[Federal Register Volume 77, Number 171 (Tuesday, September 4, 2012)]
[Notices]
[Pages 53923-53935]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2012-21545]


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NUCLEAR REGULATORY COMMISSION

[NRC-2012-0205]


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 immediately effective any amendment to an operating license or 
combined license, as applicable, upon a determination by the Commission 
that such amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from August 8, 2012, to August 21, 2012. The last 
biweekly notice was published on August 21, 2012, (77 FR 50534).

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-
0205.
    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-2012-0205. 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-2012-0205 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 http://www.regulations.gov and search for Docket ID NRC-2012-0205.
     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

[[Page 53924]]

White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2012-0205 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 in comment submissions that you do not want to be publicly 
disclosed. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into 
ADAMS, and the NRC does not 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 in their comment submissions 
that they do not want to be publicly disclosed. Your request should 
state that the NRC will not 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 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 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

[[Page 53925]]

would take place before the issuance of any amendment.
    All documents filed in the 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 NRC guidance 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC's Web site 
at http://www.nrc.gov/site-help/e-submittals.html, by email at 
[email protected], or by a toll-free call at 1-866-672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 
20852, Attention: Rulemaking and Adjudications Staff. Participants 
filing a document in this manner are responsible for serving the 
document on all other participants. Filing is considered complete by 
first-class mail as of the time of deposit in the mail, or by courier, 
express mail, or expedited delivery service upon depositing the 
document with the provider of the service. A presiding officer, having 
granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as Social Security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 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. Non-timely filings 
will not be entertained absent a determination by the presiding officer 
that the petition or request should be granted or the contentions 
should be admitted, based on a balancing of the factors specified in 10 
CFR 2.309(c)(1)(i)-(viii).
    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

[[Page 53926]]

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].

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

    Date of amendments request: July 2, 2012.
    Description of amendments request: The amendment would revise 
Technical Specification (TS) 5.5.16 by increasing the calculated peak 
containment internal pressure (Pa) from 49.4 pounds per 
square inch gauge (psig) to 49.7 psig for the design basis loss-of-
coolant accident (LOCA). In support of the revised Pa, the 
amendment would also revise the initial internal containment pressure 
limit in TS 3.6.4 by decreasing the upper bound initial pressure limit 
from 1.8 psig to 1.0 psig.
    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 Pa and the initial containment 
pressure limit 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. The change in Pa 
and the initial containment pressure limit 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 Pa. The Title 10 of the 
Code of Federal Regulations (10 CFR) Part 50, Appendix J containment 
leak rate testing program will continue to ensure that containment 
leakage remains within the leakage assumed in the offsite dose 
consequence analyses.
    Therefore, operation of the facility in accordance with the 
proposed change to TSs 3.6.4 and 5.5.16 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 type of accident from any accident previously evaluated?
    Response: No.
    The proposed change provides a higher Pa than 
currently described in the TS. This change is a result of an 
increase in the mass and energy release input for the LOCA 
containment response analysis. The Pa remains below the 
containment design pressure of 50 psig because of the change in the 
initial containment pressure limit, which is an initial condition of 
the peak pressure calculation. 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.
    Therefore, operation of the facility in accordance with the 
proposed change to TSs 3.6.4 and 5.5.16 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 Pa remains below the containment design pressure 
of 50 psig. 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 does 
not represent a significant change in the margin of safety.
    Therefore, operation of the facility in accordance with the 
proposed change to TSs 3.6.4 and 5.5.16 does not involve a 
significant reduction in the margin of safety.

    The Nuclear Regulatory Commission (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 amendments request involves no 
significant hazards consideration.
    Attorney for licensee: Steven L. Miller, General Counsel, 
Constellation Energy Nuclear Group, LLC, 100 Constellation Way, Suite 
200c, Baltimore, MD 21202.
    NRC Branch Chief: George Wilson.

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

    Date of amendment request: July 31, 2012.
    Description of amendment request: The proposed amendment would 
revise the Millstone Power Station, Unit 2 (MPS2) Technical 
Specification requirements regarding steam generator tube inspections 
and reporting as described in TSTF-510, Revision 2, ``Revision to Steam 
Generator Program Inspection Frequencies and Tube Sample Selection;'' 
however, Dominion Nuclear Connecticut, Inc. is proposing minor 
variations and deviations from TSTF-510.
    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 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 the 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 of a SGTR is not 
increased. The consequences of a SGTR are bounded by the 
conservative assumptions in the design basis accident analysis. The 
proposed change will not cause the consequences of a SGTR to exceed 
those assumptions. The proposed change to reporting requirements and 
clarifications of the existing requirements have no affect on the 
probability or consequences of SGTR.
    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 accident 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 
the 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 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

[[Page 53927]]

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.

    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 Street, RS-2, Richmond, VA 
23219.
    NRC Branch Chief: George A. Wilson.

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

    Date of amendment request: July 31, 2012.
    Description of amendment request: The proposed amendment would 
revise the Millstone Power Station, Unit 3 (MPS3) Technical 
Specification requirements regarding steam generator tube inspections 
and reporting as described in TSTF-510, Revision 2, ``Revision to Steam 
Generator Program Inspection Frequencies and Tube Sample Selection;'' 
however, Dominion Nuclear Connecticut, Inc. is proposing minor 
variations and deviations from TSTF-510.
    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 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 the 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 of a SGTR is not 
increased. The consequences of a SGTR are bounded by the 
conservative assumptions in the design basis accident analysis. The 
proposed change will not cause the consequences of a SGTR to exceed 
those assumptions. The proposed change to reporting requirements and 
clarifications of the existing requirements have no affect on the 
probability or consequences of SGTR.
    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 accident 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 
the 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 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.

    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 Street, RS-2, Richmond, VA 
23219.
    NRC Branch Chief: George A. Wilson.

Exelon Generation Company, LLC (EGC), Docket Nos. STN 50-456 and STN 
50-457, Braidwood Station, Units 1 and 2 (Braidwood), Will County, 
Illinois; Docket Nos. STN 50-454 and STN 50-455, Byron Station, Units 1 
and 2 (Byron), Ogle County, Illinois

    Date of amendment request: June 6, 2012.
    Description of amendment request: The proposed amendment would 
modify Braidwood and Byron Technical Specifications (TS) to add a Note 
to Surveillance Requirements (SR) 3.3.1.7, 3.3.1.8, and 3.3.1.12 in TS 
3.3.1, ``Reactor Trip System (RTS) Instrumentation,'' and SRs 3.3.2.2 
and 3.3.2.6 in TS 3.3.2, ``Engineered Safety Features Actuation System 
(ESFAS) Instrumentation,'' to exclude the Solid State Protection System 
input relays from the Channel Operational Test Surveillance for RTS and 
ESFAS Functions with installed bypass capability which the U.S. Nuclear 
Regulatory Commission (NRC) approved by letters dated March 30, 2012, 
and April 9, 2012.
    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 Reactor Protection System (RPS) and ESFAS provide plant 
protection and are part of the accident mitigating response. The RTS 
and ESFAS functions do not themselves act at a precursor or an 
initiator for any transient or design basis accident. Therefore, the 
proposed change does not significantly increase the probability of 
any accident previously evaluated.
    The proposed change does not alter the design assumptions, 
conditions, or configuration of the facility. The structural and 
functional integrity of the RTS and ESFAS, and any other plant 
system, is unaffected. The proposed change does not alter or prevent 
the ability of any structures, systems, and components from 
performing their intended function to mitigate the consequences of 
an initiating event within the applicable acceptance criteria. 
Surveillance testing in the bypass condition will not cause any 
design or analysis acceptance criteria to be exceeded
    The impact of using bypass testing capability upon nuclear 
safety have been previously evaluated by the NRC and determined to 
be acceptable in [Westinghouse Atomic Power] WCAP 10271-P-A, 
Revision 1, WCAP 14333-P-A, Revision 1, and WCAP 15376-P-A, Revision 
1. Thus, testing in bypass does not involve

[[Page 53928]]

a significant increase in the probability or consequences of an 
accident previously evaluated.
    Implementation of the bypass testing capability does not affect 
the integrity of the fission product barriers utilized for the 
mitigation of radiological dose consequences as a result of an 
accident. The plant response as modeled in the safety analyses is 
unaffected by this change. Hence, the release used as input to the 
dose calculations are unchanged from those previously assumed.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not result in a change in the manner in 
which the RTS and ESFAS provide plant protection. The RTS and ESFAS 
will continue to have the same setpoints after the proposed change 
in implemented. In addition, no new failure modes are being created 
for any plant equipment. The change does not result in the creation 
of any changes to the existing accident scenarios nor do they create 
any new or different accident scenarios. The types of accidents 
defined in the UFSAR [Updated Final Safety Analysis Report] continue 
to represent the credible spectrum of events to be analyzed which 
determine safe operation.
    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.
    No safety analyses are changed or modified as a result of the 
proposed TS change to reflect installed bypass testing capability. 
The proposed change does not alter the manner in which the safety 
limits, limiting safety system setpoints, of limiting conditions for 
operation are determined. Margins associated with the current 
applicable safety analyses acceptance criteria are unaffected. The 
current safety analyses remain bounding since their conclusions are 
not affected by performing surveillance testing in bypass. The 
safety systems credited in the safety analyses will continue to be 
available to perform their mitigation functions.
    Redundant RTS and ESFAS trains are maintained, and diversity 
with regard of the signals that provide reactor trip and engineered 
safety features actuation is also maintained. All signals credited 
as primary or secondary, and all operator actions credited in the 
accident analyses will remain the same. The proposed change will not 
result in plant operation in a configuration outside the design 
basis. Although there was no attempt to quantify any positive human 
factors benefit due to excluding the relays from the [Channel 
Operational Text] COT Surveillance for those RTS and ESFAS Functions 
that have installed bypass test capability, it is expected that 
there would be a new benefit due to a reduced potential for spurious 
reactor trips and actuations associated with testing.
    Implementation of the proposed change is expected to result in 
an overall improvement of safety, as reduced testing will result in 
fewer inadvertent reactor trips, less frequent actuation of ESFAF 
components, less frequent distraction of operations personnel with 
significant affecting RTS and ESFAS reliability.
    Therefore, the proposed change does not result in a significant 
reduction in the margin of safety.
    Based on the above evaluation, EGC concludes that the proposed 
amendments do not involve a 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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear 4300 Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: Michael I. Dudek.
    Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 
50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, 
York and Lancaster Counties, Pennsylvania
    Date of application for amendments: July 18, 2012.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) for Peach Bottom Atomic Power 
Station (PBAPS), Units 2 and 3 to change the operability requirements 
for the normal heat sink (NHS). The NHS for PBAPS is the Susquehanna 
River. Currently, in accordance with TS 3.7.2, the NHS is considered 
operable with a maximum water temperature of 90[emsp14][deg]F. However, 
TS 3.7.2 also currently contains provisions to allow plant operation to 
continue if the NHS water temperature exceeds the 90[emsp14][deg]F 
limit. Specifically, the NHS is still considered operable as long as 
the NHS temperature: (1) does not exceed 92[emsp14][deg]F and; (2) is 
verified at least once per hour to be less than or equal to 
90[emsp14][deg]F when averaged over the previous 24-hour period. The 
proposed amendment would change the NHS water temperature limit such 
that the NHS would be considered operable as long as the maximum water 
temperature was less than or equal to 92[emsp14][deg]F.
    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 allows plant operation to continue if the 
Normal Heat Sink (NHS) temperature does not exceed 92[emsp14][deg]F. 
The water temperature limit imposed for the NHS exists to ensure the 
ability of safety systems to mitigate the consequences of an 
accident and does not involve the prevention or identification of 
any precursors of an accident. The water temperature of the NHS 
cannot adversely affect the initiator of any accident previously 
evaluated. This change does not affect the normal operation of the 
plant to the extent that any accident previously evaluated would be 
more likely to occur.
    The safety objective of the water temperature limit for the NHS 
is to ensure that the heat removal capability of the Emergency 
Service Water (ESW) and High Pressure Service Water (HPSW) Systems 
is adequate to allow safety related equipment that is relied upon to 
mitigate the consequences of an accident or operational transient to 
perform its design function. The design basis heat removal 
capability of the affected components and systems is maintained at 
the NHS temperature limit, thus ensuring that the affected safety 
related components continuously perform their safety related 
function at the NHS temperature limit. The limits for equipment 
degradation ensure that the affected components continue to perform 
their design basis function. Consequently, the affected components 
maintain their design basis capability as previously assumed in 
[the] plant safety analyses.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequence of a previously evaluated 
accident.
    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 allows plant operation to continue if the 
Normal Heat Sink (NHS) temperature does not exceed 92[emsp14][deg]F. 
The method of operation of components (heat exchangers, coolers, 
etc.), which rely on the NHS for cooling, is not altered by this 
activity. The water temperature limit imposed for the NHS exists to 
ensure the ability of plant safety equipment to mitigate the 
consequences of an accident and does not have the potential to 
create an accident initiator. This activity does not involve a 
physical change to any plant structure, system or component that is 
considered an accident initiator. The design basis heat removal 
capability of the affected components is maintained.
    This license amendment request does not involve any changes to 
the operation, testing, or maintenance of any safety-related, or

[[Page 53929]]

otherwise important to safety systems. All systems important to 
safety will continue to be operated and maintained within their 
design bases.
    Therefore, no new failure modes are introduced and the 
possibility of a new or different kind of accident is not created.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Operation of PBAPS, Units 2 and 3 under the NHS temperature 
limit (92[emsp14][deg]F) does not reduce the margin of safety as 
defined in the basis for any Technical Specification. Technical 
Specification Surveillance Requirement (SR) 3.7.2.2 defines the 
value for satisfying the Limiting Condition for Operation for the 
temperature of the NHS. A portion of the Technical Specification 
Bases for SR 3.7.2.2 states:
    Verification of the Normal Heat Sink temperature ensures that 
the heat removal capability of the ESW and HPSW Systems is within 
the DBA [design-basis accident] analysis.
    The basis for SR 3.7.2.2 has not changed as a result of the 
proposed [change]. The heat removal capability of the components 
that rely on the ESW and HPSW Systems for cooling is based on the 
Technical Specification temperature limit (92[emsp14][deg]F) of the 
NHS and the performance capability of the equipment. Periodic 
testing and cleaning are required to verify and ensure that the 
assumed degree of degradation is not reached. The limits for 
equipment degradation ensure that affected components continue to 
perform their design basis function.
    Therefore, since the design basis capability of the affected 
components is maintained at the NHS temperature limit 
(92[emsp14][deg]F), this change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for Licensee: Mr. J. Bradley Fewell, Assistant General 
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett 
Square, PA 19348.
    NRC Branch Chief: Meena K. Khanna.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Units 1 and 2, Beaver County, 
Pennsylvania

    Date of amendment request: July 25, 2012.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) 3.1.3 to allow the normally 
required near-end of life Moderator Temperature Coefficient (MTC) 
measurement to not be performed under certain conditions. If these 
specified conditions are met, the MTC measurement would be replaced by 
a calculated value.
    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, with NRC edits in brackets:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This amendment request would change the near-end of life (EOL) 
moderator temperature coefficient (MTC) surveillance requirement 
(SR) to allow [] the required MTC measurement [to be eliminated] 
under certain conditions. This change would not result in physical 
alteration of a plant structure, system or component, or 
installation of new or different types of equipment. Modification of 
the surveillance requirement under certain conditions would not 
affect the probability of accidents previously evaluated in the 
Updated Final Safety Analysis Report (UFSAR) or cause a change to 
any of the dose analyses associated with the UFSAR accidents because 
accident mitigation functions would remain unchanged. Existing MTC 
TS limits would remain unchanged and would continue to be satisfied.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This amendment request would change the near EOL MTC SR to allow 
[] the required MTC measurement [to be eliminated] under certain 
conditions. No new accident scenarios, failure mechanisms, or 
limiting single failures are introduced as a result of the proposed 
change. No physical plant alterations are made as a result of the 
proposed change. The proposed change does not challenge the 
performance or integrity of any safety related system. MTC is a 
variable that must remain within limits but is not an accident 
initiator.
    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.
    This amendment request would change the near EOL MTC SR to allow 
[] the required MTC measurement to be eliminated under certain 
conditions. The margin of safety associated with the acceptance 
criteria of accidents previously evaluated in the UFSAR is 
unchanged. The proposed change would have no affect on the 
availability, operability, or performance of the safety-related 
systems and components. A change to a surveillance is proposed based 
on an alternate method of confirming that the surveillance 
requirement is met. The Technical Specification limiting condition 
for operation (LCO) limits for MTC remain unchanged.
    The Technical Specifications establish limits for the moderator 
temperature coefficient based on assumptions in the UFSAR accident 
analyses. Applying the conditional [elimination of] the moderator 
temperature coefficient measurement changes the method of meeting 
the surveillance requirement; however this change does not modify 
the TS values and ensures adherence to the current TS limits. The 
basis for derivation of the moderator temperature coefficient limits 
from the moderator density coefficient assumed in the accident 
analysis would not change.
    Therefore, the margin of safety as defined in the TS is not 
reduced and 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, with the edits noted above, 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: David W. Jenkins, FirstEnergy Nuclear 
Operating Company, FirstEnergy Corporation, 76 South Main Street, 
Akron, OH 44308.
    NRC Branch Chief: Meena Khanna.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Miami-Dade County, Florida

    Date of amendment request: July 16, 2012, as supplemented by letter 
dated August 10, 2012.
    Description of amendment request: The proposed amendments would 
modify Technical Specification (TS) requirements regarding steam 
generator tube inspections and reporting as described in Technical 
Specification Task Force Traveler 510, Revision 2, ``Revision to Steam 
Generator Program Inspection Frequencies and Tube Sample Selection.'' 
The proposed changes would revise TS 3/4.4.5, ``Steam Generator (SG) 
Tube Integrity,'' TS 6.8.4.j, ``Steam Generator (SG) Program.'' and TS 
6.9.1.8, ``Steam Generator Tube Inspection Report.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change 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

[[Page 53930]]

and SG tube sample selection. A steam generator tube rupture (SGTR) 
event is one of the 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 of a SGTR is 
not increased. The consequences of a SGTR are bounded by the 
conservative assumptions in the design basis 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 accident previously evaluated?
    Response: No.
    The proposed changes to the Steam Generator 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 system pressure boundary and, as such, are 
relied upon to maintain the primary system's pressure and inventory. 
As part of the reactor coolant system 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.

    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: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Acting Branch Chief: Jessie F. Quichocho.

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

    Date of amendment request: July 5, 2012.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) Limiting Condition for Operation 
(LCO) for the plant service water (PSW) and ultimate heat sink (UHS). 
Specifically, the surveillance requirement (SR) for the minimum water 
level in each PSW pump well of the intake structure would be revised 
from the existing value to a lower value. This change is based on 
updated design basis analyses that demonstrate that at the new minimum 
level sufficient water inventory remains available from the Altamaha 
River for PSW and residual heat removal service water (RHRSW) to handle 
Loss of Coolant Accident (LOCA) cooling requirements for 30 days post-
accident with no additional makeup water source available.
    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 TS change revises the minimum water level in the 
PSW pump well, as required by SR 3.7.2.1, from 60.7 [feet] ft [mean 
sea level] MSL to 60.5 ft MSL. TS SR 3.7.2.1 verifies that the UHS 
is OPERABLE by ensuring the water level in the PSW pump well of the 
intake structure is sufficient for the PSW, RHRSW and standby 
service water pumps to supply post-LOCA cooling requirements for 30 
days. The safety function of the UHS is to mitigate the impact of an 
accident. The proposed TS change does not result in or require any 
physical changes to HNP systems, structures, and components, 
including those intended for the prevention of accidents. The 
potential impact of the lower PSW pump well minimum water level on 
pump operation requirements, supply of water for 30 days post-LOCA, 
and potential environmental impact have been evaluated and found to 
be acceptable.
    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 TS change revises the minimum water level in the 
PSW pump well, as required by SR 3.7.2.1, from 60.7 ft MSL to 60.5 
ft MSL. TS SR 3.7.2.1 verifies that the UHS is OPERABLE by ensuring 
the water level in the PSW pump well of the intake structure is 
sufficient for the PSW, RHRSW and standby service water pumps to 
supply post-LOCA cooling requirements for 30 days. The proposed TS 
change does not result in or require any physical changes to HNP 
systems, structures, and components. The potential impact of the 
lower PSW pump well minimum water level on pump operation 
requirements, supply of water for 30 days post-LOCA, and potential 
environmental impact have been evaluated and found to be acceptable.
    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.
    The proposed TS change revises the minimum water level in the 
PSW pump well, as required by SR 3.7.2.1, from 60.7 1t MSL to 60.5 
1t MSL. TS SR 3.7.2.1 verifies that the UHS is OPERABLE by ensuring 
the water level in the PSW pump well of the intake structure is 
sufficient for the PSW, RHRSW and standby service water pumps to 
supply post-LOCA cooling requirements for 30 days. The proposed TS 
change does not result in or require any physical changes to HNP 
systems, structures, and components. The potential impact of the 
lower PSW pump well minimum water level on pump operation 
requirements, supply of water for 30 days post-LOCA, and potential 
environmental impact have been evaluated and found to be acceptable.
    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: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Branch Chief: Nancy L. Salgado.

[[Page 53931]]

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

    Date of amendment request: May 2, 2012.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 3.6.6, ``Containment Spray and Cooling 
Systems,'' to replace the 10-year surveillance frequency for testing 
the containment spray nozzles as required by TS Surveillance 
Requirement 3.6.6.8 with an event-based frequency.
    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 Containment Spray System and its spray nozzles are not 
accident initiators and therefore, the proposed change does not 
involve a significant increase in the probability of an accident. 
The revised surveillance requirement will require event-based 
Frequency verification in lieu of fixed Frequency verification. The 
proposed change does not have a detrimental impact on the integrity 
of any plant structure, system, or component that may initiate an 
analyzed event. The proposed change will not alter the operation or 
otherwise increase the failure probability of any plant equipment 
that can initiate an analyzed accident.
    This change does not affect the plant design. There is no 
increase in the likelihood of formation of significant corrosion 
products. Due to their location at the top of the containment, 
introduction of foreign material into the spray headers is unlikely. 
Foreign material introduced during maintenance activities would be 
the most likely source for obstruction, and verification following 
such maintenance would confirm the nozzles remain unobstructed. 
Since the Containment Spray System will continue to be available to 
perform its accident mitigation function, the consequences of 
accidents previously evaluated are not significantly increased.
    Therefore, the consequences of an accident previously evaluated 
are not significantly affected by the proposed change.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change will not physically alter the plant (no new 
or different type of equipment will be installed) or change the 
methods governing normal plant operation. The proposed change does 
not introduce new accident initiators or impact assumptions made in 
the safety analysis. Testing requirements continue to demonstrate 
that the Limiting Conditions for Operation are met and the system 
components are functional.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The system is not susceptible to corrosion-induced obstruction 
or obstruction from sources external to the system. Maintenance 
activities that could introduce foreign material into the system 
would require subsequent verification to ensure there is no nozzle 
blockage. The spray header nozzles are expected to remain unblocked 
and available in the event that the safety function is required. 
Therefore, the capacity of the system would remain unaffected. The 
proposed change does not relax any criteria used to establish safety 
limits and will not relax any safety system settings. The safety 
analysis acceptance criteria are not affected by this change.
    Therefore the proposed change does not involve a reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street NW., Washington, DC 20037.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: April 26, 2012.
    Description of amendment request: The amendment would revise the 
Technical Specifications (TSs) to adopt Technical Specification 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 program (CLIIP). The NRC 
staff issued a notice of availability of the model for referencing in 
license amendment applications in the Federal Register on October 27, 
2011 (76 FR 66763).
    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 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 the 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 of a SGTR is not 
increased. The consequences of a SGTR are bounded by the 
conservative assumptions in the design basis 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 accident 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.


[[Page 53932]]


    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: 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].

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

    Date of application for amendments: August 8, 2011, as supplemented 
by letters dated January 11, May 7, and July 18, 2012.
    Brief description of amendments: The amendments would modify 
Technical Specification (TS) 3.8.1, ``AC Sources--Operating,'' 
Surveillance Requirement (SR) 3.8.1.11 by revising the required power 
factor value to be achieved by the diesel generators (DGs) during 
conduct of the surveillance test. The proposed change would also modify 
the existing note in SR 3.8.1.11 to allow the DG to not achieve the 
required power factor if the grid conditions do not permit and the test 
is performed with DG synchronized with offsite power.
    Date of issuance: August 22, 2012.
    Effective date: As of the date of issuance to be implemented within 
90 days.
    Amendment Nos.: 302 and 279.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the License and TSs.
    Date of initial notice in Federal Register: November 29, 2011 (76 
FR 73729).
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated August 22, 2012.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 17, 2011, as supplemented 
by two letters dated August 9, 2012.
    Brief description of amendment: The amendment revises Final Safety 
Analysis Report (FSAR) Section 9.7.2.1.2, and Appendix B to provide 
additional operating margin for measurement of the Ultimate Heat Sink 
(UHS) temperature. The proposed change to Appendix B is to remove a 
license condition that is no longer needed.
    Date of issuance: August 10, 2012.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 311.
    Renewed Facility Operating License No. DPR-65: Amendment revised 
the License and Appendix B.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated August 
10, 2012.
    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.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of application for amendment: October 28, 2011, as 
supplemented on May 16, 2012.
    Brief description of amendment: This amendment request would revise 
the Technical Specifications (TSs) to increase the condensate storage 
tank low water level setpoint for the interlock to the high pressure 
coolant injection pump suction valves. Additionally, the amendment 
would correct typographical errors in TS numbering and referencing made 
in prior license amendment nos. 223 and 228.
    Date of issuance: August 7, 2012.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 237.
    Facility Operating License No. DPR-35: The amendment revised the 
License and Technical Specifications.
    Date of initial notice in Federal Register: January 10, 2012 (77 FR 
1517).
    The supplemental letter dated May 16, 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 this amendment is contained 
in a Safety Evaluation dated August 7, 2012.
    No significant hazards consideration comments received: No.

[[Page 53933]]

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: July 26, 2011.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs). Specifically, the change revised the 
minimum indicated nitrogen cover pressure specified for the 
accumulators in TS surveillance requirement (SR) 3.5.1.3 from 617 psig 
(pounds per square inch, gauge) to 626 psig. The amendments also 
correct a typographical error in the text associate with SR 3.6.2.1 
changing the word ``rage'' to ``rate.''
    Date of issuance: August 14, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: Unit 1-166 and Unit 2-148.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: September 6, 2011.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 14, 2012.
    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 23, 2011.
    Brief description of amendments: The amendments revised the 
application of Risk-Managed Technical Specifications (RMTS) to 
Technical Specification (TS) 3.7.7, ``Control Room Makeup and Cleanup 
Filtration System.'' The amendments corrected a potential 
misapplication of the Configuration Risk Management Program (CRMP) that 
is currently allowed by the TSs.
    Date of issuance: August 14, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: Unit 1-199; Unit 2-187.
    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: November 1, 2011 (76 FR 
67490).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 14, 2012.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Combined Licenses and Final Determination of No Significant Hazards 
Consideration and Opportunity for a Hearing (Exigent Public 
Announcement or Emergency Circumstances)

    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 for the 
amendment 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.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual notice of 
consideration of issuance of amendment, proposed no significant hazards 
consideration determination, and opportunity for a hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective 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 51.12(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 
application for amendment, (2) the amendment to Facility Operating 
License or Combined License, as applicable, 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].

[[Page 53934]]

    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. 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, and electronically on the 
Internet at the NRC's Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are problems in accessing the document, 
contact the PDR's Reference staff at 1-800-397-4209, 301-415-4737, or 
by email to [email protected]. If a request for a hearing or 
petition for leave to intervene is filed by the above date, the 
Commission or a presiding officer designated by the Commission or by 
the Chief Administrative Judge of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to those specific sources and 
documents of which the petitioner is aware and on which the 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.\1\ 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 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.
---------------------------------------------------------------------------

    \1\ To the extent that the applications contain attachments and 
supporting documents that are not publicly available because they 
are asserted to contain safeguards or proprietary information, 
petitioners desiring access to this information should contact the 
applicant or applicant's counsel and discuss the need for a 
protective order.
---------------------------------------------------------------------------

    Each contention shall be given a separate numeric or alpha 
designation within one of the following groups:
    1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the 
applications.
    2. Environmental--primarily concerns/issues relating to matters 
discussed or referenced in the environmental analysis for the 
applications.
    3. Miscellaneous--does not fall into one of the categories outlined 
above.
    As specified in 10 CFR 2.309, if two or more petitioners/requestors 
seek to co-sponsor a contention, the petitioners/requestors shall 
jointly designate a representative who shall have the authority to act 
for the petitioners/requestors with respect to that contention. If a 
requestor/petitioner seeks to adopt the contention of another 
sponsoring requestor/petitioner, the requestor/petitioner who seeks to 
adopt the contention must either agree that the sponsoring requestor/
petitioner shall act as the representative with respect to that 
contention, or jointly designate with the sponsoring requestor/
petitioner a representative who shall have the authority to act for the 
petitioners/requestors with respect to that contention.
    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. Since the Commission has made a final determination that the 
amendment involves no significant hazards consideration, if a hearing 
is requested, it will not stay the effectiveness of the amendment. Any 
hearing held would take place while the amendment is in effect.
    All documents filed in the 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 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 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

[[Page 53935]]

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 NRC guidance 
available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC`s Web site 
at http://www.nrc.gov/site-help/e-submittals.html, by email at 
[email protected], or by a toll-free call at 1-866-672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, 
Maryland, 20852, Attention: Rulemaking and Adjudications Staff. 
Participants filing a document in this manner are responsible for 
serving the document on all other participants. Filing is considered 
complete by first-class mail as of the time of deposit in the mail, or 
by courier, express mail, or expedited delivery service upon depositing 
the document with the provider of the service. A presiding officer, 
having granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 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.

    Dated at Rockville, Maryland, this 24th day of August 2012.
    For The Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2012-21545 Filed 8-31-12; 8:45 am]
BILLING CODE 7590-01-P