon inoperability of the CREVS in MODE 1, 
2, 3, or 4 due to actions taken as a result of a tornado warning. As 
such, there is no impact on the safety analysis for the CREVS. The 
proposed changes do not alter the permanent plant design, including 
instrument set points, that is the basis of the assumptions 
contained in the safety analyses.
    Therefore, the proposed amendment does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Jessie F. Quichocho.

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 [email protected].

[[Page 16887]]

Carolina Power and Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of application for amendments: March 6, 2012, as supplemented 
by letters dated August 29, 2012, September 21, 2012, November 29, 
2012, and January 22, 2013.
    Brief Description of amendments: The amendments revise Technical 
Specification (TS) 5.6.5.b by replacing AREVA Topical Report ANF-
524(P)(A), ANF Critical Power Methodology for Boiling Water Reactors 
with AREVA Topical Report ANP-I 0307PA, Revision 0, AREVA MCPR Safety 
Limit Methodology for Boiling Water Reactors, June 2011, in the list of 
analytical methods that have been reviewed and approved by the U.S. 
Nuclear Regulatory Commission for determining core operating limits, 
(2) revise TS 2.1.1, ``Reactor Core SLs [Safety Limits],'' by 
incorporating revised Safety Limit Minimum Critical Power Ratio 
(SLMCPR) values, and (3) revise the license condition in Appendix B, 
``Additional Conditions,'' of the operating licenses regarding an 
alternate method for evaluating SLMCPR values.
    Date of issuance: March 1, 2013.
    Effective date: Date of issuance, to be implemented prior to the 
startup from the 2014 Unit 1 refueling outage for Unit 1 changes, and 
prior to the startup from the 2013 Unit 2 refueling outage for Unit 2 
changes.
    Amendment Nos.: Unit 1--262 and Unit 2--290.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: July 3, 2012 (77 FR 
39524). The supplements dated August 29, 2012, September 21, 2012, 
November 29, 2012, and January 22, 2013, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 1, 2013.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 6, 2012.
    Brief Description of amendment: The amendment allows a delay time 
for entering a supported system Technical Specification (TS) when the 
inoperability is due solely to an inoperable snubber, if risk is 
assessed and managed consistent with the program in place for complying 
with the requirements of 10 CFR 50.65(a)(4). Limiting Condition for 
Operation (LCO) 3.0.8 is added to the TS to provide this allowance and 
define the requirements and limitations for its use.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-372, Revision 4. The NRC staff 
issued a notice of opportunity for comment in the Federal Register on 
November 24, 2004 (69 FR 68412), on possible amendments concerning 
TSTF-372, including a model safety evaluation and model no significant 
hazards consideration (NSHC) determination, using the consolidated line 
item improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on May 4, 2005 (70 FR 23252).
    Date of issuance: February 26, 2013.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 232.
    Renewed Facility Operating License No. DPR-23: Amendment changed 
the license and TSs.
    Date of initial notice in Federal Register: October 16, 2012 (77 FR 
63347).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 26, 2013.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power 
Station, Unit 2, New London County, Connecticut

    Date of amendment request: April 13, 2012.
    Description of amendment request: The proposed amendment would 
revise the Millstone Power Station, Unit 2 (MPS2) Technical 
Specification (TS) requirements related to diesel fuel oil testing 
consistent with NUREG-1432, Rev. 3.1, ``Standard Technical 
Specifications, Combustion Engineering Plants,'' December 1, 1995, and 
NRC approved Technical Specification Task Force (TSTF) TSTF-374, 
``Revision to TS 5.5.13 and Associated TS Bases for Diesel Fuel Oil,'' 
Revision 0.
    Date of issuance: March 5, 2013.
    Effective date: As of the date of issuance, and shall be 
implemented within 120 days. Amendment No.: 313.
    Renewed Facility Operating License No. DPR-65: Amendment revised 
the License and Technical Specifications.
    Date of initial notice in Federal Register: June 12, 2012 (77 FR 
35072). The supplemental letter dated May 7, 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.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 5, 2013.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point 
Nuclear Generating Unit 3, Westchester County, New York

    Date of application for amendment: August 14, 2012, as supplemented 
by letters dated October 25, November 14, and December 13, 2012, and 
February 15, 2013.
    Brief description of amendment: The amendment revises Technical 
Specification 3.5.4, ``Refueling Water Storage Tank,'' to permit non-
seismically qualified piping of the Spent Fuel Pool purification system 
to be connected to the Refueling Water Storage Tank seismic piping 
under administrative controls for a limited period of time in order to 
purify the contents of the Refueling Water Storage Tank.
    Date of issuance: February 22, 2013.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 250.
    Facility Operating License No. DPR-64: The amendment revised the 
License and the Technical Specifications.
    Date of initial notice in Federal Register: October 16, 2012 (77 FR 
63350). The letters dated October 25, November 14, and December 13, 
2012, and February 15, 2013, 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 February 22, 2013.
    No significant hazards consideration comments received: No.

[[Page 16888]]

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear 
Plant, Van Buren County, Michigan

    Date of application for amendment: February 28, 2012, supplemented 
by letters dated September 6, 2012, November 7, 2012, November 29, 
2012, February 21, 2013 and February 25, 2013.
    Brief description of amendment: The amendment revises the PNP TSs 
to support the replacement of the Region I main spent fuel (SFP) 
storage racks and the storage racks in the north tilt pit portion of 
the SFP, with new neutron absorber Metamic-equipped racks. The 
replacement of the SFP storage racks will allow recovery of the 
currently unusable storage locations in the SFP.
    Date of issuance: February 28, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 250.
    Facility Operating License No. DPR-20: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 5, 2012 (77 FR 
33246). The supplemental letters dated September 6, 2012, November 7, 
2012, November 29, 2012, February 21, 2013 and February 25, 2013, 
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 February 28, 2013.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-374, LaSalle County 
Station, Unit 2, LaSalle County, Illinois

    Date of application for amendments: October 11, 2012, as 
supplemented by letters dated January 17, February 20, and February 26, 
2013.
    Brief description of amendments: The amendment request proposed 
changes to the Technical Specifications (TSs) to revise Section 2.1.1, 
``Reactor Core SLs,'' minimum critical power ratio safety limit (MCPR 
SL) from >= 1.11 to >= 1.14 for two-loop recirculation operation and 
from >= 1.12 to >=1.17 for a single-loop recirculation operation.
    Date of issuance: February 27, 2013.
    Effective date: As of the date of issuance and shall be implemented 
after Cycle 14 is completed and prior to the operation of Cycle 15.
    Amendment No.: 192.
    Facility Operating License Nos. NPF-18: The amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: November 5, 2012 (77 FR 
66489).
    The January 17, February 20, and February 26, 2013, supplements 
contained clarifying information and did not change the NRC staff's 
initial proposed finding of no significant hazards consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 27, 2013.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: September 22, 2011, as supplemented by 
letters dated March 30, September 10 and 28, 2012, and January 3, 2013.
    Brief description of amendment: The amendment revised the curves in 
Technical Specification (TS) 3.4.9, ``RCS [Reactor Coolant System] 
Pressure and Temperature (P/T) Limits,'' to replace the 28 Effective 
Full Power Years (EFPY) restriction in TS Figures 3.4.9-1, 3.4.9-2, and 
3.4.9-3 and the minimum temperature in Surveillance Requirement (SR) 
3.4.9.5, SR 3.4.9.6, and SR 3.4.9.7. The amendment would include a set 
of updated P/T curves for pressure test, core not critical, and core 
critical conditions for 32 EFPY based on a fluence evaluation performed 
using NRC-approved fluence methodology. The new curves would show a 
shift of minimum operating temperature which allows the bolt-up and 
minimum temperatures specified for SR 3.4.9.5, SR 3.4.9.6, and SR 
3.4.9.7 to be changed from 80 degrees Fahrenheit ([deg]F) to 
70[emsp14][deg]F.
    Date of issuance: February 22, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 245.
    Renewed Facility Operating License No. DPR-46: Amendment revised 
the Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: March 6, 2012 (77 FR 
13372). The supplemental letters dated March 30, September 10 and 28, 
2012, and January 3, 2013, 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 February 22, 2013.
    No significant hazards consideration comments received: No.

Northern States Power Company--Minnesota (NSPM), Docket No. 50-263, 
Monticello Nuclear Generating Plant (MNGP), Wright County, Minnesota

    Date of application for amendment: January 20, 2012, as 
supplemented on December 7, 2012.
    Brief description of amendment: The amendment revises the MNGP 
Technical Specifications (TS) Section 1.0, ``Definitions,'' Section 
3.4.9, ``RCS [Reactor Coolant System] Pressure and Temperature (P-T) 
Limits,'' and Section 5.6, ``Administrative Controls.'' The amendment 
revises the P-T limits based on a methodology documented in the SIR-05-
044-A report, ``Pressure-Temperature Limits Report [PTLR] Methodology 
for Boiling Water Reactors,'' and relocates the revised P-T limits from 
the TS to the MNGP PTLR.
    Date of issuance: February 27, 2013.
    Effective date: This license amendment is effective as of the date 
of its date of issuance and shall be implemented within 180 days after 
start-up from the 2013 Refueling Outage.
    Amendment No.: 172.
    Renewed Facility Operating License No. DPR-22: Amendment revises 
the Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: April 17, 2012 (77 FR 
22815). The licensee's December 7, 2012, supplemental letter did not 
change the scope of the original amendment request, did not change the 
NRC staff's initial proposed finding of no significant hazards 
consideration determination, and did not expand the scope of the 
original Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 27, 2013.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: February 10, 2012, as supplemented by 
letters dated October 1, 2012, and January 22, 2013.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) to establish the limiting condition for operation 
(LCO)

[[Page 16889]]

requirements for the reactor protective system (RPS) actuation circuits 
in TS 2.15, ``Instrumentation and Control Systems.'' Specifically, the 
TS changes renumbered LCOs 2.15(1) through 2.15(4) to 2.15.1(1) through 
2.15.1(4), renumbered LCO 2.15(5) to LCO 2.15.3 with an associated 
Table 2-6, ``Alternate Shutdown and Auxiliary Feedwater Panel 
Functions,'' and implemented a new LCO 2.15.2 for the RPS logic and 
trip initiation channels. The amendment also revised the TS Table of 
Contents to reflect the renumbering and addition of the LCO for the RPS 
logic and trip initiation channels and the new Table 2-6.
    Date of issuance: February 28, 2013.
    Effective date: As of its date of issuance and shall be implemented 
within 180 days from the date of issuance.
    Amendment No.: 270.
    Renewed Facility Operating License No. DPR-40: The amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 7, 2012 (77 FR 
47128). The supplemental letters dated October 1, 2012, and January 22, 
2013, 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 February 28, 2013.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of application for amendment: March 1, 2012, as supplemented 
by letter dated December 21, 2012.
    Brief description of amendments: The proposed amendment would make 
miscellaneous changes to the Technical Specifications (TS) and Facility 
Operating License (FOL) including: (1) Correction of typographical 
errors; (2) deletion of historical requirements that have expired; (3) 
corrections of errors or omissions from previous license amendment 
requests; and (4) updating of components lists to reflect current plant 
design.
    Date of issuance: February 25, 2013.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 193.
    Renewed Facility Operating License No. NPF-57: The amendment 
revised the TSs and the Facility Operating License.
    Date of initial notice in Federal Register: April 3, 2012 (77 FR 
20075). The letter dated December 21, 2012, provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination or expand the application.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 25, 2013.
    No significant hazards consideration comments received: No.

South Carolina Electric and Gas. Docket Nos. 52-027 and 52-028, Virgil 
C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield County, 
South Carolina

    Date of amendment request: January 15, 2013.
    Brief description of amendment: The amendment authorizes a 
departure from the Virgil C. Summer Nuclear Station Units 2 and 3 
plant-specific Design Control Document (DCD) Tier 2* material 
incorporated into the Updated Final Safety Analysis Report (UFSAR) to 
revise the requirements for shear reinforcement spacing in the nuclear 
island basemat below the auxiliary building.
    Date of issuance: February 26, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: Unit 2--1, and Unit 3--1.
    Facility Combined Licenses No. NPF-93 and NPF-94: Amendment revised 
the Facility Combined Licenses.
    Date of initial notice in Federal Register: January 25, 2013 (78 FR 
5511).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 26, 2013.
    No significant hazards consideration comments received: No.

South Carolina Electric and Gas. Docket Nos. 52-027 and 52-028, Virgil 
C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield County, 
South Carolina

    Date of amendment request: January 18, 2013.
    Brief description of amendment: The amendment authorizes a 
departure from the VCSNS Units 2 and 3 plant-specific Design Control 
Document (DCD) Tier 2* material incorporated into the Updated Final 
Safety Analysis Report (UFSAR) by revising the structural criteria code 
for anchoring of headed shear reinforcement bar within the nuclear 
island basemat.
    Date of issuance: March 1, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: Unit 2--2, and Unit 3--2.
    Facility Combined Licenses No. NPF-93 and NPF-94: Amendment revised 
the Facility Combined Licenses.
    Date of initial notice in Federal Register: January 29, 2013 (78 FR 
6145).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 1, 2013.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc. Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant (VEGP) Units 3 and 4, Burke County, 
Georgia

    Date of amendment request: January 15, 2013.
    Brief description of amendment: The proposed amendment would depart 
from VEGP Units 3 and 4 plant-specific Design Control Document (DCD) 
Tier 2* material incorporated into the Updated Final Safety Analysis 
Report (UFSAR) to clarify the requirements for shear reinforcement 
spacing in the nuclear island basemat below the auxiliary building. The 
proposed change would modify the provisions for maximum spacing of the 
shear reinforcement in the basemat below the auxiliary building.
    Date of issuance: February 26, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: Unit 3--4, and Unit 4--4.
    Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised 
the Facility Combined Licenses.
    Date of initial notice in Federal Register: January 25, 2013 (78 FR 
5508).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 26, 2013.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 11th day of March 2013.

    For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2013-06164 Filed 3-18-13; 8:45 am]
BILLING CODE 7590-01-P