Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 28248-28258 [2013-11272]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 78, Number 93 (Tuesday, May 14, 2013)]
[Notices]
[Pages 28248-28258]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2013-11272]


-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION

[NRC-2013-0084]


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 (the 
Commission or NRC) is publishing this regular biweekly notice. The Act 
requires the

[[Page 28249]]

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 April 18, 2013 to May 1, 2013. The last 
biweekly notice was published on April 30, 2013 (78 FR 25310).

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-2013-0084. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; 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: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
     Fax comments to: RADB at 301-492-3446.
    For additional direction on accessing information and submitting 
comments, see ``Accessing Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

SUPPLEMENTARY INFORMATION:

I. Accessing Information and Submitting Comments

A. Accessing Information

    Please refer to Docket ID NRC-2013-0084 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-2013-0084.
     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-2013-0084 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

[[Page 28250]]

Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held. If 
the final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards consideration, then any hearing 
held would take place before the issuance of any amendment.
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC 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

[[Page 28251]]

contacting the NRC Meta System Help Desk through the ``Contact Us'' 
link located on the NRC's Web site at https://www.nrc.gov/site-help/e-submittals.html, by email at MSHD.Resource@nrc.gov, or by a toll-free 
call at 1-866 672-7640. The NRC Meta System Help Desk is available 
between 8 a.m. and 8 p.m., Eastern Time, Monday through Friday, 
excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, 
Maryland, 20852, Attention: Rulemaking and Adjudications Staff. 
Participants filing a document in this manner are responsible for 
serving the document on all other participants. Filing is considered 
complete by first-class mail as of the time of deposit in the mail, or 
by courier, express mail, or expedited delivery service upon depositing 
the document with the provider of the service. A presiding officer, 
having granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
https://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, a request to intervene will require including information on 
local residence in order to demonstrate a proximity assertion of 
interest in the proceeding. With respect to copyrighted works, except 
for limited excerpts that serve the purpose of the adjudicatory filings 
and would constitute a Fair Use application, participants are requested 
not to include copyrighted materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day deadline will not 
be entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the following three 
factors in 10 CFR 2.309(c)(1): (i) The information upon which the 
filing is based was not previously available; (ii) the information upon 
which the filing is based is materially different from information 
previously available; and (iii) the filing has been submitted in a 
timely fashion based on the availability of the subsequent information.
    For further details with respect to this license amendment 
application, see the application for amendment which is available for 
public inspection at the NRC's PDR, located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. Publicly available documents created or received at the NRC are 
accessible electronically through ADAMS in the NRC Library at https://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to 
ADAMS or who encounter problems in accessing the documents located in 
ADAMS, should contact the NRC PDR Reference staff at 1-800-397-4209, 
301-415-4737, or by email to pdr.resource@nrc.gov.

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 would revise 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.
    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.
    Divider barrier integrity is necessary to minimize bypassing of 
the ice condenser by the hot steam and air mixture released into the 
lower compartment during a Design Basis Accident (DBA). This ensures 
that most of the gases pass through the ice bed, which condenses the 
steam and limits pressure and temperature during the accident 
transient. Limiting the pressure and temperature reduces the release 
of fission product radioactivity from containment to the environment 
in the event of a DBA.
    Conducting periodic physical property tests on divider barrier 
seal test coupons provides assurance that the seal material has not 
degraded in the containment environment, including the effects of 
irradiation with the reactor at power. The proposed change to 
Technical Specification Surveillance Requirement 3.6.14.4 results in 
the correct tensile strength units being listed in this surveillance 
requirement. This is considered an editorial change to the Technical 
Specifications.
    Thus, based on the above, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve a change in the operational 
limits or the design capabilities of the containment or containment 
systems. The proposed change does not change the function or 
operation of plant equipment or introduce any new failure 
mechanisms. The technical evaluation that supports this License 
Amendment Request included a review of the containment divider 
barrier seal capability to which this change is bounded. The 
proposed change does not introduce any new or different types of 
failure mechanisms; plant equipment will continue to respond as 
designed and analyzed.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their design functions 
during and following an accident situation. These barriers include 
the fuel cladding, the reactor coolant system, and the containment 
system. The performance of the fuel cladding, the reactor coolant 
system and the containment system will not be adversely impacted by 
the proposed change since the ability of the divider barrier to 
mitigate an analyzed accident has not been adversely impacted by the 
proposed change.
    Thus, it is concluded that 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

[[Page 28252]]

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.

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

    Date of amendment request: April 9, 2013.
    Description of amendment request: The proposed amendment would 
delete certain reporting requirements contained in the Technical 
Specifications.
    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 changes do not involve the modification of any 
plant equipment or affect plant operation. The proposed changes will 
have no impact on any safety related structures, systems, or 
components. The reporting requirements proposed for deletion are not 
required because the requirements are adequately addressed by 10 CFR 
50.72 and 10 CFR 50.73, or other regulatory requirements, or are 
available on site for NRC review, and are no longer warranted.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes have no impact on the design, function or 
operation of any plant structure, system or component. The proposed 
changes do not affect plant equipment or accident analyses. The 
reporting requirements proposed for deletion are not required 
because the requirements are adequately addressed by 10 CFR 50.72 
and 10 CFR 50.73, or other regulatory requirements, or are available 
on site for NRC review, and are no longer warranted.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes do not adversely affect existing plant 
safety margins or the reliability of the equipment assumed to 
operate in the safety analyses. There is no change being made to 
safety analysis assumptions, safety limits or limiting safety system 
settings that would adversely affect plant safety as a result of the 
proposed changes. Margins of safety are unaffected by deletion of 
the reporting requirements.
    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: J. Bradley Fewell, Esquire, Associate 
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Meena K. Khanna.

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

    Date of amendment request: November 30, 2012.
    Description of amendment request: The proposed amendment would 
revise the Emergency Plan to remove references to the backup plant vent 
extended range noble gas radiation monitoring (R45) indication, 
recording, and alarm capability in the emergency response facilities. 
The R45 indicators have become obsolete and unreliable. The R45 is a 
backup to the R41 for plant vent intermediate and high range noble gas 
radiation monitoring indicators. The accident sampling function of the 
R45 will be maintained.
    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 plant vent noble gas indicators are not an initiator of or a 
precursor to any accident or transient. The proposed change to the 
Emergency Plan to delete the backup (R45) noble gas indicators does 
not impact any design function of the Salem Radiation Monitoring 
System. The backup (R45) plant vent radiation monitors do not 
perform any accident mitigating function. The modification of the 
R45 noble gas indicators does not alter or modify the function of 
systems used to mitigate the consequences of any design basis 
accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes to the Emergency Plan to delete the backup 
plant vent noble gas indicators (R45) does not introduce any new 
accident precursors and does not involve any physical plant 
alterations or changes in the methods governing normal plant 
operation that could initiate a new or different kind of accident. 
The R45 noble gas indicators only provide indication of the effluent 
release through the plant vent release path.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Margin of safety is related to the ability of the fission 
product barriers (fuel cladding, reactor coolant system, and primary 
containment) to perform their design functions during and following 
postulated accidents. The proposed amendment does not alter 
setpoints or limits established or assumed by the accident analyses. 
The R45 plant vent radiation monitor provides indication only. The 
elimination of the backup R45 noble gas indicator does not reduce 
the margin of safety since the remaining R41 noble gas indicator 
will continue to provide the accident indication capability. The 
accident sampling capability of the R45 will remain.
    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: Jeffrie J. Keenan, PSEG Nuclear LLC--N21, 
P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Meena K. Khanna.

R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of amendment request: February 28, 2013.
    Description of amendment request: The proposed amendment would 
revise Technical Specification Section 3.6.5 by adding a different 
limitation on the

[[Page 28253]]

containment average air temperature. The revised Technical 
Specification Section 3.6.5 would read as follows:
    ``Containment average air temperature shall be 
<125[emsp14][deg]F.''
    To support this proposed change, the licensee revised the accident 
analyses that were impacted by the increase in initial containment air 
temperature or increase in safety injection accumulator temperature, 
which are located in the Ginna containment, and are expected to be at 
the same temperature as containment air. The impact of the change in 
the containment air temperature was addressed by revising the Loss of 
Coolant Accident (LOCA) and a Main Steam Line break containment 
response analyses. The change in SI accumulator temperature was 
reflected in the re-evaluated core response to a large break LOCA 
(LBLOCA) and a small break LOCA. The combined impact on the post-LOCA 
long term cooling analyses was also re-assessed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to increase the containment average air 
temperature limit to 125[emsp14][deg]F, from 120[emsp14][deg]F, does 
not alter the assumed initiators to any analyzed event. The 
probability of an accident previously evaluated will not be 
increased by this proposed change. This proposed change will not 
affect radiological dose consequence analyses. The radiological dose 
consequence analyses assume a certain containment atmosphere leak 
rate based on the maximum allowable containment leakage rate, which 
is not affected by the change in allowable average containment air 
temperature resulting in a higher calculated peak containment 
pressure. The 10 CFR Part 50, Appendix J containment leak rate 
testing program will continue to ensure that containment leakage 
remains within the leakage rate assumed in the offsite dose 
consequence analyses. The acceptable leakage corresponds to the peak 
allowable containment pressure of 60 psig. The radiological dose 
consequence analyses assume a certain source term, which is not 
affected by the change in allowable average containment air 
temperature. All core limitations set forth in 10 CFR 50.46 continue 
to be met. The consequences of an accident previously evaluated will 
not be increased by this proposed change.
    Therefore, operation of the facility in accordance with the 
proposed change to the containment average air temperature limit 
will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change provides for a higher allowable containment 
average air temperature to that currently in the TS Section 3.6.5. 
The calculated peak containment temperature and pressure remain 
below the containment design temperature and pressure of 
286[emsp14][deg]F and 60 psig. This change does not involve any 
alteration in the plant configuration (no new or different type of 
equipment will be installed) or make changes in the methods 
governing normal plant operation. The change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Therefore, operation of the facility in accordance with the 
proposed change to TS Section 3.6.5 would not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The calculated peak containment pressure and temperature remain 
below the containment design pressure and temperature of 60 psig and 
286[emsp14][deg]F, respectively. The penalties applied to the BE 
[best estimate] LBLOCA analysis result in the limitations set forth 
in 10 CFR 50.46 continuing to be met. Since the radiological 
consequence analyses are based on the maximum allowable containment 
leakage rate, which is not being revised, the change in the 
calculated peak containment pressure and temperature and changes in 
core response do not represent a significant change in the margin of 
safety. The longterm impact of the peak containment temperature 
following a design basis accident exceeding the EQ profile by 
2[emsp14][deg]F with respect to the current licensing basis is 
negligible.
    Therefore, operation of the facility in accordance with the 
proposed change to increase the allowable containment average air 
temperature from 120[emsp14][deg]F to 125[emsp14][deg]F 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: Carey Fleming, Sr. Counsel--Nuclear 
Generation, Constellation Group, LLC, 750 East Pratt Street, 17 Floor, 
Baltimore, MD 21202.
    NRC Branch Chief: Sean Meighan, Acting.

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

    Date of amendment request: January 23, 2013.
    Description of amendment request: The proposed change would revise 
Technical Specification (TS) Section 5.5.9, ``Steam Generator (SG) 
Program,'' 5.6.10, ``Steam Generator Tube Inspection Report,'' and the 
Steam Generator Tube Integrity specification (LCO 3.4.17). The proposed 
changes are needed to address implementation issues associated with the 
inspection periods, and address other administrative changes and 
clarifications.
    The proposed amendment is consistent with TSTF-510, Revision 2, 
``Revision to Steam Generator Program Inspection Frequencies and Tube 
Sample Selection.''
    In addition, this proposed amendment corrects the indenting for FNP 
TS Section 5.5.9.a at the top of page 5.5-6. This change is purely 
administrative, and has no technical impact on the TS.
    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 license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response: No.
    The proposed change revises the Steam Generator (SG) Program to 
modify the frequency of verification of SG tube integrity and SG 
tube sample selection. A steam generator tube rupture (SGTR) event 
is one of design basis accidents that are analyzed as part of a 
plant's licensing basis. The proposed SG tube inspection frequency 
and sample selection criteria will continue to ensure that the SG 
tubes are inspected such that the probability a SGTR is not 
increased. The consequences of a SGTR are bounded by the 
conservative assumptions in the design accident analysis. The 
proposed change will not cause the consequences of a SGTR to exceed 
those assumptions.
    Therefore, it is concluded that this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No.
    The proposed changes to the SG Program will not introduce any 
adverse changes to the plant design basis or postulated accidents 
resulting from potential tube degradation. The proposed change does 
not affect the

[[Page 28254]]

design of the SGs or their method of operation. In addition, the 
proposed change does not impact any other plant system or component.
    Therefore, it is concluded that this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The SG tubes in pressurized water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
also isolate the radioactive fission products in the primary coolant 
from the secondary system. In summary, the safety function of a SG 
is maintained by ensuring the integrity of its tubes. Steam 
generator tube integrity is a function of the design, environment, 
and the physical condition of the tube. The proposed change does not 
affect tube design or operating environment. The proposed change 
will continue to require monitoring of the physical condition of the 
SG tubes such that there will not be a reduction in the margin of 
safety compared to the current requirements.
    Therefore, it is concluded that the proposed change does not 
involve a significant reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed change 
presents no significant hazards consideration under the standards 
set forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no 
significant hazards consideration'' is justified.

    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: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Branch Chief: Robert J. Pascarelli.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, 
Georgia

    Date of amendment request: January 23, 2013.
    Description of amendment request: The proposed change would revise 
Technical Specification Section 5.5.9, ``Steam Generator (SG) 
Program,'' 5.6.10, ``Steam Generator Tube Inspection Report,'' and the 
Steam Generator Tube Integrity specification (LCO 3.4.17). The proposed 
changes are needed to address implementation issues associated with the 
inspection periods, and address other administrative changes and 
clarifications.
    The proposed amendment is consistent with TSTF-510, Revision 2, 
``Revision to Steam Generator Program Inspection Frequencies and Tube 
Sample Selection.''
    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 license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response: No.
    The proposed change revises the Steam Generator (SG) Program to 
modify the frequency of verification of SG tube integrity and SG 
tube sample selection. A steam generator tube rupture (SGTR) event 
is one of design basis accidents that are analyzed as part of a 
plant's licensing basis. The proposed SG tube inspection frequency 
and sample selection criteria will continue to ensure that the SG 
tubes are inspected such that the probability a SGTR is not 
increased. The consequences of a SGTR are bounded by the 
conservative assumptions in the design accident analysis. The 
proposed change will not cause the consequences of a SGTR to exceed 
those assumptions.
    Therefore, it is concluded that this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No.
    The proposed changes to the SG Program will not introduce any 
adverse changes to the plant design basis or postulated accidents 
resulting from potential tube degradation. The proposed change does 
not affect the design of the SGs or their method of operation. In 
addition, the proposed change does not impact any other plant system 
or component.
    Therefore, it is concluded that this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The SG tubes in pressurized water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
also isolate the radioactive fission products in the primary coolant 
from the secondary system. In summary, the safety function of a SG 
is maintained by ensuring the integrity of its tubes. Steam 
generator tube integrity is a function of the design, environment, 
and the physical condition of the tube. The proposed change does not 
affect tube design or operating environment. The proposed change 
will continue to require monitoring of the physical condition of the 
SG tubes such that there will not be a reduction in the margin of 
safety compared to the current requirements.
    Therefore, it is concluded that the proposed change does not 
involve a significant reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed change 
presents no significant hazards consideration under the standards 
set forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no 
significant hazards consideration'' is justified.

    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. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street NE., Atlanta, 
Georgia 30308-2216.
    NRC Branch Chief: Robert Pascarelli.

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: March 25, 2013.
    Description of amendment request: The proposed change would amend 
Combined Licenses Nos.: NPF-91 and NPF-92 for Vogtle Electric 
Generating Plant (VEGP) Units 3 and 4 by departing from VEGP Units 3 
and 4 Updated Final Safety Analysis Report (UFSAR) Tier 2* material by 
revising reference document APP-OCS-GEH-120, ``AP1000 Human Factors 
Design Engineering Verification Plan,'' from Revision B to Revision 0. 
APP-OCS-GEH-120 is incorporated by reference in the updated UFSAR as a 
means to implement the activities associated with the human factors 
engineering verification and validation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?

[[Page 28255]]

    Response: No.
    Design verification provides a final check of the adequacy of 
the Human System Interface (HSI) Resources and Operation and Control 
Centers System (OCS) design. The changes do not affect the plant 
itself, and so there is no change to the probability or consequences 
of an accident previously evaluated. Changing the design 
verification 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 simply to 
verify implementation of design criteria. The Probabilistic Risk 
Assessment is not affected. 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 safety-related SSC or function used to 
mitigate an accident, the consequences of the accidents evaluated in 
the UFSAR are not affected.
    Therefore, there is no significant increase in the probability 
or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    Design verification provides a final check of the adequacy of 
the HSI Resources and Operation and Control Centers System design. 
The changes do not affect the plant itself, and so there is no new 
or different kind of accident from any accident previously 
evaluated. Therefore, the changes do not affect safety-related 
equipment, nor does it affect equipment which, if it failed, could 
initiate an accident or a failure of a fission product barrier. No 
analysis is adversely affected. No system or design function or 
equipment qualification is adversely affected by the changes. This 
activity will not allow for a new fission product release path, nor 
will it result in a new fission product barrier failure mode, nor 
create a new sequence of events that would result in significant 
fuel cladding failures. In addition, the changes do not result in a 
new failure mode, malfunction or sequence of events that could 
affect safety or safety-related equipment.
    Therefore, this activity does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The changes to the design verification plan provide a final 
check of the adequacy of the HSI Resources and Operation and Control 
Centers System design. 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/
criterion is challenged or exceeded by the requested change.
    Therefore, there is no significant reduction in a margin of 
safety.

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

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

    Date of amendment request: March 25, 2013.
    Description of amendment request: The proposed change would amend 
Combined Licenses Nos.: NPF-91 and NPF-92 for Vogtle Electric 
Generating Plant (VEGP) Units 3 and 4 by departing from VEGP Units 3 
and 4 Updated Final Safety Analysis Report (UFSAR) Tier 2* material by 
revising reference document APP-OCS-GEH-220, ``AP1000 Human Factors 
Engineering Task Support Verification Plan,'' from Revision B to 
Revision 0. APP-OCS-GEH-220 is incorporated by reference in the updated 
final safety analysis report (UFSAR) as a means to implement the 
activities associated with the human factors engineering verification 
and validation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The HFE Task Support Verification Plan is one of several 
verification and validation (V&V) activities performed on human-
system interface (HSI) resources and the Operation and Control 
Centers System (OCS), where applicable. The Task Support 
Verification Plan is used to assess and verify displays and 
activities related to normal and emergency operation. The changes 
are to the Task Support Verification Plan to clarify the scope and 
amend the details of the methodology. The Task Support Verification 
Plan does not affect the plant itself. Changing the Plan does not 
affect prevention and mitigation of abnormal events, e.g., 
accidents, anticipated operational occurrences, earthquakes, floods 
and turbine missiles, or their safety or design analyses. The 
Probabilistic Risk Assessment is not affected. 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 safety-related SSC 
or function used to mitigate an accident, the consequences of the 
accidents evaluated in the UFSAR are not affected.
    Therefore, there is no significant increase in the probability 
or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The changes to the Task Support Verification Plan change 
information related to validation and verification on Human System 
Interface and Operational Control Centers. Therefore, the changes do 
not affect the safety-related equipment itself, nor do they affect 
equipment which, if it failed, could initiate an accident or a 
failure of a fission product barrier. No analysis is adversely 
affected. No system or design function or equipment qualification 
will be adversely affected by the changes. This activity will not 
allow for a new fission product release path, nor will it result in 
a new fission product barrier failure mode, nor create a new 
sequence of events that would result in significant fuel cladding 
failures. In addition, the changes do not result in a new failure 
mode, malfunction or sequence of events that could affect safety or 
safety-related equipment.
    Therefore, this activity does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The changes to the Task Support Verification Plan affect the 
validation and verification on the Human System Interface and the 
Operational Control Centers. Therefore, the changes do not affect 
the plant itself. These changes do not affect the design or 
operation of safety-related equipment or equipment whose failure 
could initiate an accident, nor does it adversely interface with 
safety-related equipment or fission product barriers. No safety 
analysis or design basis acceptance limit/criterion is challenged or 
exceeded by the requested change.
    Therefore, 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.

[[Page 28256]]

    Attorney for licensee: Mr. M. Stanford Blanton, Blach & Bingham 
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    Acting NRC Branch Chief: Lawrence Burkhart.

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

    Date of amendment request: April 5, 2013.
    Description of amendment request: The proposed change would amend 
Combined Licenses Nos.: NPF-91 and NPF-92 for Vogtle Electric 
Generating Plant (VEGP) Units 3 and 4 by departing from VEGP Units 3 
and 4 Updated Final Safety Analysis Report (UFSAR) Tier 2* material by 
revising reference document APP-OCS-GEH-420, ``AP1000 Human Factors 
Engineering Discrepancy Resolution Process,'' from Revision B to 
Revision 0. APP-OCS-GEH-420 is incorporated by reference in the UFSAR 
as a means to implement the activities associated with the human 
factors engineering verification and validation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The HFE Discrepancy Resolution Process is used to capture and 
resolve Human Engineering Discrepancies (HEDs) identified during the 
Human Factors Engineering (HFE) verification and validation (V&V) 
activities. These discrepancy resolution process activities are used 
to support the final check of the adequacy of the HFE design of the 
Human-System Interface (HSI) resources and the Operation and Control 
Centers Systems (OCS) design. The discrepancy resolution process 
activities are performed as part of the V&V activities against the 
final configuration and control documentation, simulator or 
installed target system. The changes are to the Discrepancy 
Resolution Process to clarify the scope and amend the details of the 
methodology. The Discrepancy Resolution Process does not affect the 
plant itself. Changing the Discrepancy Resolution Process does not 
affect prevention and mitigation of abnormal events, e.g., 
accidents, anticipated operational occurrences, earthquakes, floods 
and turbine missiles, or their safety or design analyses. No safety-
related structure, system, component (SSC) or function is adversely 
affected. The document revision does not involve nor interface with 
any SSC accident initiator or initiating sequence of events, and 
thus the probabilities of the accidents evaluated in the Updated 
Final Safety Analysis Report (UFSAR) are not affected. Because the 
changes do not involve any safety-related SSC or function used to 
mitigate an accident, the consequences of the accidents evaluated in 
the UFSAR are not affected.
    Therefore, there is no significant increase in the probability 
or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The changes to the Discrepancy Resolution Process information 
are related to discrepancy resolution of HEDs during the HFE V&V 
activities on the HSI and the OCS. Therefore, the changes do not 
affect the safety-related equipment itself, nor do they affect 
equipment which, if it failed, could initiate an accident or a 
failure of a fission product barrier. No analysis is adversely 
affected. No system or design function or equipment qualification 
will be adversely affected by the changes. This activity will not 
allow for a new fission product release path, nor will it result in 
a new fission product barrier failure mode, nor create a new 
sequence of events that would result in significant fuel cladding 
failures. In addition, the changes do not result in a new failure 
mode, malfunction or sequence of events that could affect safety or 
safety-related equipment.
    Therefore, this activity does not create the possibility of a 
new or different kind of accident than any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The changes to the Discrepancy Resolution Process affect 
discrepancy resolution of HEDs during the HFE V&V activities on the 
HSI and the OCS. Therefore, the changes do not affect the 
assessments or the plant itself. These changes do not affect the 
design or operation of safety-related equipment or equipment whose 
failure could initiate an accident, nor does it adversely interface 
with safety-related equipment or fission product barriers. No safety 
analysis or design basis acceptance limit/criterion is challenged or 
exceeded by the requested change.
    Therefore, the changes do not involve a significant reduction in 
a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. M. Stanford Blanton, Blach & Bingham 
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    Acting NRC Branch Chief: Lawrence Burkhart.
Notice of Issuance of Amendments to Facility Operating Licenses and 
Combined Licenses
    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    A notice of consideration of issuance of amendment to facility 
operating license or combined license, as applicable, proposed no 
significant hazards consideration determination, and opportunity for a 
hearing in connection with these actions, was published in the Federal 
Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the NRC's Public Document Room (PDR), located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. Publicly available documents created or received at the NRC are 
accessible electronically through the Agencywide Documents Access and 
Management System (ADAMS) in the NRC Library at https://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
PDR's Reference staff at 1-800-397-4209, 301-415-4737 or by email to 
pdr.resource@nrc.gov.

[[Page 28257]]

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

    Date of amendment request: April 17, 2012.
    Brief description of amendment: The amendment revised the VYNPS 
Technical Specification (TS) 3.5.A.5 and TS 4.5.A.5 to change the 
normal position of the recirculation pump discharge bypass valves from 
``open'' to ``closed''; and therefore, the valves' safety function to 
close in support of accident mitigation is eliminated. The amendment 
also revised the TSs to require the valves to remain closed and their 
position to be verified once per operating cycle.
    Date of Issuance: April 26, 2013.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 257.
    Facility Operating License No. DPR-28: The amendment revised the 
License and TSs.
    Date of initial notice in Federal Register: October 2, 2012 (77 FR 
60150).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated April 26, 2013.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: February 22, 2012, and 
supplemented by letter dated.
    March 8, 2013.
    Brief description of amendment: FirstEnergy Nuclear Operating 
Company, the licensee for the Perry Nuclear Power Plant Unit 1 (PNPP), 
requested a license amendment to revise PNPP's Technical Specifications 
(TS) 3.10.1, and the associated TS Bases, to expand its scope to 
include provisions for temperature excursions greater than 200 degrees 
Fahrenheit ([deg]F) as a consequence of inservice leak and hydrostatic 
testing, and as a consequence of scram time testing initiated in 
conjunction with an inservice leak or hydrostatic test, while 
considering operational conditions to be in MODE 4. This change is 
consistent with the U.S. Nuclear Regulatory Commission (NRC)-approved 
Revision 0 to Technical Specification Task Force (TSTF) Improved 
Standard TS Change Traveler, TSTF-484, ``Use of TS 3.10.1 for Scram 
Time Testing Activities.''
    Date of issuance: April 18, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 163.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: July 24, 2012 (77 FR 
43377). The March 8, 2013 supplement contained clarifying information 
and did not change the NRC staff's initial proposed finding of no 
significant hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 18, 2013.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 14, 2010, as supplemented by letters 
dated August 24, 2010, September 16, 2011, March 15, 2012, July 2, 2012 
and January 31, 2013.
    Description of amendment request: The changes revise the Seabrook 
Station Technical Specifications (TSs) governing the Containment 
Enclosure Emergency Air Cleanup System (CEEACS). The amendment changes 
TS Surveillance Requirement (SR) 4.6.5.1.d.4 so that it will 
demonstrate integrity of the containment enclosure building rather than 
operability of CEEACS. The amendment relocates SR 4.6.5.1.d.4 with 
modifications to new SR 4.6.5.2.b. Additionally, the amendment makes 
some minor wording changes, deletes a definition, and removes a moot 
footnote.
    Date of issuance: April 23, 2013.
    Effective date: As of its date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 136.
    Facility Operating License No. NPF-86: The amendment revised the 
Technical Specifications and the License.
    Date of initial notice in Federal Register: July 13, 2010 (75 FR 
39979). The notice was reissued in its entirety to include a revised 
description of the amendment request on April 17, 2012 (77 FR 22815). 
The notice was reissued again in its entirety to include a revised 
description of the amendment request on July 24, 2012 (77 FR 43378). 
The supplement dated January 31, 2013, provided additional information 
that clarified the application, did not expand the scope of the 
application as noticed, and did not change the NRC staff's proposed no 
significant hazards consideration determination as published in the 
Federal Register on July 24, 2012.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 23, 2013.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, 
Georgia

    Date of application for amendments: August 31, 2012, as 
supplemented on December 6, 2012.
    Brief description of amendments: The amendments revise Technical 
Specifications (TSs) 3.6.6, 3.7.5, 3.8.1, 3.8.9, and TS Example 1.3-3 
by eliminating second completion times from the TSs in accordance with 
TS Task Force Traveler (TSTF)-439, ``Eliminate Second Completion Times 
Limiting Time from Discovery of Failure to Meet an LCO [Limiting 
Condition for Operation].'' In addition, the amendment makes an 
administrative change to TS 3.6.6 by removing an obsolete note 
associated with Condition A.
    Date of issuance: April 24, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 169 and 151.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the licenses and the TSs.
    Date of initial notice in Federal Register: December 11, 2012 (77 
FR 73690). The supplemental letter dated December 6, 2012, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 24, 2013.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 1, 2012, as supplemented by 
letter dated April 15, 2013.
    Brief description of amendment: The amendments revised Technical 
Specification (TS) Table 3.3-10, ``Accident Monitoring 
Instrumentation,'' with respect to the required actions and

[[Page 28258]]

allowed outage times for inoperable instrumentation for Neutron Flux 
(Extended Range) and Neutron Flux--Startup Rate (Extended Range) 
(Instrument Nos. 19 and 23). The required actions have been revised to 
enhance plant reliability by reducing exposure to unnecessary shutdowns 
and increase operational flexibility by allowing more time to implement 
required repairs for inoperable instrumentation. The changes are 
consistent with requirements generically approved as part of NUREG-
1431, Standard Technical Specifications, Westinghouse Plants, Revision 
4 (TS 3.3.3, ``Post Accident Monitoring (PAM) Instrumentation'').
    Date of issuance: April 25, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: Unit 1--200; Unit 2--198.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: October 2, 2012 (77 FR 
60154). The supplemental letter dated April 15, 2013, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 25, 2013.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 6th day of May 2013.
    For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2013-11272 Filed 5-13-13; 8:45 am]
BILLING CODE 7590-01-P