[Federal Register Volume 78, Number 122 (Tuesday, June 25, 2013)]
[Notices]
[Pages 38078-38087]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2013-14880]


=======================================================================
-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION

[NRC-2013-0134]


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 May 30, 2013 to June 12, 2013. The last 
biweekly notice was published on June 11, 2013 (78 FR 35058).

ADDRESSES: You may submit comment by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0134. Address 
questions about NRC dockets to Carol

[[Page 38079]]

Gallagher; telephone: 301-492-3668; email: [email protected]. For 
technical questions, contact the individual(s) listed in the FOR 
FURTHER INFORMATION CONTACT section of this document.
     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.
    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-2013-0134 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-2013-0134.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may access publicly-available documents online in the NRC 
Library at http://www.nrc.gov/reading-rm/adams.html. To begin the 
search, select ``ADAMS Public Documents'' and then select ``Begin Web-
based ADAMS Search.'' For problems with ADAMS, please contact the NRC's 
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to [email protected]. Documents may be viewed in 
ADAMS by performing a search on the document date and docket number.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2013-0134 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 you do not want to be publicly disclosed in your 
comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment 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 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

[[Page 38080]]

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

[[Page 38081]]

11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking 
and Adjudications Staff. Participants filing a document in this manner 
are responsible for serving the document on all other participants. 
Filing is considered complete by first-class mail as of the time of 
deposit in the mail, or by courier, express mail, or expedited delivery 
service upon depositing the document with the provider of the service. 
A presiding officer, having granted an exemption request from using E-
Filing, may require a participant or party to use E-Filing if the 
presiding officer subsequently determines that the reason for granting 
the exemption from use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, a request to intervene will require including information on 
local residence in order to demonstrate a proximity assertion of 
interest in the proceeding. With respect to copyrighted works, except 
for limited excerpts that serve the purpose of the adjudicatory filings 
and would constitute a Fair Use application, participants are requested 
not to include copyrighted materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day deadline will not 
be entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the following three 
factors in 10 CFR 2.309(c)(1): (i) The information upon which the 
filing is based was not previously available; (ii) the information upon 
which the filing is based is materially different from information 
previously available; and (iii) the filing has been submitted in a 
timely fashion based on the availability of the subsequent information.
    For further details with respect to this license amendment 
application, see the application for amendment which is available for 
public inspection at the NRC's PDR, located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. Publicly available documents created or received at the NRC are 
accessible electronically through ADAMS in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to 
ADAMS or who encounter problems in accessing the documents located in 
ADAMS, should contact the NRC's PDR Reference staff at 1-800-397-4209, 
301-415-4737, or by email to [email protected].

Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power 
Station, Unit 3 (MPS-3), New London County, Connecticut

    Date of amendment request: April 25, 2013.
    Description of amendment request: The amendments would revise the 
peak calculated containment internal pressure (Pa) for the 
design basis loss of coolant accident (LOCA) described in Technical 
Specification (TS) 6.8.4.f, ``Containment Leakage Rate Testing 
Program'' for MPS-3.
    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 to Pa does not alter the assumed 
initiators to any analyzed event. The probability of an accident 
previously evaluated will not be significantly increased by this 
proposed change.
    The change in Pa will not affect radiological dose 
consequence analyses. MPS-3 radiological dose consequence analyses 
assume a certain containment atmosphere leak rate based on the 
maximum allowable containment leakage rate, which is not affected by 
the change in peak calculated containment internal pressure. The 
Appendix J containment leakage rate testing program will continue to 
ensure that containment leakage remains within the leakage assumed 
in the offsite dose consequence analyses. The consequences of an 
accident previously evaluated will not be significantly increased by 
this proposed change.
    Therefore, operation of the facility in accordance with the 
proposed change to Pa will 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 provides a higher Pa than 
currently described in TS 6.8.4.f. This change is a result of an 
increase in the M&E [mass and energy] release input for the LOCA 
containment response analysis. The [Pa] remains below the 
containment design pressure of 45 psig [pounds per square inch 
gauge]. This change does not involve any alteration in the plant 
configuration (no new or different type of equipment will be 
installed) or make changes in the methods governing normal plant 
operation. The change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Therefore, operation of the facility in accordance with the 
proposed change to TS 6.8.4.f would not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The [Pa] remains below the containment design 
pressure of 45 psig. Since the MPS3 radiological consequence 
analyses are based on the maximum allowable containment leakage 
rate, which is not being revised, the change in the [Pa] 
does not represent a significant change 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: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 
23219.
    Acting NRC Branch Chief: Robert Beall.

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: April 16, 2013.
    Description of amendment request: The proposed amendments would 
remove superseded Technical Specification (TS) requirements McGuire 
Nuclear Station (MNS), Units 1 and 2. By letter dated May 28, 2010, 
Duke Energy submitted a license amendment request (LAR) to modify TS to 
allow the manual operation of the Containment Spray System in lieu of 
automatic actuation, and revise the minimum volume and low level 
setpoint on the Refueling Water Storage Tank. Because the associated 
modifications were implemented on a staggered basis for each MNS Unit 
during refueling outages, the deletion or modification of these TS 
requirements was accomplished via the use of temporary footnotes. This 
allowed the

[[Page 38082]]

requirements to be either applicable or non-applicable, depending upon 
whether the modifications had not been implemented or implemented, 
respectively. The LAR contained a commitment for MNS to submit a 
follow-up administrative license amendment request to delete the 
superseded temporary TS requirements within 180 days of the 
installation of the associated modifications for the final MNS Unit. By 
letter dated September 12, 2011, the NRC issued amendments regarding 
the TS changes requested in the May 28, 2010 LAR. Installation of the 
associated modifications on the final MNS Unit was completed on October 
18, 2012. This LAR satisfies the MNS commitment to delete the 
superseded temporary TS requirements described in the May 28, 2010 LAR. 
In addition, this LAR makes an administrative non-technical editorial 
correction by relocating NOTE 1 on TS page 3.3.2-15 to TS page 3.3.2-
14. Relocating NOTE 1 back to TS page 3.3.2-14 is consistent with the 
reference to this NOTE in TS Table 3.3.2-1, Engineered Safety Feature 
Actuation System (ESFAS) Instrumentation, Function 9, Containment 
Pressure Control 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:

    Criterion 1:
    Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This LAR proposes administrative non-technical changes only. 
These proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, or 
configurations of the facility. The proposed changes do not alter or 
prevent the ability of structures, systems and components (SSCs) to 
perform their intended function to mitigate the consequences of an 
initiating event within the assumed acceptance limits.
    Given the above discussion, it is concluded the proposed 
amendment does not significantly increase the probability or 
consequences of an accident previously evaluated.
    Criterion 2:
    Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This LAR proposes administrative non-technical changes only. The 
proposed changes will not alter the design requirements of any SSC 
or its function during accident conditions. No new or different 
accidents result from the changes proposed. The changes do not 
involve a physical alteration of the plant or any changes in methods 
governing normal plant operation. The changes do not alter 
assumptions made in the safety analysis.
    Given the above discussion, it is concluded the proposed 
amendment does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    Criterion 3:
    Does the proposed amendment involve a significant reduction in 
the margin of safety?
    Response: No.
    This LAR proposes administrative non-technical changes only. The 
proposed changes do 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 these changes. The proposed changes will not result in 
plant operation in a configuration outside the design basis. The 
proposed changes do not adversely affect systems that respond to 
safely shutdown the plant and to maintain the plant in a safe 
shutdown condition.
    Given the above discussion, it is concluded 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: Lara S. Nichols, Associate General Counsel, 
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC 
28202.
    NRC Branch Chief: Robert J. Pascarelli.

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

    Date of amendment request: April 15, 2013.
    Description of amendment request: The proposed change would revise 
Technical Specification 3.5.4, ``Refueling Water Storage Tank (RWST)'' 
such that the non-seismically qualified piping of the temporary Boric 
Acid Recovery System (BARS) may be connected to the seismic piping of 
the RWST. Operation of the BARS from the RWST will be under 
administrative controls for a limited period of time (i.e., 30 days for 
RWST filtration prior to each fuel cycle). This change will only be 
applicable until Refueling Outage R22 ends (Spring 2016).
    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 use of the non seismic Boric Acid Recovery System (BARS) to 
recirculate and filter the Refueling Water Storage Tank (RWST) water 
does not involve any changes or create any new interfaces with the 
reactor coolant system or main steam system piping. Therefore, the 
connection of the BARS Purification Loop to the RWST would not 
affect the probability of these accidents occurring. The BARS is not 
credited for safe shutdown of the plant or accident mitigation. 
Administrative controls ensure that the BARS can be isolated as 
necessary and in sufficient time to assure that the RWST volume will 
be adequate to perform the safety function as designed. Since the 
RWST will continue to perform its safety function and overall system 
performance is not affected, the consequences of the accident are 
not increased.
    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 design of the RWST and the SFP [spent fuel pool] 
Purification Loop has been revised to allow recirculation and 
purification using the BARS for a short period of time (not to 
exceed 30 days per fuel cycle) for the next two fuel cycles. The 
added BARS takes RWST water in and processes it out without 
additional connections that could affect other systems and without 
an impact from its installation. Procedures for the operation of the 
plant, including BARs, will not create the possibility of a new or 
different type of accident. Contingent upon manual operator action, 
a BARS line break will not result in a loss of the RWST safety 
function. Similarly, an active or passive failure in the BARS will 
not affect safety related structures, systems or components.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The SFP Purification Loop and recirculation and purification of 
the RWST water using the BARS is not credited for safe shutdown of 
the plant or accident mitigation. RWST volume will be maximized 
prior to purification and timely operator action can be taken to 
isolate the non seismic system from the RWST to assure it can 
perform its function. This will result in no significant reduction 
in the margin of safety.
    Therefore the proposed change does not significantly reduce the 
margin of safety.

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

[[Page 38083]]

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.
    Acting NRC Branch Chief: Sean Meighan.

National Institute of Standards and Technology (NIST), Docket No. 50-
184, Center for Neutron Research (NBSR), Montgomery County, Maryland

    Date of amendment request: July 12, 2012, as supplemented on May 
14, 2013.
    Description of amendment request: The proposed amendments would 
revise NIST NBSR's Technical specifications, Sections 3.7, 4.7, and 
6.8, pertaining to the environmental monitoring requirements and 
records retention which clarifies environmental sampling procedure and 
record retention processes.
    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 corrects a deficiency in the license 
issued in 2009 that created a disagreement in the periodicity of 
environmental sampling within the license Technical Specifications. 
Additionally, the proposed amendment aligns the record retention 
requirement (section 6.8) of the license technical specifications 
with the consensus standard ANSI/ANS 15.1. This standard has been 
endorsed by the NRC under Regulatory Guide 2.2. Neither of these 
proposed changes will have any influence or impact on reactor 
operations or previously analyzed accidents. There are no physical 
changes to the facility as a result of these administrative changes.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    No accident of any kind would be created by the proposed 
administrative changes. The sample periodicity will not change from 
the sampling periodicity used by the facility for over 40 years. 
Records are maintained and summarized in facility annual reports and 
there would be no loss of information. There are no physical changes 
to the facility as a result of these administrative changes.
    Therefore, the changes would not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Plant safety margins are established through limiting conditions 
of operation, limiting safety system settings, and safety limits 
specified in the Technical Specifications. The proposed changes do 
not alter any of the established safety margins and are 
administrative in nature.
    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: Melissa J. Lieberman, Deputy Chief Counsel 
for NIST, National Institute of Standard and Technology, 100 Bureau 
Drive, Gaithersburg, MD 20899.
    NRC Branch Chief: Alexander Adams, Jr.

South Carolina Electric and Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 1, 
Fairfield County, South Carolina

    Date of amendment request: April 2, 2013, as supplemented by a 
letter dated May 16, 2013.
    Description of amendment request: The proposed amendments would 
revise the technical specification requirements regarding steam 
generator tube inspection and reporting as described in Technical 
Specification Task Force (TSTF)-510, ``Revision to Steam Generator 
Program Inspection Frequencies and Tube Sample Selection,'' Revision 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. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the Steam Generator (SG) Program to 
modify the frequency of verification of SG tube integrity and SG 
tube sample selection. A steam generator tube rupture (SGTR) event 
is one of the design basis accidents that are analyzed as part of a 
plant's licensing basis. The proposed SG tube inspection frequency 
and sample selection criteria will continue to ensure that the SG 
tubes are inspected such that the probability of a SGTR is not 
increased. The consequences of a SGTR are bounded by the 
conservative assumptions in the design basis accident analysis. The 
proposed change will not cause the consequences of a SGTR to exceed 
those assumptions.
    Therefore, it is concluded that this 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 changes to the Steam Generator Program will not 
introduce any adverse changes to the plant design basis or 
postulated accidents resulting from potential tube degradation. The 
proposed change does not affect the design of the SGs or their 
method of operation. In addition, the proposed change does not 
impact any other plant system or component.
    Therefore, it is concluded that this 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 SG tubes in pressurized water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
also isolate the radioactive fission products in the primary coolant 
from the secondary system. In summary, the safety function of a SG 
is maintained by ensuring the integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical condition of the tube. The proposed 
change does not affect tube design or operating environment. The 
proposed change will continue to require monitoring of the physical 
condition of the SG tubes such that there will not be a reduction in 
the margin of safety compared to the current requirements.
    Therefore, it is concluded that 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: J. Hagood Hamilton, Jr., South Carolina 
Electric &

[[Page 38084]]

Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Branch Chief: Robert J. Pascarelli.

South Carolina Electric and Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 1, 
Fairfield County, South Carolina

    Date of amendment request: April 3, 2013.
    Description of amendment request: The proposed amendment would 
allow for the extension of the frequency of the containment leak rate 
test per Technical Specification 6.8.4(g) from 130-months (10.9 years) 
to 15 years.
    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 exemption involves a permanent 15-year extension to 
the current interval for Type A containment testing. The current 
test interval of 130 months (10.9 years) would be extended to a 
permanent 15-year frequency from the last Type A test. The proposed 
extension does not involve a physical change to the plant or a 
change in the manner in which the plant is operated or controlled. 
The containment is designed to provide an essentially leak tight 
barrier against the uncontrolled release of radioactivity to the 
environment for postulated accidents. As such, the reactor 
containment itself and the testing requirements invoked to 
periodically demonstrate the integrity of the reactor containment 
exist to ensure the plant's ability to mitigate the consequences of 
an accident, and do not involve the prevention or identification of 
any precursors of an accident. Therefore, this proposed extension 
does not involve a significant increase in the probability of an 
accident previously evaluated nor does it create the possibility of 
a new or different kind of accident.
    The integrity of the reactor containment is subject to two types 
of failure mechanisms which can be categorized as (1) activity based 
and (2) time based. Activity based failure mechanisms are defined as 
degradation due to system and/or component modifications or 
maintenance. Local leak rate test requirements and administrative 
controls such as configuration management and procedural 
requirements for system restoration ensure that containment 
integrity is not degraded by plant modifications or maintenance 
activities. The design and construction requirements of the 
containment itself combined with the containment inspections 
performed in accordance with ASME, Section XI, the Maintenance Rule, 
and Licensing commitments serve to provide a high degree of 
assurance that the containment will not degrade in a manner that is 
detectable only by a Type A test.
    Based on the above, the proposed extension does not involve a 
significant increase in the 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 revision to the TS involves a 15-year permanent 
extension to the current interval for Type A containment testing. 
The reactor containment and the testing requirements invoked to 
periodically demonstrate the integrity of the reactor containment 
exist to ensure the plant's ability to mitigate the consequences of 
an accident and do not involve the prevention or identification of 
any precursors of an accident. The proposed TS change does not 
involve a physical change to the plant or the manner in which the 
plant is operated or controlled.
    Therefore, the proposed TS 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 to the TS involves a 15-year permanent 
extension to the current interval for Type A containment testing. 
The proposed TS change does not involve a physical change to the 
plant or a change in the manner in which the plant is operated or 
controlled. The specific requirements and conditions of the Primary 
Containment Leak Rate Testing Program, as defined in the TS, exist 
to ensure that the degree of reactor containment structural 
integrity and leak-tightness that is considered in the plant safety 
analysis is maintained. The overall containment leak rate limit 
specified by TS is maintained. The proposed change involves only the 
extension of the interval between Type A containment leak rate 
tests. The proposed surveillance interval extension is bounded by 
the 15-year permanent extension currently authorized within NEI 94-
01, Revision 3-A. Type B and C containment leak rate tests will 
continue to be performed at the frequency currently required by TS. 
Industry experience supports the conclusion that Type B and C 
testing detects a large percentage of containment leakage paths and 
that the percentage of containment leakage paths that are detected 
only by Type A testing is small. The containment inspections 
performed in accordance with ASME, Section Xl and the Maintenance 
Rule serve to provide a high degree of assurance that the 
containment will not degrade in a manner that is detectable only by 
Type A testing.
    The combination of these factors ensures that the margin of 
safety that is in plant safety analysis is maintained. The design, 
operation, testing methods and acceptance criteria for Type A, B, 
and C containment leakage tests specified in applicable codes and 
standards will continue to be met, with the acceptance of this 
proposed change, since these are not affected by changes to the Type 
A test interval. Therefore, the proposed TS 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: J. Hagood Hamilton, Jr., South Carolina 
Electric & Gas Company, Post Office Box 764, Columbia, South Carolina 
29218.
    NRC Branch Chief: Robert J. Pascarelli.

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: May 10, 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 VEGP Units 3 
and 4 Updated Final Safety Analysis Report (UFSAR) Tier 2* material by 
revising reference document APP-OCS-GEH-520, ``AP1000 Plant Startup 
Human Factors Engineering Design Verification Plan,'' from Revision B 
to Revision 1. APP-OCS-GEH-520 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 APP-OCS-GEH-520, document confirms aspects of the human 
system interface (HSI) and Operation and Control Centers Systems 
(OCS) design features that could not be evaluated in other Human 
Factors Engineering (HFE) verification and validation (V&V) 
activities. It also confirms that the as-built in the plant HSIs, 
procedures, and training conform to the design that resulted from 
the HFE program. Additionally, it confirms that all HFE-related 
issues (including human error discrepancies (HEDs)) documented in 
the SmartPlant Foundation (SPF) Human Factors (HF)

[[Page 38085]]

Tracking System are verified as adequately addressed or resolved. 
Finally, it confirms the HFE adequacy for risk-important human 
actions in the local plant, including the ability for the tasks to 
be completed within the time window according to the Probabilistic 
Risk Assessment (PRA). The changes to the plan are to clarify the 
scope and amend the details of the methodology. The plan does not 
affect the plant itself. Changing the 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. The PRA is not 
affected. No safety-related Structure, System, or Component (SSC) or 
function is adversely affected. The document revision change does 
not involve nor interface with any SSC accident initiator or 
initiating sequence of events, and thus, the probabilities of the 
accidents evaluated in the Updated Final Safety Analysis Report 
(UFSAR) are not affected. Because the changes to the plan 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.
    APP-OCS-GEH-520, ``AP1000 Plant Startup Human Factors 
Engineering Design Verification Plan'' is the plan to confirm 
aspects of the HSI and OCS design features that could not be 
evaluated in other HFE V&V activities. The plan also confirms that 
the as-built in the plant HSIs, procedures, and training conform to 
the design that resulted from the HFE program. Additionally, it 
confirms that all HFE-related issues (including HEDs) documented in 
the SPF HF Tracking System are verified as adequately addressed or 
resolved. Finally, it confirms the HFE adequacy for risk-important 
human actions in the local plant, including the ability for the 
tasks to be completed within the time window according to the PRA. 
These functions support evaluating the HSI and OCS. 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.
    APP-OCS-GEH-520, ``AP1000 Plant Startup Human Factors 
Engineering Design Verification Plan'' is the plan to confirm 
aspects of the HSI and OCS design features that could not be 
evaluated in other HFE V&V activities. The plan also confirms that 
the as-built in the plant HSIs, procedures, and training conform to 
the design that resulted from the HFE program. Additionally, it 
confirms that all HFE-related issues (including HEDs) documented in 
the SPF HF Tracking System are verified as adequately addressed or 
resolved. Finally, it confirms the HFE adequacy for risk-important 
human actions in the local plant, including the ability for the 
tasks to be completed within the time windows in the PRA. These 
functions support evaluating the HSI and OCS. The proposed changes 
to the plan do not affect the design or operation of safety-related 
equipment or equipment whose failure could initiate an accident, nor 
does the plan adversely affect the interfaces with safety-related 
equipment or fission product barriers. No safety analysis or design 
basis acceptance limit/criterion is challenged or exceeded by the 
requested changes.
    Therefore, the 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, Blach & Bingham 
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    Acting NRC Branch Chief: Lawrence Burkhart.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: April 25, 2013.
    Description of amendment request: The amendments would revise 
Technical Specification (TS) 5.1, ``Site,'' Figures 5.1-1 through 5.1-4 
for South Texas Project (STP), Units 1 and 2, to remove identification 
of a Visitor's Center building, which has been demolished. The 
amendments also would revise Figures 5.1-1, 5.1-3, and 5.1-4 to remove 
references to the Emergency Operations Facility (EOF) within the 
Nuclear Training Facility, since the EOF was relocated to Center of 
Energy Development building located in Bay City, Texas, approximately 
12.5 air miles from the plant site in 2009. The EOF was relocated 
offsite with an emergency plan change made by the licensee under 10 CFR 
50.54(q), ``Emergency plans,'' by concluding that the change did not 
represent a decrease in effectiveness of the emergency plan. The 
amendments to remove references to the Visitor's Center Building and 
EOF from the TSs are administrative in nature.
    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 is an administrative change to STP TS design 
features to remove reference to the Visitor's Center and onsite EOF. 
The design function of structures, systems and components (SSC) 
important to safety are not impacted by the proposed change. The 
proposed change will not initiate an event. The proposed change does 
not alter or prevent the ability of SSCs from performing their 
intended function to mitigate the consequences of an initiating 
event.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change is an administrative change to STP TS design 
features to remove reference to the Visitor's Center and onsite EOF. 
The proposed change does not impact create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. There are no new failure modes or mechanisms associated 
with the proposed change. This change does not involve any 
modification in operational limits or physical design of equipment 
important to safety.
    Therefore, the proposed change does not involve a different kind 
of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change is an administrative change to STP TS design 
features to remove reference to the Visitor's Center and onsite EOF. 
The proposed change does not impact TS safety limits, TS limiting 
safety system set points, or the results of any of the 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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that

[[Page 38086]]

the request for amendments involves no significant hazards 
consideration.
    Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue NW., Washington, DC 20004.
    NRC Branch Chief: Michael T. Markley.

Previously Published Notices 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 following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

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

    Date of amendment request: May 22, 2013.
    Brief description of amendment request: The proposed amendment 
would revise the WBN Unit 1 Technical Specifications (TSs) to allow a 
one-time extension to the Completion Time for TS Limiting Condition for 
Operation (LCO) 3.6.6 Required Action A.1 from 72 hours to 7 days for 
an inoperable Containment Spray (CS) Train B. This change is necessary 
to provide sufficient time to replace a leaking mechanical seal on CS 
Pump 1B-B. The pump repair is currently scheduled for the week of June 
24, 2013. TVA requested this TS change under exigent circumstances and 
that the NRC expedites the review to support approval by June 22, 2013.
    Date of publication of individual notice in Federal Register: June 
3, 2013 (78 FR 33117).
    Expiration date of individual notice: June 17, 2013 (public 
comments); August 2, 2013 (hearing requests).

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 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 http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
PDR's Reference staff at 1-800-397-4209, 301-415-4737 or by email to 
[email protected].

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

    Date of application for amendment: September 6, 2012, as 
supplemented by letter dated December 7, 2012.
    Brief Description of amendment: The amendment revised the Technical 
Specifications (TSs) to eliminate Function 14, Steam Generator Water 
Level-Low Coincident with Steam Flow/Feedwater Flow Mistmatch, from the 
HBRSEP TS Table 3.3.1-1, ``Reactor Protection System Instrumentation.''
    Date of issuance: May 29, 2013.
    Effective date: As of date of issuance and shall be implemented 
prior exiting the scheduled fall 2013 refueling outage.
    Amendment No.: 234.
    Renewed Facility Operating License No. DPR-23: Amendment changed 
the license and TSs.
    Date of initial notice in Federal Register: November 27, 2012 (77 
FR 70840). The supplement dated December 7, 2012, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 29, 2013.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 21, 2010.
    Description of amendment request: The proposed amendment revised 
the Technical Specification (TS) 3/4.9.3.1, ``Decay Time'' for 
Millstone Power Station, Unit 2 (MPS2). The proposed change revises 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. The 
licensee requested a reduction in the minimum decay time requirement 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.
    Date of issuance: June 4, 2013.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 315.
    Renewed Facility Operating License No. DPR-65: Amendment revised 
the License and Technical Specifications.
    Date of initial notice in Federal Register: April 2, 2013 (78 FR 
19749). The supplemental letter dated July 19, 2011, 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

[[Page 38087]]

significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 4, 2013.
    No significant hazards consideration comments received: No.

South Carolina Electric and Gas. Docket Nos. 52-027 and 52-028, Virgil 
C. Summer Nuclear Station (VCSNS), Units 3 and 4, Fairfield County, 
South Carolina

    Date of amendment request: February 14, 2013.
    Brief description of amendment: The amendment authorizes a 
departure from the Virgil C. Summer Nuclear Station, Units 2 and 3 
plant-specific Design Control Document (DCD) material incorporated into 
the Updated Final Safety Analysis Report (UFSAR) to revise Figure 
3.8.8-1, Sheet 1, Note 2.
    Date of issuance: May 23, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: Unit 2-3, and Unit 3-3.
    Facility Combined Licenses No. NPF-93 and NPF-94: Amendment revised 
the Facility Combined Licenses.
    Date of initial notice in Federal Register: March 4, 2013 (78 FR 
14126).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 23, 2013.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 14th day of June 2013.

    For The Nuclear Regulatory Commission.
John D. Monninger,
Deputy Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2013-14880 Filed 6-24-13; 8:45 am]
BILLING CODE 7590-01-P