[Federal Register Volume 79, Number 33 (Wednesday, February 19, 2014)]
[Notices]
[Pages 9490-9501]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2014-03494]


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NUCLEAR REGULATORY COMMISSION

[NRC-2014-0028]


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 to 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 January 22, 2014 to February 5, 2014. The 
last biweekly notice was published on January 21, 2014 (79 FR 3412).

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2014-0028. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: [email protected]. For technical questions, contact 
the individual(s) listed in the FOR FURTHER INFORMATION CONTACT section 
of this document.
     Mail comments to: Cindy Bladey, Chief, Rules, 
Announcements, and Directives Branch (RADB), Office of Administration, 
Mail Stop: 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:

I. Accessing Information and Submitting Comments

A. Accessing Information

    Please refer to Docket ID NRC-2014-0028 when contacting the NRC 
about the availability of information regarding this document. You may 
access publicly-available information related to this document by any 
of the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2014-0028.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may access publicly available documents online in the NRC 
Library at http://www.nrc.gov/reading-rm/adams.html. To begin the 
search, select ``ADAMS Public Documents'' and then select ``Begin Web-
based ADAMS Search.'' For problems with ADAMS, please contact the NRC's 
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to [email protected]. The ADAMS accession number 
for each document referenced in this document (if that document is 
available in ADAMS) is provided the first time that a document is 
referenced.
     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-2014-0028 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 you 
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment 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

[[Page 9491]]

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's regulations are accessible electronically from the NRC Library on 
the NRC's Web site at http://www.nrc.gov/reading-rmdoc-collections/cfr/
. If a request for a hearing or petition for leave to intervene is 
filed by the above date, the Commission or a presiding officer 
designated by the Commission or by the Chief Administrative Judge of 
the Atomic Safety and Licensing Board Panel, will rule on the request 
and or petition; and the Secretary or the Chief Administrative Judge of 
the Atomic Safety and Licensing Board will issue a notice of a hearing 
or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's petitioner's interest. The petition must 
also identify the specific contentions which the requestor petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the requestor 
petitioner shall provide a brief explanation of the bases for the 
contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor 
petitioner intends to rely in proving the contention at the hearing. 
The requestor petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor petitioner intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the reques to petitioner to relief. A request or 
petitioner who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held. If 
the final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards consideration, then any hearing 
held would take place before the issuance of any amendment.
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at [email protected], or by 
telephone at 301-415-1677, to request (1) a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a request or 
petition for hearing (even in instances in which the participant, or 
its counsel or representative, already holds an NRC-issued digital ID 
certificate). Based upon this information, the Secretary will establish 
an electronic docket for the hearing in this proceeding if the 
Secretary has not already established an electronic docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the NRC's public Web 
site at http://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support

[[Page 9492]]

unlisted software, and the NRC Meta System Help Desk will not be able 
to offer assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with the NRC guidance 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC's Web site 
at http://www.nrc.gov/site-help/e-submittals.html, by email at 
[email protected], or by a toll-free call at 1-866 672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 
20852, Attention: Rulemaking and Adjudications Staff. Participants 
filing a document in this manner are responsible for serving the 
document on all other participants. Filing is considered complete by 
first-class mail as of the time of deposit in the mail, or by courier, 
express mail, or expedited delivery service upon depositing the 
document with the provider of the service. A presiding officer, having 
granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, a request to intervene will require including information on 
local residence in order to demonstrate a proximity assertion of 
interest in the proceeding. With respect to copyrighted works, except 
for limited excerpts that serve the purpose of the adjudicatory filings 
and would constitute a Fair Use application, participants are requested 
not to include copyrighted materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day deadline will not 
be entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the 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 http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to 
ADAMS or who encounter problems in accessing the documents located in 
ADAMS, should contact the NRC's PDR Reference staff at 1-800-397-4209, 
301-415-4737, or by email to [email protected].

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

    Date of amendment request: September 26, 2013.
    Description of amendment request: The amendments requests 
transition of the fire protection licensing basis at MNS, Units 1 and 
2, from Sec. Sec.  50.48(b) and 50.48(c) of Title 10 of the Code of 
Federal Regulations (10 CFR), National Fire Protection Association 
(NFPA) 805.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Criterion 1: Does the proposed amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response: No.
    Operation of MNS in accordance with the proposed amendment does 
not increase the probability or consequences of accidents previously 
evaluated. The Updated Final Safety Analysis Report documents the 
analyses of design basis accidents at MNS. The proposed amendment 
does not adversely affect accident initiators nor alter design 
assumptions, conditions, or configurations of the facility and does 
not adversely affect the ability of structures, systems, and 
components to perform their design function. Structures, systems, 
and components required to safely shut down the reactor and to 
maintain it in a safe shutdown condition will remain capable of 
performing their design functions.
    One purpose of this amendment is to permit MNS to adopt a new 
fire protection licensing basis which complies with the requirements 
in 10 CFR 50.48(a) and (c) and

[[Page 9493]]

the guidance in Regulatory Guide (RG) 1.205. The NRC considers that 
NFPA 805 provides an acceptable methodology and performance criteria 
for licensees to identify Fire Protection system and features that 
are an acceptable alternative to the Appendix R fire protection 
features (69 FR 33536; June 16, 2004). Engineering Analyses, in 
accordance with NFPA 805, have been performed to demonstrate that 
the risk-informed performance-based requirements for NFPA 805 have 
been met.
    The NFPA 805, taken as a whole, provides an acceptable 
alternative to 10 CFR 50.48(b) and satisfies 10 CFR 50.48(a) and 
General Design Criterion 3 of Appendix A to 10 CFR Part 50 and meets 
the underlying intent of the NRC's existing fire protection 
regulations and guidance, and achieves defense-in-depth and the 
goals, performance objectives, and performance criteria specified in 
Chapter 1 of the standard. The increases in core damage frequency 
associated with the LAR submittal are acceptable within the guidance 
of RG 1.174, therefore this allows self approval of the fire 
protection program changes post-transition. If there are any 
increases post-transition in core damage frequency or risk, the 
increase will be small and consistent with the intent of the 
Commission's Safety Goal Policy.
    Based on this, the implementation of this proposed amendment 
does not significantly increase the probability of any accident 
previously evaluated. Equipment required to mitigate an accident 
remains capable of performing the assumed function.
    Therefore, the consequences of any accident previously evaluated 
are not significantly increased with the implementation of the 
amendment.
    Criterion 2: Does the proposed amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response: No.
    Operation of MNS in accordance with the proposed amendment does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated. Any scenario or previously 
analyzed accident with offsite dose was included in the evaluation 
of design basis accidents documented in the Updated Final Safety 
Analysis Report. The proposed change does not alter the requirements 
or function for systems required during accident conditions. 
Implementation of the new Fire Protection licensing basis which 
complies with the requirements in 10 CFR 50.48(a) and (c) and the 
guidance in RG 1.205 will not result in new or different accidents.
    The proposed amendment does not adversely affect accident 
initiators nor alter design assumptions, conditions, or 
configurations of the facility. The proposed amendment does not 
adversely affect the ability of structure, systems, and components 
to perform their design function. Structure, systems, and components 
required to safely shut down the reactor and maintain it in a safe 
shutdown condition remain capable of performing their design 
functions.
    The purpose of this amendment is to permit MNS to adopt a new 
Fire Protection licensing basis which complies with the requirements 
in 10 CFR 50.48(a) and (c) and the guidance in RG 1.205. The NRC 
considers that NFPA 805 provides an acceptable methodology and 
performance criteria for licensees to identify Fire Protection 
systems and features that are an acceptable alternative to the 
Appendix R Fire Protection features (69 FR 33536; June 16, 2004).
    The requirements in NFPA 805 address only Fire Protection and 
the impacts of fire on the plant have already been evaluated. Based 
on this, the implementation of this proposed amendment does not 
create the possibility of a new or different kind of accident from 
any kind of accident previously evaluated. The proposed changes do 
not involve new failure mechanisms or malfunctions that can initiate 
a new accident.
    Therefore, the possibility of a new or different kind of 
accident from any kind of accident previously evaluated is not 
created with the implementation of this amendment.
    Criterion 3: Does the proposed amendment involve a significant 
reduction in the margin of safety?
    Response: No.
    Operation of MNS in accordance with the proposed amendment does 
not involve a significant reduction in the margin of safety. The 
proposed amendment does not alter the manner in which safety limits, 
limiting safety system settings or limiting conditions for operation 
are determined. The safety analysis acceptance criteria are not 
affected by this change. The proposed amendment does not adversely 
affect existing plant safety margins or the reliability of equipment 
assumed to mitigate accidents in the Updated Final Safety Analysis 
Report. The proposed amendment does not adversely affect the ability 
of Structure, Systems, and Components to perform their design 
function. Structure, Systems, and Components required to safely shut 
down the reactor and to maintain it in a safe shutdown condition 
remain capable of performing their design functions.
    One purpose of this amendment is to permit MNS to adopt a new 
fire protection licensing basis which complies with the requirements 
in 10 CFR 50.48(a) and (c) and the guidance in RG 1.205. The NRC 
considers that NFPA 805 provides an acceptable methodology and 
performance criteria for licensees to identify Fire Protection 
systems and features that are an acceptable alternative to the 
McGuire Nuclear Station's existing fire protection requirements. 
Engineering analyses, which may include engineering evaluations, 
probabilistic safety assessments, and fire modeling calculations, 
have been performed to demonstrate that the performance-based 
methods do not result in a significant reduction in the margin of 
safety.
    Based on this, the implementation of this proposed amendment 
does not significantly reduce the margin of safety. The proposed 
changes are evaluated to ensure that risk and safety margins are 
kept within acceptable limits. Therefore, the transition does not 
involve a significant reduction in the margin of safety.
    The NFPA 805 continues to protect public health and safety 
because the overall approach of NFPA 805 is consistent with the key 
principles for evaluating license basis changes, as described in RG 
1.174, is consistent with the defense-in-depth philosophy, and 
maintains sufficient safety margins.
    Margins previously established for the MNS Fire Protection 
program in accordance with existing fire protection requirements are 
not significantly reduced.
    Therefore, this proposed amendment does not result in a 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lara S. Nichols, Associate General Counsel, 
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC 
28202.
    NRC Branch Chief: Robert J. Pascarelli.

Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, 
Oconee Nuclear Station (ONS), Units 1, 2, and 3, Oconee County, South 
Carolina

    Date of amendment request: October 24, 2013.
    Description of amendment request: The proposed amendments would 
revise Section 3.1.1.1 of the Updated Final Safety Analysis Report 
(UFSAR) for ONS Units 1, 2, and 3 to clarify quality requirements of 
the Standby Shutdown Facility (SSF) and interconnected systems.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change involves no change to the plant design and 
is intended to ensure a consistent interpretation of wording 
previously included in the UFSAR regarding the QA classification of 
certain Structures, Systems, and Components (SSCs) relied upon to 
address a postulated Turbine Building flood event. The proposed 
change will help to ensure the design of the SSF is maintained 
consistent with the licensed design. The proposed UFSAR change does 
not involve operating any installed equipment in a new or different 
manner or a change to any set points for parameters which initiate 
protective or mitigation action. There is no adverse impact on 
containment integrity, radiological release pathways, fuel design, 
filtration systems, main steam relief valve set

[[Page 9494]]

points, or radwaste systems. No new radiological release pathways 
are created. Because this correction and clarification to the UFSAR 
design description does not alter the SSF design as licensed, the 
proposed change does not involve a significant increase in the 
probability or consequences of any event requiring operation of the 
SSF.
    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 requests approval to modify and clarify a 
UFSAR design description to ensure the described design of the ONS 
units and the SSF is maintained consistent with the licensed design. 
In accordance with this revision, replacement equipment is 
functionally equivalent to the existing and is designed to the 
appropriate pressure, temperature, and environmental parameters. The 
proposed change does not change the design function or operation of 
the SSF or of the interconnecting seismic induced turbine building 
flood equipment. Further, the proposed change does not create a new 
or different kind of accident since the proposed changes do not 
introduce credible new failure mechanisms, malfunctions, or accident 
initiators not considered in the design and licensing bases.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change requests approval to modify and clarify a 
UFSAR design description to ensure a consistent understanding of the 
licensed design of the plant, including the SSF. The proposed change 
does not change the design function or operation of the SSF. The 
proposed change does not involve operating any installed equipment 
in a new or different manner; a change to any set points for 
parameters which initiate protective or mitigation action; or any 
impact on the fission product barriers or safety limits.
    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: Lara S. Nichols, Deputy General Counsel, 
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC 
28202-1802.
    NRC Branch Chief: Robert J. Pascarelli.

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

    Date of amendment request: October 31, 2013.
    Description of amendment request: The licensee has indicated their 
intent to submit certifications pursuant to Title 10 of the Code of 
Federal Regulations (10 CFR) 50.82(a)(1)(i) and (ii) along with 10 CFR 
50.82(a)(2) committing to the permanent cessation of operations and the 
permanent removal of fuel from the reactor vessel. Following these 
certifications, the 10 CFR part 50 operating license will no longer 
permit operation of the reactor or placement of fuel in the reactor 
vessel. The proposed amendment includes a number of changes to revise 
or eliminate current requirements found in Section 6.0, Administrative 
Controls, of the Vermont Yankee Technical Specifications to support a 
defueled reactor, the new organization, and the permanent shutdown of 
the facility. Proposed changes include (1) elimination of the 
Mitigating Strategies License Condition in the operating license, (2) 
revisions to Section 6.1, Responsibility, regarding control room 
command function and delegation of authority, (3) revisions to Section 
6.2, Organization, to reflect emphasis on the safe handling and storage 
of spent nuclear fuel as opposed to nuclear plant operations along with 
the conversion of license reactor operators to certified fuel handlers, 
(4) elimination of Section 6.3, Actions to be Taken if a Safety Limit 
is Exceeded, (5) revision to Section 6.4, Procedures, to reflect a 
permanently defueled reactor vessel, (6) revision to Section 6.6, 
Reporting Requirements, to eliminate the Core Operating Limits Report, 
and (7) revision to Section 6.7, Programs and Manuals to eliminate the 
Integrity of Systems Outside Containment program, eliminate the Plant 
Offsite Review Committee review of changes to the Offsite Dose 
Calculation Manual, and eliminate the Primary Containment Leakage Rate 
Testing Program.
    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 examined?
    Response: No.
    The proposed amendment would not take effect until VY [Vermont 
Yankee Nuclear Power Station] has permanently ceased operation and 
entered a permanently defueled condition. The proposed amendment 
would modify the VY OL [operating license] and TS [technical 
specifications] by deleting the portions of the OL and TS that are 
no longer applicable to a permanently defueled facility, while 
modifying the other sections to correspond to the permanently 
defueled condition.
    The deletion and modification of provisions of the 
administrative controls do not directly affect the design of 
structures, systems, and components (SSCs) necessary for safe 
storage of irradiated fuel or the methods used for handling and 
storage of such fuel in the fuel pool. The changes to the 
administrative controls are administrative in nature and do not 
affect any accidents applicable to the safe management of irradiated 
fuel or the permanently shutdown and defueled condition of the 
reactor. The deletion of the Mitigation Strategy License Condition 
is also administrative in nature as the sections of the Order 
requiring implementation of the condition have been rescinded and 
the controlling regulation in which the mitigation strategies have 
been codified, 10 CFR 50.54(hh), specifies that these requirements 
are not applicable in the permanently defueled condition.
    In a permanently defueled condition, the only credible accident 
is the fuel handling accident.
    The probability of occurrence of previously evaluated accidents 
is not increased, since extended operation in a defueled condition 
will be the only operation allowed, and therefore bounded by the 
existing analyses. Additionally, the occurrence of postulated 
accidents associated with reactor operation is no longer credible in 
a permanently defueled reactor. This significantly reduces the scope 
of applicable accidents.
    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 have no impact on facility SSCs affecting 
the safe storage of irradiated fuel, or on the methods of operation 
of such SSCs, or on the handling and storage of irradiated fuel 
itself. The administrative removal of an OL condition [* * *] or 
modifications of the TS that are related only to administration of 
facility cannot result in different or more adverse failure modes or 
accidents than previously evaluated because the reactor will be 
permanently shutdown and defueled and VY will no longer [be] 
authorized to operate the reactor.
    The proposed deletion of requirements of the VY OL and TS do not 
affect systems credited in the accident analysis for the fuel 
handling accident at VY. The proposed OL and TS will continue to 
require proper control and monitoring of safety significant 
parameters and activities.
    The proposed amendment does not result in any new mechanisms 
that could initiate damage to the remaining relevant safety barriers 
for defueled plants (fuel cladding and spent fuel cooling). Since 
extended operation in a defueled condition will be the only 
operation allowed, and therefore bounded by the existing analyses, 
such a condition does not create the possibility of a new or 
different kind of accident.

[[Page 9495]]

    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Because the 10 CFR Part 50 license for VY will no longer 
authorize operation of the reactor or emplacement or retention of 
fuel into the reactor vessel once the certifications required by 10 
CFR 50.82(a)(1) are submitted, as specified in 10 CFR 50.82(a)(2), 
the occurrence of postulated accidents associated with reactor 
operation is no longer credible. The only remaining credible 
accident is a fuel handling accident (FHA). The proposed amendment 
does not adversely affect the inputs or assumptions of any of the 
design basis analyses that impact the FHA.
    The proposed changes are limited to those portions of the OL and 
TS that are not related to the safe storage of irradiated fuel. The 
requirements that are proposed to be revised or deleted from the VY 
OL and TS are not credited in the existing accident analysis for the 
remaining applicable postulated accident; and as such, do not 
contribute to the margin of safety associated with the accident 
analysis. Postulated DBAs involving the reactor are no longer 
possible because the reactor will be permanently shutdown and 
defueled and VY will no longer be authorized to operate the reactor.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

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

Indiana Michigan Power Company (IandM), Docket No. 50-315, Donald C. 
Cook Nuclear Plant, Unit 1, Berrien County, Michigan

    Date of amendment request: October 8, 2013.
    Description of amendment request: The proposed amendment would 
increase the normal reactor coolant system (RCS) temperature and 
pressure at the Donald C. Cook Nuclear Plant, Unit 1, consistent with 
the previously licensed conditions. The proposed amendment would modify 
the Unit 1 technical specifications and license basis associated with 
this change.
    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.
     SR 3.4.14.1 RCS [Pressure Isolation Valve (PIV)] 
Leakage--Surveillance Requirements
    The proposed change to the RCS PIV RCS pressure range does not 
significantly increase the probability or consequences of an 
accident previously evaluated in the [Updated Final Safety Analysis 
Report (UFSAR)]. The analytical and evaluation efforts performed for 
the [Normal Operating Pressure/Normal Operating Temperature (NOP/
NOT)] conditions were shown to be acceptable. The systems and 
components (including interface systems and control systems) will 
function as designed and all performance requirements for these 
systems remain acceptable. There are no physical changes being made 
to the fuel cladding, the RCS pressure boundary, or the containment. 
No significant increase in the consequences has been identified. The 
NOP/NOT conditions do not introduce the possibility of a change in 
the frequency of an accident because the parameter changes are not 
an initiator of any accident previously considered and no new 
failure modes have been introduced.
    Therefore, neither the probability nor the consequences of an 
accident previously evaluated has been significantly increased.
     SR 3.5.5.1 Seal Injection Flow--Surveillance 
Requirements
    The proposed change to the pressurizer pressure range and the 
elimination of the low pressure operation does not significantly 
increase the probability or consequences of an accident previously 
evaluated in the UFSAR. The analytical and evaluation efforts 
performed for the NOP/NOT conditions were shown to be acceptable. 
The systems and components (including interface systems and control 
systems) will function as designed and all performance requirements 
for these systems remain acceptable. There are no physical changes 
being made to the fuel cladding, the RCS pressure boundary, or the 
containment. No significant increase in the consequences has been 
identified. The NOP/NOT conditions do not introduce the possibility 
of a change in the frequency of an accident because the parameter 
changes are not an initiator of any accident previously considered 
and no new failure modes have been introduced.
    Therefore, neither the probability nor the consequences of an 
accident previously evaluated has been significantly increased.
     SR 3.6.10.1 Containment Air Recirculation/Hydrogen 
Skimmer (CEQ) System--Surveillance Requirements
    The proposed change to the containment air recirculation fan 
delay/start times does not significantly increase the probability or 
consequences of an accident previously evaluated in the UFSAR. The 
analytical and evaluation efforts performed for the NOP/NOT 
conditions were shown to be acceptable. The systems and components 
(including interface systems and control systems) will function as 
designed and all performance requirements for these systems remain 
acceptable. There are no physical changes being made to the fuel 
cladding, the RCS pressure boundary, or the containment. No 
significant increase in the consequences has been identified. The 
NOP/NOT conditions do not introduce the possibility of a change in 
the frequency of an accident because the parameter changes are not 
an initiator of any accident previously considered and no new 
failure modes have been introduced.
    Therefore, neither the probability nor the consequences of an 
accident previously evaluated has been significantly increased.
     UFSAR Section 6.3.2, Containment Spray Systems [CTSs], 
System Design
    The proposed revision to UFSAR Section 6.3.2 specifically 
recognizes use of the CTS pump time delay relay in mitigating the 
consequences of postulated accidents. Previously, the setting of 
this relay was established to support proper [emergency diesel 
generator] bus loading and it was accounted for as an input to 
accident analyses. Use of the time delay relay setting to mitigate 
the consequences of an accident does not significantly increase the 
probability or consequences of an accident previously evaluated in 
the UFSAR. The analytical and evaluation efforts performed for the 
NOP/NOT conditions were shown to be acceptable. The systems and 
components (including interface systems and control systems) will 
function as designed and all performance requirements for these 
systems remain acceptable. There are no physical changes being made 
to the fuel cladding, the RCS pressure boundary, or the containment. 
No significant increase in the consequences has been identified. The 
NOP/NOT conditions do not introduce the possibility of a change in 
the frequency of an accident because the parameter changes are not 
an initiator of any accident previously considered and no new 
failure modes have been introduced.
    Therefore, neither the probability nor the consequences of an 
accident previously evaluated has been significantly increased.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
     SR 3.4.14.1 RCS PIV Leakage--Surveillance Requirements
    The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated in 
the UFSAR. No new accident scenarios, failure mechanisms, or 
limiting single failures are introduced as a result of the proposed 
change. This proposed change has no adverse effects on any safety 
related system and does not challenge the performance or integrity 
of any safety related system. The specified RCS pressure functions 
support meeting the accident analyses criteria.
    Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated is not created.

[[Page 9496]]

     SR 3.5.5.1 Seal Injection Flow--Surveillance 
Requirements
    The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated in 
the UFSAR. No new accident scenarios, failure mechanisms, or 
limiting single failures are introduced as a result of the proposed 
changes. This proposed change has no adverse effects on any safety 
related system and does not challenge the performance or integrity 
of any safety related system. The specified pressurizer pressure 
range supports meeting all of the accident analyses criteria.
    Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated is not created.
     SR 3.6.10.1 Containment Air Recirculation/Hydrogen 
Skimmer (CEQ) System--Surveillance Requirements
    The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated in 
the UFSAR. No new accident scenarios, failure mechanisms, or 
limiting single failures are introduced as a result of the proposed 
change. This proposed change has no adverse effects on any safety 
related system and does not challenge the performance or integrity 
of any safety related system. The delay/start time functions support 
meeting all of the accident analyses criteria.
    Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated is not created.
     UFSAR Section 6.3.2, Containment Spray Systems, System 
Design
    The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated in 
the UFSAR because this change simply recognizes potential use of the 
existing CTS pump time delay relay setting to mitigate the 
consequences of an accident. No new accident scenarios, failure 
mechanisms or limiting single failures are introduced as a result of 
the proposed change. This proposed change has no adverse effects on 
any safety related system and does not challenge the performance or 
integrity of any safety related system. The delay/start time 
functions support meeting all of the accident analyses criteria.
    Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated is not created.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
     SR 3.4.14.1 RCS PIV Leakage--Surveillance Requirements
    The proposed change does not involve a significant reduction in 
a margin of safety. Analyses and evaluations supporting the Return 
to NOP/NOT Program conditions demonstrate that all acceptance 
criteria continue to be met. There are no changes to the design, 
material, and construction standards that are applicable to any 
System, Structure, or Component (SSC). There are no physical changes 
being made to the fuel cladding, the RCS pressure boundary, or the 
containment. Also, there is no change to a Design Basis Limit for 
Fission Product Barriers (DBLFPB).
    Therefore, the proposed change does not involve a significant 
reduction in margin of safety.
     SR 3.5.5.1 Seal Injection Flow--Surveillance 
Requirements
    The proposed change does not involve a significant reduction in 
a margin of safety. Analyses and evaluations supporting the Return 
to NOP/NOT Program demonstrate that all acceptance criteria continue 
to be met. There are no changes to the design, material, and 
construction standards that are applicable to any SSC. There are no 
physical changes being made to the fuel cladding, the RCS pressure 
boundary, or the containment. Also, there is no change to a DBLFPB.
    Therefore, the proposed change does not involve a significant 
reduction in margin of safety.
     SR 3.6.10.1 Containment Air Recirculation/Hydrogen 
Skimmer (CEQ) System--Surveillance Requirements
    The proposed change does not involve a significant reduction in 
a margin of safety. Analyses and evaluations supporting the Return 
to NOP/NOT Program conditions demonstrate that all acceptance 
criteria continue to be met. There are no changes to the design, 
material, and construction standards that are applicable to the CEQ 
System. There are no physical changes being made to the fuel 
cladding, the RCS pressure boundary, or the containment. Also, there 
is no change to a DBLFPB.
    Therefore, the proposed changes do not involve a significant 
reduction in margin of safety.
     UFSAR Section 6.3.2, Containment Spray Systems, System 
Design
    The proposed change does not involve a significant reduction in 
a margin of safety. There are no changes to the design, material, 
and construction standards that are applicable to the Containment 
Spray System. There are no physical changes being made to the fuel 
cladding, the RCS pressure boundary, or the containment. Also, there 
is no change to a DBLFPB.
    Therefore, the proposed changes do 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: Robert B. Haemer, Senior Nuclear Counsel, 
One Cook Place, Bridgman, MI 49106.
    NRC Branch Chief: Robert D. Carlson.

Indiana Michigan Power Company (IandM), Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of amendment request: November 6, 2013.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3.6.13, Divider Barrier Integrity, 
concerning the divider barrier seal inspection requirements for the 
Donald C. Cook Nuclear Plant, Units 1 and 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not involve changes to the installed 
structures, systems or components of the facility. The affected 
component (divider barrier seal) is not an accident initiator and 
therefore, this change does not involve a significant increase in 
the probability of an accident. The proposed change is considered 
adequate to ensure continued operability of the divider barrier. 
Since the divider barrier will continue to be available to perform 
its accident mitigation function, the consequences of accidents 
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.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not introduce a new mode of plant 
operation and does not involve physical modification to the plant. 
The change does not introduce new accident initiators or impact 
assumptions made in the safety analysis. Testing requirements 
continue to demonstrate that the Limiting Conditions for Operation 
are met and the system components are functional.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not exceed or alter a design basis or 
safety limit, so there is no significant reduction in the margin of 
safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Robert B. Haemer, Senior Nuclear Counsel, 
One Cook Place, Bridgman, MI 49106.
    NRC Branch Chief: Robert D. Carlson.

[[Page 9497]]

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

    Date of amendment request: December 20, 2013.
    Description of amendment request: The proposed amendments would 
revise the Prairie Island Nuclear Generating Plant, Units 1 and 2, 
Emergency Plan to increase the staff augmentation times for certain 
Emergency Response Organization functions from 30 minutes and 60 
minutes to 90 minutes.
    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 increase in staff augmentation times has no effect 
on normal plant operation or on any accident initiator or precursors 
and does not impact the function of plant structures, systems, or 
components (SSCs). The proposed change does not alter or prevent the 
ability of the Emergency Response Organization to perform their 
intended functions to mitigate the consequences of an accident or 
event. The ability of the emergency response organization to respond 
adequately to radiological emergencies has been demonstrated as 
acceptable through a staffing analysis as required by 10 CFR Part 
50, Appendix E.IV.A.9.
    Therefore, the proposed Emergency Plan 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 change does not impact the accident analysis. The 
change does not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed), a change in 
the method of plant operation, or new operator actions. The proposed 
change does not introduce failure modes that could result in a new 
accident, and the change does not alter assumptions made in the 
safety analysis. This proposed change increases the staff 
augmentation response times in the Emergency Plan, which are 
demonstrated as acceptable through a staffing analysis as required 
by 10 CFR Part 50, Appendix E.IV.A.9. The proposed change does not 
alter or prevent the ability of the Emergency Response Organization 
to perform their intended functions to mitigate the consequences of 
an accident or event.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel cladding, reactor coolant 
system pressure boundary, and containment structure) to limit the 
level of radiation dose to the public. The proposed change is 
associated with the Emergency Plan staffing and does not impact 
operation of the plant or its response to transients or accidents. 
The change does not affect the Technical Specifications. The 
proposed change does not involve a change in the method of plant 
operation, and no accident analyses will be affected by the proposed 
change. Safety analysis acceptance criteria are not affected by this 
proposed change. The revised Emergency Plan will continue to provide 
the necessary response staff with the proposed change. A staffing 
analysis and a functional analysis were performed for the proposed 
change on the timeliness of performing major tasks for the 
functional areas of Emergency Plan. The analysis concluded that an 
increase in staff augmentation times, with the addition of two on-
shift positions, would not significantly affect the ability to 
perform the required Emergency Plan tasks. Therefore, the proposed 
change is determined to not adversely affect the ability to meet 10 
CFR 50.54(q)(2), the requirements of 10 CFR Part 50, Appendix E, and 
the emergency planning standards as described in 10 CFR 50.47 (b).
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: November 26, 2013.
    Description of amendment request: The proposed change would amend 
Combined License Nos. NPF-93 and NPF-94 for VCSNS Units 2 and 3 by 
departing from approved AP1000 Design Control Document (DCD) Tier 2 
information as incorporated into the Updated Final Safety Analysis 
Report (UFSAR) to allow use of a new methodology to determine the 
effective thermal conductivity resulting from oxidation of the 
inorganic zinc (IOZ) used in the containment vessel coating system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Implementation of a methodology which specifies an effective 
thermal conductivity and oxidation progression for the inorganic 
zinc coating of the containment vessel is used to eliminate non-
mechanistic modeling of inorganic zinc thermal conductivity in the 
containment integrity analyses to show that the value for inorganic 
zinc thermal conductivity used in the containment integrity analyses 
is conservative, but is not used to change any of the parameters 
used in those analyses. There is no change to any accident initiator 
or condition of the containment that would affect the probability of 
any accident. The containment peak pressure analysis as reported in 
the UFSAR is not affected; therefore, the previously reported 
consequences are not affected.
    Therefore, there is no significant increase in the probability 
or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment to implement a methodology which 
specifies an effective thermal conductivity and oxidation 
progression and effects for the inorganic zinc coating of the 
containment vessel is used to eliminate non-mechanistic modeling of 
inorganic zinc thermal conductivity in the containment integrity 
analyses to show that the value for inorganic zinc thermal 
conductivity used in the containment integrity analyses is 
conservative, but is not used to change any of the parameters used 
in the containment peak pressure analysis. The change in methodology 
does not change the condition of containment; therefore, no new 
accident initiator is created. The containment peak pressure 
analysis as currently evaluated is not affected, and the 
consequences previously reported are not changed. The new 
methodology does not change the containment; therefore, no new fault 
or sequence of events that could lead to containment failure or 
release of radioactive material is created.
    Therefore, this activity 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.

[[Page 9498]]

    The proposed implementation of a methodology which specifies an 
effective thermal conductivity and oxidation progression and effects 
for the inorganic zinc coating of the containment vessel is used to 
eliminate non-mechanistic modeling of inorganic zinc thermal 
conductivity in the containment integrity analyses to show that the 
value for inorganic zinc thermal conductivity used in the 
containment integrity analyses is conservative, but is not used to 
change any of the parameters used in the containment peak pressure 
analysis. The change in methodology does not change the condition of 
the containment and the integrity of the containment vessel is not 
affected. The containment peak pressure analysis as currently 
evaluated is not affected, and the consequences previously reported 
are not changed. No safety analysis or design basis acceptance 
limit/criterion is changed by the proposed change, thus no margin of 
safety is reduced.
    Therefore, the changes do not involve a significant reduction in 
a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
    NRC Branch Chief: Lawrence Burkhart.

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: December 17, 2013.
    Description of amendment request: The proposed amendment would 
revise the VCSNS Units 2 and 3 Emergency Plan to facilitate compliance 
with the Final Rule for Emergency Planning and Preparedness published 
on November 23, 2011. These proposed changes include the addition of 
text that (1) clarifies the distance of the Emergency Operations 
Facility from the site, (2) updates the content of exercise scenarios 
to be performed at least once each exercise cycle, and (3) requires the 
Evacuation Time Estimate to be updated annually between decennial 
censuses. This amendment request also proposes a new license condition 
to ensure the completion of a staffing analysis of on-shift personnel 
responsibilities no later than 180 days before fuel load.
    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 VCSNS Units 2 and 3 Emergency Plan provides assurance that 
the requirements of emergency preparedness regulations are met. The 
changes do not affect the design, construction, or operation of the 
nuclear plant, so there is no change to the probability or 
consequences of an accident previously evaluated.
    Adding a license condition related to an emergency preparedness 
staffing analysis and changing the VCSNS Units 2 and 3 Emergency 
Plan does not affect prevention and mitigation of abnormal events, 
e.g., accidents, anticipated operational occurrences, earthquakes, 
floods and turbine missiles, or their safety or design analyses as 
the purpose of the plan is to implement emergency preparedness 
regulations. No safety-related structure, system, component (SSC) or 
function is adversely affected. The change does not involve nor 
interface with any SSC accident initiator or initiating sequence of 
events, and thus, the probabilities of the accidents evaluated in 
the UFSAR are not affected. Because the changes do not involve any 
SSC or function used to mitigate an accident, the consequences of 
the accidents evaluated in the UFSAR are not affected.
    Therefore, the requested 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 VCSNS Units 2 and 3 Emergency Plan provides assurance that 
the requirements of emergency preparedness regulations are met. The 
changes do not affect the design, construction, or operation of the 
nuclear plant, so there is no new or different kind of accident from 
any accident previously evaluated. The changes do not affect safety-
related equipment, nor do they affect equipment which, if it failed, 
could initiate an accident or a failure of a fission product 
barrier. In addition, the changes do not result in a new failure 
mode, malfunction, or sequence of events that could affect safety or 
safety-related equipment.
    Therefore, the requested 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 VCSNS Units 2 and 3 Emergency Plan provides assurance that 
the requirements of emergency preparedness regulations are met. The 
changes do not affect the assessments or the plant itself. The 
changes do not affect safety-related equipment or equipment whose 
failure could initiate an accident, nor does it adversely interface 
with safety-related equipment or fission product barriers. No safety 
analysis or design basis acceptance limit or criterion is challenged 
or exceeded by the requested change.
    Therefore, the requested amendment does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
    NRC Branch Chief: Lawrence Burkhart.

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

[[Page 9499]]

Assessment as indicated. All of these items are available for public 
inspection at the NRC's Public Document Room (PDR), located at One 
White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland 20852. Publicly available documents created or 
received at the NRC are accessible electronically through the 
Agencywide Documents Access and Management System (ADAMS) in the NRC 
Library at http://www.nrc.gov/reading-rm/adams.html. If you do not have 
access to ADAMS or if there are problems in accessing the documents 
located in ADAMS, contact the PDR's Reference staff at 1-800-397-4209, 
301-415-4737 or by email to [email protected].

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

    Date of application for amendments: January 28, 2013, as 
supplemented by letter dated April 1, 2013.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 1.3, ``Completion Times'' Example 1.3-3, TS 3.6.6, 
``Containment Spray and Cooling Systems,'' TS 3.7.3, ``Auxiliary 
Feedwater (AFW) System,'' TS 3.8.1, ``AC [Alternating Current] Sources-
Operating,'' and TS 3.8.9, ``Distribution Systems-Operating'' by 
eliminating the second completion time in accordance with TS Task Force 
(TSTF)-439-A, Revision 2, ``Eliminate Second Completion Times Limiting 
Time from Discovery of Failure to Meet an LCO [limiting condition for 
operation].''
    Date of issuance: January 29, 2014.
    Effective date: As of the date of issuance to be implemented within 
90 days.
    Amendment Nos.: 304 and 282.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the License and Technical Specifications.
    Date of initial notice in Federal Register: May 28, 2013 (78 FR 
31981).
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated January 29, 2014.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power 
Station, Unit 2 (MPS2), New London County, Connecticut

    Date of amendment request: March 21, 2013.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 3.1.3.7--Control Rod Drive Mechanisms to 
provide consistency with the operability requirements of TS Table 3.3-
1, Reactor Protective Instrumentation, when control rod drive 
mechanisms are energized and capable of withdrawal for MPS2.
    Date of issuance: January 30, 2014.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 317.
    Renewed Facility Operating License No. DPR-65: Amendment revised 
the License and Technical Specifications.
    Date of initial notice in Federal Register: June 11, 2013 (78 FR 
35061).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 30, 2014.
    No significant hazards consideration comments received: No.

DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of application for amendment: January 11, 2013.
    Brief description of amendment: The amendment revises the Fermi 2 
Technical Specifications (TSs) to risk-inform requirements regarding 
selected Required Action end states. Additionally, it would modify the 
TSs Required Actions with a Note prohibiting the use of limiting 
condition for operation 3.0.4.a when entering the preferred end state 
(Mode 3) on startup. The changes are consistent with the NRC's 
Technical Specification Task Force traveler TSTF-423, Revision 1, 
``Technical Specifications End States, NEDC-32988-A,'' dated December 
22, 2009 (ADAMS Accession No. ML093570241).
    Date of issuance: January 17, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 194.
    Facility Operating License No. NPF-43: Amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: April 16, 2013 (78 FR 
22565).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 17, 2014.
    No significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of application for amendment: January 31, 2012, as 
supplemented by letters dated July 31, August 22, October 5, and 
November 12, 2012, and January 7, April 11, May 9, and August 6, 2013.
    Brief description of amendment: The amendment allows the licensee 
to expand the operating domain by the implementation of Average Power 
Range Monitor/Rod Block Monitor/Technical Specifications/Power Range 
Neutron Monitoring/Maximum Extended Load Line Limit Analysis (ARTS/
PRNM/MELLLA). The Neutron Monitoring System will be modified by 
replacing the Average Power Range Monitor (APRM) subsystem with the 
Nuclear Measurement Analysis and Control (NUMAC) Power Range Neutron 
Monitoring (PRNM) System. The modification of the PRNM system replaces 
analog technology with digital technology to improve the management and 
maintenance of the system. The licensee will expand the operating 
domain to Maximum Extended Load Line Limit Analysis (MELLLA) and make 
changes to certain allowable values and limits and to the Technical 
Specifications (TSs). The changes to the TSs include the adoption of 
Technical Specifications Task Force (TSTF) Change Traveler TSTF-493, 
``Clarify Application of Setpoint Methodology for LSSS [Limiting Safety 
System Setting] Functions,'' Option A surveillance notes. Furthermore, 
the amendment allows a change in the licensing basis to support 
Anticipated Transient without Scram accident mitigation with one 
Standby Liquid Control pump instead of two.
    Date of Issuance: January 31, 2014.
    Effective Date: The license amendment is effective as of its date 
of issuance and shall be implemented within 60 days thereafter. The 
Technical Specification revisions will be applicable following 
completion of the refueling outage (R22) scheduled to begin May 8, 
2015.
    Amendment No.: 226.
    Renewed Facility Operating License No. NPF-21: The amendment 
revised the Facility Operating License and Technical Specifications.
    Date of Initial Notice in Federal Register: September 11, 2012 (77 
FR 55867). The supplemental letters dated July 31, August 22, October 
5, and November 12, 2012, and January 7, April 11, May 9, and August 6, 
2013, provided additional information that clarified the application, 
did not expand the scope of the application as originally noticed, and 
did not change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a

[[Page 9500]]

Safety Evaluation dated January 31, 2014.
    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: September 18, 2012, as 
supplemented on March 12, 2013, July 17, 2013, and November 15, 2013.
    Brief description of amendment: The amendment revises the MNGP 
Renewed Facility Operating License and Technical Specification (TS) 
3.8.3, ``Diesel Fuel Oil, Lube Oil, and Starting Air,'' by removing the 
current stored diesel fuel oil, and lube oil numerical volume 
requirements from the TSs and replacing them with duration-based 
numerical requirements consistent with TSTF-501, Revision 1.
    Date of issuance: January 28, 2014.
    Effective date: This license amendment is effective as of the date 
of issuance and shall be implemented within 60 days from date of 
issuance.
    Amendment No.: 178.
    Renewed Facility Operating License No. DPR-22: Amendment revises 
the Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: December 11, 2012 (77 
FR 73689). The licensee's supplements dated March 12, 2013, July 17, 
2013, and November 15, 2013, did not change the scope of the original 
amendment request, did not change the NRC staff's initial proposed 
finding of no significant hazards consideration determination, and did 
not expand the scope of the original Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 28, 2014.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 13, 2012, as 
supplemented by letters dated June 21, 2013, and July 23, 2013.
    Brief description of amendments: The amendments made changes to the 
Prairie Island Nuclear Generating Plant Emergency Plan emergency action 
level initiating conditions for the classification of liquid effluent 
releases and for the determination of fuel clad barrier loss.
    Date of issuance: January 25, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: Unit 1--210; Unit 2--198.
    Renewed Facility Operating License Nos. DPR-42 and DPR-60: 
Amendments revised the Prairie Island Nuclear Generating Plant 
Emergency Plan.
    Date of initial notice in Federal Register: March 4, 2013 (78 FR 
14134). The supplemental letters dated June 21, 2013, and July 23, 
2013, provided additional information that clarified the application, 
did not expand the scope of the application as originally noticed, and 
did not change the NRC staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 25, 2014.
    No significant hazards consideration comments received: No.

Northern States Power Company--Minnesota, Docket Nos. 50-263, 
Monticello Nuclear Generating Plant, Wright County, Minnesota

    Date of amendment request: December 21, 2012, as supplemented by 
letter dated May 16, 2013.
    Brief description of amendments: The amendment made changes to the 
Monticello Nuclear Generating Plant Emergency Plan by revising the 
Emergency Action Level (EAL) setpoint for the Turbine Building Normal 
Waste Sump (TBNWS) Monitor. The change to the EAL restores indication 
of an Alert classification of a liquid effluent release via the TBNWS 
pathway to within the indication range of the applicable 
instrumentation.
    Date of issuance: January 28, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment No.: 177.
    Renewed Facility Operating License No. DPR-22: Amendment revised 
the Renewed Facility Operating License.
    Date of initial notice in Federal Register: March 4, 2013 (78 FR 
14133). The supplemental letter dated May 16, 2013, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the NRC 
staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 28, 2014.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: April 27, 2012, as supplemented by 
letter dated June 27, 2013.
    Brief description of amendment: The amendment revised the Fort 
Calhoun Station, Unit 1 (FCS) Technical Specification (TS) Limiting 
Condition for Operation 2.16, ``River Level,'' and TS Surveillance 
Requirement 3.2, ``Equipment and Sampling Tests,'' and a related change 
to the FCS Radiological Emergency Response Plan to revise two emergency 
action levels related to high water level in the Missouri River.
    Date of issuance: January 28, 2014.
    Effective date: As of its date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment No.: 274.
    Renewed Facility Operating License No. DPR-40: The amendment 
revised the Renewed Facility Operating License and Technical 
Specifications.
    Date of initial notice in Federal Register: December 26, 2012 (77 
FR 76082). The supplemental letter dated June 27, 2013, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a safety evaluation dated January 28, 2014.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 3, 2013.
    Brief description of amendment: This amendment allows for the 
extension of the 130-month frequency of the VCSNS containment 
integrated leak rate test (ILRT) or Type A test, that is required by TS 
6.8.4(g) to 15 years on a permanent basis.
    Date of issuance: February 5, 2014.
    Effective date: This license amendment is effective as of the date 
of its issuance.
    Amendment No.: 194.

[[Page 9501]]

    Renewed Facility Operating License No. NPF-12: Amendment revises 
the License.
    Date of initial notice in Federal Register: June 25, 2013 (78 FR 
38084).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 5, 2014.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 15, 2013, as supplemented by a 
letter dated November 15, 2013.
    Brief description of amendment: The proposed amendment modified 
design details related to the construction of Module CA03 which forms 
the west wall of the in-containment refueling water storage tank. The 
changes sought to clarify the materials used in fabrication of the 
module, as well as the design details related to the horizontal 
stiffeners used to support the in-containment refueling water storage 
tank, and module legs used to anchor the module in place.
    Date of issuance: January 28, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: Unit 3-17, and Unit 4-17.
    Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised 
the Facility Combined Licenses.
    Date of initial notice in Federal Register: September 3, 2013 (78 
FR 54288).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 28, 2014.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 10th day of February 2014.

    For the Nuclear Regulatory Commission.
Michele. G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2014-03494 Filed 2-18-14; 8:45 am]
BILLING CODE 7590-01-P