[Federal Register Volume 77, Number 238 (Tuesday, December 11, 2012)]
[Notices]
[Pages 73684-73694]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2012-29612]


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NUCLEAR REGULATORY COMMISSION

[NRC-2012-0292]


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 Commission publish notice of any amendments issued, or 
proposed to be issued and grants the Commission the authority to issue 
and make

[[Page 73685]]

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 November 15 to November 28, 2012. The last 
biweekly notice was published on November 27, 2012 (77 FR 70837).

ADDRESSES: You may access information and comment submissions related 
to this document, which the NRC possesses and are publicly available, 
by searching on http://www.regulations.gov under Docket ID NRC-2012-
0292. You may submit comments by any of the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0292. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; email: [email protected].
     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-2012-0292 when contacting the NRC 
about the availability of information regarding this document. You may 
access information related to this document by any of the following 
methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0292.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may access publicly available documents online in the NRC 
Library at http://www.nrc.gov/reading-rm/adams.html. To begin the 
search, select ``ADAMS Public Documents'' and then select ``Begin Web-
based ADAMS Search.'' For problems with ADAMS, please contact the NRC's 
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to [email protected]. Documents may be viewed in 
ADAMS by performing a search on the document date and docket number.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2012-0292 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 that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment submissions into ADAMS.

Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses and Combined Licenses, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 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 ``Rules of 
Practice for Domestic Licensing Proceedings'' in 10 CFR part 2. 
Interested person(s) should consult a current copy of 10 CFR 2.309, 
which is available at the NRC's PDR, located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. The NRC's regulations are accessible electronically from the NRC 
Library on the NRC's Web site at http://www.nrc.gov/reading-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

[[Page 73686]]

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

[[Page 73687]]

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

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

    Date of amendment request: October 4, 2012.
    Description of amendment request: The proposed amendment would 
modify Technical Specifications by relocating specific surveillance 
frequencies to a licensee controlled program with the adoption of 
Technical Specification Task Force (TSTF)-425, Revision 3, ``Relocate 
Surveillance Frequencies to Licensee Control--Risk-Informed Technical 
Specification Task Force (RITSTF) Initiative 5b.'' Additionally, the 
change would add a new program, the Surveillance Frequency Control 
Program (SFCP), to TS Section 6, Administrative Controls.
    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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes relocate the specified frequencies for 
periodic surveillance requirements to licensee control under a new 
Surveillance Frequency Control Program. Surveillance frequencies are 
not an initiator to any accident previously evaluated. As a result, 
the probability of any accident previously evaluated is not 
significantly increased. The systems and components required by the 
TSs for which the surveillance frequencies are relocated are still 
required to be operable, meet the acceptance criteria for the 
surveillance requirements, and be capable of performing any 
mitigation function assumed in the accident analysis. As a result, 
the consequences of any accident previously evaluated are not 
significantly increased.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    No new or different accidents result from utilizing the proposed 
changes. The changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements. The changes do not alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumptions and current plant operating practice.
    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 change involve a significant reduction in 
the margin of safety?
    Response: No.
    The design, operation, testing methods, and acceptance criteria 
for systems, structures, and components (SSCs), specified in 
applicable codes and standards (or alternatives approved for use by 
the NRC) will continue to be met as described in the plant licensing 
basis (including the final safety analysis report and bases to TS), 
since these are not affected by changes to the surveillance 
frequencies. Similarly, there is no impact to safety analysis 
acceptance criteria as described in the plant licensing basis. To 
evaluate a change in the relocated surveillance frequency, Dominion 
will perform a probabilistic risk evaluation using the guidance 
contained in NRC approved NEI 04-10, Rev. 1, in accordance with the 
TS SFCP. NEI 04-10, Rev. 1, methodology provides reasonable 
acceptance guidelines and methods for evaluating the risk increase 
of proposed changes to surveillance frequencies consistent with 
Regulatory Guide 1.177.
    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

[[Page 73688]]

proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 
23219.
    NRC Branch Chief: George A. Wilson.

Duke Energy Carolinas, LLC, Docket Nos. 50-270 and 50-287, Oconee 
Nuclear Station, Units 2 and 3 (ONS2 and ONS3), Oconee County, South 
Carolina

    Date of amendment request: October 5, 2012.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to authorize a one-time, 19 
month extension to the integrated leak rate test (ILRT) of the reactor 
containment building (also known as the containment). The ILRT is 
normally performed every 10 years. The upcoming ILRT for ONS2 is 
currently due by May 29, 2014, and for ONS3 is due by December 21, 
2014.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed exemption involves a one-time extension to the 
current interval for ONS Unit 2 and Unit 3 Type A containment 
testing. The current test interval of 120 months (10 years) would be 
extended on a one-time basis to no longer than approximately 139 
months from the last Type A test. The proposed extension does not 
involve either a physical change to the plant or a change in the 
manner in which the plant is operated or controlled. The containment 
is designed to provide an essentially leak tight barrier against the 
uncontrolled release of radioactivity to the environment for 
postulated accidents. As such, the containment and the testing 
requirements invoked to periodically demonstrate the integrity of 
the containment exist to ensure the plant's ability to mitigate the 
consequences of an accident, and do not involve the prevention or 
identification of any precursors of an accident. Therefore, this 
proposed extension does not involve a significant increase in the 
probability of an accident previously evaluated.
    This proposed extension is for the next ONS Unit 2 and Unit 3 
Type A containment leak rate test only. The Type B and C containment 
leak rate tests would continue to be performed at the frequency 
currently required by the ONS TS [Technical Specification]. As 
documented in NUREG-1493, Type B and C tests have identified a very 
large percentage of containment leakage paths and the percentage of 
containment leakage paths that are detected only by Type A testing 
is very small. The ONS Unit 2 and Unit 3 Type A test history 
supports this conclusion.
    The integrity of the containment is subject to two types of 
failure mechanisms that can be categorized as (1) activity based and 
(2) time based. Activity based failure mechanisms are defined as 
degradation due to system and/or component modifications or 
maintenance. Local leak rate test requirements and administrative 
controls such as configuration management and procedural 
requirements for system restoration ensure that containment 
integrity is not degraded by plant modifications or maintenance 
activities. The design and construction requirements of the 
containment combined with the containment inspections performed in 
accordance with ASME [American Society of Mechanical Engineers] 
Section Xl, the Maintenance Rule, and TS requirements serve to 
provide a high degree of assurance that the containment would not 
degrade in a manner that is detectable only by a Type A test.
    Based on the above, the proposed extension does not 
significantly increase the consequences of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment to the TS involves a one-time extension 
to the current interval for the ONS Unit 2 and Unit 3 Type A 
containment test. The containment and the testing requirements to 
periodically demonstrate the integrity of the containment exist to 
ensure the plant's ability to mitigate the consequences of an 
accident do not involve any accident precursors or initiators. The 
proposed change does not involve a physical change to the plant 
(i.e., no new or different type of equipment will be installed) or a 
change to the manner in which the plant is operated or controlled.
    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 amendment to the TS involves a one-time extension 
to the current interval for the ONS Unit 2 and Unit 3 Type A 
containment test. This amendment does not alter the manner in which 
safety limits, limiting safety system set points, or limiting 
conditions for operation are determined. The specific requirements 
and conditions of the TS Containment Leak Rate Testing Program exist 
to ensure that the degree of containment structural integrity and 
leak-tightness that is considered in the plant safety analysis is 
maintained. The overall containment leak rate limit specified by TS 
is maintained.
    The proposed change involves only the extension of the interval 
between Type A containment leak rate tests for ONS Unit 2 and Unit 
3. The proposed surveillance interval extension is bounded by the 15 
year ILRT Interval currently authorized within NEI [Nuclear Energy 
Institute] 94-01, Revision 2A. Type B and C containment leak rate 
tests would continue to be performed at the frequency currently 
required by TS. Industry experience supports the conclusion that 
Type B and C testing detects a large percentage of containment 
leakage paths and that the percentage of containment leakage paths 
that are detected only by Type A testing is small. The containment 
inspections performed in accordance with ASME Section Xl, TS and the 
Maintenance Rule serve to provide a high degree of assurance that 
the containment would not degrade in a manner that is detectable 
only by Type A testing. The combination of these factors ensures 
that the margin of safety in the plant safety analysis is 
maintained. The design, operation, testing methods and acceptance 
criteria for Type A, B, and C containment leakage tests specified in 
applicable codes and standards would continue to be met, with the 
acceptance of this proposed change, since these are not affected by 
changes to the Type A test interval.
    Therefore, the proposed 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: Lara S. Nichols, Associate General 
Counsel, Duke Energy Corporation, 526 South Church Street--EC07H, 
Charlotte, NC 28202-1802.
    NRC Branch Chief: Robert J. Pascarelli.

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

    Date of amendment request: September 18, 2012.
    Description of amendment request: The amendment proposes to 
revise Technical Specification (TS) Sections 3.1.6, ``Rod Pattern 
Control,'' and 3.3.2.1, ``Control Rod Block Instrumentation,'' to 
allow MNGP to reference an optional improved 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 
Specification Task Force (TSTF) Improved Standard Technical 
Specifications Change Traveler, TSTF-476, Revision 1, ``Improved 
BPWS Control Rod Insertion Process (NEDO-33091).''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided 
its analysis of the issue of no significant hazards consideration. 
Consistent with the consolidated line item improvement process 
(CLIIP), the licensee referenced the no

[[Page 73689]]

significant hazards consideration published in the Federal Register 
on May 23, 2007 (72 FR 29004), which is provided below:
    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated
    The proposed changes modify the TS to allow the use of the 
improved banked position withdrawal sequence (BPWS) during shutdowns 
if the conditions of NEDO-33091-A, Revision 2, ``Improved BPWS 
Control Rod Insertion Process,'' July 2004, have been satisfied. The 
staff finds that the licensee's justifications to support the 
specific TS changes are consistent with the approved topical report 
and TSTF-476, Revision 1. Since the change only involves changes in 
control rod sequencing, the probability of an accident previously 
evaluated is not significantly increased, if at all. The 
consequences of an accident after adopting TSTF-476 are no different 
than the consequences of an accident prior to adopting TSTF-476. 
Therefore, the consequences of an accident previously evaluated are 
not significantly affected by this change. Therefore, this change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Criterion 2--The Proposed Change Does Not Create the Possibility 
of a New or Different Kind of Accident From any Previously Evaluated
    The proposed change will not introduce new failure modes or 
effects and will not, in the absence of other unrelated failures, 
lead to an accident whose consequences exceed the consequences of 
accidents previously evaluated. The control rod drop accident (CRDA) 
is the design basis accident for the subject TS changes. This change 
does not create the possibility of a new or different kind of 
accident from an accident previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety
    The proposed change, TSTF-476, Revision 1, incorporates the 
improved BPWS, previously approved in NEDO-33091-A, into the 
improved TS. The control rod drop accident (CRDA) is the design 
basis accident for the subject TS changes. In order to minimize the 
impact of a CRDA, the BPWS process was developed to minimize control 
rod reactivity worth for BWR plants. The proposed improved BPWS 
further simplifies the control rod insertion process, and in order 
to evaluate it, the staff followed the guidelines of Standard Review 
Plan Section 15.4.9, and referred to General Design Criterion 28 of 
Appendix A to 10 CFR Part 50 as its regulatory requirement. The TSTF 
stated the improved BPWS provides the following benefits: (1) Allows 
the plant to reach the all-rods-in condition prior to significant 
reactor cool down, which reduces the potential for re-criticality as 
the reactor cools down; (2) reduces the potential for an operator 
reactivity control error by reducing the total number of control rod 
manipulations; (3) minimizes the need for manual scrams during plant 
shutdowns, resulting in less wear on control rod drive (CRD) system 
components and CRD mechanisms; and, (4) eliminates unnecessary 
control rod manipulations at low power, resulting in less wear on 
reactor manual control and CRD system components. The addition of 
procedural requirements and verifications specified in NEDO-33091-A, 
along with the proper use of the BPWS will prevent a control rod 
drop accident (CRDA) from occurring while power is below the low 
power setpoint (LPSP). The net change to the margin of safety is 
insignificant. Therefore, this 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: Peter M. Glass, Assistant General 
Counsel, Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, 
MN 55401.
    NRC Branch Chief: Robert D. Carlson.

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

    Date of amendment request: September 18, 2012.
    Description of amendment request: The amendment proposes to revise 
Technical Specification (TS) 3.8.3, ``Diesel Fuel Oil, Lube Oil, and 
Starting Air,'' by relocating the current stored diesel fuel oil and 
lube oil numerical volume requirements from the TS to the TS Bases so 
that they may be modified under licensee control. The TS are modified 
so that the stored diesel fuel oil and lube oil inventory will require 
that a 7-day supply be available for operation of one emergency diesel 
generator, and the stored lube oil inventory will also continue to 
require that a 7-day supply be available for each diesel generator. The 
changes are consistent with NRC-approved Technical Specification Task 
Force (TSTF) Improved Standard Technical Specifications Change Traveler 
(TSTF-501), Revision 1, ``Relocate Stored Fuel Oil and Lube Oil Volume 
Values to Licensee Control.''
    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 relocates the volume of diesel fuel oil 
required to support 7-day operation of a[n] emergency diesel 
generator (EDG), and the volume equivalent to a 6-day supply, to 
licensee control. The proposed change also relocates the volume of 
diesel lube oil required to support 7-day operation of each onsite 
EDG, and the volume [of fuel oil] equivalent to a 6-day supply, to 
licensee control. The specific volume of fuel oil equivalent to a 7-
day and 6-day supply is calculated using the NRC-approved 
methodology described in Regulatory Guide 1.137, ``Fuel-Oil Systems 
for Standby Diesel Generators,'' and ANSI N195-1976, ``Fuel Oil 
Systems for Standby Diesel-Generators.'' The specific volume of lube 
oil equivalent to a 7-day and 6-day supply is based on the diesel 
generator manufacturer's consumption values for the run time of the 
diesel generator. Because the requirement to maintain a 7-day supply 
of diesel fuel oil and lube oil is not changed and is consistent 
with the assumptions in the accident analyses, and the actions taken 
when the volume of fuel oil and lube oil are less than a 6-day 
supply have not changed, neither the probability nor the 
consequences of any accident previously evaluated will be affected.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The change does not involve a physical alteration of the plant 
(i.e., no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. The change 
does not alter assumptions made in the safety analysis but ensures 
that the diesel generator operates as assumed in the accident 
analysis. The proposed change is consistent with the safety analysis 
assumptions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change relocates the volume of diesel fuel oil 
required to support 7-day operation of a[n] emergency diesel 
generator, and the volume equivalent to a 6-day supply, to licensee 
control. The proposed change also relocates the volume of diesel 
lube oil required to support 7-day operation of each onsite 
emergency diesel generator, and the volume equivalent to a 6-day 
supply, to licensee control. As the bases for the existing limits on 
diesel fuel oil and lube oil are not changed, no change is made to 
the accident analysis assumptions and no margin of safety is reduced 
as part of the change.
    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

[[Page 73690]]

amendment request involves no significant hazards consideration.
    Attorney for licensee: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: Robert D. Carlson.

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: August 31, 2012.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 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. 
These changes are consistent with NRC-approved Industry/Technical 
Specification Task Force (TSTF) Traveler TSTF-439-A, Revision 2, 
``Eliminate Second Completion Times Limiting Time from Discovery of 
Failure to Meet an LCO.'' Additionally, the proposed LAR will make an 
administrative revision to TS 3.6.6 by removing an obsolete note 
associated with Condition 3.6.6.A.
    Basis for proposed no significant hazards consideration 
determination: As required 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration 
(NSHC).

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The change proposed by incorporating TSTF-439-A, Revision 2, 
eliminates certain Completion Times from the Technical 
Specifications. Completion Times are not an initiator to any 
accident previously evaluated. As a result, the probability of an 
accident previously evaluated is not affected. The consequences of 
an accident during the revised Completion Time are no different than 
the consequences of the same accident during the existing Completion 
Times. As a result, the consequences of an accident previously 
evaluated are not affected by this change. 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 described above does not affect the source 
term, containment isolation, or radiological release assumptions 
used in evaluating the radiological consequences of an accident 
previously evaluated. Further, the proposed change does not increase 
the types or amounts of radioactive effluent that may be released 
offsite, nor significantly increase individual or cumulative 
occupational/public radiation exposures. 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.
    Additionally, the proposed change to delete the note from TS 
Condition 3.6.6.A is administrative in nature and does not impact 
the operation, physical configuration, or function of plant SSCs. 
The proposed change does not impact the initiators or assumptions of 
analyzed events, nor does the proposed change impact the mitigation 
of accidents or transient events.
    Therefore, this proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not involve a physical alteration of the 
plant (i.e. no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. The 
proposed changes do not alter any assumptions made in the safety 
analysis.
    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 change to delete the second Completion Time does 
not alter the manner in which safety limits, limiting safety system 
settings or limiting conditions for operation are determined. The 
safety analysis acceptance criteria are not affected by this change. 
The proposed change will not result in plant operation in a 
configuration outside of the design basis. Therefore, the proposed 
change does not involve a significant reduction in a margin of 
safety.
    The proposed change to delete the note from TS Condition 3.6.6.A 
is administrative in nature and does not involve any physical 
changes to plant SSCs, or the manner in which SSCs are operated, 
maintained, modified, tested, or inspected. The proposed change does 
not involve a change to any safety limits, limiting safety system 
settings, limiting conditions of operation, or design parameters for 
any SSC. The proposed change does not impact any safety analysis 
assumptions and do not involve a change in initial conditions, 
system response times, or other parameters affecting any accident 
analysis. The proposed change will not result in plant operation in 
a configuration outside of the design basis.
    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. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Branch Chief: Robert J. Pascarelli.

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

    Date of amendment request: October 17, 2012.
    Description of amendment request: The proposed change would amend 
Combined License Nos.: NPF-91 and NPF-92 for Vogtle Electric Generating 
Plant (VEGP) Units 3 and 4 in regard to the Turbine Building structures 
and layout by: (1) Changing the door location on the motor-driven fire 
pump room in the Turbine Building, (2) clarifying the column line 
designations for the southwest and southeast walls of the Turbine 
Building first bay, (3) changing the floor to ceiling heights at three 
different elevations in the Turbine Building main area, and (4) 
increasing elevations and wall thickness in certain walls of the 
Turbine Building first Bay.
    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 to the Turbine Building configuration do 
not alter the assumed initiators to any analyzed event. Changing the 
door location does not affect the operation of any systems or 
equipment inside or outside the Turbine Building that could initiate 
an analyzed accident. Clarifying the column line designations does 
not affect the operation of any systems or equipment inside or 
outside the Turbine Building that could initiate an analyzed 
accident. The changes in elevation and wall thickness do not affect 
the operation of any systems or equipment inside or outside the 
Turbine Building that could initiate an analyzed accident. In 
preparing this license amendment, it was considered if the changes 
to the Turbine Building door location, column line designations, 
wall thickness, and floor elevations would have an adverse impact on 
the ability of the Turbine Building structure to perform its design 
function to protect the systems, equipment, and components within 
this building. It was concluded that there was no adverse impact, 
because design of this structure, including the redesigned first bay 
wall heights and thicknesses, will continue to be in accordance with 
the same codes and

[[Page 73691]]

standards as stated in the VEGP Units 3 and 4 Updated Final Safety 
Analysis Report (UFSAR). The Turbine Building first bay continues to 
maintain its seismic Category II rating. Based on the above, the 
probability of an accident previously evaluated will not be 
increased by these proposed changes.
    The proposed Turbine Building configuration changes will not 
affect radiological dose consequence analysis. The affected portions 
of the Turbine Building are unrelated to radiological analyses. 
Therefore, no accident source term parameter or fission product 
barrier is impacted by these changes. Structures, systems, and 
components (SSCs) required for mitigation of analyzed accidents are 
not affected by these changes, and the function of the Turbine 
Building to provide weather protection for SSCs inside the building 
is not adversely affected by these changes. Mitigation of a high 
energy line break (HELB) in the Turbine Building first bay is not 
adversely affected by this change, because additional vent area will 
be added to the south wall of the first bay above the Auxiliary 
Building roof. This additional vent area will exceed the vent area 
that is blocked by the change to the Turbine Building main area 
elevations. Consequently, this activity will not increase the 
consequences of any analyzed accident, including the main steam line 
limiting break.
    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 Turbine Building configuration changes to the 
location of a door leading to the Motor-Driven Fire Pump room, 
column line designations, floor elevations in the main area, and 
wall heights and thicknesses in the first bay do not change the 
design function of the Turbine Building or any of the systems or 
equipment in the Turbine Building or in any other Nuclear Island 
structures. In assessing the proposed changes, it was considered if 
they would lead to a different type of possible accident than those 
previously evaluated. The proposed changes do not adversely affect 
any system design functions or methods of operation. The proposed 
changes do not introduce any new equipment or components or change 
the operation of any existing systems or equipment in a manner that 
would result in a new failure mode, malfunction, or sequence of 
events that could affect safety-related or nonsafety-relate 
equipment. This activity will not create a new sequence of events 
that would result in significant fuel cladding failures. With the 
implementation of these changes to the design of this structure, 
including the redesigned first bay wall heights and thicknesses, the 
structure will continue to be in accordance with the same codes and 
standards as stated in the VEGP Units 3 and 4 UFSAR. The Turbine 
Building First Bay continues to maintain its seismic Category II 
rating. Based on the above, it was concluded that the proposed 
changes would not lead to a different type of possible accident than 
those previously considered.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The margin of safety for the design of the Turbine Building, 
including the seismic Category II Turbine Building first bay, is 
determined by the use of the current codes and standards and 
adherence to the assumptions used in the analyses of this structure 
and the events associated with this structure. The relocated door to 
the motor-driven fire pump room will continue to meet the current 3-
hour fire rating requirements. The revised column line designations 
do not represent a physical plant modification, and have no adverse 
impact on plant construction or operation. The design of the Turbine 
Building, including the increased elevations in the main area and 
the increased height and thickness of the redesigned first bay 
walls, will continue to be in accordance with the same codes and 
standards as stated in the UFSAR. The increased elevation of the 
first bay roof to allow the installation of blow-out panels will 
provide additional gross vent area for the first bay, which more 
than compensates for the current vent area that will be blocked by 
the change in the Turbine Building main area elevations. 
Consequently, this activity will not adversely affect the first 
bay's ability to relieve pressure in the event of the limiting main 
steam line break, and consequently this activity will not reduce the 
current margin of safety associated with this event to the design 
pressure limits for Wall 11 of the Nuclear Island and the walls of 
the first bay. The first bay will continue to maintain a seismic 
Category II rating. Adhering to the same codes and standards for the 
Turbine Building structural design and maintaining a seismic 
Category II rating for the Turbine Building first bay preserves the 
current structural safety margins.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: September 27, 2012.
    Description of amendment request: The proposed amendment changes 
the applicable Emergency Action Level for North Anna to include a 15-
minute threshold for reactor coolant system leaks.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Criterion 1:
    Does the proposed amendment involve a significant increase in 
the probability or consequence of an accident previously evaluated?
    Response: No.
    The change affects the North Anna [and Surry Power Station] 
Emergency Action Levels, but does not alter any of the requirements 
of the Operating License or the Technical Specifications. The 
proposed change does not modify any plant equipment and does not 
impact any failure modes that could lead to an accident. 
Additionally, the proposed change has no effect on the consequences 
of any analyzed accident since the change does not affect any 
equipment related to accident mitigation. Based on this discussion, 
the proposed amendment does not increase the probability or 
consequence of an accident previously evaluated.
    Criterion 2:
    Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The change affects the North Anna [and Surry Power Station] 
Emergency Action Levels, but does not alter any of the requirements 
of the Operating License or the Technical Specifications. It does 
not modify any plant equipment and there is no impact on the 
capability of the existing equipment to perform their intended 
functions. No system setpoints are being modified. No new failure 
modes are introduced by the proposed change. The proposed amendment 
does not introduce any accident initiators or malfunctions that 
would cause a new or different kind of accident.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Criterion 3:
    Does the proposed amendment involve a significant reduction in a 
margin of safety?
    Response: No.
    The change affects the North Anna [and Surry Power Station] 
Emergency Action Levels, but does not alter any of the requirements 
of the Operating License or the Technical Specifications. The 
proposed change does not affect any of the assumptions used in the 
accident analysis, nor does it affect any operability requirements 
for equipment important to plant safety.
    Therefore, the proposed change will not result in a significant 
reduction in the margin of safety in operation of the facility as 
discussed in this license amendment request.


[[Page 73692]]


    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 
23219.
    NRC Branch Chief: Robert J. Pascarelli.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Units 1 and 2, Surry County, Virginia

    Date of amendment request: September 27, 2012.
    Description of amendment request: The proposed amendment changes 
the applicable Emergency Action Level for Surry Power Station (SPS) to 
include a 15-minute threshold for reactor coolant system leaks.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Criterion 1:
    Does the proposed amendment involve a significant increase in 
the probability or consequence of an accident previously evaluated?
    Response: No.
    The change affects the [North Anna and] Surry Power Station 
Emergency Action Levels, but does not alter any of the requirements 
of the Operating License or the Technical Specifications. The 
proposed change does not modify any plant equipment and does not 
impact any failure modes that could lead to an accident. 
Additionally, the proposed change has no effect on the consequences 
of any analyzed accident since the change does not affect any 
equipment related to accident mitigation. Based on this discussion, 
the proposed amendment does not increase the probability or 
consequence of an accident previously evaluated.
    Criterion 2:
    Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The change affects the [North Anna and] Surry Power Station 
Emergency Action Levels, but does not alter any of the requirements 
of the Operating License or the Technical Specifications. It does 
not modify any plant equipment and there is no impact on the 
capability of the existing equipment to perform their intended 
functions. No system setpoints are being modified. No new failure 
modes are introduced by the proposed change. The proposed amendment 
does not introduce any accident initiators or malfunctions that 
would cause a new or different kind of accident.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Criterion 3:
    Does the proposed amendment involve a significant reduction in a 
margin of safety?
    Response: No.
    The change affects the [North Anna and] Surry Power Station 
Emergency Action Levels, but does not alter any of the requirements 
of the Operating License or the Technical Specifications. The 
proposed change does not affect any of the assumptions used in the 
accident analysis, nor does it affect any operability requirements 
for equipment important to plant safety. Therefore, the proposed 
change will not result in a significant reduction in the margin of 
safety in operation of the facility as discussed in this license 
amendment request.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219.
    NRC Branch Chief: Robert J. Pascarelli.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: October 18, 2012.
    Description of amendment request: The amendment would revise 
Paragraph 2.C(5)(a) of the renewed facility operating license and the 
fire protection program as described in the Updated Safety Analysis 
Report (USAR) to allow a deviation from the separation requirements of 
10 CFR Part 50, Appendix R, Section III.G.2, as documented in Appendix 
9.5E of the Wolf Creek Generating Station USAR, for the volume control 
tank outlet valves.
    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 structures, systems and components (SSCs) 
are not impacted by the proposed change. An evaluation of not 
maintaining the 10 CFR Part 50, Appendix R, Section III.G.2, 
separation requirements for the volume, control tank outlet valves 
and associated circuits determined that the fire protection features 
provided in fire area A-8 as well as the low fixed combustible 
loading provides reasonable assurance that at least one valve will 
respond to a close signal from the control room following a credible 
fire in the area. The proposed change does not alter or prevent the 
ability of SSCs from performing their intended function to mitigate 
the consequences of an initiating event within the assumed 
acceptance limits. Therefore, the probability of any accident 
previously evaluated is not increased. Equipment required to 
mitigate an accident remains capable of performing the assumed 
function.
    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 change will not alter the requirements or function 
for systems required during accident conditions. An evaluation of 
not maintaining the 10 CFR Part 50, Appendix R, Section llI.G.2, 
separation requirements for the volume control tank outlet valves 
and associated circuits determined that the fire protection features 
provided in fire area A-8 as well as the low fixed combustible 
loading provides reasonable assurance that at least one valve will 
respond to a close signal from the control room following a credible 
fire in the area. The design function of structures, systems and 
components are not impacted by the proposed change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    There will be no effect on the manner in which safety limits or 
limiting safety system settings are determined nor will there be any 
effect on those plant systems necessary to assure the accomplishment 
of protection functions. There will be no impact on departure from 
nuclear boiling ratio (DNBR) limits, heat flux hot channel factor 
(FQ(Z)) limits, nuclear enthalpy rise hot channel factor 
(F\N\[Delta]H) limits, peak centerline temperature (PCT) 
limits, peak local power density or any other 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

[[Page 73693]]

amendment request involves no significant hazards consideration.
    Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw 
Pittman, LLP., 2300 N Street NW., Washington, DC 20037.
    NRC Branch Chief: Michael T. Markley.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Combined Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    A notice of consideration of issuance of amendment to facility 
operating license or combined license, as applicable, proposed no 
significant hazards consideration determination, and opportunity for a 
hearing in connection with these actions, was published in the Federal 
Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) The 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the NRC's Public Document Room (PDR), located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. Publicly available documents created or received at the NRC are 
accessible electronically through the Agencywide Documents Access and 
Management System (ADAMS) in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
PDR's Reference staff at 1-800-397-4209, 301-415-4737 or by email to 
[email protected].

NextEra Energy Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of application for amendments: June 18, 2012, as supplemented 
on September 17, 2012.
    Brief description of amendments: The amendments approved a change 
in scope of Cyber Security Plan Implementation Milestone 6, and revise 
License Condition 4.D, ``Physical Protection,'' of the Renewed Facility 
Operating Licenses for the Point Beach Nuclear Plant, Units 1 and 2.
    Date of issuance: November 23, 2012.
    Effective date: As of the date of issuance and shall be implemented 
by December 31, 2012.
    Amendment Nos.: 247 (Unit 1) and 251 (Unit 2).
    Renewed Facility Operating License Nos. DPR-24 and DPR-27: 
Amendments revised the Renewed Facility Operating License.
    Date of initial notice in Federal Register: September 11, 2012 (77 
FR 55873).
    The licensee's September 17, 2012, supplemental letter contained 
clarifying information, did not change the scope of the original 
amendment request, did not change the NRC staff's initial proposed 
finding of no significant hazards consideration determination, and did 
not expand the scope of the original Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 23, 2012.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 2, 2012.
    Brief description of amendment: The amendments revised the 
Technical Specifications (TSs) by deleting the Steam Generator Water 
Level Low Coincident with Steam Flow/Feedwater Flow Mismatch Reactor 
Trip Function from the TS Table 3.3.1-1 Item 15.
    Date of issuance: November 20, 2012.
    Effective date: As of the date of issuance and shall be implemented 
during Fall 2013 refueling outage for Unit 1 and during Spring 2013 
refueling outage for Unit 2.
    Amendment Nos.: Unit 1--268 and Unit 2--249.
    Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments 
changed the licenses and the technical specifications.
    Date of initial notice in Federal Register: June 12, 2012 (77 FR 
35076).
    The supplement dated August 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.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 20, 2012.
    No significant hazards consideration comments received: No.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: April 26, 2012.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) to adopt NRC-approved Technical Specifications 
Task Force (TSTF) Change Traveler TSTF-510, Revision 2, ``Revision to 
Steam Generator Program Inspection Frequencies and Tube Sample 
Selection,'' using the consolidated line item improvement process 
(CLIIP). Specifically, the amendment revised TS 3.4.17, ``Steam 
Generator (SG) Tube Integrity,'' TS 5.5.9, ``Steam Generator (SG) 
Program,'' and TS 5.6.10, ``Steam Generator Tube Inspection Report,'' 
and included TS Bases changes that summarize and clarify the purpose of 
the TS.
    Date of issuance: November 19, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 199.
    Renewed Facility Operating License No. NPF-42. The amendment 
revised the Operating License and Technical Specifications.
    Date of initial notice in Federal Register: September 4, 2012 (77 
FR 53931).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 19, 2012.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 30th day of November 2012.


[[Page 73694]]


    For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2012-29612 Filed 12-10-12; 8:45 am]
BILLING CODE 7590-01-P