Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 47785-47795 [2013-18851]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 78, Number 151 (Tuesday, August 6, 2013)]
[Notices]
[Pages 47785-47795]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2013-18851]


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NUCLEAR REGULATORY COMMISSION

[NRC-2013-0175]


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 July 11, 2013, to July 23, 2013. The last 
biweekly notice was published on July 23, 2013, (78 FR 44167).

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-2013-0175. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422;

[[Page 47786]]

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, 06A44M, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001.
    For additional direction on accessing information and submitting 
comments, see ``Accessing Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

SUPPLEMENTARY INFORMATION: 

I. Accessing Information and Submitting Comments

A. Accessing Information

    Please refer to Docket ID NRC-2013-0175 when contacting the NRC 
about the availability of information regarding this document. You may 
access information related to this document, which the NRC possesses 
and is publicly-available, by the following methods:
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2013-0175.
     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-2013-0175 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 Section 50.92 of Title 10 of the Code 
of Federal Regulations (10 CFR), this means that operation of the 
facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license or 
combined license. Requests for a hearing and a petition for leave to 
intervene shall be filed in accordance with the Commission's ``Agency 
Rules of Practice and Procedure'' in 10 CFR Part 2. Interested 
person(s) should consult a current copy of 10 CFR 2.309, which is 
available at the NRC's PDR, located at One White Flint North, Room O1-
F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The 
NRC regulations are accessible electronically from the NRC Library on 
the NRC's Web site at https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is 
filed by the above date, the Commission or a presiding officer 
designated by the Commission or by the Chief Administrative Judge of 
the Atomic Safety and Licensing Board Panel, will rule on the request 
and/or petition; and the Secretary or the Chief Administrative Judge of 
the Atomic Safety and Licensing Board will issue a notice of a hearing 
or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific

[[Page 47787]]

contentions which the requestor/petitioner seeks to have litigated at 
the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held. If 
the final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards consideration, then any hearing 
held would take place before the issuance of any amendment.
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at hearing.docket@nrc.gov, or by 
telephone at 301-415-1677, to request (1) a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a request or 
petition for hearing (even in instances in which the participant, or 
its counsel or representative, already holds an NRC-issued digital ID 
certificate). Based upon this information, the Secretary will establish 
an electronic docket for the hearing in this proceeding if the 
Secretary has not already established an electronic docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the agency's public Web 
site at https://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with the NRC guidance 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC 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:

[[Page 47788]]

Rulemaking and Adjudications Staff. Participants filing a document in 
this manner are responsible for serving the document on all other 
participants. Filing is considered complete by first-class mail as of 
the time of deposit in the mail, or by courier, express mail, or 
expedited delivery service upon depositing the document with the 
provider of the service. A presiding officer, having granted an 
exemption request from using E-Filing, may require a participant or 
party to use E-Filing if the presiding officer subsequently determines 
that the reason for granting the exemption from use of E-Filing no 
longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
https://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, a request to intervene will require including information on 
local residence in order to demonstrate a proximity assertion of 
interest in the proceeding. With respect to copyrighted works, except 
for limited excerpts that serve the purpose of the adjudicatory filings 
and would constitute a Fair Use application, participants are requested 
not to include copyrighted materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day deadline will not 
be entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the three factors in 10 
CFR 2.309(c)(1)(i)(iii).
Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc., 
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, 
Louisiana
    Date of amendment request: May 28, 2013.
    Description of amendment request: The amendment would modify safety 
limits (SL) in Technical Specification (TS) 2.1.1, ``Reactor Core 
SLs,'' to reduce the minimum reactor dome pressure associated with the 
critical power correlation from 785 pounds per square inch gauge (psig) 
to 685 psig. The RBS has evaluated the critical power correlation for 
the General Electric Nuclear Energy advanced fuel designs (i.e., GE14 
and GNF2 fuels) used at the facility which will allow for a lower-bound 
pressure. The change will provide a greater pressure margin such that 
the reactor remains above the proposed low SL of 685 psig in the event 
of a Pressure Regulator Maximum Demand Open transient.
    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.
    Decreasing the reactor dome pressure limit in TS Safety Limits 
2.1.1 for reactor Rated Thermal Power range effectively expands the 
validity range for the GEXL 14 and GEXL 17 correlations and the 
calculation of Minimum Critical Power Ratio Safety Limit (MCPR). The 
MCPR rises during the pressure reduction following the scram that 
terminates the Pressure Regulator Failure Open (PRFO) transient. Since 
the change does not involve a modification of any plant hardware, the 
probability and consequence of the PRFO transient are essentially 
unchanged. The reduction in the reactor dome pressure safety limit from 
785 psig to 685 psig provides greater margin to accommodate the 
pressure reduction during the transient within the revised TS limit.
    The proposed change will continue to support the validity range for 
the GEXL correlations applied at RBS and the calculation of MCPR as 
approved. The proposed TS revision involves no significant changes to 
the operation of any systems or components in normal, accident or 
transient operating conditions.
    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 reduction in the reactor dome pressure safety limit 
from 785 psig to 685 psig is a change based upon previously approved 
documents and does not involve changes to the plant hardware or its 
operating characteristics. As a result, no new failure modes are being 
introduced.
    Therefore, the change does not introduce a new or different kind of 
accident from those 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, and through the parameters for 
safe operation and setpoints for the actuation of equipment relied upon 
to respond to transients and design basis accidents. The proposed 
change in reactor dome pressure enhances the safety margin, which 
protects the fuel cladding integrity during a depressurization 
transient, but does not change the requirements governing operation or 
availability of safety equipment assumed to operate to preserve the 
margin of safety. The change does not alter the behavior of plant 
equipment, which remains unchanged. The available pressure range is 
expanded by the change, thus offering greater margin for pressure 
reduction during the transient.
    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: Joseph A. Aluise, Associate General 
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Michael T. Markley.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts
    Date of amendment request: April 5, 2013.
    Description of amendment request: The proposed amendment would 
revise the Pilgrim Technical Specifications (TSs) to reduce the reactor 
steam dome pressure from 785 pounds per square inch, gauge (psig) to 
685 psig specified in TS Reactor Core Safety Limits 2.1.1 and 2.1.2. 
The proposed amendment is intended to address the potential to exceed 
the low pressure TS safety limit associated with a pressure regulator 
failure open (PRFO)--maximum

[[Page 47789]]

demand abnormal operation occurrence, as identified by General Electric 
Nuclear Energy in its report, ``10 CFR 21 Reportable Condition 
Notification: Potential to Exceed Low Pressure Technical Specification 
Safety Limit,'' MFN 05-021, dated March 29, 2005.
    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, along with the NRC's edits in 
square brackets:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    Decreasing the reactor dome pressure in Technical Specification 
Safety Limits 2.1.1 and 2.1.2 for reactor Rated Thermal Power ranges 
effectively expands the validity range for GEXL [GE critical quality-
boiling length correlation] and the calculation of Minimum Critical 
Power Ratio Safety Limit (MCPR). MCPR rises during the pressure 
reduction following the scram that terminates the PRFO transient. Since 
the change does not involve a modification of any plant hardware, the 
probability and consequence of the PRFO transient are essentially 
unchanged. The reduction in the reactor dome pressure value in the 
safety limit from 785 psig to 685 psig provides adequate margin to 
accommodate the pressure reduction during the transient within the 
revised TS limit.
    The expanded GEXL correlation range supports Pilgrim's revised low 
pressure safety limit of 685 psig. The proposed TS revision involves no 
significant changes to the operation of any systems or components in 
normal or accident or transient operating conditions.
    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 reduction in the reactor dome pressure value in the 
safety limit from 785 psig to 685 psig reflects a wider range of 
applicability for the GEXL correlation which is approved by the NRC for 
fuels in use at Pilgrim and does not involve changes to the plant 
hardware or its operating characteristics. As a result, no new failure 
modes are being introduced.
    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, and through the parameters for 
safe operation and setpoints for the actuation of equipment relied upon 
to respond to transients and design basis accidents. The proposed 
change in reactor dome pressure restores the safety margin, which 
protects the fuel cladding integrity during a depressurization 
transient, but does not change the requirements governing operation or 
availability of safety equipment assumed to operate to preserve the 
margin of safety. The change does not alter the behavior of plant 
equipment, which remains unchanged. The reduction in the reactor dome 
pressure value in the safety limit from 785 psig to 685 psig provides 
adequate margin to accommodate the pressure reduction during the 
transient within the revised TS limit.
    Therefore, the proposed change does not involve a significant 
reduction in [a] margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Acting Branch Chief: Robert Beall.
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: May 14, 2013.
    Description of amendment request: The proposed amendment would 
revise the Vermont Yankee Technical Specifications (TSs) to reduce 
reactor pressure associated with the fuel cladding integrity safety 
limits (SLs) from 800 pounds per square inch, absolute (psia) to 700 
psia in SLs 1.1.A and 1.1.B. The proposed change is intended to address 
the potential to exceed the low pressure TS SL associated with a 
pressure regulator failure-maximum demand open (PRFO) transient as 
reported by General Electric Nuclear Energy in its Part 21 
Communication, ``Potential to Exceed Low Pressure Technical 
Specification Safety Limit,'' SC05-03, dated March 29, 2005.
    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 reactor pressure in Fuel Cladding 
Integrity Safety Limits 1.1.A and 1.1.B does not alter the use of the 
analytical methods used to determine the safety limits that have been 
previously reviewed and approved by the NRC. The proposed change is in 
accordance with NRC approved critical power correlation methodologies 
and as such maintains required safety margins. The proposed change does 
not adversely affect accident initiators or precursors nor does it 
alter the design assumptions, conditions, or configuration of the 
facility or the manner in which the plant is operated and maintained.
    The proposed change does not alter or prevent the ability of 
structures, systems, and components (SSCs) from performing their 
intended function to mitigate the consequences of an initiating event 
within the assumed acceptance limits. The proposed change does not 
require any physical change to any plant SSCs nor does it require any 
change in systems or plant operations. The proposed change is 
consistent with the safety analysis assumptions and resultant 
consequences.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    There are no hardware changes nor are there any changes in the 
method which any plant systems perform a safety function. No new 
accident scenarios, failure mechanisms, or limiting single failures are 
introduced as a result of the proposed change.
    The proposed change does not introduce any new accident precursors, 
nor does it involve any physical plant alterations or changes in the 
methods governing normal plant operation. Also, the change does not 
impose any new or

[[Page 47790]]

different requirements or eliminate any existing requirements. The 
change does not alter assumptions made in the safety analysis.
    Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction in a 
margin of safety?
    Response: No.
    Margin of safety is related to confidence in the ability of the 
fission product barriers (fuel cladding, reactor coolant system, and 
primary containment) to perform their design functions during and 
following postulated accidents. Evaluation of the 10 CFR Part 21 issue 
by General Electric determined that the PRFO transient provides 
additional margin to the Minimum Critical Power Ratio Safety Limit and 
is not a threat to fuel cladding integrity.
    The proposed change to Fuel Integrity Cladding Safety Limits 1.1.A 
and 1.1.B is consistent with, and within the capabilities of the 
applicable NRC approved critical power correlations, and thus continues 
to ensure that valid critical power calculations are performed. No 
setpoints at which protective actions are initiated are altered by the 
proposed change. The proposed change does not alter the manner in which 
the safety limits are determined. This change is consistent with plant 
design and does not change the TS operability requirements; thus, 
previously evaluated accidents are not affected by this proposed 
change.
    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., 400 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Acting Branch Chief: Robert Beall.
Florida Power and Light Company, et al., Docket No. 50-335, St. Lucie 
Plant, Unit 1, St. Lucie County, Florida
    Date of amendment request: May 10, 2013.
    Description of amendment request: The amendment will revise the 
Technical Specifications (TSs) to allow the use of M5[supreg] fuel rod 
cladding material at St. Lucie Plant, Unit 1. The current acceptable 
fuel rod cladding material is identified in TS 5.3.1, Reactor Core, 
Fuel Assemblies. The proposed change would revise TS 5.3.1 to add 
M5[supreg] to the approved fuel rod cladding materials and TS 6.9.1.11 
to add Framatome (AREVA) topical report BAW-10240(P)(A), Revision 0, 
``Incorporation of M5[supreg] Properties in Framatome ANP Approved 
Methods,'' to the analytical methods used to determine the core 
operating limits previously reviewed and approved by the NRC.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change would allow the use of M5[supreg] fuel rod 
cladding in the St. Lucie Unit 1 reactor. The topical report BAW-
10240(P)--A prepared by Framatome, currently known as AREVA, has been 
approved by the NRC for use with M5[supreg] fuel cladding. The fuel 
cladding itself is not an accident initiator and does not affect 
accident probability. Use of M5[supreg] fuel cladding, which has 
essentially the same properties as currently licensed Zircaloy, has 
been shown to meet all 10 CFR 50.46 acceptance criteria and, therefore, 
will not increase the consequences of an accident.
    The proposed change to Technical Specification 6.9.1.11 (Core 
Operating Limits Report (COLR)) enables the use of the appropriate 
methodology to analyze accidents for cores containing fuel with 
M5[supreg] cladding to ensure that the plant continues to meet 
applicable design criteria and safety analysis acceptance criteria. The 
proposed change to the list of NRC-approved methodologies listed in 
Technical Specification 6.9.1.11 has no impact on plant operation and 
configuration. The list of methodologies in Technical Specification 
6.9.1.11 does not impact either the initiation of an accident or the 
mitigation of its consequences.
    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.
    Use of M5[supreg] clad fuel will not result in changes in the 
operation or configuration of the facility. The material properties of 
M5[supreg] are similar to those of Zircaloy. Therefore, M5[supreg] fuel 
rod cladding will perform similarly to those fabricated from Zircaloy, 
thus precluding the possibility of the fuel becoming an accident 
initiator and causing a new or different type of accident. The proposed 
change to Technical Specification 5.3.1, to add M5[supreg] as a fuel 
clad material, does not create any new accident initiators.
    The proposed change to the list of NRC-approved methodologies 
listed in Technical Specification 6.9.1.11, to add BAW-10240(P)--A, has 
no impact on any plant configuration or system performance. There is no 
change to the parameters within which the plant is normally operated, 
and thus the possibility of a new or different type of accident is not 
created.
    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 will not involve a significant reduction in the 
margin of safety because it has been demonstrated that the material 
properties of the M5[supreg] are not significantly different from those 
of Zircaloy. The M5[supreg] is expected to perform similarly to 
Zircaloy for all normal operating and accident scenarios, including 
both loss of coolant accident (LOCA) and non-LOCA scenarios. For LOCA 
scenarios, plant-specific LOCA analyses using M5[supreg] properties 
demonstrate that the acceptance criteria of 10 CFR 50.46 have been 
satisfied.
    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: James Petro, Managing Attorney--Nuclear, 
Florida Power & Light, P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Branch Chief: Jessie F. Quichocho.

[[Page 47791]]

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo County, 
California
    Date of amendment request: June 6, 2013.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3.7.10, ``Control Room Ventilation 
System (CRVS),'' and TS 5.6.5, ``Core Operating Limits Report (COLR),'' 
to incorporate editorial changes. Specifically, the proposed amendments 
delete footnote (1), which expired on December 10, 2012, and is no 
longer applicable, from TS 3.7.10 Condition A Completion Time, and 
corrects inconsistent wording between TS 5.6.5a.4 and TS 3.2.1, between 
TS 5.6.5a.5, and TS 3.2.2, and between TS 5.6.5a.9 and TS 3.4.1.
    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 change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed editorial changes do not involve any physical changes 
to structures, systems or components. The proposed editorial change to 
TS 3.7.10 deletes a footnote that is no longer applicable. The proposed 
editorial changes to TS 5.6.5 correct administrative discrepancies in 
the TS to provide consistency with the existing TS Sections 3.2.1, 
3.2.2 and 3.4.1.
    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 accident from any accident previously evaluated?
    Response: No.
    The proposed editorial changes to TS 3.7.10 and TS 5.6.5 do not 
involve an accident.
    Therefore, the proposed change does not create the possibility of a 
new or different accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed editorial changes to TS 3.7.10 and TS 5.6.5 do not 
impact accident analyses, fission product barriers, or 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 requests involve no significant hazards consideration.
    Attorney for licensee: Jennifer Post, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Branch Chief: Michael T. Markley.
South Carolina Electric and Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 1, 
Fairfield County, South Carolina
    Date of amendment request: April 3, 2013.
    Description of amendment request: The proposed amendment would add 
an exception to Technical Specification 3.0.4 in Technical 
Specification 3/4.7.6, Control Room Emergency Filtration System 
(CREFS).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change adds an exception to the provisions of 
Specification 3.0.4 in Technical Specification 3/4.7.6, ``Control Room 
Emergency Filtration System (CREFS)'' that was previously included in 
this Technical Specification prior to Amendment 180. The proposed 
change would allow entry into the applicable Modes of Technical 
Specification 3/4.7.6 Actions b.1 and b.2 (Modes 5 and 6) while relying 
on the actions. 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 does not alter the Technical Specification Limiting Condition 
for Operation, Applicability, or remedial Actions that provide for the 
safe operation of the plant when the Limiting Condition for Operation 
is not met. The Actions in Technical Specification 3/4.7.6 Action 
statement b. continue to ensure the safe operation of the plant in the 
same manner as before. In addition, the proposed change does not affect 
the Surveillance Requirements of Technical Specification 3/4.7.6. As 
such, the Surveillance Requirements continue to provide the same level 
of assurance as before that the CREFS and control room boundary will 
perform their required safety functions to mitigate the consequences of 
events within the assumed acceptance limits.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change adds an exception to the provisions of 
Specification 3.0.4 in Technical Specification 3/4.7.6, ``Control Room 
Emergency Filtration System (CREFS)'' that was previously included in 
this Technical Specification prior to Amendment 180. The proposed 
change would allow entry into the applicable Modes of Technical 
Specification 3/4.7.6 Actions b.1 and b.2 (Modes 5 and 6) while relying 
on the actions. The proposed change does not alter the operability 
requirements or remedial Actions of Technical Specification 3/4.7.6, 
nor does the change affect the CREFS or control room boundary function 
during accident conditions. The change does not involve a physical 
alteration of the plant (i.e., no new or different type of equipment 
will be installed) or a significant change in the methods governing 
normal plant operation. The change does not alter assumptions made in 
the applicable safety analyses. As such, the proposed change does not 
impact the safety analyses assumptions and is consistent with current 
plant operating practices.
    Therefore, the proposed TS change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change adds an exception to the provisions of 
Specification 3.0.4 in Technical Specification 3/4.7.6, ``Control Room 
Emergency Filtration System (CREFS)''

[[Page 47792]]

that was previously included in this Technical Specification prior to 
Amendment 180. The proposed change would allow entry into the 
applicable Modes of Technical Specification 3/4.7.6 Actions b.1 and b.2 
(Modes 5 and 6) while relying on the actions. The proposed change 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 the change. 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 shutdown the plant and to maintain the 
plant in a safe shutdown condition. As such, the CREFS and control room 
boundary will continue to provide the same level of safety as before.
    Therefore, the proposed TS change does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: J. Hagood Hamilton, Jr., South Carolina 
Electric & Gas Company, Post Office Box 764, Columbia, South Carolina 
29218.
    NRC Branch Chief: Robert J. Pascarelli.
Southern Nuclear Operating Company, Inc. Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia
    Date of amendment request: June 19, 2013.
    Description of amendment request: The proposed changes would amend 
Combined License numbers NPF-91 and NPF-92 for Vogtle Electric 
Generating Plant Units 3 and 4 by departing from the plant-specific 
design control document Tier 2 and Tier 2* material contained within 
the updated final safety analysis report (UFSAR) related to the design 
of structural wall modules used to construct containment internal 
structures and portions of the auxiliary building. The proposed changes 
would revise requirements for design spacing of shear studs and the 
design of structural elements in order to address interferences and 
obstructions other than wall openings.
    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

[[Page 47793]]

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: Mr. M. Stanford Blanton, Blach & Bingham 
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Lawrence Burkhart.
ZionSolutions LLC, Docket Nos. 50-295 and 50-304, Zion Nuclear Power 
Station (ZNPS), Units 1 and 2, Lake County, Illinois
    Date of amendment request: June 18, 2012, and supplemented June 5, 
2013.
    Description of amendment request: The proposed amendments would 
revise the Physical Security Plan associated with the transfer and 
storage of spent fuel at the Independent Spent Fuel Storage 
Installation (ISFSI).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment, which incorporates ISFSI security 
functions, does not reduce the ability of the Security organization to 
prevent attempts of radiological sabotage and, therefore, does not 
increase the probability or consequences of a radiological release 
previously evaluated. The proposed ZNPS ISFSI Physical Security Plan 
will not affect any important-to-safety systems or components, their 
mode of operation or operating strategies. The changes have no effect 
on accident initiators or mitigation.
    Therefore, the proposed amendment will not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment incorporating ISFSI security functions does 
not affect the operation of systems that are important-to-safety. The 
ZNPS ISFSI Physical Security Plan amendment does not affect any of the 
parameters or conditions that could contribute to the initiation of any 
accident. No new accident scenarios are created as a result of the ZNPS 
ISFSI Physical Security Plan. In addition, the design functions of 
equipment important to safety are not altered as a result of the 
proposed ZNPS ISFSI Physical Security Plan.
    Therefore, the proposed ISFSI Security Plan will not create the 
possibility of a new or different accident from any previously 
evaluated.
    3. Does the change involve a significant reduction in a margin of 
safety?
    Response: No.
    Implementation of the proposed amendment incorporating ISFSI 
security functions will not reduce a margin of safety as detailed in 
the Technical Specifications, as there are no Technical Specification 
requirements associated with the physical security system. 
Specifically, the proposed ZNPS ISFSI Physical Security Plan does not 
represent a change in initial conditions, system response time, or any 
other parameter affecting the course of an accident analysis supporting 
the Bases of any Technical Specification. The proposed amendment does 
not reduce the effectiveness of any security/safeguards measures 
currently in place at the ZNPS.
    Therefore, the proposed ZNPS ISFSI Physical Security Plan will 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: Russ Workman, Deputy General Counsel, 
EnergySolutions, 423 West 300 South, Suite 200, Salt Lake City, UT 
84101.
    NRC Branch Chief: Bruce Watson.
Notice of Issuance of Amendments to Facility Operating Licenses and 
Combined Licenses
    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    A notice of consideration of issuance of amendment to facility 
operating license or combined license, as applicable, proposed no 
significant hazards consideration determination, and opportunity for a 
hearing in connection with these actions, was published in the Federal 
Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland 20852. Publicly available documents created or received at the 
NRC are accessible electronically through the Agencywide Documents 
Access and Management System (ADAMS) in the NRC Library at https://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are

[[Page 47794]]

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. 50-341, Fermi 2, Monroe County, 
Michigan
    Date of application for amendment: December 21, 2012.
    Brief description of amendment:
    The amendment revises Fermi 2 operating license to change its name 
on the license to ``DTE Electric Company.'' This name change is purely 
administrative in nature. Detroit Edison is a wholly owned subsidiary 
of DTE Energy Company, and this name change is part of a set of name 
changes of DTE Energy subsidiaries to conform their names to the 
``DTE'' brand name. No other changes are contained within this 
amendment. This change does not involve a transfer of control over or 
of an interest in the license for Fermi 2.
    Date of issuance: July 12, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 193.
    Facility Operating License No. NPF-43: Amendment revised the 
operating license.
    Date of initial notice in Federal Register: March 4, 2013 (78 FR 
14131).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 12, 2013.
    No significant hazards consideration comments received: No.
Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414, 
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina; 
and Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 
2, Mecklenburg County, North Carolina
    Date of amendment request: January 21, 2013.
    Description of amendment request: The amendments revised the 
divider barrier seal test coupons' tensile strength in Technical 
Specification Surveillance Requirement 3.6.14.4 from ``> 39.7 psi'' to 
``> 39.7 lbs.'' This change is an administrative change to correct an 
error where the wrong units were used when Catawba and McGuire 
converted to Standard Technical Specifications in 1998 using NUREG-
1431, Revision 1.
    Date of issuance: July 16, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 270, 266, 270 and 250.
    Renewed Facility Operating License Nos. NPF-35, NPF-52, NPF-9 and 
NPF-17: Amendments revised the licenses and the technical 
specifications.
    Date of initial notice in Federal Register: May 14, 2013 (78 FR 
28251).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 16, 2013.
    No significant hazards consideration comments received: No.
Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, 
Comanche Peak Nuclear Power Plant, Units 1 and 2, Somervell County, 
Texas
    Date of amendment request: July 12, 2012, as supplemented by letter 
dated October 23, 2012.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 5.7.1, ``High Radiation Areas with Dose Rates not 
Exceeding 1.0 rem [roentgen equivalent man]/hour at 30 Centimeters from 
the Radiation Source or from any Surface Penetrated by the Radiation,'' 
and 5.7.2, ``High Radiation Areas with Dose Rates Greater than 1.0 rem/
hour at 30 Centimeters from the Radiation Source or from any Surface 
Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter 
from the Radiation Source or from any Surface Penetrated by the 
Radiation,'' to allow entry into high radiation areas by personnel 
continuously escorted by individuals qualified in radiation protection 
procedures and to require a pre-job briefing prior to entry into such 
areas. In addition, the amendment incorporates an editorial change to 
TS Table 3.3.3-1, ``Post Accident Monitoring Instrumentation.'' The 
typographical error in the title of TS Table 3.3.1-1 column ``CONDITION 
REFERENCED FROM REQUIRED ACTION E.1,'' is corrected to read, 
``CONDITION REFERENCED FROM REQUIRED ACTION D.1,'' to reflect that the 
Required Actions for Condition D of TS 3.3.3, ``Post Accident 
Monitoring (PAM) Instrumentation'' are listed in the table.
    Date of issuance: July 11, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: Unit 1--159; Unit 2--159.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: November 13, 2012 (77 
FR 67683).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 11, 2013.
    No significant hazards consideration comments received: No.
Northern States Power Company--Minnesota, Docket No. 50-263, Monticello 
Nuclear Generating Plant, Wright County, Minnesota
    Date of application for amendment: September 18, 2012.
    Brief description of amendment: The amendment revises the MNGP 
Technical Specifications (TS) Sections 3.1.6, ``Rod Pattern Control,'' 
and 3.3.2.1, ``Control Rod Block Instrumentation,'' to allow MNGP to 
reference an optional Banked Position Withdrawal Sequence (BPWS) 
shutdown sequence in the TS Bases. In addition, a footnote is revised 
in TS Table 3.3.2.1-1, ``Control Rod Block Instrumentation,'' to allow 
operators to bypass the rod worth minimizer if conditions for the 
optional BPWS shutdown process are satisfied. The changes are 
consistent with NRC-approved Technical Specifications Task Force (TSTF) 
Improved Standard Technical Specifications Change Traveler, TSTF-476, 
Revision 1, ``Improved BPWS Control Rod Insertion Process (NEDO-
33091).''
    Date of issuance: July 15, 2013.
    Effective date: This license amendment is effective as of the date 
of its date of issuance and shall be implemented within 180 days after 
start-up from the 2013 Refueling Outage.
    Amendment No.: 173.
    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.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 15, 2013.
    No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Units 1 and 2, Salem County, New Jersey
    Date of application for amendments: July 17, 2012, as supplemented 
on January 28, 2013, and March 22, 2013.
    Brief description of amendments: The amendment revised Salem 
Nuclear Generating Station Technical Specification 3.7.6.1 (Unit 1) and 
3.7.6 (Unit 2), ``Control Room Emergency Air Conditioning System,'' to 
eliminate the separate action statements for securing an inoperable 
Control Area Air Conditioning System and Control Room

[[Page 47795]]

Emergency Air Conditioning System isolation damper in the closed 
position and entering the actions for an inoperable control room 
envelope boundary.
    Date of issuance: July 17, 2013.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment Nos.: 304 and 286.
    Renewed Facility Operating License Nos. DPR-70 and DPR-75: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: April 2, 2013 (78 FR 
19754).
    The supplemental letter dated March 22, 2013, provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination or expand the application.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 17, 2013.
    No significant hazards consideration comments received: No.
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: August 14, 2012, as supplemented by 
letters dated February 28, April 19, and June 24, 2013.
    Brief description of amendment request: The amendments revised 
Technical Specification (TS) 5.6.5, ``Core Operating Limits Report 
(COLR),'' to reference and allow use of Westinghouse WCAP-16045-P-A, 
Addendum 1-A, ``Qualification of the NEXUS Nuclear Data Methodology,'' 
(Reference 1 of Enclosure 1) to determine core operating limits. The 
non-proprietary version is WCAP-16045-NP-A, Addendum 1-A (Reference 2 
of Enclosure 1).
    Date of issuance: July 17, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos: 191 and 187.
    Facility Operating License Nos. NPF-2 and NPF-8: The amendment 
revised the Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: October 9, 2012 (77 FR 
61440).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 17, 2013.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 29th day of July, 2013.
    For The Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2013-18851 Filed 8-5-13; 8:45 am]
BILLING CODE 7590-01-P