Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 54280-54294 [2013-21247]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 78, Number 170 (Tuesday, September 3, 2013)]
[Notices]
[Pages 54280-54294]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2013-21247]


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NUCLEAR REGULATORY COMMISSION

[NRC-2013-0201]


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 August 9, 2013, to August 21, 2013. The last 
biweekly notice was published on August 20, 2013 (78 FR 51219).

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 https://www.regulations.gov and search for Docket ID NRC-2013-0201. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: Carol.Gallagher@nrc.gov. 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: 3WFN, 06-44M, 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: 

[[Page 54281]]

I. Accessing Information and Submitting Comments

A. Accessing Information

    Please refer to Docket ID NRC-2013-0201 when contacting the NRC 
about the availability of information regarding this document. You may 
access publicly-available information related to this action by the 
following methods:
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2013-0201.
     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-2013-0201 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 https://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 Sec.  50.92 of Title 10 of the Code of 
Federal Regulations (10 CFR), this means that operation of the facility 
in accordance with the proposed amendment would not (1) involve a 
significant increase in the probability or consequences of an accident 
previously evaluated; or (2) create the possibility of a new or 
different kind of accident from any accident previously evaluated; or 
(3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license or 
combined license. Requests for a hearing and a petition for leave to 
intervene shall be filed in accordance with the Commission's ``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 basis 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

[[Page 54282]]

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 hearing.docket@nrc.gov, or by 
telephone at 301-415-1677, to request (1) a digital information (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign 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's 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's 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 to 
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

[[Page 54283]]

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 three factors in 10 
CFR 2.309(c)(1)(i)-(iii).
    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.

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

    Date of amendment request: October 15, 2012, and August 12, 2013.
    Description of amendment request: The proposed amendments would 
remove License Conditions which are no longer necessary to address an 
interim configuration of the LaSalle County Station (LSCS), Unit 2, 
spent fuel pool prior to completing installation of NETCO-SNAP-
IN[supreg] inserts. By letter dated August 12, 2013, EGC provided 
additional information and expanded the scope of the application as 
originally noticed. The August 12, 2013, letter proposed to clarify 
language in the LSCS, Units 1 and 2, Technical Specifications (TS) 
applicable to the design features for TS 4.3, `Fuel Storage.' The 
proposed amendment was initially published in the Federal Register 
Biweekly notice on April 2, 2013 (78 FR 19751).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided on 
August 12, 2013, its revised 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. The proposed change 
also revises TS Section 4.3.1 to clarify that for the Unit 2 SFP, 
spent fuel shall only be stored in storage rack cells containing a 
neutron absorbing rack insert. 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, and the revision to TS Section 4.3.1 
ensures spent fuel is stored only in cells that contain inserts. 
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. The proposed change 
also revises TS Section 4.3.1 to clarify that for the Unit 2 SFP, 
spent fuel shall only be stored in storage rack cells containing a 
neutron absorbing rack insert. 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, and the revision to TS Section 4.3.1 ensures spent fuel 
is stored only in cells that contain inserts.
    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. The proposed change 
also revises TS Section 4.3.1 to clarify that for the Unit 2 SFP, 
spent fuel shall only be stored in storage rack cells containing a 
neutron absorbing rack insert. 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, and the revision to TS Section 4.3.1 ensures 
spent fuel is stored only in cells that contain inserts.
    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.
    Acting NRC Branch Chief: Jeremy S. Bowen.

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of amendment request: June 10, 2013.
    Description of amendment request: The proposed amendment revises 
Technical Specification (TS) Surveillance Requirements (SR) 3.8.4.2 and 
3.8.4.5. The proposed change would resolve a non-cited violation (NCV) 
that was documented in an NRC's Inspection Report. Specifically, the 
NRC identified an NCV for the failure to verify that safety-related 
batteries would remain operable if all the inter-cell and terminal 
connections were at the maximum resistance value allowed by SR 3.8.4.2 
and SR 3.8.4.5 (i.e., 150 micro-ohms).

[[Page 54284]]

The proposed change maintains the existing resistance limit for inter-
cell and terminal connections, and adds new acceptance criteria for 
total battery connection resistance to ensure that the safety-related 
batteries can perform their specified safety function.
    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 revisions of SR 3.8.4.2 and SR 3.8.4.5 to add a battery 
connector resistance acceptance criterion will not challenge the 
ability of the safety-related batteries to perform their safety 
function. The total battery connection resistance is a parameter 
that is representative of overall battery performance, and ensures 
that the safety-related batteries remain capable of performing their 
specified safety function. Appropriate monitoring and maintenance 
will continue to be performed on the safety-related batteries. In 
addition, the safety-related batteries are within the scope of 10 
CFR 50.65, ``Requirements for monitoring the effectiveness of 
maintenance at nuclear power plants,'' which will ensure the control 
of maintenance activities associated with this equipment.
    Current TS requirements will not be altered and will continue to 
require that the equipment be regularly monitored and tested. Since 
the proposed change does not alter the manner in which the batteries 
are operated, there is no significant impact on reactor operation.
    The proposed change does not involve a physical change to the 
batteries, nor does it change the safety function of the batteries. 
The DC power system/batteries will retain adequate independency, 
redundancy, capacity, and testability to permit the functioning 
required of the engineered safety features. The proposed TS revision 
involves no significant changes to the operation of any systems or 
components in normal or accident operating conditions and no changes 
to existing structures, systems, or components.
    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 changes revising SR 3.8.4.2 and SR 3.8.4.5 to add 
an additional acceptance criterion for battery connector resistance 
is an increase in conservatism, without a change in system testing 
methods, operation, or control. Safety-related batteries installed 
in the plant will be required to meet criteria more restrictive and 
conservative than current acceptance criteria and standards. The 
proposed change does not affect the manner in which the batteries 
are tested and maintained; therefore, there are no new failure 
mechanisms for the system.
    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 margin of safety is established through the design of the 
plant structures, systems, and components, the parameters within 
which the plant is operated, and the setpoints for the actuation of 
equipment relied upon to respond to an event. The proposed change 
does not modify the safety limits or setpoints at which protective 
actions are initiated. The new acceptance criterion is more 
restrictive than the existing acceptance criteria for inter-cell and 
terminal connection resistance, and the proposed change ensures the 
availability and operability of safety-related battery operability 
and availability.
    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: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    Acting NRC Branch Chief: Jeremy Bowen.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit 2, Oswego County, New York

    Date of amendment request: July 5, 2013.
    Description of amendment request: The proposed amendment includes 
supporting changes to NMP2 Technical Specification (TS) 3.1.7, 
``Standby Liquid Control (SLC) System,'' to increase the isotopic 
enrichment of boron-10 in the sodium pentaborate solution utilized in 
the SLC System and decrease the SLC System tank volume. The following 
are the proposed changes to the NMP2 TS 3.1.7, ``Standby Liquid Control 
(SLC) System'':
     Revise the acceptance criterion in SR 3.1.7.10 by 
increasing the sodium pentaborate boron-10 enrichment requirement from 
>= 25 atom percent to >= 92 atom percent, and make a corresponding 
change in TS Figure 3.1.7-1, ``Sodium Pentaborate Solution Volume/
Concentration Requirements.''
     Revise TS Figure 3.1.7-1 to account for the decrease in 
the minimum volume of the SLC system tank. At a sodium pentaborate 
concentration of 13.6% the minimum volume changes from 4,558.6 gallons 
to 1,600 gallons. At a sodium pentaborate concentration of 14.4%, the 
minimum volume changes from 4,288 gallons to 1,530 gallons.
    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. Will the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The SLC System is used to mitigate the consequences of an 
Anticipated Transient Without SCRAM (ATWS) special event and is used 
to limit the radiological dose during a Loss of Coolant Accident 
(LOCA). The proposed changes do not affect the capability of the SLC 
System to perform these two functions in accordance with the 
assumptions of the associated analyses.
    A SLC System failure is not a precursor of any previously 
evaluated accident in the NMP2 Updated Safety Analysis Report 
(USAR). Consequently there is no change in the probability of an 
accident previously evaluated.
    The current ATWS analysis is not adversely affected by the 
proposed changes because the reactivity insertion rate would 
increase by a factor greater than 3 and the amount of injected 
boron-10 is not reduced. The ability of the SLC System to mitigate 
radiological dose in the event of a LOCA is not affected by these 
changes.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Will the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    Structures, systems and components (SSCs) previously required 
for the mitigation of a transient remain capable of fulfilling their 
intended design functions. The proposed changes do not adversely 
affect safety-related SSCs and do not challenge the performance or 
integrity of any safety-related SSC. The physical changes to the SLC 
System are limited to the increase in the boron-10 enrichment of the 
sodium pentaborate solution in the SLC System storage tank, the 
corresponding decrease in the net sodium pentaborate solution volume 
requirement in the SLC System storage tank, and the associated 
instrumentation changes. In addition, the effective SLC System flow 
rate utilized in the boron equivalency analysis is reduced. The 
proposed changes do not otherwise affect the design or operation of 
the SLC System.
    This change does not adversely affect any current system 
interfaces or create any new interfaces that could result in an 
accident or

[[Page 54285]]

malfunction of a different kind than was previously evaluated.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Will the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The SLC System is used to mitigate the consequences of an ATWS 
event and is used to limit the radiological dose during a LOCA. The 
proposed changes do not affect the capability of the SLC System to 
perform these two functions in accordance with the assumptions of 
the associated analyses. The current ATWS analysis is not adversely 
affected by the proposed changes because the reactivity insertion 
rate would increase by a factor greater than 3 and the amount of 
injected boron-10 is not reduced. The ability of the SLC System to 
mitigate radiological dose in the event of a LOCA by maintaining 
suppression pool pH >= 7.0 is not affected by these changes.
    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: Gautam Sen, Senior Counsel, Constellation 
Energy Nuclear Group, LLC, 100 Constellation Way, Suite 200C, 
Baltimore, MD 21202.
    Acting NRC Branch Chief: Robert Beall.

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

    Date of amendment request: April 19, 2013.
    Description of amendment request: The licensee proposed to revise 
MNGP Technical Specification (TS) 1.1, ``Definitions,'' to modify the 
definition of ``Shutdown Margin (SDM)'' to require calculation of the 
SDM at a reactor moderator temperature of 68 degrees Fahrenheit 
([deg]F), or at a higher temperature that represents the most reactive 
state throughout the operating cycle. This change is needed for newer 
boiling water reactor fuel designs which may be more reactive at 
shutdown temperatures above 68[emsp14][deg]F. The proposed change is 
consistent with Technical Specifications Task Force (TSTF) Traveler 
TSTF-535, Revision 0, ``Revise Shutdown Margin Definition to Address 
Advanced Fuel Designs.'' Notice of availability of TSTF-535 was 
published in the Federal Register on February 26, 2013 (78 FR 13100).
    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 (NSHC), 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 definition of SDM. SDM is not an 
initiator to any accident previously evaluated. Accordingly, the 
proposed change to the definition of ADM has no effect on the 
probability of any accident previously evaluated. ADM is an 
assumption in the analysis of some previously evaluated accidents 
and inadequate SDM could lead to an increase in consequences for 
those accidents. However, the proposed change revised the SDM 
definition to ensure that the correct SDM is determined for all fuel 
types at all times during the fuel cycle. As a result, the proposed 
change does not adversely affect the consequences of any accident 
previously evaluated.
    Therefore, it is concluded that these changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises the definition of SDM. The change 
does not involve a physical alteration of the plant (i.e., no new of 
different type of equipment will be installed) or a change in 
methods governing normal plant operations. The change does not alter 
assumptions made in the safety analysis regarding SDM.
    Therefore, it is concluded that these changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revised the definition of SDM. The proposed 
change does not alter the manner in which safety limits, limiting 
safety system settings or limiting conditions for operation are 
determined. The proposed change ensures that the SDM assumed in 
determining safety limits, limiting safety system settings or 
limiting conditions for operation is correct for all BWR fuel types 
at all times during the fuel cycle.
    Therefore, it is concluded that these changes do not involve a 
significant reduction in a margin of safety.

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

Pacific Gas and Electric Co., Docket No. 50-133, Humboldt Bay Power 
Plant (HBPP), Unit 3 Humboldt County, California

    Date of amendment request: May 3, 2013.
    Description of amendment request: The proposed amendment would add 
License Condition 2.C.5 that approves the License Termination Plan 
(LTP) and adds a license condition that establishes the criteria for 
determining when changes to the LTP require prior NRC approval.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The change allows for the approval of the LTP and provides the 
criteria for when changes to the LTP require prior NRC approval. 
This change does not affect possible initiating events for the 
decommissioning accidents previously evaluated in the Humboldt Bay 
Power Plant (HBPP) defueled safety analysis report (DSAR), as 
updated, appendix A, ``Implications of Decommissioning Accidents 
with Potential for Radiological Impacts to the Environment,' or 
alter the configuration or operation of the facility. Safety limits, 
limiting safety system settings, and limiting control systems are no 
longer applicable to HBPP in the permanently defueled mode, and are 
therefore not relevant.
    The proposed change does not affect the boundaries used to 
evaluate compliance with liquid or gaseous effluent limits, and has 
no impact on plant operations.
    Therefore, the proposed license amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The safety analysis for the facility remains accurate as 
described in the HBPP DSAR, as updated, appendix A. There are 
sections of the LTP that refer to the decommissioning activities 
still remaining (e.g. removal of large components, decontamination, 
etc.). However, these activities are performed in accordance with 
approved HBPP work packages/steps and undergo 10 CFR 50.59 screening 
prior to initiation. The proposed amendment merely makes mention of 
these processes and does not bring about physical changes to the 
facility. Therefore, the facility

[[Page 54286]]

conditions for which the postulated accidents have been evaluated 
are still valid and no new accident scenarios, failure mechanisms, 
or single failures are introduced by this amendment. The system 
operating procedures are not affected.
    Therefore, the proposed changes will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    There are no changes to the design or operation of the facility 
resulting from this amendment. The proposed change does not affect 
the boundaries used to evaluate compliance with liquid or gaseous 
effluent limits, and has no impact on plant shutdown operations. 
Accordingly, neither the postulated accident assumptions in the 
DSAR, as updated, appendix A, nor the Technical Specifications are 
affected.
    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: Ms. Jennifer K. Post, Pacific Gas and 
Electric Company, 77 Beale Street, B30A, San Francisco, CA.
    NRC Branch Chief: Bruce Watson.

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: July 17, 2013.
    Description of amendment request: The proposed amendment would 
depart from VCSNS Units 2 and 3 plant-specific Design Control Document 
(DCD) Tier 2 and Tier 2* material contained within the Updated Final 
Safety Analysis Report (UFSAR) to acknowledge various obstructions and 
interferences (other than wall openings and penetrations) that may 
cause a change to the design spacing of shear studs and the design and 
spacing of wall module trusses in a local area, and to acknowledge 
appropriate weld types.
    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 design function of the containment structural modules is to 
support the reactor coolant system components and related piping 
systems and equipment. The design functions of the affected 
structural modules in the auxiliary building are to provide support 
and protection for new and spent fuel and the equipment needed to 
support fuel handling, cooling, and storage in the spent fuel racks, 
and to provide support, protection, and separation for the seismic 
Category I mechanical and electrical equipment located outside the 
containment building. The design function of the shear studs is to 
enable the concrete and steel faceplates to act in a composite 
manner and transfer loads into the concrete of the structural 
modules. The structural modules are seismic Category I structures 
and are designed for dead, live, thermal, pressure, safe shutdown 
earthquake loads, and loads due to postulated pipe breaks. The loads 
and load combinations applicable to the structural modules in the 
auxiliary building are the same as for the containment internal 
structures except that there are no design basis accident loadings 
due to the automatic depressurization system or pressure loads due 
to pipe breaks. The proposed changes to the UFSAR are to include 
types of interferences other than wall openings and penetrations 
that may cause a change in the design spacing of shear studs and the 
design and spacing of wall module trusses in a local area. The 
proposed changes clarify that the stud spacing is specified as a 
design value and add the tolerance for stud spacing. The revised 
spacing including the tolerance continues to be in conformance with 
the design and analysis requirements identified in the UFSAR. The 
proposed changes also include clarification of a requirement for a 
complete joint penetration weld. The thickness, geometry, and 
strength of the structures are not adversely altered. The material 
of the steel plates is not altered. The properties of the concrete 
included in the structural modules are not altered. As a result, the 
design function of the containment structural modules is not 
adversely affected by the proposed change. There is no change to 
plant systems or the response of systems to postulated accident 
conditions. There is no change to the predicted radioactive releases 
due to postulated accident conditions. The plant response to 
previously evaluated accidents or external events is not adversely 
affected, nor does the change described create any new accident 
precursors.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes to the UFSAR acknowledge types of 
interferences (other than wall openings and penetrations) that may 
cause a change in the typical design spacing of shear studs and the 
design and spacing of wall module trusses in a local area. The 
proposed changes clarify that the stud spacing is specified as a 
design value and provide the tolerance for stud spacing. The revised 
spacing, including the tolerance, continues to be in conformance 
with the design and analysis requirements identified in the UFSAR. 
Stud spacing and sizing are evaluated to demonstrate that stud 
loadings and shear transfer capability are within acceptable limits 
and that the structural module acts in a composite manner. An 
additional proposed change is to clarify a requirement for a 
complete joint penetration weld. The thickness, geometry, and 
strength of the structures are not adversely altered. The materials 
of the steel plates are not altered. The properties of the concrete 
included in the structural modules are not altered. The changes to 
the internal design of the structural modules do not create any new 
accident precursors. As a result, the design function of the modules 
is not adversely affected by the proposed changes.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The criteria and requirements of American Concrete Institute 
(ACI) 349 and American Institute of Steel Construction (AISC) N690 
provide a margin of safety to structural failure. The design of the 
shear studs and wall trusses for the structural wall modules 
conforms to applicable criteria and requirements in ACI 349 and AISC 
N690 and, therefore, maintain the margin of safety. The proposed 
changes to the UFSAR acknowledge types of interferences (other than 
wall openings and penetrations) that may cause a change in the 
typical design spacing of shear studs and the design and spacing of 
wall module trusses in a local area. The proposed changes clarify 
that the stud spacing is specified as a design value and add the 
tolerance for stud spacing. The revised spacing including the 
tolerance continues to be in conformance with the design and 
analysis requirements identified in the UFSAR. An additional 
proposed change is to clarify a requirement for a complete joint 
penetration weld. There is no change to the capacity of the weld or 
to the design requirements of the modules. There is no change to the 
method of evaluation from that used in the design basis 
calculations.
    Therefore, the proposed amendment does not result in 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.

[[Page 54287]]

    NRC 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: March 15, 2013, and revised on July 10, 
2013, and supplemented on August 16, 2013.
    Description of amendment request: The proposed change would amend 
Combined License Nos. NPF-91 and NPF-92 for Vogtle Electric Generating 
Plant (VEGP) Units 3 and 4 by departing from the plant-specific Design 
Control Document (DCD) Tier 1 (and corresponding Combined License 
Appendix C information) and Tier 2 material by making changes to the 
Non-Class 1E dc and Uninterruptible Power Supply System (EDS) and 
Uninterruptible Power Supply System (IDS) and making changes to the 
corresponding Tier 1 information in Appendix C to the Combined License. 
The proposed changes would:

    (1) Increase EDS total equipment capacity, component ratings, 
and protective device sizing to support increased load demand,
    (2) Relocate equipment and moving Turbine Building (TB) first 
bay EDS Battery Room and Charger Room. The floor elevation increases 
from elevation 148'-0'' to elevation 148'-10'' to accommodate 
associated equipment cabling with this activity, and
    (3) Remove the Class 1E IDS Battery Back-up tie to the Non-Class 
1E EDS Battery.

    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 design function of the Turbine Building (TB) is to provide 
weather protection for the laydown and maintenance of major turbine/
generator components. The TB first bay is a seismic Category II 
structure designed to prevent the collapse under a safe shutdown 
earthquake (SSE) to protect the adjacent auxiliary building. The 
electrical system and air-handling units are designed to provide 
electrical power to plant loads and maintain acceptable temperatures 
for electrical equipment rooms and work areas. The electrical 
equipment continues to be in accordance with the same codes and 
standards stated in the Updated Final Safety Analysis Report 
(UFSAR). The proposed relocation of equipment, including the 
increase in floor elevation by 10 inches to accommodate overhead 
equipment cabling, does not impact the TB design function. The TB 
first bay continues to meet seismic Category II requirements. Based 
on this, the proposed changes would not increase the probability of 
an accident previously evaluated.
    The proposed changes do not involve any accident initiating 
event, thus the probabilities of the accidents previously evaluated 
are not affected. The relocation of equipment does not involve any 
safety-related structures, systems, or components; the affected 
rooms do not represent a radioactive material barrier; and this 
activity does not affect the containment of radioactive material. 
The radioactive material source terms and release paths used in the 
safety analyses are unchanged, thus the radiological releases in the 
accident analyses are not affected. Therefore, the consequences of 
an accident previously evaluated are not affected.
    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 proposed changes would use the same type of electrical 
equipment with higher ratings and capacity, change the source of a 
battery back-up, and relocate equipment. The electrical equipment 
will continue to perform its design functions because the same 
electrical codes and standards as stated in the UFSAR continue to be 
met. Therefore the proposed changes do not affect equipment failure 
probabilities or alter any accident initiator or initiating sequence 
of events. The proposed changes in location of equipment and 
elevation of the TB first bay floor do not affect the design 
function of the TB first bay to protect the adjacent auxiliary 
building by meeting seismic Category II structure requirements, or 
affect the operation of the relocated equipment, or the ability of 
the relocated equipment to meet its design functions. Because the 
SSCs and equipment affected by the proposed changes continue to meet 
their design functions, the structural codes and standards as stated 
in the UFSAR, the proposed changes do not introduce a different type 
of accident than those previously considered.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The current seismic requirements applicable to the seismic 
Category II TB first bay structure, including the seismic modeling 
and analysis methods, will continue to apply to the TB first bay 
floor elevation increase. The proposed changes to relocate equipment 
and the increase in the floor elevation will continue to meet the 
fire rating requirements and will be in accordance with the same 
codes and standards currently identified in the UFSAR. The proposed 
changes to the electrical equipment will continue to meet existing 
electrical equipment industry standard recommendations identified in 
the UFSAR. Because no safety analysis or design basis acceptance 
limit/criterion is challenged or exceeded by these proposed changes, 
no margin of safety is reduced.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: July 2, 2013.
    Description of amendment request: The proposed change would amend 
Combined License Nos. NPF-91, and NPF-92 for VEGP Units 3 and 4, 
respectively, by revising Tier 2* and associated Tier 2 information 
related to the design details of connections in several locations 
between the steel plate composite construction (SC) used for the shield 
building and the standard reinforced concrete (RC) walls, floors, and 
roofs of the auxiliary building and lower walls of the shield building. 
These connections are also referred to as ``RC to SC connections.'' 
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 design functions of the nuclear island structures are to 
provide support, protection, and separation for the seismic Category 
I mechanical and electrical equipment located in the nuclear island. 
The nuclear island structures are structurally designed to meet

[[Page 54288]]

seismic Category I requirements as defined in Regulatory Guide 1.29. 
The change to the detail design of connections between the RC and SC 
structures do not have an adverse impact on the response of the 
nuclear island structures to safe shutdown earthquake ground motions 
or loads due to anticipated transients or postulated accident 
conditions. The changes to the detail design do not impact the 
support, design, or operation of mechanical and fluid systems. There 
is no change to plant systems or the response of systems to 
postulated accident conditions. There is no change to the predicted 
radioactive releases due to postulated accident conditions. The 
plant response to previously evaluated accidents or external events 
is not adversely affected, nor do the changes describe create any 
new accident precursors.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes are to the detail design of connections 
between the RC and SC structures. The changes to the detail design 
of connections do not change the criteria and requirements for the 
design and analysis of the nuclear island structures. The changes to 
the detail design of connections do not change the design function, 
support, design, or operation of mechanical and fluid systems. The 
changes to the detail design of connections do not change the 
methods used to connect the RC to the SC. The changes to the detail 
design of the connections do not result in a new failure mechanism 
for the nuclear island structures or new accident precursors. As a 
result, the design functions of the nuclear island structures are 
not adversely affected by the proposed changes.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    No safety analysis or design basis acceptance limit/criterion is 
involved by the requested changes, thus, no margin of safety is 
reduced.
    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. M. Stanford Blanton, Balch & Bingham 
LLP, 1710 Sixth Avenue North, Birmington, AL 35203-2015.
    NRC Branch Chief: Lawrence J. Burkhart.

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

    Date of amendment request: July 15, 2013.
    Description of amendment request: The proposed change would amend 
Combined License Nos. NPF-91, and NPF-92 for VEGP Units 3 and 4, 
respectively, by revising Tier 2* information related to the 
construction of Module CA03. Some of these changes include the removal 
of specifically mentioned materials, increasing anchoring supports and 
allowing the use of anchor bars with hooks.
    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 design functions of the nuclear island structures are to 
provide support, protection, and separation for the seismic Category 
I mechanical and electrical equipment located in the nuclear island. 
The nuclear island structures are structurally designed to meet 
seismic Category I requirements as defined in Regulatory Guide 1.29. 
The change to the design details for the in-containment refueling 
water storage tank (IRWST) west wall does not have an adverse impact 
on the response of the nuclear island structures to safe shutdown 
earthquake ground motions or loads due to anticipated transients or 
postulated accident conditions, nor does it change the seismic 
Category I classification. The change to the design details for the 
IRWST west wall does not impact the support, design, or operation of 
mechanical and fluid systems. There is no change to plant systems or 
the response of systems to postulated accident conditions. There is 
no change to the predicted radioactive releases due to postulated 
accident conditions. The plant response to previously evaluated 
accidents or external events is not adversely affected, nor does the 
change described create any new accident precursors.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change is to revise design details for the IRWST 
west wall. The change of the design details for the IRWST west wall 
does not change the design requirements of the nuclear island 
structures, nor the seismic Category I classification. The change of 
the design details for the IRWST west wall does not change the 
design function, support, design, or operation of mechanical and 
fluid systems. The change of the design details for the IRWST west 
wall does not result in a new failure mechanism for the nuclear 
island structures or introduce any new accident precursors. As a 
result, the design function of the nuclear island structures is not 
adversely affected by the proposed change.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    No safety analysis or design basis acceptance limit/criterion is 
involved by the requested changes, thus, no margin of safety is 
reduced.
    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. M. Stanford Blanton, Balch & Bingham 
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Lawrence J. 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: August 6, 2013.
    Description of amendment request: The proposed change would amend 
Combined License Nos. NPF-91, and NPF-92 for Vogtle Electric Generating 
Plant (VEGP) Units 3 and 4 by departing from the plant-specific Design 
Control Document (DCD) Tier 1(and corresponding Combined License 
Appendix C information) and Tier 2 material by revising the safety 
function and classification of Liquid Radwaste System (WLS) drain hubs 
in the Chemical and Volume Control System and Passive Core Cooling 
System (PXS) compartments. In addition, the proposed changes would 
modify the PXS compartment drain piping connection; WLS valve types, 
and depiction of components in the WLS figures.
    Because, this proposed change requires a departure from Tier 1 
information in the Westinghouse Advanced Passive 1000 DCD, the licensee 
also requested an exemption from the requirements of the Generic

[[Page 54289]]

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 design function of the WLS is containment isolation and the 
prevention of backflow in the drain lines from the CVS compartment 
and the PXS compartment to the containment sump which prevents cross 
flooding of these compartments. The proposed changes to the WLS 
drainage function; the CVS and PXS compartment drain hubs; and the 
WLS valve types do not affect these design functions or any other 
system design function. Revising the drain hub safety 
classification, the PXS drains connection type, and the WLS valve 
types do not involve any accident initiating event or component 
failure. The changes to how components (valves, filters) are 
depicted in the figure provide consistency with the figure legend 
and do not alter any system functions. The system will utilize the 
same codes and standards previously used for the system. Since there 
are no impacts on accident initiating events or component failures, 
the probability of an accident previously evaluated is not affected. 
The radioactive material source terms and release paths used in the 
safety analyses are unchanged, thus the radiological releases in the 
Updated Final Safety Analysis Report (UFSAR) accident analyses are 
not affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes to the WLS system do not adversely affect 
the design or quality of any structure, system or component. 
Revising the WLS safety functions and re-classifying the drain hubs 
as nonsafety-related does not create a new fault or sequence of 
events that could result in a radioactive material release nor do 
the changes to the WLS piping connections, valve types and the 
depiction of components on the figure have any impact on any 
accident previously evaluated.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes to the WLS system drain hubs, piping 
connection, valve type, and Tier 1 figure depiction would not affect 
any radioactive material barrier. No safety analysis or design basis 
acceptance limit/criterion is challenged or exceeded by the proposed 
change, thus no margin of safety is reduced.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

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

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348, and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama

    Date of amendment request: December 21, 2012, as supplemented on 
May 21, 2013.
    Description of amendment request: The proposed amendments would 
revise the Joseph M. Farley Nuclear Plant (FNP) Facility Operating 
Licenses (FOL), Appendix C, to require Southern Nuclear Operating 
Company (SNC) to fully implement and maintain in effect the Degraded 
Voltage Protection modification schedule.
    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 change to the FNP FOL that incorporates the 
Degraded Voltage Protection modification implementation schedule is 
administrative in nature. This proposed change does not alter 
accident analysis assumptions, add any initiators, or affect the 
function of plant systems or the manner in which systems are 
operated, maintained, modified, tested or inspected.
    Therefore, this 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 change to the FNP FOL that incorporates the 
Degraded Voltage Protection modification implementation schedule is 
administrative in nature. This proposed change does not alter 
accident analysis assumptions, add any initiators, or affect the 
function of plant systems or the manner in which systems are 
operated, maintained, modified, tested or inspected.
    Therefore, this proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Plant safety margins are established through limiting conditions 
for operation, limiting safety system settings, and safety limits 
specified in the technical specifications. The proposed change to 
the FNP FOL is administrative in nature. Because there is no change 
to these established safety margins as a result of this change, the 
proposed change does not involve a significant reduction in a margin 
of safety.
    Therefore, the proposed change does not involve a significant 
reduction in 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: Leigh D. Perry, SVP & General Counsel, 
Southern Nuclear Operating Company, 40 Inverness Center Parkway, 
Birmingham, AL 35242.
    NRC Branch Chief: Robert J. Pascarelli.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321, and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of amendment request: December 21, 2012, as supplemented June 
21, 2013.
    Description of amendment request: The proposed License Amendment 
Request (LAR) would revise the Edwin I. Hatch Nuclear Plant (HNP) 
Facility Operating Licenses to require Southern Nuclear Operating 
Company (SNC) to implement modifications that will eliminate the need 
for administrative controls with regard to protection of the plant from 
degraded grid voltage conditions for HNP.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


[[Page 54290]]


    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to the HNP FOL that incorporates the 
Degraded Voltage Protection modification implementation schedule is 
administrative in nature. This proposed change does not alter 
accident analysis assumptions, add any initiators, or affect the 
function of plant systems or the manner in which systems are 
operated, maintained, modified, tested or inspected.
    Therefore, this 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 change to the HNP FOL that incorporates the 
Degraded Voltage Protection modification implementation schedule is 
administrative in nature. This proposed change does not alter 
accident analysis assumptions, add any initiators, or affect the 
function of plant systems or the manner in which systems are 
operated, maintained, modified, tested or inspected.
    Therefore, this proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Plant safety margins are established through limiting conditions 
for operation, limiting safety system settings, and safety limits 
specified in the technical specifications. The proposed change to 
the HNP FOL is administrative in nature. Because there is no change 
to these established safety margins as a result of this change, the 
proposed change does not involve a significant reduction in a margin 
of safety.
    Therefore, the proposed change does not involve a significant 
reduction in 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: Leigh D. Perry, SVP & General Counsel, 
Southern Nuclear Operating Company, 40 Inverness Center Parkway, 
Birmingham, AL 35242.
    NRC Branch Chief: Robert J. Pascarelli.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321, and 50-366, Edwin I. Hatch 
Nuclear Plant (HNP), Units 1 and 2, Appling County, Georgia

    Date of amendment request: July 23, 2013.
    Description of amendment request: The proposed amendments would 
modify Technical Specification (TS) requirements related to control 
room envelope (CRE) habitability in accordance with the Nuclear 
Regulatory Commission (NRC)-approved Revision 3 of Technical 
Specification Task Force (TSTF) Standard Technical Specifications (STS) 
Change Traveler TSTF-448, ``Control Room Habitability.''
    The NRC staff published a notice of opportunity for comment in the 
Federal Register on October 17, 2006 (71 FR 61075), on possible license 
amendments adopting TSTF-448 using the NRC's consolidated line-item 
improvement process (CLIIP) for amending licensees' TSs, which included 
a model safety evaluation (SE) and model no significant hazards 
consideration (NSHC) determination. The NRC staff subsequently issued a 
notice of availability of the models for referencing in license 
amendment applications in the Federal Register on January 17, 2007 (72 
FR 2022), which included the resolution of public comments on the model 
SE and model NSHC determination. The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated July 23, 2013.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC is presented below:

Criterion 1

    The Proposed Change Does Not Involve a Significant Increase in 
the Probability or Consequences of an Accident Previously Evaluated.
    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does 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. The proposed change revises the TS for the CRE emergency 
ventilation system, which is a mitigation system designed to 
minimize unfiltered air leakage into the CRE and to filter the CRE 
atmosphere to protect the CRE occupants in the event of accidents 
previously analyzed. An important part of the CRE emergency 
ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated. Therefore, the probability of any accident 
previously evaluated is not increased. Performing tests to verify 
the operability of the CRE boundary and implementing a program to 
assess and maintain CRE habitability ensure that the CRE emergency 
ventilation system is capable of adequately mitigating radiological 
consequences to CRE occupants during accident conditions, and that 
the CRE emergency ventilation system will perform as assumed in the 
consequence analyses of design basis accidents. Thus, the 
consequences of any accident previously evaluated are not increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

Criterion 2

    The Proposed Change Does Not Create the Possibility of a New or 
Different Kind of Accident from any Accident Previously Evaluated.
    The proposed change does not impact the accident analysis. The 
proposed change does not alter the required mitigation capability of 
the CRE emergency ventilation system, or its functioning during 
accident conditions as assumed in the licensing basis analyses of 
design basis accident radiological consequences to CRE occupants. No 
new or different accidents result from performing the new 
surveillance or following the new program. The proposed change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a significant 
change in the methods governing normal plant operation. The proposed 
change does not alter any safety analysis assumptions and is 
consistent with current plant operating practice.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.

Criterion 3

    The Proposed Change Does Not Involve a Significant Reduction in 
the Margin of Safety.
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation as determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.

[[Page 54291]]

    Attorney for licensee: Leigh D. Perry, SVP & General Counsel, 
Southern Nuclear Operating Company, 40 Inverness Center Parkway, 
Birmingham, AL 35242.
    NRC Branch Chief: Robert Pascarelli.

Tennessee Valley Authority, Docket Nos. 50-327, and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: July 3, 2013 (SQN-TS-12-04).
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TSs) 3/4.6.5, ``Ice Condenser.'' 
The proposed changes would revise TS Limiting Condition for Operation 
3.6.5.1.d and TS Surveillance Requirement 4.6.5.1.d.2 to raise the 
overall ice condenser ice weight from 2,225,880 pounds (lbs) to 
2,540,808 lbs and to raise the minimum TS ice basket weight from 1145 
lbs to 1307 lbs, respectively. These changes are necessary to address 
the issues raised in Nuclear Safety Advisory Letter (NSAL) 11-5, 
``Westinghouse LOCA [Loss-of-Coolant Accident] Mass and Energy Release 
Calculation Issues.'' The issues identified in NSAL-11-5 affected 
plant-specific LOCA mass and energy release calculation results that 
are used as input to the containment integrity response analyses. The 
basis for the proposed changes is provided in WCAP-12455, Revision 1, 
Supplement 2R, ``Tennessee Valley Authority [TVA] Sequoyah Nuclear 
Plant [SQN] Units 1 and 2 Containment Integrity Reanalyses Engineering 
Report.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequence of an accident previously evaluated?
    Response: No.
    The analyzed accidents of consideration in regards to changes 
affecting the ice condenser are a loss of coolant accident (LOCA) 
and a main steam line break (MSLB) inside containment. The ice 
condenser is a passive system and is not postulated as being the 
initiator of any LOCA or MSLB and is designed to remain functional 
following a design basis earthquake. In addition, the ice condenser 
does not interconnect or interact with any systems that have an 
interface with the reactor coolant or main steam systems.
    For SQN, the LOCA is the more severe accident in terms of 
containment pressure and ice bed melt out, and is therefore the more 
limiting accident. The revised SQN LOCA containment integrity 
analysis determined that the post-LOCA peak containment pressure is 
below the containment design pressure and that the margin to ice 
meltout is maintained. The analysis assumes an ice weight that 
ensures sufficient heat removal capability is available from the ice 
condenser to limit the accident peak pressure inside containment.
    TVA has evaluated the effects of the increased ice condenser ice 
weight and determined that the increase in ice weight does not 
invalidate the ice condenser seismic qualification, does not 
adversely affect the capacity of the ice bed to absorb iodine during 
a LOCA, and does not diminish the boron concentration of the 
recirculated primary coolant during a LOCA.
    TVA has also evaluated differences between the as-built plant 
and the assumptions of the revised analysis and determined that the 
results of the revised analysis remain valid for Model 57AG steam 
generators and for AREVA Advanced W17 High Thermal Performance (HTP) 
fuel.
    The proposed changes reflect the ice weight assumed in the 
containment integrity analysis including conservative allowances for 
sublimation and weighing instrument systematic error. Accordingly, 
the proposed changes ensure that ice weight values maintain margin 
between the calculated peak containment accident pressure and the 
containment design pressure. The results of the analysis and the 
margins are maintained; therefore, the consequences of a previously 
evaluated accident are not adversely affected by the proposed 
changes.
    Because (1) the ice condenser is not an accident initiator, (2) 
the results of the revised analysis remain valid for Model 57AG 
steam generators and for AREVA Advanced W17 High Thermal Performance 
(HTP) fuel, and (3) the proposed changes to the TSs are limited to 
revision of the ice weight values to reflect the revised containment 
integrity analysis, there is no change in the probability of an 
accident previously evaluated in the SQN Updated Final Safety 
Analysis Report (UFSAR).
    Based on the above discussions, the proposed changes do not 
involve an 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 ice condenser serves to limit the peak pressure inside 
containment following a LOCA or MSLB. The proposed changes are 
limited to the revision of the minimum ice weights specified in the 
TSs. The revised containment pressure analysis determined that 
sufficient ice would be present to maintain the peak containment 
pressure below the containment design pressure. No new modes of 
operation, accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of this proposed change.
    TVA has evaluated the effects of the increased ice condenser ice 
weight and determined that the increase in ice weight does not 
invalidate the ice condenser seismic qualification, does not 
adversely affect the capacity of the ice bed to absorb iodine during 
a LOCA, and does not diminish the boron concentration of the 
recirculated primary coolant during a LOCA. TVA has also evaluated 
differences between the as-built plant and the assumptions of the 
revised analysis and determined that the results of the revised 
analysis remain valid for Model 57AG steam generators and for AREVA 
Advanced W17 High Thermal Performance (HTP) fuel. Because sufficient 
ice weight is available to maintain the peak containment pressure 
below the containment design pressure, the results of the revised 
analysis remain valid for Model 57AG steam generators and for AREVA 
Advanced W17 High Thermal Performance (HTP) fuel, and the increase 
in ice weight does not invalidate the ice condenser seismic 
qualification, the increased ice weight does not create the 
possibility of an accident that is different than any already 
evaluated in the SQN UFSAR.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The operability of the ice bed ensures that the required ice 
inventory will (1) be distributed evenly through the containment 
bays, (2) contain sufficient boron to preclude dilution of the 
containment sump following the LOCA and (3) contain sufficient heat 
removal capability to condense the reactor system volume released 
during a LOCA. These conditions are consistent with the assumptions 
used in the accident analyses.
    The revised analysis demonstrates that the ice condensers will 
continue to preclude over-pressurizing the lower containment and 
continue to absorb sufficient heat energy to assist in precluding 
containment vessel failure. TVA has evaluated the effects of the 
increased ice condenser ice weight and determined that the increase 
in ice weight does not invalidate the ice condenser seismic 
qualification, does not adversely affect the capacity of the ice bed 
to absorb iodine during a LOCA, and does not diminish the boron 
concentration of the recirculated primary coolant during a LOCA.
    The proposed changes are required to resolve non-conservative 
TSs currently addressed by administrative controls established in 
accordance with Nuclear Regulatory Commission (NRC) Administrative 
Letter 98-10. The revised containment integrity response analysis 
requires an increase in the required ice weight to ensure that the 
post-LOCA peak containment pressure remains within the design 
limits. As a result, the proposed changes restore margin between the 
accident peak pressure and the containment design pressure and 
resolve non-conservative TSs ice weight values currently under 
administrative controls. Accordingly, 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

[[Page 54292]]

standards of 50.92(c) are satisfied. Therefore, the NRC staff proposes 
to determine that the amendment request involves no significant hazards 
consideration. Attorney for licensee: General Counsel, Tennessee Valley 
Authority, 400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 
37902. Acting NRC Branch Chief: Douglas A. Broaddus.

Virginia Electric and Power Company, Docket Nos. 50-338, and 50-339, 
North Anna Power Station, Units 1 and 2, Louisa County, Virginia

Virginia Electric and Power Company, Docket No. 50-280, and 50-281, 
Surry Power Station, Units 1 and 2, Surry County, Virginia

    Date of amendment request: June 26, 2013.
    Description of amendment request: The proposed license amendment 
(Agencywide Documents Access and Management System (ADAMS) Accession 
No. ML13179A014) requests the approval of (1) generic application of 
Appendix D, ``Qualification of the ABB-NV and WLOP Critical Heat Flux 
(CHF) Correlations in the Dominion VIPRE-D Computer Code,'' to Fleet 
Report DOM-NAF-2-A, ``Reactor Core Thermal-Hydraulics Using the VIPRE-D 
Computer Code,'' (2) the plant-specific application of Appendix D to 
DOM-NAF-2-A to North Anna and Surry Power Stations (in accordance with 
Section 2.1 of DOM-NAF-2-A), and (3) an increase in the Surry Power 
Station Technical Specification Minimum Temperature for Criticality.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The first and second proposed changes would allow Dominion to 
use the VIPRE-D/ABB-NV and VIPRE-D/WLOP code/correlation pairs to 
perform licensing calculations for North Anna and Surry, using the 
DDLs documented in Appendix D of Fleet Report DOM-NAF-2. Neither 
code/correlation pair methodology makes any contribution to the 
potential accident initiators and thus cannot increase the 
probability of any accident. Further, since the DDLs for ABB-NV and 
WLOP meet the required design basis of avoiding departure from 
nucleate boiling (DNB) with 95% probability at a 95% confidence 
level, the use of the new code/correlations does not increase the 
potential consequences of any accident. The pertinent evaluations 
that need to be performed as part of the cycle specific reload 
safety analysis to confirm that the existing safety analyses remain 
applicable have been performed and determined to be acceptable. The 
use of a different code/correlation pair will not increase the 
probability of an accident because plant systems will not be 
operated in a different manner, and system interfaces will not 
change. The use of the VIPRE-D/ABB-NV and VIPRE-D/WLOP code/
correlation pairs to perform licensing calculations for North Anna 
and Surry will not result in a measurable impact on normal operating 
plant releases and will not increase the predicted radiological 
consequences of accidents postulated in the Updated Final Safety 
Analysis Report (UFSAR). Therefore, neither the probability of 
occurrence nor the consequences of any accident previously evaluated 
is significantly increased.
    The third proposed change, an increase of the Surry Minimum 
Temperature for Criticality limit from 522 [deg]F to 538 [deg]F, 
would provide Dominion with increased flexibility during loading 
pattern development as well as improved design margins when coupled 
with the second proposed change. The Minimum Temperature for 
Criticality is used within the reload verification process to ensure 
the assumptions made in the safety analysis remain bounding for the 
given cycle design. With implementation of the proposed change, the 
reload design and licensing requirements will remain in place and 
continue to be met at the increased Minimum Temperature for 
Criticality limit.
    The increase in the Surry Minimum Temperature for Criticality 
limit will not increase the probability of an accident because plant 
systems will not be operated in a different manner, and system 
interfaces will not change. Should the reactor coolant system (RCS) 
temperature fall below the proposed limit, the unit would be in an 
abnormal condition requiring operator action. The operator actions 
are not changing as a result of the increased Minimum Temperature 
for Criticality limit. The increase in the Surry Minimum Temperature 
for Criticality will not result in a measurable impact on normal 
operating plant releases and will not increase the predicted 
radiological consequences of accidents postulated in the UFSAR. 
Therefore, neither the probability of occurrence nor the 
consequences of any accident previously evaluated is significantly 
increased.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not involve a physical alteration of the 
plant (no new or different type of equipment will be installed).
    The use of the VIPRE-D/ABB-NV and VIPRE-D/WLOP code/correlation 
pairs and the applicable fuel design limits for DNB ratio (DNBR) 
does not impact any of the applicable design criteria and the 
pertinent licensing basis criteria will continue to be met. 
Demonstrated adherence to these standards and criteria precludes new 
challenges to components and systems that could introduce a new type 
of accident. Setpoint safety analysis evaluations have demonstrated 
that the use of the VIPRE-D/ABB-NV and VIPRE-D/WLOP code/correlation 
pairs is acceptable. Design and performance criteria will continue 
to be met, and no new single failure mechanisms will be created. The 
use of the VIPRE-D/ABB-NV and VIPRE-D/WLOP code/correlation pairs 
does not involve any alteration to plant equipment or procedures 
that would introduce any new or unique operational modes or accident 
precursors.
    The increase in the Surry Minimum Temperature for Criticality 
does not result in any plant design changes. In addition, the 
minimum temperature at which the reactor is taken critical is not an 
accident initiator. The nominal average reactor coolant system 
temperature during an approach to criticality is several degrees 
higher than the limit proposed for the Minimum Temperature for 
Criticality.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously analyzed.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The first two proposed changes would allow Dominion to use the 
VIPRE-D/ABB-NV and VIPRE-D/WLOP code/correlation pairs to perform 
licensing calculations for North Anna and Surry using the DDLs 
documented in Appendix D of Fleet Report DOM-NAF-2. North Anna TS 
2.1, ``Safety Limits,'' states that, ``the departure from nucleate 
boiling ratio (DNBR) shall be maintained greater than or equal to 
the 95/95 DNBR criterion for the DNB correlations and methodologies 
specified in Section 5.6.5 [COLR].'' The DNBR limits meet the design 
basis of avoiding DNB with 95% probability at a 95% confidence 
level. Surry TS 2.1, ``Safety Limits, Reactor Core,'' specifies that 
``for transients analyzed using the deterministic methodology, the 
DNBR shall be maintained greater than or equal to the applicable DNB 
correlation limit.'' The required DNBR margin of safety for North 
Anna and Surry, which in this case is the margin between the 95/95 
DNBR limit and clad failure, is therefore not reduced. Therefore, 
the proposed TS changes do not involve a significant reduction in a 
margin of safety.
    The increased Minimum Temperature for Criticality in conjunction 
with the appropriate core designs will ensure the current TS limits 
for the most positive moderator temperature coefficient will 
continue to be satisfied. The current analyses are bounding and 
remain applicable with the increased Minimum Temperature for 
Criticality. 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 no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 
23219.

[[Page 54293]]

    NRC Branch Chief: Robert Pascarelli.

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 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.

Detroit Edison Company, Docket No. 50-16, Enrico Fermi Atomic Power 
Plant, Unit 1, (Fermi 1) Monroe County, Michigan.

    Date of amendment request: December 21, 2012 (ML13002A037).
    Brief description of amendment: This amendment revised the Fermi 1 
license to change the licensee's name on the license to ``DTE Electric 
Company.'' This name change is purely administrative in nature. Detroit 
Edison is a wholly owned subsidiary of DTE Energy Company, and this 
name change is part of a set of name changes of DTE Energy subsidiaries 
to conform their names to the ``DTE'' brand name. No other changes are 
contained within this amendment. This change does not involve a 
transfer of control over or of an interest in the license for Fermi 1.
    Date of issuance: August 8, 2013.
    Effective date: On the date of issuance of this amendment and must 
be fully implemented no later than 60-calendar days from the date of 
issuance.
    Amendment No.: 21.
    Facility Operating License No. DPR-9: Amendment revised the License 
by replacing ``the Detroit Edison'' with ``DTE Electric'' on pages 1, 
2, 4, and 5.
    Date of initial notice in Federal Register: March 19, 2013 (78 FR 
16876).
    The NRC's related evaluation of the amendment is contained in a 
Safety Evaluation dated August 8, 2013.
    No significant hazards consideration comments: None received.

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

    Date of application for amendments: November 22, 2011, as 
supplemented by letters dated July 9, 2012, November 12, 2012, January 
28, 2013, and May 15, 2013.
    Brief description of amendments: The amendments revised the 
Technical Specifications to allow single discharge header operation of 
the nuclear service water system for a time period of 14 days.
    Date of issuance: August 9, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 271 and 267.
    Renewed Facility Operating License Nos. NPF-35 and NPF-52: 
Amendments revised the licenses and the Technical Specifications.
    Date of initial notice in Federal Register: May 15, 2012 (77 FR 
28630). The supplements dated July 9, 2012, November 12, 2012, January 
28, 2013, and May 15, 2013, provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the staff's original proposed 
no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 9, 2013.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 21, 2012, as supplemented on 
March 19, April 29, May 7, May 14, and June 26, 2013.
    Brief description of amendment: The amendment revised the VYNPS 
licensing basis relative to how the station satisfies the requirements 
of 10 CFR 50.63, ``Loss of all alternating current power,'' by 
replacing the Vernon Hydroelectric Station with an onsite diesel 
generator as the alternate alternating current power source that would 
provide acceptable capability to withstand a station blackout under 10 
CFR 50.63(c)(2). The change involves revisions to the VYNPS facility 
and procedures described in the Updated Final Safety Analysis Report.
    Date of Issuance: August 15, 2013.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 258.
    Facility Operating License No. DPR-28: The amendment revised the 
License.
    Date of initial notice in Federal Register: March 19, 2013 (78 FR 
16881).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated August 15, 2013.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 6, 2012.
    Brief description of amendment: The amendment revises the MNGP 
Technical Specifications (TS) Section 3.10.1. Specifically, the 
amendment revises Limiting Condition for Operation 3.10.1 and the 
associated TS Bases to expand its scope to include provisions for 
temperature excursions

[[Page 54294]]

greater than 212[emsp14][deg]F as a consequence of inservice leak and 
hydrostatic testing, and as a consequence of scram time testing 
initiated in conjunction with an inservice or hydrostatic test, while 
considering operation conditions to be in Mode 4. The changes are 
consistent with NRC-approved Technical Specifications Task Force (TSTF) 
Improved Standard Technical Specifications Change Traveler, TSTF-484, 
Revision 0, ``Use of TS 3.10.1 for Scram Time Testing Activities.''
    Date of issuance: August 9, 2013.
    Effective date: This license amendment is effective as of the date 
of its date of issuance and will be implemented within 120 days of 
issuance.
    Amendment No.: 174.
    Renewed Facility Operating License No. DPR-22: Amendment revises 
the Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: March 4, 2013.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 9, 2013.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 23rd day of August 2013.

    For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2013-21247 Filed 8-30-13; 8:45 am]
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