Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 19746-19757 [2013-07467]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 78, Number 63 (Tuesday, April 2, 2013)]
[Notices]
[Pages 19746-19757]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2013-07467]


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NUCLEAR REGULATORY COMMISSION

[NRC-2013-0060]


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 (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 March 7, 2013, to March 20, 2013. The last 
biweekly notice was published on March 19, 2013 (78 FR 16876).

ADDRESSES: You may access information and comment submissions related 
to this document, which the NRC possesses and is publicly available, by 
searching on https://www.regulations.gov under Docket ID NRC-2013-0060. 
You may submit comments by any of the following methods:
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2013-0060. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; email: Carol.Gallagher@nrc.gov.
     Mail comments to: Cindy Bladey, Chief, Rules, 
Announcements, and Directives Branch (RADB), Office of Administration, 
Mail Stop: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
     Fax comments to: RADB at 301-492-3446.
    For additional direction on accessing information and submitting 
comments, see ``Accessing Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

SUPPLEMENTARY INFORMATION: 

I. Accessing Information and Submitting Comments

A. Accessing Information

    Please refer to Docket ID NRC-2013-0060 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 https://www.regulations.gov and search for Docket ID NRC-2013-0060.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may access publicly available documents online in the NRC 
Library at https://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 NRC-2013-0060 in the subject line of your 
comment submission, in order to ensure that the NRC is able to make 
your comment submission available to the public in this docket.
    The NRC cautions you not to include identifying or contact 
information that that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at https://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC

[[Page 19747]]

does not routinely edit comment submissions to remove such information 
before making the comment submissions available to the public or 
entering the comment submissions into ADAMS.
Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses and Combined Licenses, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing
    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in section 50.92 of Title 10 of the Code 
of Federal Regulations (10 CFR), this means that operation of the 
facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license or 
combined license. Requests for a hearing and a petition for leave to 
intervene shall be filed in accordance with the Commission's ``Agency 
Rules of Practice and Procedure'' in 10 CFR part 2. Interested 
person(s) should consult a current copy of 10 CFR 2.309, which is 
available at the NRC's PDR, located at One White Flint North, Room O1-
F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The 
NRC regulations are accessible electronically from the NRC Library on 
the NRC's Web site at https://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's 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

[[Page 19748]]

the Secretary by email at hearing.docket@nrc.gov, or by telephone at 
301-415-1677, to request (1) a digital identification (ID) certificate, 
which allows the participant (or its counsel or representative) to 
digitally sign documents and access the E-Submittal server for any 
proceeding in which it is participating; and (2) advise the Secretary 
that the participant will be submitting a request or petition for 
hearing (even in instances in which the participant, or its counsel or 
representative, already holds an NRC-issued digital ID certificate). 
Based upon this information, the Secretary will establish an electronic 
docket for the hearing in this proceeding if the Secretary has not 
already established an electronic docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at https://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 https://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with the NRC guidance 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC's Web site 
at https://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 
https://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 https://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to 
ADAMS or who encounter problems in accessing the documents located in 
ADAMS, should contact the NRC's PDR Reference staff at 1-800-397-4209, 
301-415-4737, or by email to pdr.resource@nrc.gov.

[[Page 19749]]

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

    Date of amendment request: July 21, 2010.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) \3/4\.9.3.1, ``Decay Time'' for 
Millstone Power Station Unit 2 (MPS2). The proposed change would revise 
TS \3/4\.9.3.1 by reducing the minimum decay time for irradiated fuel 
prior to movement in the reactor vessel from 150 hours to 100 hours. A 
reduction in the minimum decay time requirement is requested to provide 
additional flexibility in outage planning such that irradiated fuel can 
be moved from the reactor vessel to the spent fuel pool earlier in an 
outage.
    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:

Criterion 1

    Will operation of the facility in accordance with the proposed 
change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The accident of concern related to the proposed change is the 
FHA [fuel handling accident]. This accident assumes a dropped fuel 
assembly with resulting damage and release of the gap activity from 
the entire assembly. The FHA assumes that fuel movement is delayed 
for some time period after shutdown to accommodate for radioactive 
decay of the short-lived fission products. The probability of a FHA 
occurrence is dependent on moving fuel not when the fuel movement 
occurs. Reducing the decay time required by TS \3/4\.9.3.1 from 150 
hours to 100 hours does not increase the probability of a FHA since 
the timing of fuel movement in the reactor pressure vessel does not 
alter/impact the manner in which fuel assemblies are handled.
    Reducing the decay time requirement in TS \3/4\.9.3.1 from 150 
hours to 100 hours does not change the consequences of the offsite 
dose and control room dose projections for the currently approved 
design basis FHA analysis. The current FHA analysis presented in 
FSAR [final safety analysis report] Section 14.7.4 and approved in 
License Amendment 298 assumes a minimum 100 hour decay time. 
Therefore, the dose results of this FHA analysis are unchanged, and 
remain within applicable regulatory limits.
    Based on the reasons presented above, operation of the facility 
in accordance with the proposed amendment would not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.

Criterion 2

    Will operation of the facility in accordance with the proposed 
change create the possibility of a new or different kind of accident 
from any accident previously evaluated?
    Response: No.
    The proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
No new accident will be created as a result of reducing the decay 
time requirement in TS \3/4\.9.3.1. Plant operation, including fuel 
handling, will not be affected by the proposed change, as to when 
fuel is moved and no new failure modes will be created.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.

Criterion 3

    Will operation of the facility in accordance with the proposed 
change involve a significant reduction in the margin of safety?
    Response: No.
    The proposed change does not significantly reduce the margin of 
safety. The current analysis of record for the FHA already accounts 
for irradiated fuel with at least 100 hours of decay. This approved 
analysis has shown that the projected doses will remain within 
applicable regulatory limits; therefore, the margin of safety is 
unchanged.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 
23219.
    NRC Acting Branch Chief: Sean C. Meighan.

Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point 
Nuclear Generating, Unit 2, Westchester County, New York

    Date of amendment request: January 28, 2013.
    Description of amendment request: Nuclear Safety Advisory Letter 
11-5 identified Westinghouse methodology errors in the long-term mass 
and energy releases during a large break loss-of-coolant accident. 
These impacted the containment integrity analysis for Indian Point, 
Unit 2. A re-analysis of the large break loss-of-coolant accident for 
the limiting single failure concluded that four, rather than three 
containment fan cooler units would need to be credited. The proposed 
change will revise Technical Specification Bases Sections 3.6.4, 
``Containment Pressure,'' 3.6.5, ``Containment Air Temperature,'' and 
3.6.6, ``Containment Spray System and Containment Fan Cooler Unit (FCU) 
System.''
    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 
identified?
    Response: No.
    The proposed change would not change the current limiting EDG 
[emergency diesel generator] failure but would credit four rather 
than three can cooler units for containment heat removal. Four fan 
cooler units are available after the single failure. The fan cooler 
units are not accident initiators so the probability of an accident 
does not increase. Crediting all four fan cooler units will keep the 
post accident containment pressure within current limits and 
therefore does not increase the probability or consequences of a 
previously evaluated accident, but is a change from the analyses 
approved by the NRC [Nuclear Regulatory Commission] during stretch 
power uprate.
    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 of different 
kind of accident from any accident previously evaluated?
    Response: No.
    There are no changes to design, no changes to operating 
procedures, and the revised licensing basis change is consistent 
with the available equipment following the postulated worst case 
single failure.
    Therefore the proposed change does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The change reflects the credit for equipment that was always 
available but not previously credited (as a conservatism) in the 
licensing basis analyses. With credit for four fan cooler units, the 
post accident containment pressure remains within current limits and 
there is no reduction in a margin of safety.
    Therefore the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440

[[Page 19750]]

Hamilton Avenue, White Plains, NY 10601.
    NRC Acting Branch Chief: Sean Meighan.

Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point 
Nuclear Generating, Unit 2, Westchester County, New York

    Date of amendment request: February 6, 2013.
    Description of amendment request: The proposed amendment will 
revise the Reactor Heatup and Cooldown curves and Low Temperature 
Overpressure Protection Requirements in Technical Specifications (TSs) 
3.4.3, ``RCS [reactor coolant system] Pressure and Temperature (P/T) 
Limits,'' 3.4.6, ``RCS Loops--MODE 4,'' 3.4.7, ``RCS Loops--MODE 5, 
Loops Filled,'' 3.4.10, ``Pressurizer Safety Valves,'' and 3.4.12, 
``Low Temperature Overpressure Protection (LTOP).''
    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. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence of consequences of an accident previously 
evaluated.
    The proposed TS changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
Except for a setpoint change for automatic PORV [power-operated 
relief valve] actuation, there are no physical changes to the plant 
being introduced by the proposed changes to the heatup and cooldown 
limitation curves. The proposed changes do not modify the RCS 
pressure boundary. That is, there are no changes in operating 
pressure, materials, or seismic loading. The proposed changes do not 
adversely affect the integrity of the RCS pressure boundary such 
that its function in the control of radiological consequences is 
affected. The proposed heatup and cooldown limitation curves were 
generated in accordance with the fracture toughness requirements of 
10CFR50 [10 CFR 50] Appendix G, and ASME B&PV code [American Society 
of Mechanical Engineers Boiler and Pressure Vessel Code], Section 
XI, Appendix G edition with 2000 Addenda. The proposed heatup and 
cooldown limitation curves were established in compliance with the 
methodology used to calculate and predict effects of radiation on 
embrittlement of RPV [reactor pressure vessel] beltline materials. 
Use of this methodology provides compliance with the intent of 
10CFR50 [10 CFR 50] Appendix G and provides margins of safety that 
ensure non-ductile failure of the RPV will not occur. The proposed 
heatup and cooldown limitation curves prohibit operation in regions 
where it is possible for non-ductile failure of carbon and low alloy 
RCS materials to occur. Hence, the primary coolant pressure boundary 
integrity will be maintained throughout the limit of applicability 
of the curves, 48 EFPY [Effective Full Power Years].
    Operation within the proposed LTOP limits ensures that 
overpressurization of the RCS at low temperatures will not result in 
component stresses in excess of those allowed by the ASME B&PV Code 
Section XI Appendix G.
    Consequently, the proposed changes do not involve a significant 
increase in the probability or the consequences of an accident 
previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed TS changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. No new modes of operation are introduced by the proposed 
changes. The proposed changes will not create any failure mode not 
bounded by previously evaluated accidents. Further, the proposed 
changes to the heatup and cooldown limitation curves and the LTOP 
limits do not affect any activities or equipment other than the RCS 
pressure boundary and do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Consequently, the proposed changes do not create the possibility 
of a new or different kind of accident, from any accident previously 
evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in the margin of 
safety.
    The Proposed TS changes do not involve a significant reduction 
in the margin of safety. The revised heatup and cooldown limitation 
curves and LTOP limits are established in accordance with current 
regulations and the ASME B&PV Code 1998 edition with 2000 Addenda. 
These proposed changes are acceptable because the ASME B&PV Code 
maintains the margin of safety required by 10CFR50.55(a) [10 CFR 
50.55(a)]. Because operation will be within these limits, the RCS 
materials will continue to behave in a non-brittle manner consistent 
with the original design bases.
    The proposed changes to the allowable operation of charging and 
safety injection pumps when LTOP is required to be operable is 
consistent with the IP2 licensing bases as established in TS 
Amendment 262.
    Therefore, Entergy has concluded that the proposed changes do 
not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, and with the changes noted above in square brackets, it 
appears that the three standards of 10 CFR 50.92(c) are satisfied. 
Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Acting Branch Chief: Sean Meighan.

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

    Date of amendment request: January 28, 2013.
    Description of amendment request: Nuclear Safety Advisory Letter 
(NSAL) 11-5 identified Westinghouse methodology errors in the long-term 
mass and energy releases during a large break loss-of-coolant accident. 
These impacted the containment integrity analysis for Indian Point Unit 
No. 3 and required revisions to limiting initial operating conditions 
(i.e., containment temperature, containment pressure, and refueling 
water storage tank temperature) and require revisions to Technical 
Specifications (TSs) 3.5.4, ``Refueling Water Storage Tank (RWST),'' 
and 3.6.4, ``Containment Pressure.'' In addition, revisions are 
proposed for TS 3.6.3, ``Containment Isolation Valves,'' to delete a 
redundant surveillance requirement and TS 5.5.15, ``Containment Leakage 
Rate Testing Program,'' to reflect a slightly higher calculated 
containment peak pressure.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change would not change the current EDG [emergency 
diesel generator] failure but limits the RWST temperature to 
<=105[emsp14][deg]F and containment pressure to <=1.5 psig [pounds 
per square inch gauge] (when RWST temperature is >95[emsp14][deg]F 
or containment/accumulator temperature is >125[emsp14][deg]F). The 
proposed change also removes a redundant TS for Containment testing 
and corrects the peak pressure in the containment testing program. 
The initial conditions assumed in accident analysis are not accident 
initiators so the probability of an accident does not increase. The 
change in initial conditions compensates for the error corrections 
and maintains the post accident containment pressure within 0.38 
psig of the current value and within Containment testing limits and 
therefore does not increase the probability or consequences of a 
previously evaluated accident. Therefore the proposed change does 
not involve a significant increase in the

[[Page 19751]]

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 change to the initial conditions assumed in the analysis for 
peak containment pressure, the removal of a redundant Technical 
Specification and the correction to the peak pressure limit in the 
Containment testing program do not create the possibility of a new 
or different accident. There are no changes to design or operating 
procedures that could create a new or different kind of accident 
since the changes only affect the initiating conditions. The revised 
analysis is consistent with the available equipment following the 
postulate worst case single failure.
    Therefore the proposed change does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The change in peak containment pressure is from 42 psig to 42.38 
psig as a result of the error corrections of NSAL-11-5 and change to 
the initial conditions for the RWST temperature and containment 
pressure. There is an insignificant impact on other programs due to 
change in peak containment pressure, which remains well below the 
containment design pressure of 47 psig. Therefore there is not 
significant reduction in margin.
    Therefore the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Acting Branch Chief: Sean Meighan.

Exelon Generation Company (EGC), LLC, Docket No. 50-374, LaSalle County 
Station (LSCS), Unit 2, LaSalle County, Illinois

    Date of amendment request: October 15, 2012.
    Description of amendment request: The proposed amendment would 
remove License Conditions which are no longer necessary to address an 
interim configuration of the LaSalle County Station, Unit 2, spent fuel 
pool prior to completed installation of NETCO-SNAP-IN[supreg] inserts.
    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 removes License Conditions within the LSCS 
Unit 2 Operating License related to interim configurations of the 
SFP during the installation of the NETCO-SNAP-IN[supreg] inserts and 
the required completion date for installation. All changes proposed 
by EGC in this license amendment request are administrative in 
nature because they remove License Conditions that have either been 
satisfied or that are no longer applicable. There are no physical 
changes to the facilities, nor any changes to the station operating 
procedures, limiting conditions for operation, or limiting safety 
system settings.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change removes License Conditions within the LSCS 
Unit 2 Operating License related to interim configurations of the 
SFP during the installation of the NETCO-SNAP-IN[supreg] inserts and 
the required completion date for installation. There are no changes 
to the SFP criticality analysis associated with the proposed change. 
No physical changes to the plant are proposed, and there are no 
changes to the manner in which the plant is operated. Rather, the 
proposed change is administrative because it involves removing 
License Conditions that have either been satisfied or that are no 
longer applicable.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change removes License Conditions within the LSCS 
Unit 2 Operating License related to interim configurations of the 
SFP during the installation of the NETCO-SNAP-IN[supreg] inserts and 
the required completion date for installation. Plant safety margins 
are established through limiting conditions for operation, limiting 
safety system settings, and safety limits specified in Technical 
Specifications. The proposed change does not alter these established 
safety margins. The proposed change does not alter the criticality 
analysis for the SFP and does not affect the SFP criticality safety 
margin. The proposed change is administrative because it involves 
removing License Conditions that have either been satisfied or that 
are no longer applicable.
    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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Ms. Tamra Domeyer, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Acting Branch Chief: Jeremy S. Bowen.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: July 12, 2012.
    Description of amendment request: The proposed amendments would 
modify Technical Specification 3.7.3, ``Ultimate Heat Sink,'' by 
establishing controls which allow for the increase of cooling water 
temperature from 104[emsp14][deg]F to 107[emsp14][deg]F for plant 
safety systems.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change makes no physical changes to the plant, nor 
does it alter any of the assumptions or conditions upon which the 
UHS is designed. These assumptions and conditions as described in 
the LSCS UFSAR include failure of the cooling lake dike, a loss of 
offsite power, and a DBA LOCA on one unit and a normal shutdown of 
the other unit.
    The accidents analyzed in the UFSAR are assumed to be initiated 
by the failure of plant structures, systems, or components (SSCs). 
An inoperable UHS is not an initiator of any analyzed events as 
described in the UFSAR. The impact on the structural integrity of 
the UHS due to a potential increase water temperature prior to and 
during the UHS design basis event has been evaluated, and does not 
increase the probability of the failure of the cooling lake dike. 
The proposed temperature limit for cooling water supplied to the 
plant from the CSCS Pond could reduce the commercial capability of 
the LSCS units; however, it does not result in an increase in the 
probability of occurrence for any of the events described in the 
UFSAR.
    The basis provided in Regulatory Guide 1.27, ``Ultimate Heat 
Sink for Nuclear Power Plants,'' Revision 1, dated March 1974, was 
employed for the temperature analysis of the LSCS UHS to implement 
General Design Criteria 2, ``Design bases for protection

[[Page 19752]]

against natural phenomena,'' and 44, ``Cooling water,'' of Appendix 
A to 10 CFR Part 50. This Regulatory Guide was employed for both the 
original design and licensing basis of the LSCS UHS and a subsequent 
evaluation which investigated the potential for changing the average 
water temperature of the cooling water supplied to the plant from 
the CSCS Pond from a fixed temperature limit to a limit based on the 
time of day. The meteorological conditions chosen for the LSCS UHS 
analysis utilized a 31-day period consisting of the most severe one 
day, combined with the most severe 30 days based on historical data. 
The heat loads selected for the UHS analysis considered failure of 
the cooling lake dike, a loss of offsite power, and a DBA LOCA on 
one unit and a normal shutdown of the other unit. The LSCS cooling 
lake is conservatively assumed to be unavailable at the start of the 
event.
    The analysis shows that with an initial UHS temperature less 
than or equal to the proposed time-of-day-based limit, the required 
safety-related heat loads can be adequately cooled for 30 days while 
continuing to ensure safety-related cooling water temperature 
remains less than the design temperature for LSCS, Units 1 and 2.
    Based on the above, it has been demonstrated that the change of 
the initial temperature limit for cooling water supplied to the 
plant from the CSCS Pond to less than or equal to a temperature 
based on the time of day will not impede the ability of the 
equipment and components cooled by the UHS during a UHS design basis 
event to perform their safety functions.
    There is no impact of this change on LSCS safety analyses 
including the consequences of all postulated events since all 
required safety-related equipment continues to perform as designed. 
The effects of the proposed change on the ability of the UHS to 
assure that a 30-day supply of water is available considering losses 
due to evaporation, seepage, and firefighting have been considered. 
Sufficient inventory remains available to mitigate the design basis 
event for the LSCS UHS for the required 30-day period.
    Therefore, the proposed activity does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not physically alter the operation, 
testing, or maintenance of any plant SSCs beyond operating with a 
UHS temperature limit based on the time of day. The proposed change 
is bounded by existing design analyses. Moreover, the UHS 
temperature does not initiate accident precursors. The impact of 
increased UHS temperature can affect the commercial operation of the 
plant, but the proposed change would not create any accident not 
considered in the LSCS UFSAR.
    This proposed change will not alter the manner in which 
equipment operation is initiated, nor will the functional demands on 
credited equipment be changed. No alteration in the procedures that 
ensure the LSCS units remain within analyzed limits is proposed, and 
no change is being made to procedures relied upon to respond to an 
off-normal event.
    As such, no new failure modes are being introduced. The proposed 
change does not alter assumptions made in the LSCS safety analysis.
    Changing the temperature of cooling water supplied to the plant 
from the CSCS Pond (i.e., the UHS) as proposed has no impact on 
plant accident response. The proposed temperature limits do not 
introduce new failure mechanisms for SSCs. An engineering analysis 
performed to support the change in temperature of cooling water 
supplied to the plant from the CSCS Pond provides the basis to 
conclude that the equipment is adequately designed for operation as 
proposed.
    All systems that are important to safety will continue to be 
operated and maintained within their design bases, and the proposed 
change will continue to ensure that all associated systems and 
components are operated reliably within their design capabilities.
    The proposed change will ensure the maximum temperature of the 
cooling water supplied to the plant during the UHS design basis 
event remains less than the current safety-related cooling water 
design temperature for LSCS, Units 1 and 2. Therefore, there is no 
impact of this change on the LSCS safety analyses including 
inventory and cooling requirements for safety-related systems using 
the UHS as their cooling water supply.
    All systems will continue to be operated within their design 
capabilities, no new failure modes are introduced, nor is there any 
adverse impact on plant equipment; therefore, the proposed change 
does not result in the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is determined by the design and 
qualification of the plant equipment, the operation of the plant 
within analyzed limits, and the point at which protective or 
mitigative actions are initiated. The proposed change does not 
impact any of these factors. There are no required design changes or 
equipment performance parameter changes associated with the proposed 
change. No protection setpoints are affected as a result of this 
change. The proposed change in the limit for the temperature of 
cooling water supplied to the plant from the CSCS Pond will not 
change the operational characteristics of the design of any 
equipment or system. All accident analysis assumptions and 
conditions will continue to be met.
    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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Ms. Tamra Domeyer, Associate General 
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Acting Branch Chief: Jeremy S. Bowen.

NextEra Energy Seabrook, LLC., Docket No. 50-443, Seabrook Station, 
Unit 1, Rockingham County, New Hampshire

    Date of amendment request: March 1, 2013.
    Description of amendment request: The proposed amendment will 
revise the Seabrook Technical Specifications (TSs). The proposed 
amendment will make administrative changes and corrections to the TSs. 
The proposed changes delete TS Index and make corrections to TS 3.4.8, 
``Reactor Coolant System Specific Activity,'' and TS 6.8.1.6.a, ``Core 
Operating Limits Report.''
    Basis for proposed NSHC 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 changes (1) remove the index from the TS, (2) 
correct an error in the units of activity for 100/E in TS 3.4.8, 
Reactor Coolant System Specific Activity, and (3) remove an 
incorrect, non-applicable reference in TS 6.8, Core Operating Limits 
Report. The proposed changes are all administrative in nature. The 
administrative changes are not initiators of any accident previously 
evaluated, and, consequently, the probability and consequences of an 
accident previously evaluated is not significantly increased.
    Therefore, the proposed change does 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 previously evaluated.
    The proposed changes are administrative in nature so no new or 
different accidents result from the proposed changes. The changes do 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed), a significant change 
in the method of plant operation, or new operator actions. The 
changes do not alter 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 accident previously 
evaluated.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.

[[Page 19753]]

    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel cladding, reactor coolant 
system pressure boundary, and containment structure) to limit the 
level of radiation dose to the public. The proposed administrative 
changes do not involve a change in the method of plant operation, do 
not affect any accident analyses, and do not relax any safety system 
settings.
    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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves NSHC.
    Attorney for licensee: Mr. James Petro, Managing Attorney, Florida 
Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Meena Khanna.

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

    Date of amendment request: September 28, 2012, as supplemented on 
November 8, 2012 and December 18, 2012.
    Description of amendment request: The proposed amendment requests 
U.S. Nuclear Regulatory Commission (NRC) approval to adopt a new fire 
protection licensing basis which complies with the requirements in 10 
CFR 50.48(a), 10 CFR 50.48(c), and the guidance in Regulatory Guide 
(RG) 1.205, Revision 1, ``Risk-Informed, Performance Based Fire 
Protection for Existing Light-Water Nuclear Power Plants.'' This 
amendment request also follows the guidance in Nuclear Energy Institute 
(NEI) 04-02, Revision 2, ``Guidance for Implementing a Risk-Informed, 
Performance-Based Fire Protection Program Under 10 CFR 50.48(c).'' If 
approved, the PINGP fire protection program would transition to a new 
Risk-Informed, Performance-Based alternative in accordance with 10 CFR 
50.48(c), which incorporates by reference National Fire Protection 
Association Standard 805 (NFPA 805). The NFPA 805 fire protection 
program would supersede the current fire protection program licensing 
basis in accordance with 10 CFR Part 50, Appendix R.
    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.
    Operation of the Prairie Island Nuclear Generating Plant (PINGP) 
in accordance with the proposed amendment does not increase the 
probability or consequences of accidents previously evaluated. 
Engineering analyses, which may include engineering evaluations, 
probabilistic safety assessments, and fire modeling evaluations, 
have been performed to demonstrate that the performance-based 
requirements of National Fire Protection Association Standard 805 
(NFPA 805) have been satisfied. The PINGP Updated Safety Analysis 
Report (USAR) documents the analyses of design basis accidents 
(DBAs) at PINGP. The proposed amendment does not adversely affect 
accident initiators nor alter design assumptions, conditions, or 
configurations of the facility that would increase the probability 
or consequences of accidents previously evaluated. Further, the 
changes to be made for fire hazard protection and mitigation do not 
adversely affect the ability of structures, systems, and components 
(SSCs) to perform their design functions, nor do they affect the 
postulated initiators or assumed failure modes for accidents 
described and evaluated in the USAR. SSCs required to safely shut 
down the reactor and to maintain it in a safe shutdown condition 
will remain capable of performing their design functions.
    The purpose of this proposed amendment is to permit PINGP to 
adopt a new fire protection licensing basis which complies with the 
requirements in 10 CFR 50.48(a) and (c) and the guidance in Revision 
1 of Regulatory Guide (RG) 1.205. The NRC considers that NFPA 805 
provides an acceptable methodology and performance criteria for 
licensees to identify fire protection systems and features that are 
an acceptable alternative to the 10 CFR Part 50, Appendix R fire 
protection features (69 FR 33536; June 16, 2004). Engineering 
analyses, in accordance with NFPA 805, have been performed to 
demonstrate that the risk-informed, performance-based (RI-PB) 
requirements per NFPA 805 have been met.
    NFPA 805, taken as a whole, provides an acceptable alternative 
to 10 CFR 50.48(b), satisfies 10 CFR 50.48(a) and General Design 
Criterion (GDC) 3 of Appendix A to 10 CFR Part 50, and meets the 
underlying intent of the NRC's existing fire protection regulations 
and guidance, and provides for defense-in-depth. The goals, 
performance objectives, and performance criteria specified in 
Chapter 1 of NFPA 805 ensure that if there are any increases in the 
net core damage frequency (CDF) or risk associated with this license 
amendment request (LAR) submittal, the increase will be small and 
consistent with the Commission's Safety Goal Policy.
    Based on this, the implementation of this amendment does not 
significantly increase the probability of any accident previously 
evaluated. Equipment required to mitigate an accident remains 
capable of performing the assumed function(s). The proposed 
amendment will not affect the source term, containment isolation, or 
radiological release assumptions used in evaluating the radiological 
consequences of any accident previously evaluated.
    Therefore, the consequences of any accident previously evaluated 
are not significantly increased with the implementation of the 
proposed amendment.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    Operation of PINGP in accordance with the proposed amendment 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated. Any scenario or 
previously analyzed accident with offsite dose was included in the 
evaluation of DBAs documented in the USAR. The proposed change does 
not alter the requirements or function for systems required during 
accident conditions. Implementation of the new fire protection 
licensing basis which complies with the requirements in 10 CFR 
50.48(a) and (c) and the guidance in Revision 1 of RG 1.205 will not 
result in new or different accidents.
    The proposed amendment does not introduce new or different 
accident initiators nor alter design assumptions or conditions of 
the facility. The proposed amendment does not adversely affect the 
ability of SSCs to perform their design function. SSCs required to 
safely shut down the reactor and maintain it in a safe shutdown 
condition remain capable of performing their design functions.
    The purpose of this amendment is to permit PINGP to adopt a new 
fire protection licensing basis which complies with the requirements 
in 10 CFR 50.48(a) and (c) and the guidance in Revision 1 of RG 
1.205. The NRC considers that NFPA 805 provides an acceptable 
methodology and performance criteria for licensees to identify fire 
protection systems and features that are an acceptable alternative 
to the 10 CFR Part 50, Appendix R fire protection features (69 FR 
33536, June 16, 2004). The requirements in NFPA 805 address only 
fire protection and the impacts of fire on the plant that have 
already been evaluated. Based on this, the implementation of this 
amendment does not create the possibility of a new or different kind 
of accident from any kind of accident previously evaluated. The 
proposed amendment does not introduce any new accident scenarios, 
transient precursors, failure mechanisms, malfunctions, or limiting 
single failures that could initiate a new accident. There will be no 
adverse effect or challenges imposed on a safety related system as a 
result of this proposed amendment.
    Therefore, the possibility of a new or different kind of 
accident from any kind of accident previously evaluated is not 
created with the implementation of this amendment.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Operation of PINGP in accordance with the proposed amendment 
does not involve a significant reduction in a margin of safety.

[[Page 19754]]

The proposed amendment does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not affected by this change. The proposed amendment does not 
adversely affect existing plant safety margins or the reliability of 
equipment assumed to mitigate accidents in the USAR. The proposed 
amendment does not adversely affect the ability of SSCs to perform 
their design function. SSCs required to safely shut down the reactor 
and to maintain it in a safe shutdown condition remain capable of 
performing their design function.
    The purpose of this amendment is to permit PINGP to adopt a new 
fire protection licensing basis which complies with the requirements 
in 10 CFR 50.48(a) and (c) and the guidance in Revision 1 of RG 
1.205. The NRC considers that NFPA 805 provides an acceptable 
methodology and performance criteria for licensees to identify fire 
protection systems and features that are an acceptable alternative 
to the 10 CFR Part 50, Appendix R fire protection features (69 FR 
33536; June 16, 2004). Engineering analyses, which may include 
engineering evaluations, probabilistic safety assessments, and fire 
modeling evaluations, have been performed to demonstrate that the 
performance-based methods do not result in a significant reduction 
in a margin of safety.
    Based on this, the implementation of this amendment does not 
significantly reduce a margin of safety. The proposed changes are 
evaluated to ensure that the risk and safety margins are kept within 
acceptable limits.
    Therefore, the transition to NFPA 805 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 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.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, (SSES) Units 1 and 2, Luzerne County, Pennsylvania

    Date of amendment requests: December 19, 2012.
    Description of amendment requests: The proposed amendments would 
modify the SSES Unit 1 and SSES Unit 2 Technical Specifications (TS) 
Section 2.1.1 to reflect a revised Low Pressure Safety Limit. The 
change to TS Section 2.1.1 became necessary as a result of General 
Electric (GE) PART 21 REPORT, SC05-03, ``Potential to Exceed Low 
Pressure Technical Specification Safety Limit.''
    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 changes the low pressure safety limit in 
Technical Specification (TS) 2.1.1 from 785 psig [pounds per square 
inch gauge] to 557 psig based on the capabilities of the current 
critical power correlation used by Susquehanna (SPCB). The SPCB 
correlation is approved for CPR [critical power ratio] calculations 
by the NRC for reactor pressures > 571.4 psia [pounds per square 
inch absolute] and is listed as an approved analytical method in TS 
5.6.5.b.
    The proposed changes will not alter existing Final Safety 
Analysis Report (FSAR) design basis accident analysis assumptions, 
add any accident initiators, or affect the function of the plant 
safety-related structures, systems, or components (SSCs) as to how 
they are operated, maintained, modified, tested, or inspected.
    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 change to the Low Pressure Safety Limits does not result in 
the need for any new or different FSAR design basis accident 
analysis. The inclusion does not introduce new equipment that could 
create a new or different kind of accident, and no new equipment 
failure modes are created. In addition, the proposed change does not 
affect the function of any safety-related SSC as to how they are 
operated, maintained, modified, tested or inspected. As a result, no 
new accident scenarios, failure mechanisms, or limiting single 
failures are introduced as a result of this proposed amendment.
    Therefore, the proposed amendment does not create a possibility 
for an accident of a new or different type than those previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The margin of safety is associated with the confidence in the 
ability of the fission product barriers (i.e., fuel cladding, 
reactor coolant pressure boundary, and containment structure) to 
limit the level of radiation to the public. Evaluation of the 10 CFR 
Part 21, ``Reporting of Defects and Noncompliance'' issue that 
identified the need for the proposed change determined that there 
was no decrease in the safety margin and therefore no threat to fuel 
cladding integrity. The proposed changes to the Low Pressure Safety 
Limits would not alter the way safety-related SSCs function and 
would not alter the way PPL Susquehanna Units 1 and 2 are operated. 
The proposed changes to the safety limit are within the capabilities 
of the existing NRC approved CPR correlation and ensure valid CPR 
calculations for the Anticipated Operational Occurrences (AOOs) 
defined in the FSAR. The proposed amendment would have no impact on 
the structural integrity of the fuel cladding, reactor coolant 
pressure boundary, or containment structure. Based on the above 
considerations, the proposed amendment would not degrade the 
confidence in the ability of the fission product barriers to limit 
the level of radiation to the public.
    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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Branch Chief: Meena K. Khanna.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit 1 and Unit 2 (Salem), Salem County, New Jersey

    Date of amendment request: July 17, 2012, as supplemented on 
January 28, 2013.
    Description of amendment request: The proposed amendments would 
revise Salem Technical Specifications (TS) 3.7.6.1 (Unit 1) and 3.7.6 
(Unit 2), ``Control Room Emergency Air Conditioning System,'' to 
eliminate the separate action statements for securing an inoperable 
Control Area Air Conditioning System and Control Room Emergency Air 
Conditioning System isolation damper in the closed position and 
entering the actions for an inoperable control room envelope boundary.
    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 Control Room Emergency Air Conditioning System (CREACS) is 
not an initiator of or a precursor to any accident or transient. The 
CREACS system is in standby during normal operation and initiates in 
the

[[Page 19755]]

event of a safety injection signal or control room radiation 
monitoring actuation in response to a design basis accident to 
pressurize the Control Room Envelope (CRE) and provide filtration of 
the CRE atmosphere to maintain the control room operator doses 
within the limits of General Design Criteria (GDC) 19. The system 
also operates in recirculation mode to mitigate the consequences of 
a fire or toxic gas release that occurs outside of the CRE.
    The design of plant equipment is not being modified by the 
proposed amendment. The elimination of the action to secure the 
isolation dampers between the normal Control Area Air Conditioning 
System (CAACS) and the CREACS when these dampers are inoperable and 
entering the actions for the inoperable control room boundary will 
ensure operation of the plant within the limits of the radiological, 
smoke and chemical hazard analyses. The intent of the original 
action for securing the inoperable isolation damper in the closed 
position was to maintain the boundary of the CRE. The actions for an 
inoperable control room boundary ensure that mitigating actions are 
implemented that maintain the CRE boundary within the limits of the 
radiological, smoke and chemical hazard analyses.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes to the TS to implement the actions for an 
inoperable control room boundary when a normal CAACS and CREACS 
isolation damper is inoperable do not introduce any new accident 
precursors and do not involve any physical plant alterations or 
changes in the methods governing normal plant operation that could 
initiate a new or different kind of accident. The proposed amendment 
does not alter the function of the system to initiate and pressurize 
the control room envelope in the event of a DBA nor alter the 
ability to initiate CREACS in the recirculation mode in response to 
a fire or chemical release that occurs outside of the CRE.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Margin of safety is related to the ability of the fission 
product barriers (fuel cladding, reactor coolant system, and primary 
containment) to perform their design functions during and following 
postulated accidents. The proposed amendment does not alter 
setpoints or limits established or assumed by the accident analyses. 
The control room envelope is considered a barrier for the control 
room operators during a design basis accident radiological release 
and a barrier in the event of a fire or chemical hazard that occurs 
outside of the CRE. Implementing the actions for an inoperable 
control room boundary in the event of an inoperable isolation damper 
between the normal CAACS and CREACS ensure operation of the plant 
within the limits of the radiological, smoke and chemical hazard 
analysis. The actions for an inoperable control room boundary ensure 
that mitigating actions are implemented that maintain the CRE 
boundary within the limits of the radiological, smoke and chemical 
hazard analyses. Therefore the plant will continue to be operated 
consistent with the plant safety analyses.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, 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: Jeffrie J. Keenan, PSEG Nuclear LLC--N21, 
P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Meena K. Khanna.

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: March 13, 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 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 Part 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 amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.

[[Page 19756]]

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

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

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

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

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

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

[[Page 19757]]

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 NRC'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 https://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.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: November 14, 2012.
    Brief description of amendments: The amendments relocate the 
Technical Specification (TS) requirements for motor-operated valve 
thermal overload protection from the TSs to the Technical Requirements 
Manual.
    Date of issuance: March 19, 2013.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendments Nos.: 209 and 170.
    Facility Operating License Nos. NPF-39 and NPF-85: The amendments 
revised the License and TSs.
    Date of initial notice in Federal Register: January 8, 2013 (78 FR 
1270).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 19, 2013.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry 
Nuclear Power Plant (PNPP), Unit 1, Lake County, Ohio

    Date of application for amendment: July 3, 2012, supplemented by 
letter dated January 7, 2013.
    Brief description of amendment: The proposed amendment would modify 
PNPP's Technical Specifications (TS) 3.8.1, ``AC [alternating current] 
Sources--Operating.'' Specifically, the proposed amendment will modify 
nine surveillance requirements (SRs) by excluding Division 3 from the 
current mode restrictions, thus allowing performance of the subject SRs 
in any mode of plant operation. The proposed amendment also deletes 
expired TS 3.8.1 provisions regarding use of a delayed access circuit.
    Date of issuance: March 5, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 162.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: November 13, 2012 (77 
FR 67682). The January 7, 2013 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 amendment is contained 
in a Safety Evaluation dated March 7, 2013.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: January 18, 2013.
    Brief description of amendment: The proposed amendment would depart 
from VEGP Units 3 and 4 plant-specific Design Control Document (DCD) 
Tier 2 material incorporated into the Updated Final Safety Analysis 
Report (UFSAR) by revising the structural criteria code for anchoring 
of headed shear reinforcement bar within the nuclear island basemat.
    Date of issuance: March 1, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: Unit 3-5, and Unit 4-5.
    Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised 
the Facility Combined Licenses.
    Date of initial notice in Federal Register: January 29, 2013 (78 FR 
6142).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 1, 2013.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 25th day of March 2013.
    For The Nuclear Regulatory Commission.
John D. Monninger,
Deputy Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2013-07467 Filed 4-1-13; 8:45 am]
BILLING CODE 7590-01-P