[Federal Register Volume 78, Number 42 (Monday, March 4, 2013)]
[Notices]
[Pages 14126-14141]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2013-04885]


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NUCLEAR REGULATORY COMMISSION

[NRC-2013-0045]


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 7, 2013, to February 20, 2013. The 
last biweekly notice was published on February 19, 2013 (78 FR 11688).

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

SUPPLEMENTARY INFORMATION: 

I. Accessing Information and Submitting Comments

A. Accessing Information

    Please refer to Docket ID  when contacting the NRC 
about the availability of information regarding this document. You may 
access information related to this document, which the NRC possesses 
and is publicly available, by the following methods:
     Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for Docket ID .
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may access publicly available documents online in the NRC 
Library at http://www.nrc.gov/reading-rm/adams.html. To begin the 
search, select ``ADAMS Public Documents'' and then select ``Begin Web-
based ADAMS Search.'' For problems with ADAMS, please contact the NRC's 
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to pdr.resource@nrc.gov. Documents may be viewed in 
ADAMS 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  in the subject line of 
your comment submission, in order to ensure that the NRC is able to 
make your comment submission available to the public in this docket.
    The NRC cautions you not to include identifying or contact 
information in comment submissions that you do not want to be publicly 
disclosed. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into 
ADAMS, and the NRC does not edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information in their comment submissions 
that they do not want to be publicly disclosed. Your request should 
state that the NRC will not edit comment submissions to remove such 
information before making the comment submissions available to the 
public or entering the comment submissions into ADAMS.

[[Page 14127]]

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in Title 10 of the Code of Federal 
Regulations (10 CFR), Section 50.92, this means that operation of the 
facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license or 
combined license. Requests for a hearing and a petition for leave to 
intervene shall be filed in accordance with the Commission's ``Rules of 
Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2. 
Interested person(s) should consult a current copy of 10 CFR 2.309, 
which is available at the NRC's PDR, located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. NRC regulations are accessible electronically from the NRC 
Library on the NRC 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 hearing.docket@nrc.gov, or by 
telephone at 301-415-1677, to request (1) a digital information (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign

[[Page 14128]]

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 Electronic Information Exchange System, users 
will be required to install a Web browser plug-in from the NRC 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 NRC guidance 
available on the NRC 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 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 Web site 
at http://www.nrc.gov/site-help/e-submittals.html, by email at 
MSHD.Resource@nrc.gov, or by a toll-free call at 1-866 672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, 
Maryland, 20852, Attention: Rulemaking and Adjudications Staff. 
Participants filing a document in this manner are responsible for 
serving the document on all other participants. Filing is considered 
complete by first-class mail as of the time of deposit in the mail, or 
by courier, express mail, or expedited delivery service upon depositing 
the document with the provider of the service. A presiding officer, 
having granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
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 PDR Reference staff at 1-800-397-4209, 
301-415-4737, or by email to pdr.resource@nrc.gov.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, 
and 3, Maricopa County, Arizona

    Date of amendment request: December 12, 2012.

[[Page 14129]]

    Description of amendment request: The amendments would change the 
Technical Specifications (TSs) by replacing the current limits on 
primary coolant gross specific activity with limits on primary coolant 
noble gas activity. The noble gas activity would be based on DOSE 
EQUIVALENT XE-133 and would take into account only the noble gas 
activity in the primary coolant. The changes are consistent with NRC-
approved Industry/Technical Specification Task Force (TSTF) Standard 
Technical Specification Change Traveler, TSTF-490, Revision 0, 
``Deletion of E-Bar Definition and Revision to RCS [Reactor Coolant 
System] Specific Activity Technical Specifications,'' with deviations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The license concluded that the no significant hazards 
consideration determination published in the Federal Register on March 
19, 2007 (72 FR 12838), is applicable, and is presented below:

    1. The Proposed Change Does Not Involve a Significant Increase 
in the Probability or Consequences of an Accident Previously 
Evaluated
    Response: Reactor coolant specific activity is not an initiator 
for any accident previously evaluated. The Completion Time when 
primary coolant gross activity is not within limit is not an 
initiator for any accident previously evaluated. The current 
variable limit on primary coolant iodine concentration is not an 
initiator to any accident previously evaluated. As a result, the 
proposed change does not significantly increase the probability of 
an accident. The proposed change will limit primary coolant noble 
gases to concentrations consistent with the accident analyses. The 
proposed change to the Completion Time has no impact on the 
consequences of any design basis accident since the consequences of 
an accident during the extended Completion Time are the same as the 
consequences of an accident during the Completion Time. As a result, 
the consequences of any accident previously evaluated are not 
significantly increased.
    2. The Proposed Change Does Not Create the Possibility of a New 
or Different Kind of Accident from any Accident Previously Evaluated
    Response: The proposed change in specific activity limits does 
not alter any physical part of the plant nor does it affect any 
plant operating parameter. The change does not create the potential 
for a new or different kind of accident from any previously 
calculated.
    3. The Proposed Change Does Not Involve a Significant Reduction 
in the Margin of Safety
    Response: The proposed change revises the limits on noble gase 
[sic] radioactivity in the primary coolant. The proposed change is 
consistent with the assumptions in the safety analyses and will 
ensure the monitored values protect the initial assumptions in the 
safety analyses.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Michael G. Green, Senior Regulatory Counsel, 
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695, 
Phoenix, Arizona 85072-2034.
    NRC Branch Chief: Michael T. Markley.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, 
and 3, Maricopa County, Arizona

    Date of amendment request: December 26, 2012.
    Description of amendment request: The amendments would adopt 
Technical Specifications Task Force (TSTF) Traveler TSTF-500, Revision 
2, ``DC Electrical Rewrite--Update to TSTF-360,'' with one variation. 
The amendments would revise the TS requirements related to direct 
current (DC) electrical systems in TS Limiting Condition for Operation 
(LCO) 3.8.4, ``DC Sources--Operating,'' LCO 3.8.5, ``DC Sources--
Shutdown,'' and LCO 3.8.6, ``Battery Parameters.'' In addition, new TS 
5.5.19, ``Battery Monitoring and Maintenance Program,'' is being 
proposed for Section 5.5, ``Administrative Controls--Programs and 
Manuals.''
    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 any accident previously 
evaluated?
    Response: No.
    The proposed changes restructure the Technical Specifications 
(TS) for the direct current (DC) electrical power system and are 
consistent with TSTF-500, Revision 2. The proposed changes modify TS 
Actions relating to battery and battery charger inoperability. The 
DC electrical power system, including associated battery chargers, 
is not an initiator of any accident sequence analyzed in the Updated 
Final Safety Analysis Report (UFSAR). Rather, the DC electrical 
power system supports equipment used to mitigate accidents. The 
proposed changes to restructure TS and change surveillances for 
batteries and chargers to incorporate the updates included in TSTF-
500, Revision 2, will maintain the same level of equipment 
performance required for mitigating accidents assumed in the UFSAR. 
Operation in accordance with the proposed TS would ensure that the 
DC electrical power system is capable of performing its specified 
safety function as described in the UFSAR. Therefore, the mitigating 
functions supported by the DC electrical power system will continue 
to provide the protection assumed by the analysis. The relocation of 
preventive maintenance surveillances, and certain operating limits 
and actions, to a licensee-controlled Battery Monitoring and 
Maintenance Program will not challenge the ability of the DC 
electrical power system to perform its design function. Appropriate 
monitoring and maintenance that are consistent with industry 
standards will continue to be performed. In addition, the DC 
electrical power system is within the scope of 10 CFR 50.65, 
Requirements for monitoring the effectiveness of maintenance at 
nuclear power plants, which will ensure the control of maintenance 
activities associated with the DC electrical power system.
    The integrity of fission product barriers, plant configuration, 
and operating procedures as described in the UFSAR will not be 
affected by the proposed changes. Therefore, the consequences of 
previously analyzed accidents will not increase by implementing 
these changes. Therefore, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No.
    The proposed changes involve restructuring the TS for the DC 
electrical power system. The DC electrical power system, including 
associated battery chargers, is not an initiator to any accident 
sequence analyzed in the UFSAR. Rather, the DC electrical power 
system supports equipment used to mitigate accidents. The proposed 
changes to restructure the TS and change surveillances for batteries 
and chargers to incorporate the updates included in TSTF-500, 
Revision 2, will maintain the same level of equipment performance 
required for mitigating accidents assumed in the UFSAR. 
Administrative and mechanical controls are in place to ensure the 
design and operation of the DC systems continues to meet the plant 
design basis described in the UFSAR. Therefore, operation of the 
facility in accordance with this proposed change will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?

[[Page 14130]]

    Response: No.
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The equipment margins will be maintained in 
accordance with the plant-specific design bases as a result of the 
proposed changes. The proposed changes will not adversely affect 
operation of plant equipment. These changes will not result in a 
change to the setpoints at which protective actions are initiated. 
Sufficient DC capacity to support operation of mitigation equipment 
is ensured. The changes associated with the new Battery Maintenance 
and Monitoring Program will ensure that the station batteries are 
maintained in a highly reliable manner. The equipment fed by the DC 
electrical sources will continue to provide adequate power to 
safety-related loads in accordance with analysis assumptions.
    TS changes made in accordance with TSTF-500, Revision 2, 
maintain the same level of equipment performance stated in the UFSAR 
and the current TSs. Therefore, the proposed changes do not involve 
a significant reduction of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Michael G. Green, Senior Regulatory Counsel, 
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695, 
Phoenix, Arizona 85072-2034.
    NRC Branch Chief: Michael T. Markley.

Calvert Cliffs Nuclear Power Plant, LLC, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, 
Maryland

    Date of amendments request: October 2, 2012, as supplemented by 
letter dated November 26, 2012.
    Description of amendments request: The amendments would revise 
Technical Specification (TS) 3.8.3 ``Diesel Fuel Oil'' by relocating 
the current stored diesel fuel oil numerical volume requirements from 
the TS to the TS Bases and TS 3.8.1 ``AC Sources-Operating'' by 
relocating the specific numerical value for the day tank fuel oil 
volume from the TS to the TS Bases. The changes would be consistent 
with Nuclear Regulatory Commission (NRC)-approved Industry Technical 
Specification Task Force Standard Technical Specification Change 
Traveler, TSTF-501-A, Revision 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. Involve a significant increase in the probability or 
consequences of an accident previously evaluated; or
    No.
    The proposed change relocates the volume of diesel fuel oil 
required to support 7-day operation of an onsite diesel generator, 
and the volume equivalent to a 6-day supply, to licensee control. 
The specific volume of fuel oil equivalent to a 7- and 6-day supply 
is calculated using the limiting energy content of the fuel, the 
required diesel generator output and the corresponding fuel oil 
consumption rate. Because the requirement to maintain a 7-day supply 
of diesel fuel oil is not changed and is consistent with the 
assumptions in the accident analysis, and the actions taken with the 
volume of fuel oil is less than a 6-day supply have not changed, 
neither the probability nor the consequences of any accident 
previously evaluated will be affected.
    The proposed change also relocates the volume of diesel fuel oil 
required to support one hour of diesel generator operation at full 
load in the day tank. The specific volume and time is not changed 
and is consistent with the existing plant design basis to support a 
diesel generator under accident load conditions.
    Therefore, operation of the facility in accordance with the 
proposed amendment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different type of accident 
from any accident previously evaluated; or
    No.
    The change does not involve a physical alteration of the plant 
(i.e., no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. The change 
does not alter assumptions made in the safety analysis but ensures 
that the diesel generator operates as assumed in the accident 
analysis. The proposed change is consistent with the safety analysis 
assumptions.
    The proposed change also relocates the volume of diesel fuel oil 
required to support one hour of diesel generator operation at full 
load in the day tank. The change does not alter assumptions made in 
the safety analysis but ensures that the diesel generator operates 
as assumed in the accident analysis. The proposed change is 
consistent with the safety analysis assumptions. Therefore, the 
proposed amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    No.
    The proposed change relocates the volume of diesel fuel oil 
required to support 7-day operation of an onsite diesel generator, 
and the volume equivalent to a 6-day supply, and one hour day tank 
supply to licensee control. As the basis for the existing limits on 
diesel fuel oil are not changed, no change is made to the accident 
analysis assumptions and no margin of safety is reduced as part of 
this change.
    The proposed change also relocates the volume of diesel fuel oil 
required to support one hour of diesel generator operation at full 
load in the day tank. As the basis for the existing limits on diesel 
fuel oil are not changed, no change is made to the accident analysis 
assumptions and no margin of safety is reduced as part of this 
change.
    Therefore, the proposed amendment would not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Steven L. Miller, General Counsel, 
Constellation Energy Nuclear Group, LLC, 100 Constellation Way, Suite 
200c, Baltimore, MD 21202.
    NRC Branch Chief: George Wilson.

Calvert Cliffs Nuclear Power Plant, LLC, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, 
Maryland

    Date of amendments request: October 16, 2012.
    Description of amendments request: The amendments would revise 
Surveillance Requirements (SRs) 3.8.1.8, 3.8.1.11, and 3.8.2.1 and add 
SR 3.8.1.17 of Technical Specification (TS) 3.8.1 ``AC Sources--
Operating.''
    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.
    This amendment request proposes to add or modify certain [TS 
SRs] for the diesel generators. This proposed amendment will provide 
additional assurance that the AC Sources relied upon to ensure the 
availability of necessary power to the Engineered Safety Features 
systems are capable of performing their specified safety function if 
needed. The diesel generators and their associated emergency loads 
are accident mitigating features, not accident initiators. This 
proposed amendment does not change the design function of the diesel 
generators or any of their required loads, and does not change the 
way the systems and plant are operated or maintained. This proposed 
amendment does not impact any plant systems that are accident 
initiators and does not adversely impact any accident mitigating 
systems.

[[Page 14131]]

    The proposed amendment does not affect the operability 
requirements for the diesel generators, as verification of such 
operability will continue to be performed as required. Continued 
verification of operability supports the capability of the diesel 
generators to perform their required design functions of providing 
emergency power to the Engineered Safety Features systems, 
consistent with the plant safety analyses as described in the 
Updated Final Safety Analysis Report (UFSAR).
    Adding or modifying [TS SRs] for the diesel generators will not 
significantly increase the probability of an accident previously 
evaluated because the diesel generators and their emergency loads 
are accident mitigation features, not accident initiators. Adding or 
modifying [TS SRs] for the diesel generators will not change any of 
the dose analyses associated with the UFSAR Chapter 14 accidents 
because accident mitigation functions and requirements remain 
unchanged.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    This amendment request proposes to add or modify certain [TSs 
SRs] for the diesel generators. This proposed amendment does not 
change the design function of the diesel generators or any required 
loads, and does not change the way the systems and plant are 
operated or maintained. This proposed amendment does not impact any 
plant systems that are accident initiators and does not adversely 
impact any accident mitigating systems. Performance of these 
surveillances tests will provide additional assurance that the AC 
Sources relied upon to ensure the availability of necessary power to 
the Engineered Safety Features systems are capable of performing 
their specified safety function if needed.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    Response: No.
    This amendment request proposes to add or modify certain [TS 
SRs] for the diesel generators. This proposed amendment will provide 
additional assurance that the AC Sources relied upon to ensure the 
availability of necessary power to the Engineered Safety Features 
systems are capable of performing their specified safety function if 
needed. 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. This proposed amendment does not involve or 
affect fuel cladding, the reactor coolant system, or the primary 
containment. Performance of these surveillances tests will provide 
continued assurance that the AC Sources relied upon to ensure the 
availability of necessary power to the Engineered Safety Features 
systems are capable of performing their specified safety function if 
needed.
    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: Steven L. Miller, General Counsel, 
Constellation Energy Nuclear Group, LLC, 100 Constellation Way, Suite 
200c, Baltimore, MD 21202.
    NRC Branch Chief: George Wilson.

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

    Date of amendment request: December 21, 2012.
    Description of amendment request: The proposed amendment would 
revise the Fermi 2 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 2.
    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 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: Robert D. Carlson.

Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, 
Comanche Peak Nuclear Power Plant, Units 1 and 2, Somervell County, 
Texas

    Date of amendment request: December 19, 2012.
    Brief description of amendments: The amendments would revise 
Technical Specification (TS) 3.8.1, ``AC [Alternating Current] 
Sources--Operating,'' to revise the Completion Time (CT) for Required 
Action A.3, ``Restore required offsite circuit to OPERABLE status,'' on 
one-time basis from 72 hours to 14 days for Comanche

[[Page 14132]]

Peak Nuclear Power Plant (CPNPP), Units 1 and 2. The CT extension from 
72 hours to 14 days will be used twice while completing the plant 
modification to install alternate startup transformer (ST) XST1A and 
will expire on March 31, 2014. After completion of this modification, 
if ST XST1 should require maintenance or if failure occurs, the 
alternate ST XST1A can be aligned to the Class 1E buses well within the 
current CT of 72 hours. Installation of alternate ST will result in 
improved plant design and will improve the long-term reliability of the 
138 kiloVolt (kV) offsite circuit ST.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change will revise the CT for the loss of one 
offsite source from 72 hours to 14 days to allow two, one-time, 14-
day CTs. The proposed two, one-time extensions of the CT for the 
loss of one offsite power circuit does not significantly increase 
the probability of an accident previously evaluated. The TS will 
continue to require equipment that will power safety related 
equipment necessary to perform any required safety function. The 
two, one-time extensions of the CT to 14 days does not affect the 
design of the STs, the interface of the STs with other plant 
systems, the operating characteristic of the STs, or the reliability 
of the STs.
    The consequence of a LOOP [loss-of-offsite power] event has been 
evaluated in the CPNPP Final Safety Analysis Report (Reference 8.1 
[of application dated December 19, 2012]) and the Station Blackout 
evaluation. Increasing the CT for one offsite power source twice on 
a one-time basis from 72 hours to 14 days does not increase the 
consequences of a LOOP event nor change the evaluation of LOOP 
events.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not result in a change in the manner in 
which the electrical distribution subsystems provide plant 
protection. The proposed change will only affect the time allowed to 
restore the operability of the offsite power source through a ST. 
The proposed change does not affect the configuration, or operation 
of the plant. The proposed change to the CT will facilitate 
installation of a plant modification which will improve plant design 
and will eliminate the necessity to shut down both Units if XST1 
fails or requires maintenance that goes beyond the current TS CT of 
72 hours. This change will improve the long-term reliability of the 
138kV offsite circuit ST which is common to both CPNPP Units.
    There are no changes to the STs or the supporting systems 
operating characteristics or conditions. The change to the CT does 
not change any existing accident scenarios, nor create any new or 
different accident scenarios. In addition, the change does not 
impose any new or different requirements or eliminate any existing 
requirements. The change does not alter any of the assumptions made 
in the safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not affect the acceptance criteria for 
any analyzed event nor is there a change to any safety limit. The 
proposed change does not alter the manner in which safety limits, 
limiting safety system settings, or limiting conditions for 
operation are determined. Neither the safety analyses nor the safety 
analysis acceptance criteria are affected by this change. The 
proposed change will not result in plant operation in a 
configuration outside the current design basis. The proposed 
activity only increases, for two, one-time pre-planned occurrences, 
the period when the plant may operate with one offsite power source. 
The margin of safety is maintained by maintaining the ability to 
safely shut down the plant and remove residual heat.
    Therefore, the proposed change does not involve a 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: Timothy P. Matthews, Esq., Morgan, Lewis and 
Bockius, 1111 Pennsylvania Avenue NW, Washington, DC 20004.
    NRC Branch Chief: Michael T. Markley.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee 
Atomic Power Station, Lincoln County, Maine

    Date of amendment request: January 3, 2012.
    Description of amendment request: The amendment proposes to revise 
License Condition 2.B(6)(d) ``Physical Protection.'' It is proposed to 
update the title of the Physical Security Plan, from the ``Maine Yankee 
Nuclear Power Station Physical Security Plan'', the ``Maine Yankee 
Nuclear Atomic Power Station Guard Training and Qualification Plan'', 
and the ``Maine Yankee Nuclear Power Safeguards Contingency Plan'' to 
the ``Maine Yankee Independent Spent Fuel Storage Installation Physical 
Security Plan.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment is a title change only. There is no 
reduction in commitments in the Maine Yankee Independent Spent Fuel 
Storage Installation Physical Security Plan therefore; the proposed 
amendment does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment is a title change only. There is no 
reduction in commitments in the Maine Yankee Independent Spent Fuel 
Storage Installation Physical Security Plan therefore; the proposed 
amendment does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment is a title change only. There is no 
reduction in commitments in the Maine Yankee Independent Spent Fuel 
Storage Installation Physical Security Plan 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: Joseph Fay, Maine Yankee Atomic Power 
Company, 362 Injun Hollow Road, East Hampton, Connecticut, 06424-3099.
    NRC Branch Chief: Michele M. Sampson.

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

    Date of amendment request: December 6, 2012.
    Description of amendment request: The amendment proposes to revise 
the

[[Page 14133]]

Monticello Nuclear Generating Plant (MNGP) Technical Specification (TS) 
Limiting Condition for Operation 3.10.1, ``Inservice Leak and 
Hydrostatic Testing Operation,'' and the associated Bases, to expand 
its scope to include provisions for temperature excursions greater than 
212[emsp14][deg]F as a consequence of inservice leak and hydrostatic 
testing, and as a consequence of scram time testing initiated in 
conjunction with an inservice leak or hydrostatic test, while 
considering operational conditions to be in MODE 4. The change is 
consistent with NRC-approved Technical Specification Task Force (TSTF) 
Improved Standard Technical Specifications Change Traveler, TSTF-484, 
Revision 0, ``Use of TS 3.10.1 for Scram Time Testing Activities.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided 
its analysis of the issue of no significant hazards consideration, 
which is provided below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Technical Specifications currently allow for operation at 
greater than 200[emsp14][deg]F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. Extending the activities that can apply this allowance will not 
adversely impact the probability or consequences of an accident 
previously evaluated. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No.
    Technical Specifications currently allow for operation at 
greater than 200[emsp14][deg]F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. No new operational conditions beyond those currently allowed by 
LCO 3.10.1 are introduced. The changes do not involve a physical 
alteration of the plant (i.e., no new or different type of equipment 
will be installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements or eliminate any existing requirements. The 
changes do not alter assumptions made in the safety analysis. The 
proposed changes are consistent with the safety analysis assumptions 
and current plant operating practice. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    Technical Specifications currently allow for operation at 
greater than 200[emsp14][deg]F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. Extending the activities that can apply this allowance will not 
adversely impact any margin of safety. Allowing completion of 
inspections and testing and supporting completion of scram time 
testing initiated in conjunction with an inservice leak or 
hydrostatic test prior to power operation results in enhanced safe 
operations by eliminating unnecessary maneuvers to control reactor 
temperature and pressure. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: Robert D. Carlson.

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

    Date of amendment request: December 21, 2012.
    Description of amendment request: The amendment proposes to revise 
the Monticello Nuclear Generating Plant (MNGP) Emergency Plan by 
revising the Emergency Action Level (EAL) setpoint for the Turbine 
Building Normal Waste Sump (TBNWS) Monitor. The proposed change reduces 
the classification of a liquid effluent release via the TBNWS pathway 
to approximately 48 times the Offsite Does Calculation Manual (ODCM) 
limit from the current 200 times the ODCM limit, thus establishing a 
value within the indication capability of the radiation monitor.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided 
its analysis of the issue of no significant hazards consideration, 
which is provided below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to the emergency plan does not impact the 
physical function of plant structures, systems, or components (SSCs) 
or the manner in which SSCs perform their design function. The 
proposed change neither adversely affects accident initiators or 
precursors, nor alters design assumptions. The proposed change does 
not alter or prevent the ability of operable SSCs to perform their 
intended function to mitigate the consequences of an initiating 
event within assumed acceptance limits. No operating procedures or 
administrative controls that function to prevent or mitigate 
accidents are affected by the proposed change.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No.
    The proposed change does not impact the accident analysis. The 
change does not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed), a change in 
the method of plant operation, or new operator actions. The proposed 
change will not introduce failure modes that could result in a new 
accident, and the change does not alter assumptions made in the 
safety analysis. The proposed change revises an emergency action 
level (EAL), which establishes the threshold for placing the plant 
in an emergency classification. EALs are not initiators of any 
accidents.
    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 
the margin of safety?
    Response: No.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel cladding, reactor coolant 
system pressure boundary, and containment structure) to limit the 
level of radiation does to the public. The proposed change is 
associated with the EALs and does not impact operation of the plant 
or its response to transients or accidents. The change does not 
affect the technical specifications or the operating license. The 
proposed change does not involve a change in the method of plant 
operation, and no accident analyses will be affected by the proposed 
change. Additionally, the proposed change will not relax any 
criteria used to establish safety limits and will not relax any 
safety system settings. The safety analysis acceptance criteria are 
not affected by this change. The proposed change will not result in 
plant operation in a configuration outside the design basis. The 
proposed change does not adversely affect systems that respond to 
safely shutdown the plant and to maintain the plant in a safe 
shutdown condition.
    The revised EAL provides more appropriate and accurate criteria 
for determining protective measures that should be considered within 
and outside the site boundary to protect public health and safety. 
The emergency plan will continue to activate an emergency response 
commensurate with the extent of degradation of plant safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this

[[Page 14134]]

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: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: Robert D. Carlson.

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

    Date of amendment request: January 4, 2013.
    Description of amendment request: The licensee proposed to revise 
the MNGP Technical Specifications (TS) 3.6.4.3, ``Standby Gas Treatment 
(SGT) System,'' TS 3.7.4, ``Control Room Emergency Filtration (CREF) 
System,'' and TS 5.5.6, ``Ventilation Filter Testing Program (VFTP).'' 
The licensee proposed to modify the TS requirements to operate 
ventilation systems with charcoal filters from 10 hours each month to 
15 minutes in accordance with Technical Specifications Task Force 
(TSTF) Traveler TSTF-522, Revision 0, ``Revise Ventilation System 
Surveillance Requirements to Operate for 10 hours per Month.''
    Specifically, the licensee proposed to revise the surveillance 
requirements STET which currently require testing of SGT and CREF 
Systems, with heaters operating, for a continuous 10 hour period every 
31 days without the heaters operating. The associated SRs are proposed 
to be revised to require operation of these systems for 15 continuous 
minutes every 31 days. Additionally, the licensee proposed to remove 
Specification 5.5.6, Item e, under the VFTP, concerning operation of 
the SGT and CREF Systems heaters.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided 
its analysis of the issue of no significant hazards consideration, 
which is provided below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change replaces existing SRs to operate the SGT 
System and CREF System equipped with electric heaters for a 
continuous 10 hour period every 31 days with a requirement to 
operate the systems for 15 continuous minutes (without the heaters 
operating) and removes a no longer required SR under the VFTP.
    These systems are not accident initiators and, therefore, these 
changes do not involve a significant increase in the probability of 
an accident. The proposed system and filter testing changes are 
consistent with current regulatory guidance for these systems and 
will continue to assure that these systems perform their design 
function which may include mitigating accidents. Thus, the changes 
do not involve a significant increase in the consequences of an 
accident.
    Therefore, it is concluded that these 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 replaces existing SRs to operate the SGT 
System and CREF System equipped with electric heaters for a 
continuous 10 hour period every 31 days with a requirement to 
operate the systems for 15 continuous minutes (without the heaters 
operating) and removes a no longer required SR under the VFTP.
    The change proposed for these ventilation systems does not 
change any systems operations or maintenance activities. Testing 
requirements will be revised and will continue to demonstrate that 
the Limiting Conditions for Operation (LCO) are met and the system 
components are capable of performing their intended safety 
functions. The changes do not create new failure modes or mechanisms 
and no new accident precursors are generated.
    Therefore, it is concluded that these changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes replaces existing SRs to operate the SGT 
System and CREF System equipped with electric heaters for a 
continuous 10 hour period every 31 days with a requirement to 
operate the systems for 15 continuous minutes (without the heaters 
operating) and removes a no longer required SR under the VFTP. 
Testing requirements will be revised and will continue to 
demonstrate that the LCOs are met and the system components are 
capable of performing their intended safety functions.
    The proposed changes are consistent with regulatory guidance. 
Therefore, it is concluded that these 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 the licensee: Peter M. Glass, Assistant General 
Counsel, Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 
55401
    NRC Branch Chief: Robert D. Carlson.

Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue 
County, Minnesota

    Date of amendment request: December 13, 2012.
    Description of amendment request: The proposed amendments would 
revise the Prairie Island Nuclear Generating Plant Emergency Plan by 
revising certain emergency action levels described in the plan.
    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.
    This license amendment request proposes to revise Emergency Plan 
emergency action levels for classification of liquid effluent 
releases and determining fuel clad barrier loss. These changes 
propose to use installed plant radiation monitors differently but do 
not involve any physical plant changes.
    The Emergency Plan emergency action levels and installed plant 
radiation monitors are not accident initiators and therefore the 
proposed changes do not involve an increase in the probability of an 
accident. The proposed emergency action level changes do not affect 
the capability of any structures, system or components to mitigate a 
design basis accident. Thus the proposed changes do not involve a 
significant increase in the consequences of an accident.
    Therefore, the proposed Emergency Plan emergency action level 
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.
    This license amendment request proposes to revise Emergency Plan 
emergency action levels for classification of liquid effluent 
releases and determining fuel clad barrier loss. These changes 
propose to use installed plant radiation monitors differently but do 
not involve any physical plant changes.
    The proposed Emergency Plan emergency action level changes do 
not change any system operations or maintenance activities. The 
changes do not involve physical alteration of the plant, that is, no 
new or different type of equipment will be installed. The changes do 
not alter assumptions made in the safety analyses but ensures that 
the plant Emergency Plan is effectively and consistently 
implemented. These changes do not create new failure modes or 
mechanisms which are not identifiable during testing and no new 
accident precursors are generated.

[[Page 14135]]

    Therefore, the proposed Emergency Plan emergency action level 
changes do not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    This license amendment request proposes to revise Emergency Plan 
emergency action levels for classification of liquid effluent 
releases and determining fuel clad barrier loss. These changes 
propose to use installed plant radiation monitors differently but do 
not involve any physical plant changes.
    Margin of safety is provided by the ability of accident 
mitigation structures systems or components to perform at their 
analyzed capability. The changes proposed in this license amendment 
request do not affect the capability of any equipment to perform its 
accident mitigation function. Thus, no margin of safety is reduced 
as part of this change.
    Therefore, the proposed Emergency Plan emergency action level 
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 requests involve no significant hazards consideration.
    Attorney for licensee: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: Robert D. Carlson.

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: February 7, 2013.
    Description of amendment request: The proposed change would amend 
Combined License Nos.: NPF-93 and NPF-94 for Virgil C. Summer Nuclear 
Station (VCSNS) Units 2 and 3 in regard to the Primary Sampling System 
(PSS) by: (1) Replacing containment air return check valve PSS-PL-V024 
with a solenoid-operated valve, and (2) redesigning the PSS inside-
containment header and adding a PSS containment penetration.
    Because, this proposed change requires a departure from Tier 1 
information in the Westinghouse Advanced Passive 1000 design control 
document (DCD), the licensee also requested an exemption from the 
requirements of the Generic DCD Tier 1 in accordance with 52.63(b)(1).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The Primary Sampling System (PSS) provides the safety-related 
function of preserving containment integrity by isolation of the PSS 
lines penetrating containment. The proposed amendment will enhance 
the ability of the PSS to perform its nonsafety-related function of 
providing the capability to obtain reactor coolant and containment 
atmosphere samples, while maintaining the ability of the PSS to 
perform its safety-related containment isolation function. The 
replacement of a check valve with a solenoid-operated containment 
isolation valve and the redesigned inside-containment header does 
not affect the safety-related function of isolating the PSS lines 
for containment isolation. The components added by this proposed 
activity, including tubing and the solenoid-operated containment 
isolation valve, are designed to the same codes and standards as 
other components addressed in the certified design that perform 
similar functions. The additional PSS containment penetration is a 
passive extension of containment and is identical in form, fit, and 
function to other PSS sampling containment penetrations currently 
addressed in the certified AP1000 plant design. The addition of a 
new PSS containment penetration will not change the maximum 
allowable leakage rate allowed by Technical Specifications and 
verified periodically in accordance with regulations. Furthermore, 
the proposed PSS configuration changes will neither impact any 
accident source term parameter or fission product barrier nor affect 
radiological dose consequence 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 amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The additional containment penetration is similar in form, fit, 
and function to the PSS penetrations that are currently described in 
the Updated Final Safety Analysis Report. Because the PSS changes 
use valve types, piping, and a containment penetration consistent 
with those already described in the Updated Final Safety Analysis 
Report, no new failure modes or equipment failure initiators are 
introduced by these changes. Accordingly, the proposed changes do 
not create any new malfunctions, failure mechanisms, or accident 
initiators.
    Therefore, the proposed amendment 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 containment isolation function is not changed by this 
activity and is bounded by the existing design. The proposed PSS 
containment penetration is similar in form, fit, and function to 
other containment penetrations in similar applications in the 
current certified AP1000 plant design. The additional PSS 
containment penetration is an extension of containment, and, 
therefore, does not affect containment or its ability to perform its 
design function. The addition of PSS components, including the 
solenoid-operated containment isolation valve, the additional PSS 
containment penetration, and the associated tubing, do not exceed or 
alter a design basis or safety limit. Because the containment 
isolation function, containment leakage rate limit, potential 
containment leakage, and protective shielding are not changed by 
this activity and are bounded by the existing design, there is no 
change to any current margin of safety.
    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: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
    NRC Branch Chief: Lawrence Burkhart, Acting.

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: February 14, 2013.
    Description of amendment request: The proposed change would amend 
Combined License Nos.: NPF-93 and NPF-94 for Virgil C. Summer Nuclear 
Station (VCSNS) Units 2 and 3 in regard to the structural module stud 
size and spacing by increasing the carbon steel vertical stud spacing, 
decreasing the stainless steel stud diameter, and decreasing the 
stainless steel vertical and horizontal stud spacing in accordance with 
the design basis.
    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.

[[Page 14136]]

    The design function of the containment modules is to support the 
reactor coolant system components and related piping systems and 
equipment. The design functions of the affected structural module in 
the auxiliary building are to provide support and protection for new 
and spent fuel and the equipment needed to support fuel handling, 
cooling, and storage in the spent fuel racks, and to provide 
support, protection, and separation for the seismic Category I 
mechanical and electrical equipment located outside the containment 
building. The design function of the shear studs it to transfer 
loads into the concrete of the structural modules. The proposed 
change corrects a drawing note regarding shear stud size and spacing 
for structural wall modules to be consistent with the underlying 
design basis calculations, which are more conservative. The 
thickness, geometry, and strength of the structures are not 
adversely altered. The properties of the concrete included in the 
modules are not altered. As a result, the design function of the 
structural modules is not adversely affected by the proposed change. 
There is no change to plant, systems or the response of systems to 
postulated accident conditions. There is no change to the predicted 
radioactive releases due to normal operation or postulated accident 
conditions. The plant response to previously evaluated accidents or 
external events is not adversely affected, nor does the change 
described create any new accident precursors.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change corrects a drawing note regarding shear stud 
size and spacing for structural wall modules to be consistent with 
the underlying design basis calculations. Stud spacing and sizing 
are updated such that stud loadings are within acceptable limits and 
that the structural module acts in a composite manner. The 
thickness, geometry, and strength of the structures are not 
adversely altered. The properties of the concrete included in the 
modules are not altered. The change to the internal design of the 
structural modules does not create any new accident precursors. As a 
result, the design function of the modules is not adversely affected 
by the proposed change.
    Therefore, the proposed amendment 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 criteria and requirements of AISC-N690 provide a margin of 
safety to structural failure. The design of the shear studs for the 
structural wall modules conforms to criteria and requirements in 
AISC-N690 and therefore maintains the margin of safety. The proposed 
change corrects a drawing note regarding shear stud size and spacing 
for the structural wall modules so as to be consistent with the 
underlying design basis calculations. There was no change to the 
method of evaluation from that used in the design basis 
calculations.
    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: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
    NRC Branch Chief: Lawrence Burkhart, Acting.

South Carolina Electric and Gas Company 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: February 7, 2013 and revised on February 
14, 2013.
    Description of amendment request: The proposed change would amend 
Combined License Nos.: NPF-93 and NPF-94 for Virgil C. Summer Nuclear 
Station (VCSNS) Units 2 and 3 to allow the use of concentrically and 
eccentrically braced frames in the turbine building main area and 
modify the applicable design code.
    Because, this proposed change requires a departure from Tier 1 
information in the Westinghouse Advanced Passive 1000 design control 
document (DCD), the licensee also requested an exemption from the 
requirements of the Generic DCD Tier1 in accordance with 52.63(b)(1).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The turbine building bracing design is changed to a mixed 
bracing system which uses special concentric and eccentric bracing. 
The turbine building does not contain safety-related systems or 
components. The main area of the turbine building continues to meet 
its design function of preventing a turbine building collapse from 
impairing the integrity of seismic Category I structures, systems, 
or components. The first bay of the turbine building is designed to 
prevent the collapse of the main area of the Turbine Building onto 
the Nuclear Island during a seismic event. The proposed changes do 
not affect or impact this design capability. Therefore, the response 
of the safety related systems, structures, and components in the 
Nuclear Island to earthquakes and postulated accidents are not 
affected by the bracing of the turbine building. Based on the above, 
there is no change in the probability of an accident previously 
evaluated. The activity does not introduce a new fission product 
release path, result in a new fission product barrier failure mode, 
or create a new sequence of events that result in significant fuel 
cladding failures. Accordingly, there is no change in the 
consequences of an accident previously evaluated.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The turbine building bracing design is changed to a mixed 
bracing system which uses Special Concentrically Braced Framing 
(SCBF) and Eccentrically Braced Framing (EBF). The main area of the 
turbine building continues to meet its design function of preventing 
a turbine building collapse from impairing the integrity of seismic 
Category I structures, systems, or components. The design function 
of the turbine building first bay to provide the intended 
limitations to a potential collapse onto the nuclear island during a 
seismic event is retained. The turbine building structure does not 
involve any accident initiating component and therefore, changes to 
use SCBF and EBF would not introduce new accident components or 
faults.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Use of a mixed bracing system and changing the structural code 
design for the turbine building main area continue to meet the 
design function of preventing a turbine building collapse from 
impairing the integrity of seismic Category I Structures, Systems, 
and Components. In addition, the first bay of the turbine building 
continues to be designed to seismic Category II requirements to 
prevent a turbine building collapse from impairing the integrity of 
the seismic Category I nuclear island structures, systems and 
components. This portion of the turbine building and its design is 
unchanged by the proposed amendment. Maintaining the seismic 
Category II rating for the turbine building first bay, along with 
continuing to meet the design function for the non-safety, non-
seismic design of the turbine building main area preserves the 
current structural safety margins.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.


[[Page 14137]]


    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: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
    NRC Branch Chief: Lawrence Burkhart, Acting.

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: January 11, 2013.
    Description of amendment request: The proposed change would amend 
Combined License Nos. NPF-91 and NPF-92 for Vogtle Electric Generating 
Plant (VEGP) Units 3 and 4 in regard to the Chemical and Volume Control 
System (CVS) by: (1) Providing a spring-assisted check valve around the 
air-operated Reactor coolant System (RCS) Purification Return Line Stop 
Check Valve, (2) replacing the CVS zinc addition inboard containment 
isolation lift check valve with an air-operated globe valve and a 
thermal relief valve and (3) separating the zinc and hydrogen injection 
paths and relocate the zinc injection path.
    Because, this proposed change requires a departure from Tier 1 
information in the Westinghouse Advanced Passive 1000 design control 
document (DCD), the licensee also requested an exemption from the 
requirements of the Generic DCD Tier 1 in accordance with 52.63(b)(1).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The changes to provide a spring-assisted check valve located in 
the bypass line around the makeup stop check valve would continue to 
meet the existing design functions because the ASME Boiler and 
Pressure Vessel Code (ASME Code) Section III valves will maintain 
the flow isolation design function and preserve the Reactor Coolant 
System (RCS) pressure boundary safety function. The replacement of 
the Chemical and Volume Control System (CVS) zinc addition inboard 
containment isolation lift check valve with an air operated globe 
valve and addition of a pressure relief valve would continue to meet 
the containment isolation and RCS pressure boundary design functions 
because the replacement valves will be designed, analyzed, tested 
and qualified, including seismic qualification, to ASME Code Section 
III requirements. Separating the zinc and hydrogen injection paths 
and relocating the zinc injection point would continue to meet 
containment boundary requirements, including containment isolation 
and in-service testing, and preserve the RCS pressure boundary 
safety functions because the revised containment isolation 
configuration is consistent with those described in 10 CFR 50, 
Appendix A, General Design Criterion (GDC) 55, and the additional 
valves and piping will be qualified to ASME Code Section III. 
Because the proposed CVS changes would preserve the CVS safety-
related design functions, the probability of an accident previously 
evaluated is not affected.
    The CVS safety functions have been preserved, because the 
proposed CVS configuration changes, including revised valve types, 
will perform the same safety functions as the current design. The 
proposed CVS configuration changes would neither impact any accident 
source term parameter or fission product barrier nor affect 
radiological dose consequence analysis.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The additional containment penetration is similar in form, fit, 
and function to the CVS combined zinc/hydrogen containment 
penetration that is currently described in the Updated Final Safety 
Analysis Report. Because the CVS changes use valve types, piping, 
and a containment penetration consistent with those already 
described in the Updated Final Safety Analysis Report, no new 
failure modes or equipment failure initiators are introduced by 
these changes. Accordingly, the proposed changes do not create any 
new malfunctions, failure mechanisms, or accident initiators.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The containment isolation and pressure relief functions would 
not be changed by this activity and are consistent with the existing 
design. The proposed CVS containment penetration is similar in form, 
fit, and function to existing CVS combined zinc/hydrogen containment 
penetration and, therefore, does not affect containment or its 
ability to perform its design function. The addition of these CVS 
components, including piping, a spring-assisted check valve, an air-
operated containment isolation valve, a thermal relief valve and the 
additional CVS containment penetration do not impact a design basis 
or safety limit. Because the CVS design functions of controlling the 
RCS oxygen concentration, reducing radiation fields, containment 
isolation and overpressure protection within existing limits are not 
changed by this activity and are bounded by the existing design, 
there is no change to any current margin of safety.
    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: Mr. M. Stanford Blanton, Balch & Bingham 
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Lawrence Burkhart, Acting.

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 7, 2013 and revised on February 
15, 2013.
    Description of amendment request: The proposed change would amend 
Combined License Nos.: NPF-91 and NPF-92 for Vogtle Electric Generating 
Plant (VEGP) Units 3 and 4 to allow the use of concentrically and 
eccentrically braced frames in the turbine building main area and 
modify the applicable design code.
    Because this proposed change requires a departure from Tier 1 
information in the Westinghouse Advanced Passive 1000 design control 
document (DCD), the licensee also requested an exemption from the 
requirements of the Generic DCD Tier1 in accordance with 52.63(b)(1).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The turbine building bracing design is changed to a mixed 
bracing system which uses special concentric and eccentric bracing. 
The turbine building does not contain safety-related systems or 
components. The main area of the turbine building continues to meet 
its design function of preventing a turbine building collapse from 
impairing the

[[Page 14138]]

integrity of seismic Category I structures, systems, or components. 
The first bay of the turbine building is designed to prevent the 
collapse of the main area of the Turbine Building onto the Nuclear 
Island during a seismic event. The proposed changes do not affect or 
impact this design capability. Therefore, the response of the safety 
related systems, structures, and components in the Nuclear Island to 
earthquakes and postulated accidents are not affected by the bracing 
of the turbine building. Based on the above, there is no change in 
the probability of an accident previously evaluated. The activity 
does not introduce a new fission product release path, result in a 
new fission product barrier failure mode, or create a new sequence 
of events that result in significant fuel cladding failures. 
Accordingly, there is no change in the consequences of an accident 
previously evaluated.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The turbine building bracing design is changed to a mixed 
bracing system which uses Special Concentrically Braced Framing 
(SCBF) and Eccentrically Braced Framing (EBF). The main area of the 
turbine building continues to meet its design function of preventing 
a turbine building collapse from impairing the integrity of seismic 
Category I structures, systems, or components. The design function 
of the turbine building first bay to provide the intended 
limitations to a potential collapse onto the nuclear island during a 
seismic event is retained. The turbine building structure does not 
involve any accident initiating component and therefore, changes to 
use SCBF and EBF would not introduce new accident components or 
faults.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Use of a mixed bracing system and changing the structural code 
design for the turbine building main area continue to meet the 
design function of preventing a turbine building collapse from 
impairing the integrity of seismic Category I Structures, Systems, 
and Components. In addition, the first bay of the turbine building 
continues to be designed to seismic Category II requirements to 
prevent a turbine building collapse from impairing the integrity of 
the seismic Category I nuclear island structures, systems and 
components. This portion of the turbine building and its design is 
unchanged by the proposed amendment. Maintaining the seismic 
Category II rating for the turbine building first bay, along with 
continuing to meet the design function for the non-safety, non-
seismic design of the turbine building main area preserves the 
current structural safety margins.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

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

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of amendment request: December 13, 2012.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 3.7.9, ``Ultimate Heat Sink (UHS),'' to 
incorporate more restrictive UHS level and pond temperature limits 
which are specified in Surveillance Requirements (SRs) 3.7.9.1 and 
3.7.9.2, respectively. In addition, new SR 3.7.9.4 would be added to 
verify that the UHS cooling tower fans respond appropriately to 
automatic start signals.
    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.
    There are no design changes associated with the proposed 
amendment. All design, material, and construction standards that 
were applicable prior to this amendment request will continue to be 
applicable. The proposed change will not adversely affect accident 
initiators or precursors or adversely alter the design assumptions, 
conditions, and configuration of the facility or the manner in which 
the plant is operated and maintained with respect to such initiators 
or precursors. The proposed changes do not affect the way in which 
safety-related systems perform their functions.
    All accident analysis acceptance criteria will continue to be 
met with the proposed changes. The proposed changes will not affect 
the source term, containment isolation, or radiological release 
assumptions used in evaluating the radiological consequences of an 
accident previously evaluated. The proposed changes will not alter 
any assumptions or change any mitigation actions in the radiological 
consequence evaluations in the FSAR [final safety analysis report]. 
The applicable radiological dose acceptance criteria will continue 
to be met.
    The intent of the modified UHS water level and temperature 
limits for TS 3.7.9, as proposed, is to ensure that the UHS can 
perform its specified safety function for accident mitigation, 
including consideration of its 30-day mission time. The proposed 
surveillance limits are more restrictive and are based on an 
analysis that includes credit given to specific operator actions 
(with assumed completion times) not previously assumed. However, the 
operator actions are reasonable and have been established in 
accordance with NRC-approved guidance. Further, they have been 
simulator verified and proven to be capable of being met by plant 
operators under applicable accident scenarios.
    The crediting of these operator actions is consistent with the 
plant's current licensing basis which already credits operator 
action to provide long-term protection of the UHS following an 
accident. These actions, in conjunction with the more restrictive 
proposed UHS water temperature and level surveillance limits, 
support the plant's existing accident analysis such that there is no 
change in analyzed consequences. In light of these considerations, 
there is no significant increase in the consequences of any accident 
previously evaluated with regard to the assumed operator actions and 
revised UHS water level and temperature limits, as proposed. The 
proposed change adds additional controls to the Technical 
Specifications but does not physically alter safety-related systems 
or affect the way in which safety-related systems perform their 
functions per the intended plant design.
    As such, the proposed change will not alter or prevent the 
capability of structures, systems, and components (SSCs) to perform 
their intended functions for mitigating the consequences of an 
accident and meeting applicable acceptance limits. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    With respect to any new or different kind of accident, there are 
no proposed design changes nor are there any changes in the method 
by which any safety-related plant SSC performs its specified safety 
function. The proposed change will not affect the normal method of 
plant operation. No new transient precursors will be introduced as a 
result of this amendment. The reanalysis discussed herein addresses 
new large break LOCA [loss-of-coolant accident] scenarios with 
assumptions, including single failures, aimed at maximizing the UHS 
temperature and minimizing the UHS inventory.
    The proposed change adds requirements to the Technical 
Specifications. The change does not involve a physical modification 
of the plant. The UHS level and temperature limits within which the 
plant is normally operated are being changed in the

[[Page 14139]]

conservative direction. Appropriate changes have been made to the 
emergency operating procedures relied upon to mitigate a design 
basis event. The change does not have a detrimental impact on the 
manner in which plant equipment operates or responds to an actuation 
signal. The changes to the ultimate heat sink (UHS) surveillance 
limits are in the conservative direction.
    The proposed change does not, therefore, create the possibility 
of a new or different accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    There will be no effect on those plant systems necessary to 
assure the accomplishment of protection functions associated with 
reactor operation or the reactor coolant system. There will be no 
impact on the overpower limit, departure from nucleate boiling ratio 
(DNBR) limits, heat flux hot channel factor (FQ), nuclear 
enthalpy rise hot channel factor (F[Delta]H), loss of coolant 
accident peak cladding temperature (LOCA PCT), peak local power 
density, or any other limit and associated margin of safety. 
Required shutdown margins in the COLR [core operating limits report] 
will not be changed.
    The proposed change does not eliminate any surveillances or 
alter the frequency of surveillances required by the Technical 
Specifications. The proposed change would add Technical 
Specification Surveillance Requirements for assuring the automatic 
closure of the UHS cooling tower bypass valves when required and the 
automatic start of the UHS cooling tower fans and their transition 
from slow speed to fast speed when required. The extent of 
Callaway's conformance to NRC Regulatory Guide (RG) 1.27 is 
discussed in FSAR Site Addendum Table 9.2-5 (see Attachment 4 to 
this Enclosure [to the submittal]). RG 1.27 requires that the UHS be 
sized for 30 day post-LOCA operation; however, it does not specify a 
margin value above that 30-day requirement. During initial plant 
licensing (Callaway Safety Evaluation Report, NUREG-0830, Supplement 
4, Section 2.4.4) a UHS level margin of 50% was accepted in lieu of 
a more restrictive minimum Technical Specification water level of 
834 feet mean sea level (16 feet above the reference pond bottom) 
and a thermal and hydrologic analysis of the ESW [essential service 
water] and UHS. In this amendment request SR 3.7.9.1 is being 
changed to adopt the former and the supporting EF-123 analysis 
addresses the latter. The SER [safety evaluation report] Supplement 
4 discussion, copied in Section 2.2 of this Evaluation, will no 
longer be applicable upon NRC approval of this license amendment 
request.
    As such, the proposed change does not involve a significant 
reduction in a margin of safety as defined in any regulatory 
requirement or guidance document.

    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: John O'Neill, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street NW., Washington, DC 20037.
    NRC Branch Chief: Michael T. Markley.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of amendment request: December 20, 2012.
    Description of amendment request: The amendment would revise a 
methodology in the licensing basis as described in the Final Safety 
Analysis Report--Standard Plant to include damping values for the 
seismic design and analysis of the integrated head assembly that are 
consistent with the recommendations of NRC Regulatory Guide 1.61, 
``Damping Values for Seismic Design of Nuclear Power Plants,'' Revision 
1, March 2007.
    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 change would allow use of critical damping values 
consistent with the recommendations of RG [Regulatory Guide] 1.61, 
``Damping Values for Seismic Design of Nuclear Power Plants,'' 
Revision 1, dated March 2007, for the seismic design and analysis of 
the IHA [integrated head assembly].
    The RG 1.61, Revision 1, Table 1 note allowing use of a 
``weighted average'' for design-basis SSE [safe shutdown earthquake] 
damping values applicable to steel structures of different 
connection types, is also applied to determine the IHA design-basis 
OBE [operating basis earthquake] damping values. RG 1.61, Revision 
1, Table 2 for OBE damping values does not contain the same note 
found in Table 1. However use of the note for the determination of 
the OBE damping value is consistent with the use of the note for the 
determination of the SSE damping values, and a weighted average more 
realistically represents the IHA structure. RG 1.61, Revision 1, 
specifies the damping values that the NRC staff currently considers 
acceptable for complying with the agency's regulations and guidance 
for seismic analysis. Revision 1 incorporates the latest data and 
information, and reduces unnecessary conservatism in specification 
of damping values for seismic design and analysis of SSCs 
[structures, systems, and components].
    The proposed change does not change the design functions of the 
IHA or its response to design-basis events, nor does it affect the 
capability of related SSCs to perform their design or safety 
functions. The use of the proposed damping values in the seismic 
design and analysis of the IHA is related to the ability of the IHA 
to function in response to design-basis seismic events, and is 
unrelated to the probability of occurrence of those events, or other 
previously evaluated accidents. Therefore, the proposed change will 
not have any impact on the probability of an accident previously 
evaluated.
    The proposed damping values are an element of the seismic 
analyses performed to confirm the ability of the IHA to function 
under postulated seismic events while maintaining resulting stresses 
within ASME [American Society of Mechanical Engineers Boiler and 
Pressure Vessel Code] Section III allowable values. Therefore, the 
use of damping values consistent with the recommendations of RG 
1.61, Revision 1 does not result in an increase in the consequences 
of accidents previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve changes to any plant SSCs, 
nor does it involve changes to any plant operating practice or 
procedure. The damping values are an element of the seismic analyses 
performed to confirm the ability of the IHA to function under 
postulated seismic events while maintaining resulting stresses 
within ASME Section III allowable values. Therefore, no credible new 
failure mechanisms, malfunctions, or accident initiators not 
considered in the design and licensing bases are created that would 
create the possibility of a new or different kind of accident.
    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 change involve a significant reduction in a margin 
of safety?
    Response: No.
    The design basis of the plant requires structures to be capable 
of withstanding normal and accident loads including those from a 
design basis earthquake. The proposed change would allow the use of 
damping values in the IHA seismic analyses that are, in general, 
more realistic and, thus, more accurate than the damping values 
recommended in RG 1.61, Revision 0, used in the original analysis 
for the SSE, or the plant specific damping values used in the 
original analysis for the OBE. The damping values in RG 1.61, 
Revision 0, were based on limited data, expert opinion, and other 
information available in 1973. NRC and industry research since 1973 
shows that the damping values provided in the original version of RG 
1.61 may not reflect realistic damping values for SSCs. RG 1.61, 
Revision 1, therefore, provides damping values based on the updated 
research results that predict

[[Page 14140]]

and estimate damping values for seismic design of SSCs in nuclear 
power plants, and similarly should not be regarded as an arbitrary 
lowering of the margins 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: John O'Neill, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street NW., Washington, DC 20037.
    NRC Branch Chief: Michael T. Markley.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Combined Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    A notice of consideration of issuance of amendment to facility 
operating license or combined license, as applicable, proposed no 
significant hazards consideration determination, and opportunity for a 
hearing in connection with these actions, was published in the Federal 
Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland 20852. Publicly available documents created or received at the 
NRC are accessible electronically through the Agencywide Documents 
Access and Management System (ADAMS) in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR's Reference staff at 1-800-397-4209, 301-415-4737 or by 
email to pdr.resource@nrc.gov.

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

    Date of application for amendment: March 16, 2012, as supplemented 
by letter dated August 16, 2012.
    Brief Description of amendment: The amendment revised the Technical 
Specifications (TSs) to make corrections in TS Table 3.3.1-1 for 
Overtemperature Delta Temperature consistent with NUREG-1431, Revision 
3, ``Standard Technical Specifications Westinghouse Plants.''
    Date of issuance: February 13, 2013.
    Effective date: As of date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 231.
    Renewed Facility Operating License No. DPR-23: Amendment changed 
the license and TSs.
    Date of initial notice in Federal Register: April 17, 2012 (77 FR 
22811). The supplement dated August 16, 2012, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 13, 2013.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois

Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 
2, Ogle County, Illinois

    Date of application for amendment: June 6, 2012, as supplemented by 
letter dated. November 19, 2012.
    Brief description of amendment: The proposed amendment modifies 
Braidwood and Byron technical specifications (TS) to add a Note to 
surveillance requirements (SRs) 3.3.1.7, 3.3.1.8, and 3.3.1.12 in TS 
3.3.1, ``Reactor Trip System (RTS) Instrumentation,'' and SRs 3.3.2.2 
and 3.3.2.6 in TS 3.3.2, ``Engineered Safety Features Actuation System 
(ESFAS) Instrumentation,'' to exclude the Solid State Protection System 
input relays from the Channel Operational Test Surveillance for RTS and 
ESFAS functions with installed bypass capability which the U.S. Nuclear 
Regulatory Commission (NRC) approved by letters dated March 30, and 
April 9, 2012.
    Date of issuance: February 6, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 171 for Braidwood Station, Units 1 and 2, and 178 
for Byron Station, Unit Nos. 1 and 2, respectively.
    Facility Operating License Nos. NPF-72. NPF-77, NPF-37, and NPF-66: 
The amendments revised the Technical Specifications and License.
    Date of initial notice in Federal Register: September 4, 2012 (77 
FR 53927).
    The November 19, 2012, supplement contained clarifying information 
and did not change the NRC staff's initial proposed finding of no 
significant hazards consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 6, 2013.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry 
Nuclear Plant (BFN), Units 2 and 3, Limestone County, Alabama

    Date of application for amendments: February 25, 2011, as 
supplemented by letters dated September 15, 2011, July 30, 2012, and 
January 24, 2013. The enclosure to the July 30, 2012, letter 
superseded, in its entirety, the enclosure to the February 25, 2011, 
letter.
    Brief description of amendments: The amendments delete the BFN, 
Units 2 and 3, Technical Specification (TS) Surveillance Requirement 
3.5.1.12, which requires the verification of the capability to 
automatically transfer the power supply from the normal source to the 
alternate source for each Low-Pressure Coolant Injection subsystem 
inboard injection valve and each recirculation pump discharge valve on 
a 24-month frequency. In addition, these amendments approve the use of 
a modified loss-of-coolant accident

[[Page 14141]]

(LOCA) methodology that requires revising TS 5.6.5.b to include a 
reference to the modified LOCA methodology. Also, the amendments revise 
TSs 3.3.1.1, 5.6.5.a, and 5.6.5.b to include the modified LOCA 
methodology and the oscilliation power range monitor upscale function 
period based detection algorithm setpoint limits.
    Date of issuance: February 15, 2013.
    Effective date: The amendments are effective as of this date of 
issuance. For Unit 2, the amendment shall be implemented prior to 
entering Mode 3 (i.e., Hot Shutdown) from the spring 2013 refueling 
outage. For Unit 3, changes to TSs 5.6.5 and 3.3.1 shall be implemented 
within 60 days of issuance. The remaining changes shall be implemented 
prior to entering Mode 3 from the spring 2014 refueling outage.
    Amendment Nos.: Unit 1--309 and Unit 2--268.
    Renewed Facility Operating License Nos. DPR-52 and DPR-68: 
Amendments revised the licenses and TSs.
    Date of initial notice in Federal Register: The original 
application dated February 25, 2011, was noticed on May 3, 2011 (76 FR 
24930). The supplement dated July 30, 2012, was noticed on November 5, 
2012 (77 FR 66490). The supplement dated January 24, 2013, provided 
additional information that clarified the licensee's July 30, 2012, 
submittal, did not expand the scope of the application as noticed and 
did not change the NRC staff's proposed no significant hazards 
consideration determination as published in the FR on November 5, 2012 
(77 FR 66490).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 15, 2013.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket No. 50-339, North Anna 
Power Station, Unit No. 2, Louisa County, Virginia

    Date of application for amendment: May 11, 2012.
    Brief Description of amendment: The amendment would revise the 
Technical Specification (TS) 3.1.7, ``Rod Position Indication'' to 
allow two demand position indicators in one or more banks to be 
inoperable for up to 4 hours. This change is proposed as a temporary 
change to the TS for the current operating cycle and is proposed as a 
footnote to the current TS Limiting Condition for Operation (LCO) 
Section 3.1.7, Condition D.
    Date of issuance: February 14, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within the end of operating Cycle 22.
    Amendment No.: 251.
    Renewed Facility Operating License No. NPF-7: Amendment changes the 
license and the TS.
    Date of initial notice in Federal Register: June 12, 2012 (77 FR 
35077).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 14, 2013.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 25th day of February 2013.

    For the Nuclear Regulatory Commission.
Louise Lund,
Deputy Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2013-04885 Filed 3-1-13; 8:45 am]
BILLING CODE 7590-01-P