Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 35069-35079 [2012-13921]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 77, Number 113 (Tuesday, June 12, 2012)]
[Notices]
[Pages 35069-35079]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2012-13921]



[[Page 35069]]

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NUCLEAR REGULATORY COMMISSION

[NRC-2012-0131]


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 May 17, 2012 to May 30, 2012. The last 
biweekly notice was published on May 29, 2012 (77 FR 31655).

ADDRESSES: You may access information and comment submissions related 
to this document, which the NRC possesses and is publicly available, by 
searching on https://www.regulations.gov under Docket ID NRC-2012-0131. 
You may submit comments by the following methods:
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2012-0131. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; email: Carol.Gallagher@nrc.gov.
     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-2012-0131 when contacting the NRC 
about the availability of information regarding this document. You may 
access information related to this document, which the NRC possesses 
and is publicly available, by the following methods:
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2012-0131.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may access publicly available documents online in the NRC 
Library at https://www.nrc.gov/reading-rm/adams.html. To begin the 
search, select ``ADAMS Public Documents'' and then select ``Begin Web-
based ADAMS Search.'' For problems with ADAMS, please contact the NRC's 
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to pdr.resource@nrc.gov. Documents may be viewed in 
ADAMS by performing a search on the document date and docket number.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2012-0131 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 https://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 Title 10 of the Code of Federal 
Regulations (10 CFR) 50.92, 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

[[Page 35070]]

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 
https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or a presiding officer designated by the Commission or 
by the Chief Administrative Judge of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the Chief Administrative Judge of the Atomic Safety and 
Licensing Board will issue a notice of a hearing or an appropriate 
order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held. If 
the final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards consideration, then any hearing 
held would take place before the issuance of any amendment.
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC E-Filing rule (72 FR 49139; 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at hearing.docket@nrc.gov, or by 
telephone at 301-415-1677, to request (1) a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a request or 
petition for hearing (even in instances in which the participant, or 
its counsel or representative, already holds an NRC-issued digital ID 
certificate). Based upon this information, the Secretary will establish 
an electronic docket for the hearing in this proceeding if the 
Secretary has not already established an electronic docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the agency's public Web 
site at https://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through Electronic Information Exchange System, users 
will be required to install a Web browser plug-in from the NRC's Web 
site. Further information on the Web-based submission form, including 
the installation of the Web browser plug-in, is available on the NRC's 
public Web site at https://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with NRC guidance 
available on the NRC public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The

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E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC Web site 
at https://www.nrc.gov/site-help/e-submittals.html, by email at 
MSHD.Resource@nrc.gov, or by a toll-free call at 1-866 672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, 
Maryland, 20852, Attention: Rulemaking and Adjudications Staff. 
Participants filing a document in this manner are responsible for 
serving the document on all other participants. Filing is considered 
complete by first-class mail as of the time of deposit in the mail, or 
by courier, express mail, or expedited delivery service upon depositing 
the document with the provider of the service. A presiding officer, 
having granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
https://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 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 
accessible electronically through ADAMS in the NRC Library at https://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to 
ADAMS or who encounter problems in accessing the documents located in 
ADAMS, should contact the NRC's PDR Reference staff at 1-800-397-4209, 
301-415-4737, or by email to pdr.resource@nrc.gov.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, 
and 3, Maricopa County, Arizona

    Date of amendment request: March 8, 2012.
    Description of amendment request: The amendments would eliminate 
the use of the term CORE ALTERATIONS throughout the Technical 
Specifications (TSs). The proposed amendment incorporates changes 
reflected in Technical Specification Task Force (TSTF) Change Traveler 
TSTF-471-A, Revision 1, ``Eliminate use of term CORE ALTERATIONS in 
ACTIONS and Notes.'' The U.S. Nuclear Regulatory Commission (NRC) staff 
reviewed and approved TSTF-471 by letter dated December 7, 2006 (ADAMS 
Accession No. ML062860320). The changes are consistent with NUREG-1432, 
``Standard Technical Specifications--Combustion Engineering Plants,'' 
Revision 4 (Agencywide Documents Access and Management System (ADAMS) 
Accession No. ML12102A165).
    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 eliminates the use of the defined term CORE 
ALTERATIONS from the Technical Specifications. CORE ALTERATIONS are 
not an initiator of any accident previously evaluated except a fuel 
handling accident. The revised Technical Specifications that protect 
the initial conditions of a fuel handling accident also require the 
suspension of movement of irradiated fuel assemblies. Suspending 
movement of irradiated fuel assemblies protects the initial 
condition of a fuel handling accident and, therefore, suspension of 
CORE ALTERATIONS is not required. Suspension of CORE ALTERATIONS 
does not provide mitigation of any accident previously evaluated. 
Therefore, CORE ALTERATIONS do not affect the initiators of the 
accidents previously evaluated and suspension of CORE ALTERATIONS 
does not affect the mitigation of the accidents previously 
evaluated.
    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.
    No new or different accidents result from utilizing the proposed 
change. The changes do not involve a physical modification of the 
plant (i.e., no new or different type of equipment will be 
installed) or a significant change in the methods governing normal 
plant operation. In addition, the changes do not impose any new or 
different requirements. The changes do not alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumptions.
    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.
    Only two accidents are postulated to occur during plant 
conditions where CORE ALTERATIONS may be made: a fuel handling 
accident and a boron dilution

[[Page 35072]]

accident. Suspending movement of irradiated fuel assemblies prevents 
a fuel handling accident. Also requiring the suspension of CORE 
ALTERATIONS is a redundant requirement to suspending movement of 
irradiated fuel assemblies and does not increase the margin of 
safety. CORE ALTERATIONS have no effect on a boron dilution 
accident. Core components are not involved in the initiation or 
mitigation of a boron dilution accident and the SHUTDOWN MARGIN 
limit is based on assuming the worse-case configuration of the core 
components.
    Therefore, CORE ALTERATIONS have no effect on the margin of 
safety related to a boron dilution accident.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Michael G. Green, Senior Regulatory Counsel, 
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695, 
Phoenix, Arizona 85072-2034.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: April 13, 2012.
    Description of amendment request: The proposed amendment would 
revise the Millstone Power Station, Unit 2 (MPS2) Technical 
Specification (TS) requirements related to diesel fuel oil testing 
consistent with NUREG-1432, Rev. 3.1, ``Standard Technical 
Specifications, Combustion Engineering Plants,'' December 1, 1995, and 
NRC approved Technical Specification Task Force (TSTF) TSTF-374, 
``Revision to TS 5.5.13 and Associated TS Bases for Diesel Fuel Oil,'' 
Revision 0.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration, which is 
presented below:

Criterion 1

    Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes modify the TS requirements related to 
diesel fuel oil testing consistent with NRC approved TSTF-374, 
``Revision to TS 5.5.13 and Associated TS Bases for Diesel Fuel 
Oil,'' Revision 0. To adopt changes consistent with the content of 
TSTF-374 for use in the custom TS of MPS2, the existing MPS2 diesel 
fuel oil testing program will be modified. These changes replace the 
criteria of ``Water and sediment < 0.05%'' with the criteria of ``A 
clear and bright appearance with proper color or a water and 
sediment content within limits'' and remove specific American 
Society for Testing and Materials (ASTM) standard references from 
TS.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, and 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not adversely 
affect the ability of structures, systems, and components (SSCs) to 
perform their intended safety function to mitigate the consequences 
of an initiating event within the assumed acceptance limits. The 
proposed changes do not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of any accident previously evaluated. 
Further, the proposed changes do not increase the types and amounts 
of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures.
    Therefore, the changes do not involve a significant increase in 
the probability or consequences or any accident previously 
evaluated.

Criterion 2

    Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes are used to provide operational flexibility 
regarding evolving industry standards while maintaining operational 
conditions which are consistent with the design basis. Removing of 
specific details from TS, since the details are already specified in 
licensee-controlled documents, provides the flexibility needed to 
maintain state-of-the-art technology in fuel oil sampling and 
analysis methodology. The procedural details associated with the 
involved specifications that are removed from TS and residing in 
licensee-controlled documents are not required to be in the TS to 
provide adequate protection of the public health and safety, since 
the TS still retains the requirement for compliance with applicable 
standards. The changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation in the provision, maintaining, or use of diesel fuel oil. 
The requirements retained in the TS continue to require testing of 
the diesel fuel oil to ensure the proper functioning of the DGs.
    Therefore, the changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.

Criterion 3

    Does the proposed amendment involve a significant reduction in 
the margin of safety?
    Response: No.
    The proposed changes are consistent with the content of TSTF-374 
for use in the custom TS of MPS2. These changes remove specific ASTM 
standard references and a preventive maintenance cleaning 
requirement from TS since the references and requirements are 
already specified in licensee-controlled documents. The proposed 
changes provide the flexibility needed to improve fuel oil sampling 
and analysis methodologies while maintaining sufficient controls to 
ensure continued quality of the fuel oil. The margin of safety 
provided to the DGs by these detailed fuel specifications is 
unaffected by the proposed changes since there continue to be TS 
requirements to ensure fuel oil is of the appropriate quality for 
emergency DG use and DG operability is unaffected.

    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 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: March 20, 2012.
    Description of amendment request: The proposed amendment would 
modify Braidwood and Byron Technical Specifications to permanently 
exclude portions of the steam generator (SG) tube below the top of the 
SG tubesheet from periodic SG tube inspections and plugging or repair 
for Braidwood, Unit 2 and for Byron, Unit 2. In addition, the proposed 
amendment would revise TS 5.6.9 to remove reference to the previous 
temporary alternate repair criteria and provide reporting requirements 
specific to the permanent alternate repair criteria.
    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

[[Page 35073]]

consequences of an accident previously evaluated?
    Response: No.
    The previously analyzed accidents are initiated by the failure 
of plant structures, systems, or components. The proposed change 
that alters the steam generator (SG) inspection and reporting 
criteria does not have a detrimental impact on the integrity of any 
plant structure, system, or component that initiates an analyzed 
event. The proposed change will not alter the operation of, or 
otherwise increase the failure probability of any plant equipment 
that initiates an analyzed accident.
    Of the various accidents previously evaluated, the proposed 
changes only affect the steam generator tube rupture (SGTR), 
postulated steam line break (SLB), feedwater line break (FLB), 
locked rotor and control rod ejection accident evaluations. Loss-of-
coolant accident (LOCA) conditions cause a compressive axial load to 
act on the tube. Therefore, since the LOCA tends to force the tube 
into the tubesheet rather than pull it out, it is not a factor in 
this amendment request. Another faulted load consideration is a safe 
shutdown earthquake (SSE); however, the seismic analysis of Model D5 
SGs has shown that axial loading of the tubes is negligible during 
an SSE.
    During the SGTR event, the required structural integrity margins 
of the SG tubes and the tube-to-tubesheet joint over the H* distance 
will be maintained. Tube rupture in tubes with cracks within the 
tubesheet is precluded by the constraint provided by the presence of 
the tubesheet and the tube-to-tubesheet joint. Tube burst cannot 
occur within the thickness of the tubesheet. The tube-to-tubesheet 
joint constraint results from the hydraulic expansion process, 
thermal expansion mismatch between the tube and tubesheet, and from 
the differential pressure between the primary and secondary side, 
and tubesheet rotation. Based on this design, the structural margins 
against burst, as discussed in draft Regulatory Guide (RG) 1.121, 
``Bases for Plugging Degraded PWR Steam Generator Tubes,'' and TS 
5.5.9, are maintained for both normal and postulated accident 
conditions.
    The proposed change has no impact on the structural or leakage 
integrity of the portion of the tube outside of the tubesheet. The 
proposed change maintains structural and leakage integrity of the SG 
tubes consistent with the performance criteria of TS 5.5.9. 
Therefore, the proposed change results in no significant increase in 
the probability of the occurrence of a SGTR accident.
    At normal operating pressures, leakage from tube degradation 
below the proposed limited inspection depth is limited by the tube-
to-tubesheet crevice. Consequently, negligible normal operating 
leakage is expected from degradation below the inspected depth 
within the tubesheet region.
    The consequences of an SGTR event are not affected by the 
primary-to-secondary leakage flow during the event as primary-to-
secondary leakage flow through a postulated tube that has been 
pulled out of the tubesheet is essentially equivalent to a severed 
tube. Therefore, the proposed change does not result in a 
significant increase in the consequences of a SGTR.
    Primary-to-secondary leakage from tube degradation in the 
tubesheet area during operating and accident conditions is 
restricted due to contact of the tube with the tubesheet. The 
leakage is modeled as flow through a porous medium through the use 
of the Darcy equation. The leakage model is used to develop a 
relationship between operational leakage and leakage at accident 
conditions that is based on differential pressure across the 
tubesheet and the viscosity of the fluid. A leak rate ratio was 
developed to relate the leakage at operating conditions to leakage 
at accident conditions. Since the fluid viscosity is based on fluid 
temperature and it is shown that for the most limiting accident, the 
fluid temperature does not exceed the normal operating temperature 
and therefore the viscosity ratio is assumed to be 1.0. Therefore, 
the leak rate ratio is a function of the ratio of the accident 
differential pressure and the normal operating differential 
pressure.
    The leakage factor of 1.93 for Braidwood Station Unit 2 and 
Byron Station Unit 2, for a postulated SLB/FLB, has been calculated 
as shown in Table 9-7 of WCAP-17072-P, Revision 0. However, EGC 
Braidwood Station Unit 2 and Byron Station Unit 2 will apply a 
factor of 3.11 as determined by Westinghouse evaluation LTR-SGMP-09-
100 P-Attachment, Revision 1, to the normal operating leakage 
associated with the tubesheet expansion region in the condition 
monitoring (CM) and operational assessment (OA). The leakage factor 
of 3.11 applies specifically to Byron Unit 2 and Braidwood Unit 2, 
both hot and cold legs, in Table RAI24-2 of LTR-SGMP-09-100 P-
Attachment, Revision 1. Through application of the limited tubesheet 
inspection scope, the existing operating leakage limit provides 
assurance that excessive leakage (i.e., greater than accident 
analysis assumptions) will not occur. The assumed accident induced 
leak rate limit is 0.5 gallons per minute at room temperature 
(gpmRT) for the faulted SG and 0.218 gpmRT for each of the unfaulted 
SGs for accidents that assume a faulted SG. These accidents are the 
SLB and the locked rotor with a stuck open PORV. The assumed 
accident induced leak rate limit for accidents that do not assume a 
faulted SG is 1.0 gpmRT for all SGs. These accidents are the locked 
rotor and control rod ejection.
    No leakage factor will be applied to the locked rotor or control 
rod ejection transients due to their short duration, since the 
calculated leak rate ratio is less than 1.0.
    The TS 3.4.13 operational leak rate limit is 150 gallons per day 
(gpd) (0.104 gpmRT) through any one SG. Consequently, there is 
sufficient margin between accident leakage and allowable operational 
leakage. The maximum accident leak rate ratio for the Model D5 
design SGs is 1.93 as indicated in WCAP-17072-P, Revision 0, Table 
9-7. However, EGC will use the more conservative value of 3.11 
accident leak rate ratio for the most limiting SG model design 
identified in Table RAI24-2 of LTRSGMP-09-100 P-Attachment Revision 
1. This results in significant margin between the conservatively 
estimated accident leakage and the allowable accident leakage (0.5 
gpmRT).
    For the CM assessment, the component of leakage from the prior 
cycle from below the H* distance will be multiplied by a factor of 
3.11 and added to the total leakage from any other source and 
compared to the allowable accident induced leakage limit. For the 
OA, the difference in the leakage between the allowable leakage and 
the accident induced leakage from sources other than the tubesheet 
expansion region will be divided by 3.11 and compared to the 
observed operational leakage.
    Based on the above, the performance criteria of NEI-97-06, 
Revision 3, and draft RG 1.121 continue to be met and the proposed 
change does not involve a significant increase in the probability or 
consequences of the applicable accidents previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not introduce any changes or mechanisms 
that create the possibility of a new or different kind of accident. 
Tube bundle integrity is expected to be maintained for all plant 
conditions upon implementation of the permanent alternate repair 
criteria. The proposed change does not introduce any new equipment 
or any change to existing equipment. No new effects on existing 
equipment are created nor are any new malfunctions introduced.
    Therefore, based on the above evaluation, the proposed changes 
do not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change defines the safety significant portion of 
the SG tube that must be inspected and repaired. WCAP-17072-P, 
Revision 0, as modified by WCAP-17330-P, Revision 1, identifies the 
specific inspection depth below which any type tube degradation has 
no impact on the performance criteria in NEI 97-06, Revision 3, 
`Steam Generator Program Guidelines.''
    The proposed change that alters the SG inspection and reporting 
criteria maintains the required structural margins of the SG tubes 
for both normal and accident conditions. NEI 97-06, and draft RG 
1.121 are used as the bases in the development of the limited 
tubesheet inspection depth methodology for determining that SG tube 
integrity considerations are maintained within acceptable limits. 
Draft RG 1.121 describes a method acceptable to the NRC for meeting 
General Design Criteria (GDC) 14, ``Reactor Coolant Pressure 
Boundary,'' GDC 15, ``Reactor Coolant System Design,'' GDC 31, 
``Fracture Prevention of Reactor Coolant Pressure Boundary,'' and 
GDC 32, ``Inspection of Reactor Coolant Pressure Boundary,'' by 
reducing the probability and consequences of a SGTR. Draft RG 1.121 
concludes that by determining the limiting safe conditions for tube 
wall degradation, the probability and consequences of a SGTR are 
reduced. This draft RG uses safety factors on loads for tube burst 
that are consistent with

[[Page 35074]]

the requirements of Section III of the American Society of 
Mechanical Engineers (ASME) Code.
    For axially oriented cracking located within the tubesheet, tube 
burst is precluded due to the presence of the tubesheet. For 
circumferentially oriented cracking, WCAP-17072-P, Revision 0, as 
modified by WCAP-17330-P, Revision 1, defines a length of 
degradation-free expanded tubing that provides the necessary 
resistance to tube pullout due to the pressure induced forces, with 
applicable safety factors applied. Application of the limited hot 
and cold leg tubesheet inspection criteria will preclude 
unacceptable primary-to-secondary leakage during all plant 
conditions. The methodology for determining leakage as described in 
WCAP-17072-P, Revision 0, as modified by LTR-SGMP-09-100 P-
Attachment\ shows that significant margin exists between an 
acceptable level of leakage during normal operating conditions that 
ensures meeting the SLB accident-induced leakage assumption and the 
TS leakage limit of 150 gpd.
    Based on the above, it is concluded that the proposed changes do 
not result in any reduction in a margin of safety.
    Based on the above, EGC concludes that the proposed change 
presents no significant hazards consideration under the standards 
set forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no 
significant hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
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: Jacob I. Zimmerman.

Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, 
Comanche Peak Nuclear Power Plant, Units 1 and 2, Somervell County, 
Texas

    Date of amendment request: March 28, 2012.
    Brief description of amendment: The amendment would revise 
Technical Specification (TS) 5.5.9, ``Unit 1 Model D76 and Unit 2 Model 
D5 Steam Generator (SG) Program,'' to permanently exclude portions of 
the Comanche Peak Nuclear Power Plant (CPNPP), Unit 2, Model D5 SG 
tubes below the top of the SG tubesheet from periodic SG tube 
inspections. In addition, this amendment would revise TS 5.6.9, ``Unit 
1 Model D76 and Unit 2 Model D5 Steam Generator Tube Inspection 
Report,'' to provide permanent reporting requirements specific to 
CPNPP, Unit 2, that have previously been established on a one-cycle 
basis.
    The proposed amendment constitutes a redefinition of the SG tube 
primary-to-secondary pressure boundary and defines the safety 
significant portion of the tube that must be inspected or plugged. Tube 
flaws detected below the safety significant portion of the tube are not 
required to be plugged. Allowing flaws in the non-safety significant 
portion of the tube to remain in service minimizes unnecessary tube 
plugging and maintains the safety margin of the steam generators to 
perform the safety function to maintain the reactor coolant pressure 
boundary, maintain reactor coolant flow, and maintain primary to 
secondary heat transfer.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    Of the accidents previously evaluated, the limiting transients 
with consideration to the proposed change to the SG tube inspection 
and repair criteria are the steam generator tube rupture (SGTR) 
event, the steam line break (SLB), and the feed line break (FLB) 
postulated accidents.
    The required structural integrity margins of the SG tubes and 
the tube-to-tubesheet joint over the H* distance will be maintained. 
Tube rupture in tubes with cracks within the tubesheet is precluded 
by the constraint provided by the presence of the tubesheet and the 
tube-to-tubesheet joint. Tube burst cannot occur within the 
thickness of the tubesheet. The tube-to-tubesheet joint constraint 
results from the hydraulic expansion process, thermal expansion 
mismatch between the tube and tubesheet, differential pressure 
between the primary and secondary side, and tubesheet rotation. 
Based on this design, the structural margins against burst, as 
discussed in Regulatory Guide (RG) 1.121, ``Bases for Plugging 
Degraded PWR [Pressurized Water Reactor] Steam Generator Tubes,'' 
[(Agencywide Documents Access and Management System (ADAMS) 
Accession No. ML082120667)] and TS 5.5.9 are maintained for both 
normal and postulated accident conditions.
    The proposed change has no impact on the structural or leakage 
integrity of the portion of the tube outside of the tubesheet. The 
proposed change maintains structural and leakage integrity of the SG 
tubes consistent with the performance criteria in TS 5.5.9. 
Therefore, the proposed change results in no significant increase in 
the probability of the occurrence of [an] SGTR accident.
    At normal operating pressures, leakage from tube degradation 
below the proposed limited inspection depth is limited by the tube-
to-tubesheet crevice. Consequently, negligible normal operating 
leakage is expected from degradation below the inspected depth 
within the tubesheet region. The consequences of an SGTR event are 
not affected by the primary-to-secondary leakage flow during the 
event as primary-to-secondary leakage flow through a postulated tube 
that has been pulled out of the tubesheet is essentially equivalent 
to a severed tube. Therefore, the proposed change does not result in 
a significant increase in the consequences of [an] SGTR.
    The probability of [an] SLB is unaffected by the potential 
failure of a steam generator tube as the failure of tube is not an 
initiator for [an] SLB event.
    The leakage factor of 3.16 for CPNPP Unit 2, for a postulated 
SLB/FLB, has been calculated as described in Westinghouse [Electric 
Company, LLC] Letter LTR-SGMP-09-100 [N]P--Attachment, ``Response to 
NRC Request for Additional Information on H*; Model F and Model D5 
Steam Generators,'' dated August 12, 2009 [(ADAMS Accession No. 
ML101730391)], and is shown in Revised Table 9-7 of this same 
document. Specifically, for the condition monitoring (CM) 
assessment, the component of leakage from the prior cycle from below 
the H* distance will be multiplied by a factor of 3.16 and added to 
the total leakage from any other source and compared to the 
allowable accident induced leakage limit. For the operational 
assessment (OA), the difference in the leakage between the allowable 
leakage and the accident induced leakage from sources other than the 
tubesheet expansion region will be divided by 3.16 and compared to 
the observed operational leakage. The accident-induced leak rate 
limit for CPNPP Unit 2 is 1.0 gpm [gallons per minute]. The TS 
operational leak rate limit through any one steam generator is 150 
gpd [gallons per day] (0.1 gpm). Consequently, there is significant 
margin between accident leakage and allowable operational leakage. 
The SLB/FLB overall leakage factor is 3.16 resulting in significant 
margin between the conservatively estimated accident induced leakage 
and the allowable accident leakage.
    No leakage factor was applied to the locked rotor or control rod 
ejection transients due to their short duration.
    The previously analyzed accidents are initiated by the failure 
of plant structures, systems, or components. The proposed change 
that alters the SG inspection and reporting criteria does not have a 
detrimental impact on the integrity of any plant structure, system, 
or component that initiates an analyzed event. The proposed change 
will not alter the operation of, or otherwise increase the failure 
probability of any plant equipment that initiates an analyzed 
accident.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No.
    The proposed change that alters the steam generator inspection 
and reporting criteria

[[Page 35075]]

does not introduce any new equipment, create new failure modes for 
existing equipment, or create any new limiting single failures. 
Plant operation will not be altered, and all safety functions will 
continue to perform as previously assumed in accident analyses.
    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 
the margin of safety?
    Response: No.
    The proposed change that alters the steam generator inspection 
and reporting criteria maintains the required structural margins of 
the SG tubes for both normal and accident conditions. Nuclear Energy 
Institute [(NEI) document NEI] 97-06, Rev. 3, ``Steam Generator 
Program Guidelines,'' and NRC Regulatory Guide (RG) 1.121, ``Bases 
for Plugging Degraded PWR Steam Generator Tubes,'' are used as the 
bases in the development of the limited tubesheet inspection depth 
methodology for determining that SG tube integrity considerations 
are maintained within acceptable limits. RG 1.121 describes a method 
acceptable to the NRC for meeting General Design Criteria (GDC) 14, 
``Reactor coolant pressure boundary,'' GDC 15, ``Reactor coolant 
system design,'' GDC 31, ``Fracture prevention of reactor coolant 
pressure boundary,'' and GDC 32, ``Inspection of reactor coolant 
pressure boundary,'' by reducing the probability and consequences of 
a SGTR. RG 1.121 concludes that by determining the limiting safe 
conditions for tube wall degradation, the probability and 
consequences of a SGTR are reduced. RG 1.121 uses safety factors on 
loads for tube burst that are consistent with the requirements of 
Section III of the American Society of Mechanical Engineers (ASME) 
Code.
    For axially oriented cracking located within the tubesheet, tube 
burst is precluded due to the presence of the tubesheet. For 
circumferentially oriented cracking, the H* Analysis documented in 
Section 4.1 [of the application dated March 28, 2012] defines a 
length of degradation-free expanded tubing that provides the 
necessary resistance to tube pullout due to the pressure induced 
forces, with applicable safety factors applied. Application of the 
limited hot and cold leg tubesheet inspection criteria will preclude 
unacceptable primary-to-secondary leakage during all plant 
conditions. The methodology for determining leakage provides for 
large margins between calculated and actual leakage values in the 
proposed limited tubesheet inspection depth criteria.
    Therefore, the proposed change does not involve a significant 
reduction in any 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: Timothy P. Matthews, Esq., Morgan, Lewis and 
Bockius, 1800 M Street NW., Washington, DC 20036.
    NRC Branch Chief: Michael T. Markley.

NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold 
Energy Center (DAEC), Linn County, Iowa

    Date of amendment request: September 29, 2011, as supplemented by 
letter dated March 12, 2012.
    Description of amendment request: The proposed amendment would 
revise the DAEC Technical Specifications (TSs) by modifying existing 
Surveillance Requirements (SRs) regarding various modes of operation of 
the main steam safety/relief valves (SRVs). The proposed amendment 
would modify the TS requirements for testing of the SRVs by replacing 
the current requirement to manually actuate each SRV during plant 
startup with a series of overlapping tests that demonstrate the 
required functions of successive valve stages. Elimination of the 
manual actuation requirement at low reactor pressure and steam flow 
decreases the potential for SRV leakage and spurious SRV opening.
    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 any accident previously 
evaluated?
    Response: No.
    The proposed changes modify TS SR 3.4.3.2, SR 3.5.1.9, and SR 
3.6.1.5.1 to provide an alternative means for testing the main steam 
SRVs, ADS [Automatic Depressurization System] valves, and LLS [Low-
Low Set] relief valves. Accidents are initiated by the malfunction 
of plant equipment, or the catastrophic failure of plant structures, 
systems, or components. The performance of SRV testing is not a 
precursor to any accident previously evaluated and does not change 
the manner in which the valves are operated. The proposed testing 
requirements will not contribute to the failure of the SRVs nor any 
plant structure, system, or component. NextEra Energy Duane Arnold 
has determined that the proposed change in testing methodology 
provides an equivalent level of assurance that the SRVs are capable 
of performing their intended safety functions. Thus, the proposed 
changes do not affect the probability of an accident previously 
evaluated.
    The performance of SRV testing provides confidence that the 
relief valves are capable of depressurizing the reactor pressure 
vessel (RPV). This will protect the reactor vessel from 
overpressurization and allow the combination of the Low Pressure 
Coolant Injection and Core Spray Systems to inject into the RPV as 
designed. The LLS relief logic causes two LLS relief valves to be 
opened at a lower pressure than the relief mode pressure setpoints 
and causes the LLS relief valves to stay open longer, such that 
reopening of more than one valve is prevented on subsequent 
actuations. Thus, the LLS relief function prevents excessive short 
duration SRV cycling, which limits induced thrust loads on the SRV 
discharge line for subsequent actuations of the relief valve. The 
proposed changes do not affect any function related to the safety 
mode of the dual function SRVs. The proposed changes involve the 
manner in which the subject valves are tested, and have no effect on 
the types or amounts of radiation released or the predicted offsite 
doses in the event of an accident. The proposed testing requirements 
are sufficient to provide confidence that these valves are capable 
of performing their intended safety functions.
    In addition, an inadvertent opening of an SRV is an analyzed 
event in the DAEC UFSAR [Updated Final Safety Analysis Report] 
(Section 15.1.7.2), as well as the assumption of a single SRV 
failure to open on demand in other transients and accidents, as 
appropriate (e.g., one ADS valve failure in the LOCA [loss-of-
coolant accident] analysis). Since the proposed testing requirements 
do not alter the assumptions for any analyzed transient or accident, 
the radiological consequences of any accident previously evaluated 
are not increased.
    Therefore, the 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 previously evaluated?
    Response: No.
    The proposed changes do not affect the assumed accident 
performance of the main steam SRVs, nor any plant structure, system, 
or component previously evaluated. The proposed changes do not 
install any new equipment, and installed equipment is not being 
operated in a new or different manner. The proposed change in test 
methodology will ensure that the valves remain capable of performing 
their safety functions due to meeting the testing requirements of 
the American Society of Mechanical Engineers Boiler and Pressure 
Vessel Code, with the exception of opening the valve following 
installation or maintenance for which a relief request has been 
submitted (Ref. 6.1 [of the September 29, 2011, application]), 
proposing an acceptable alternative. No setpoints are being changed 
which would alter the dynamic response of plant equipment. 
Accordingly, no new failure modes are introduced.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?

[[Page 35076]]

    Response: No.
    Overpressure protection of the RCPB [reactor coolant pressure 
boundary] is based on the SRVs' setpoints and total relief capacity. 
The setpoints are verified at an offsite testing facility; this 
requirement is not altered by the proposed change. The relief 
capacity of each SRV is determined by the valve's geometry, which is 
also not altered by the proposed test methods.
    The proposed changes will allow testing of the valve actuation 
electrical circuitry, including the solenoid, and mechanical 
actuation components, without causing the SRV to open. The SRVs will 
be manually actuated prior to installation in the plant. Therefore, 
all modes of SRV operation will be tested prior to entering the mode 
of operation requiring the valves to perform their safety functions. 
The proposed changes do not affect the valve setpoint or the 
operational criteria that cause the SRVs to open during plant 
transients or accidents, either manually or automatically. There are 
no changes proposed which alter the setpoints at which protective 
actions are initiated, and there is no change to the operability 
requirements for equipment assumed to operate for accident 
mitigation.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Mitchell S. Ross, P. O. Box 14000 Juno 
Beach, FL 33408-0420.
    NRC Acting Branch Chief: Istvan Frankl.

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

    Date of amendment request: April 6, 2012, and revised on April 12 
and May 7, 2012.
    Description of amendment request: The proposed changes would amend 
Combined License Nos. NPF-91 and NPF-92 for Vogtle Electric Generating 
Plant (VEGP) Units 3 and 4, respectively, in regard to the upper 
tolerance on the Nuclear Island (NI) critical sections basemat 
thickness as identified in the plant-specific Design Control Document 
(DCD).
    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.
    As indicated in FSAR (plant-specific DCD) Subsection 3.8.5.5, 
the design function of the basemat is to provide the interface 
between the nuclear island structures and the supporting soil or 
rock. The basemat transfers the load of nuclear island structures to 
the supporting soil or rock. The basemat transmits seismic motions 
from the supporting soil or rock to the nuclear island. The revision 
of the basemat construction tolerance does not have an adverse 
impact on the response of the basemat and nuclear island structures 
to safe shutdown earthquake ground motions or loads due to 
anticipated transients or postulated accident conditions. The 
revision of the basemat construction tolerance does not impact the 
support, design, or operation of mechanical and fluid systems. There 
is no change to plant systems or the response of systems to 
postulated accident conditions. There is no change to the predicted 
radioactive releases due to normal operation or postulated accident 
conditions. The plant response to previously evaluated accidents or 
external events is not adversely affected, nor does the change 
described create any new accident precursors.
    Therefore, there is no significant increase in the probability 
or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change is to increase the construction tolerance 
for the basemat thickness. The revision of the basemat construction 
tolerance does not change the design of the basemat or nuclear 
island structures. The revision of the basemat construction 
tolerance does not change the design function, support, design, or 
operation of mechanical and fluid systems. The revision of the 
basemat construction tolerance does not result in a new failure 
mechanism for the basemat or new accident precursors. As a result, 
the design function of the basemat is not adversely affected by the 
proposed change.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The revision in the basemat thickness construction tolerance 
does not have an adverse impact on the strength of the basemat. The 
increase in the basemat thickness construction tolerance does not 
have an adverse impact on the seismic design spectra or the 
structural analysis of the basemat or other nuclear island 
structures. The revision in the basemat thickness construction 
tolerance has no impact of the analysis of the nuclear island for 
sliding or overturning. As a result, the design function of the 
basemat is not adversely affected by the proposed change.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

Virginia Electric and Power Company, Docket No. 50-338 and 50-339, 
North Anna Power Station, Units 1 and 2, Louisa County, Virginia

    Date of amendment request: April 2, 2012.
    Description of amendment request: The proposed amendment would 
delete the Steam Generator Water Level Low Coincident with Steam Flow/
Feedwater Flow Mismatch Reactor Trip Function from the Technical 
Specification (TS) Table 3.3.1-1 Item 15.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

Criterion 1--Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?

    The initiating conditions and assumptions for accidents 
described in the Updated Final Safety Analyses Report remain as 
previously analyzed. The proposed change does not introduce a new 
accident initiator nor does it introduce changes to any existing 
accident initiators or scenarios described in the Updated Final 
Safety Analyses Report. The Steam/Feedwater Flow Mismatch and Low 
Steam Generator Water Level reactor trip is not credited for 
accident mitigation in any accident analyses described in the 
Updated Final Safety Analyses Report. The Steam/Feedwater Flow 
Mismatch and Low Steam Generator Water Level trip was designed to 
meet the control and protection systems interaction criteria of 
IEEE-279. The Steam Generator Level Median Signal Selector (MSS) 
prevents adverse control and protection system interaction such that 
it replaces the need for the Steam/Feedwater Flow Mismatch and Low 
Steam Generator Water Level reactor trip to satisfy the IEEE-279 
requirements. As such, the affected control and protection systems 
will continue to perform their required functions without adverse 
interaction, and maintain the capability to shut down the reactor 
when required on Low-Low Steam Generator water level. The ability to 
mitigate a loss of heat

[[Page 35077]]

sink accident previously evaluated is unaffected. The frequency 
categories of previously evaluated accidents are not changed.
    Therefore, neither the probability of occurrence nor the 
consequences of an accident previously evaluated is significantly 
increased.

Criterion 2--Does the change create the possibility of a new or 
different kind of accident from any accident previously evaluated?

    The substitution of the MSS for the Steam/Feedwater Flow 
Mismatch and Low Steam Generator Water Level trip will not introduce 
any new failure modes to the required protection functions. The MSS 
only interacts with the feedwater control system. The Steam 
Generator Water Level Low-Low protection function is not affected by 
this change. Isolation devices upstream of the MSS circuitry ensure 
that the Steam Generator Water Level Low-Low protection function is 
not affected. The MSS is designed to reduce the frequency of system 
failures through utilization of highly reliable components in a 
configuration that relies on a minimum of additional equipment. 
Components used in the MSS are of a quality consistent with low 
failure rates and minimum maintenance requirements, and conform to 
protection system requirements. Furthermore, the design provides the 
capability for complete unit testing that provides unambiguous 
determination of credible system failures. It is through these 
features that the overall design of the MSS minimizes the occurrence 
of undetected failures that may exist between test intervals.
    Therefore, the possibility for a new or different kind of 
accident from any accident previously evaluated is not created.

Criterion 3--Does this change involve a significant reduction in a 
margin of safety?

    The proposed amendment does not involve revisions to any safety 
analysis limits or safety system settings that will adversely impact 
plant safety. The proposed amendment does not alter the functional 
capabilities assumed in a safety analysis for any system, structure, 
or component important to the mitigation and control of design bases 
accident conditions within the facility. Nor does this amendment 
revise any parameters or operating restrictions that are assumptions 
of a design basis accident. In addition, the proposed amendment does 
not affect the ability of safety systems to ensure that the facility 
can be placed and maintained in a shutdown condition for extended 
periods of time.
    The ability of the Steam Generator Water Level Low-Low reactor 
trip function credited in the safety analysis to protect against a 
sudden loss of heat sink event is not affected by the proposed 
change: Since the Steam Generator Low-Low Level trip is credited 
alone as providing complete protection for the accident transients 
that result in low steam generator level, eliminating the Steam/
Feedwater Flow Mismatch and Low Steam Generator Water Level trip 
will not change any safety analysis conclusion for any analyzed 
accident described in the Updated Final Safety Analyses Report.
    The MSS prevents adverse control and protection system 
interaction such that it replaces the need for the Steam/Feedwater 
Flow Mismatch and Low Steam Generator Water Level reactor trip and 
satisfies the IEEE-279 requirements.
    The proposed change improves the margin of safety since removal 
of the Steam/Feedwater Flow Mismatch and Low Steam Generator Water 
Level trip function decreases the potential for challenges to plant 
safety systems, decreases the plant surveillance/maintenance 
activity, and reduces plant complexity. These changes result in a 
reduction in the potential for unnecessary plant transients.
    The Technical Specifications continue to assure that the 
applicable operating parameters and systems are maintained within 
the design requirements and safety analysis assumptions. Therefore, 
the elimination of this trip function will not result in a 
significant reduction in the margin of safety as defined in the 
Updated Final Safety Analyses Report or Technical Specifications.
    Therefore, it is concluded that 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 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: Nancy L. Salgado.

Virginia Electric and Power Company, Docket No. 50-339, North Anna 
Power Station, Unit 2, Louisa County, Virginia

    Date of amendment request: May 11, 2012.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) 3.1.7, ``Rod Position 
Indication'' to allow two demand position indicators in one or more 
banks to be inoperable for up to 4 hours. This change is proposed as a 
temporary change to the TS for the current operating cycle and is 
proposed as a footnote to the current TS Limiting Condition for 
Operation (LCO) Section 3.1.7, Condition D.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

Criterion 1--Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?

    The proposed change provides a new Condition for two demand 
position indicators inoperable in one or more banks. The 
inoperability of two demand position indicators in one or more banks 
does not directly affect any accident analysis or design basis 
limits or cause any limit not to be met, because the demand position 
indicator only provides the intended demand as determined by the rod 
control system. The actual position of the control rods is 
determined by use of the Rod Position Indications (RPIs) for each 
control rod, or the movable incore detector system when the RPIs are 
inoperable.
    The inoperability of the demand position indicators does prevent 
the comparison of the RPIs to the demand position indication for 
verification of rod insertion and rod group alignment limits, which 
is conducted as a periodic surveillance to maintain the reactor 
within analyzed conditions. The use of a 4 hour Completion Time 
limit provides a restriction that limits the time that reactor 
operation can continue during this loss of the demand position 
indication. Since the loss of the demand position indication does 
not cause the rods to change position, hence the actual control rod 
positions are expected to remain within required limits. Placing the 
Rod Control System in a condition incapable of rod movement is a 
positive control to prevent rod stepping while maintenance is being 
performed.
    The proposed change to allow two demand position indicators to 
be inoperable in one or more banks does not affect the automatic or 
manual shutdown capability of the reactor protection system and no 
accident analyses are impacted by the proposed change. The 
operability of the control rods is not affected by the inoperability 
of the demand position indicators.
    Therefore, neither the probability of occurrence nor the 
consequences of an accident previously evaluated is significantly 
increased.

Criterion 2--Does the change create the possibility of a new or 
different kind of accident from any accident previously evaluated?

    The proposed change provides new requirements for two demand 
position indicators inoperable in one or more banks. No new accident 
initiators are introduced by the proposed requirements because the 
allowed condition for inoperability of the demand position 
indicators does not cause any new failure modes to be created that 
can cause an accident. The proposed change does not affect the 
reactor protection system or the reactor control system. The control 
rods should remain within the required limits because the failure of 
the demand position indicators does not cause the rods to change 
position and the RPIs remain available in the affected banks to 
verify the position of the control rods. In addition, the Rod 
Control System is placed in a condition incapable of rod movement as 
a positive control to prevent rod stepping while maintenance is

[[Page 35078]]

being performed. Hence, no new failure modes or accident sequences 
are created that would cause a new or different kind of accident 
from any accident previously evaluated.
    Therefore, the possibility for a new or different kind of 
accident from any accident previously evaluated is not created.

Criterion 3--Does this change involve a significant reduction in a 
margin of safety?

    The operability of the RPIs is required to determine control rod 
positions and thereby ensure compliance with the control rod 
alignment and insertion limits. The proposed change does not alter 
the requirement to determine rod position, but provides a new 
Condition for two demand position indicators inoperable in one or 
more banks. The inoperability of two demand position indicators for 
one or more banks results in the reduced ability to periodically 
verify that RPIs are operable and within expected limits. The 
condition does prevent the comparison of the RPIs to the demand 
position indication for verification of rod insertion and rod group 
alignment limits, which is conducted as periodic surveillance to 
maintain the reactor within analyzed conditions. The loss of the 
demand position indication does not cause the rods to change 
position, hence the actual control rod positions are expected to 
remain within required limits. The use of a 4 hour Completion Time 
limit provides a restriction that limits the time that reactor 
operation can continue during this loss of the demand position 
indication. This ensures the condition is promptly corrected or the 
reactor shutdown in accordance with the applicable Technical 
Specifications action statements. Thus, the proposed change 
maintains the operation of the reactor within the applicable margins 
of safety because the inoperability will be corrected or the unit 
will be shutdown prior to any significant reduction in the ability 
to verify control rod position by the use analog RPIs.
    Therefore, it is concluded that 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 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: Nancy L. Salgado.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas
    Date of amendment request: November 30, 2011.
    Description of amendment request: The proposed amendment would 
revise the Wolf Creek Generating Station Technical Specification (TS) 
3.8.1, ``AC Sources--Operating,'' Surveillance Requirements related to 
Diesel Generator test loads, voltage, and frequency. The proposed 
changes will correct non-conservative Diesel Generator load values that 
are currently under administrative controls.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The diesel generators are required to be OPERABLE in the event 
of a design basis accident coincident with a loss of offsite power 
to mitigate the consequences of the accident. The diesel generators 
are not accident initiators and therefore these changes do not 
involve a significant increase in the probability of an accident 
previously evaluated.
    The accident analyses assume that at least one engineered safety 
feature bus is provided with power either from the offsite circuits 
or the diesel generators. The Technical Specification change 
proposed in this license amendment request will continue to assure 
that the diesel generators have the capacity and capability to 
assume their maximum design basis accident loads. The proposed 
change does not significantly change how the plant would mitigate an 
accident previously evaluated.
    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, and configuration of the facility or the manner in which 
the plant is operated and maintained. The proposed change does not 
adversely affect the ability of structures, systems, and components 
(SSC) to perform their intended safety function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change does not affect the source term, 
containment isolation, or radiological release assumptions used in 
evaluating the radiological consequences of any accident previously 
evaluated. Further, the proposed change does not increase the types 
and amounts of radioactive effluent that may be released offsite, 
nor significantly increase individual or cumulative occupational/
public radiation exposure.
    Therefore, the proposed change does not represent a significant 
increase 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 Technical Specification change does not involve a 
change in the plant design, system operation, or the use of the 
diesel generators. The proposed change requires the diesel 
generators to be tested at increased loads which envelope the actual 
power demand requirements for the diesel generators during design 
basis conditions. These revised loads continue to demonstrate the 
capability and capacity of the diesel generators to perform their 
required functions. There are no new failure modes or mechanisms 
created due to testing the diesel generators at the proposed test 
loading. Testing of the emergency diesel generators at the proposed 
test loadings does not involve any modification in the operational 
limits or physical design of plant systems. There are no new 
accident precursors generated due to the proposed test loadings.
    Therefore, it is concluded that the proposed change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed Technical Specification change will continue to 
demonstrate that the diesel generators meet the Technical 
Specification definition of OPERABILITY, that is, the proposed tests 
will demonstrate that the diesel generators will perform their 
safety function and the necessary diesel generator attendant 
instrumentation, controls, cooling, lubrication and other auxiliary 
equipment required for the emergency diesel generators to perform 
their safety function loads are also tested at these proposed 
loadings. The proposed testing will also continue to demonstrate the 
capability and capacity of the diesel generators to supply their 
required loads for mitigating a design basis accident.
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not impacted by this change. The proposed change will not result 
in plant operation in a configuration outside the design basis.
    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: 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

[[Page 35079]]

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 
available online in the NRC Library at https://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems 
in accessing the documents located in ADAMS, contact the PDR's 
Reference staff at 1-800-397-4209, 301-415-4737 or by email to 
pdr.resource@nrc.gov.

Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc., 
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, 
Louisiana

    Date of amendment request: July 27, 2011, as supplemented by 
letters dated September 16, 2011, and February 7, February 24, and 
April 3, 2012.
    Brief description of amendment: The amendment modified River Bend 
Station's (RBS) Technical Specification (TS) 3.3.6.1, ``Primary 
Containment and Drywell Isolation Instrumentation,'' to revise the 
allowable value (AV) and related setpoints for the Main Steam Tunnel 
Temperature functions 1.e, 3.f, and 4.h in TS Table 3.3.6.1-1. In 
addition, the RBS's Emergency Action Levels will be revised to reflect 
the changes to the AV and related setpoints in TS 3.3.6.1.
    Date of issuance: May 30, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 174.
    Facility Operating License No. NPF-47: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: February 7, 2012 (77 FR 
6147). The supplemental letters dated September 16, 2011, and February 
7, February 24, and April 3, 2012, provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the staff's original proposed 
no significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 30, 2012.
    No significant hazards consideration comments received: No.

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: June 2, 2011, as supplemented 
by letter dated November 10, 2011.
    Brief description of amendments: The amendments modify Technical 
Specification (TS) 3.1.2, ``Reactivity Anomalies,'' to change the 
method used to perform the reactivity anomaly surveillance. 
Specifically, the amendments allow performance of the surveillance 
based on the difference between the monitored (i.e., actual) core 
reactivity and the predicted core reactivity. The surveillance was 
previously performed based on the difference between the monitored 
control rod density and the predicted control rod density.
    Date of issuance: May 25, 2012.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendments Nos.: 284 and 287.
    Renewed Facility Operating License Nos. DPR-44 and DPR-56: The 
amendments revised the License and TSs.
    Date of initial notice in Federal Register: September 6, 2011 (76 
FR 55129).
    The letter dated November 10, 2011, provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination or expand the application beyond the scope 
of the original Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 25, 2012.
    No significant hazards consideration comments received: No.

South Carolina Electric and Gas Company, Docket No. 50-395, Virgil C. 
Summer, Nuclear Station (VCSNS), Unit 1, Jenkinsville, South Carolina

    Date of application for amendment: August 11, 2011.
    Brief description of amendment: This amendment revised the VCSNS 
Technical Specification (TS) to allow an updating of the applicable 
topical report in TS 6.9.1.11, ``Core Operating Limits Report'' to use 
the three-dimensional Advanced Nodal Code neutronic model.
    Date of Issuance: May 30, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No: 190.
    Renewed Facility Operating License No. NPF-12: Amendment revises 
the License and Technical Specifications.
    Date of initial notice in Federal Register: October 11, 2011 (76 FR 
62864).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 30, 2012.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 1st day of June, 2012.
    For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2012-13921 Filed 6-11-12; 8:45 am]
BILLING CODE 7590-01-P