[Federal Register Volume 77, Number 84 (Tuesday, May 1, 2012)]
[Notices]
[Pages 25753-25760]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2012-10195]



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NUCLEAR REGULATORY COMMISSION

[NRC-2012-0096]


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 April 5, 2012 to April 18, 2012. The last 
biweekly notice was published on April 17, 2012 (77 FR 22808).

ADDRESSES: You may access information and comment submissions related 
to this document, which the NRC possesses and is publicly available, by 
searching on http://www.regulations.gov under Docket ID NRC-2012-0096.
    You may submit comments by the following methods:
    <bullet> Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0096. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; email: <a href="/cdn-cgi/l/email-protection#511230233e3d7f16303d3d3036393423113f23327f363e27"><span class="__cf_email__" data-cfemail="76351704191a5831171a1a17111e13043618041558111900">[email protected]</span></a>.
    <bullet> 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.
    <bullet> 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-2012-0096 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:
    <bullet> Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0096.
    <bullet> 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 <a href="/cdn-cgi/l/email-protection#225246500c5047514d57504147624c50410c454d54"><span class="__cf_email__" data-cfemail="94e4f0e6bae6f1e7fbe1e6f7f1d4fae6f7baf3fbe2">[email protected]</span></a>. Documents may be viewed in 
ADAMS by performing a search on the document date and docket number.
    <bullet> 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-2012-0096 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 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; (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 25754]]

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. 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 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's E-Filing rule (72 FR 49139, 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the Internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at <a href="/cdn-cgi/l/email-protection#aac2cfcbd8c3c4cd84cec5c9c1cfdeeac4d8c984cdc5dc"><span class="__cf_email__" data-cfemail="2149444053484f460f454e424a4455614f53420f464e57">[email protected]</span></a>, or by 
telephone at 301-415-1677, to request (1) a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a request or 
petition for hearing (even in instances in which the participant, or 
its counsel or representative, already holds an NRC-issued digital ID 
certificate). Based upon this information, the Secretary will establish 
an electronic docket for the hearing in this proceeding if the 
Secretary has not already established an electronic docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the agency's public Web 
site at http://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with the NRC 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

[[Page 25755]]

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 
<a href="/cdn-cgi/l/email-protection#89c4dac1cda7dbecfae6fcfbeaecc9e7fbeaa7eee6ff"><span class="__cf_email__" data-cfemail="7d302e3539532f180e12080f1e183d130f1e531a120b">[email protected]</span></a>, 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 
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 <a href="/cdn-cgi/l/email-protection#cbbbafb9e5b9aeb8a4beb9a8ae8ba5b9a8e5aca4bd"><span class="__cf_email__" data-cfemail="2b5b4f5905594e58445e59484e6b455948054c445d">[email protected]</span></a>.

Detroit Edison, Docket No. 50-341, Fermi 2, Monroe County, Michigan

    Date of amendment request: January 10, 2012.
    Description of amendment request: The proposed amendment would 
modify Fermi 2 Plant Operating License, Appendix A, Technical 
Specifications (TS) to revise the Residual Heat Removal (RHR) 
Suppression Pool Cooling Surveillance Requirement (SR) 3.6.2.3.2, flow 
requirement from greater than or equal to 10,000 gallons per minute 
(gpm) to greater than or equal to 9,250 gpm. This change is consistent 
with the RHR suppression cooling rate associated with RHR heat 
exchanger minimum thermal performance requirements. Additionally, the 
proposed license amendment clarifies that SR 3.6.2.3.2 applies only to 
pumps required for meeting the Limiting Condition of Operation.
    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. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed TS SR 3.6.2.3.2 minimum flow of greater than or 
equal to 9,250 gpm is consistent with that assumed for accident 
extrapolation calculations of measured thermal performance obtained 
during RHR heat exchanger testing. This testing is performed to 
periodically demonstrate that the actual heat exchanger thermal 
performance exceeds that assumed for establishing the maximum post-
accident bulk average suppression pool temperature. Therefore, the 
change in required RHR suppression pool cooling flow will not result 
in any increase in post-accident suppression pool temperature above 
that already evaluated for demonstrating adequate Net Pump Suction 
Head (NPSH) for any Emergency Core Cooling System (ECCS) pump. The 
change in the applicability of the surveillance to each required RHR 
pump provides consistency with the design of the system and 
maintains full capability of each RHR suppression pool cooling 
subsystem to provide post accident design basis cooling.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change revises TS SR 3.6.2.3.2 for RHR suppression 
pool cooling flow to be consistent with that assumed for evaluating 
measured heat exchanger thermal performance against the minimum 
requirements of the plant safety analysis. Changing the 
applicability of the surveillance to each required RHR pump is 
consistent with the system design requirement and maintains full 
capability of each RHR suppression pool cooling subsystem to provide 
the post accident cooling function. No physical changes are being 
made to the installed RHR system or the manner in which it is 
operated. No new or different accident scenarios are created by this 
change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The RHR system has historically been capable of meeting TS SR 
3.6.2.3.2. This Surveillance requires demonstration of a system 
flow, in conjunction with a prescribed RHR heat exchanger capacity 
that ensures the overall suppression pool cooling capacity meets the 
requirements of the safety

[[Page 25756]]

analysis. However, the lack of available operating margin inherent 
in the design orifices of the RHR suppression pool cooling test 
return line and identification of a non-conservative bias in the 
test flow instrument calibration have eroded the flow test margin 
such that it is possible that the TS SR may not be satisfied in the 
future even though a large margin is maintained compared to the 
minimum performance assumed in the containment safety analyses. The 
proposed change makes the margin between TS SR 3.6.2.3.2 and the 
performance assumed in the plant safety analyses available as a 
design and operating margin. This is ensured by establishing a 
higher level of required heat exchanger performance, where ample 
margin is available. Heat exchanger testing is conducted in 
accordance with existing testing standards as prescribed by EPRI TR-
107397, Service Water Heat Exchanger Testing Guidelines. The minimum 
required flow rate necessary to satisfy RHR suppression pool cooling 
TS SR 3.6.2.3.2 will be documented in the plant design basis with 
the minimum required flow adjusted upward as necessary to account 
for instrument uncertainty and bias as well as differences between 
assumed accident and actual test operating conditions.
    The change in the applicability of the surveillance to each 
required RHR pump is consistent with the design basis of the plant 
and maintains full capability of the system to provide its safety 
related cooling function following a design basis accident.
    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: Bruce R. Masters, DTE Energy, General 
Council--Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226-1279.
    NRC Acting Branch Chief: Shawn A. Williams.

Detroit Edison, Docket No. 50-341, Fermi 2, Monroe County, Michigan

    Date of amendment request: January 10, 2012.
    Description of amendment request: The proposed amendment would 
modify Fermi 2 Plant Operating License, Appendix A, Technical 
Specifications (TS) to modify Surveillance Requirement (SR) 3.4.3.2, in 
TS 3.4.3, ``Safety Relief Valves (SRVs)'', SR 3.5.1.13, in TS 3.5.1, 
``ECCS-Operating,'' and SR 3.6.1.6.1, in TS 3.6.1.6, ``Low-Low Set 
(LLS) Valves.'' This proposed amendment replaces the current 
requirement in these TS SRs to verify the SRV opens when manually 
actuated with an alternate requirement that verifies the SRV is capable 
of being opened. The verification of that capability would be satisfied 
by a series of overlapping tests, performed during a refueling outage, 
that demonstrate the required functions of successive valve stages.
    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. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change does not modify the method of demonstrating 
the operability of the Safety Relief Valves (SRVs) in both the 
safety and relief modes of operation. As currently stated in the 
Technical Specification (TS) Bases ``* * * valve OPERABILITY and the 
setpoints for overpressure protection are verified, per ASME Code 
requirements, prior to valve installation.'' The proposed change 
does modify the method for demonstrating the proper mechanical 
functioning of the SRVs. The SRVs are required to function in the 
safety mode to prevent overpressurization of the reactor vessel and 
reactor coolant system pressure boundary during various analyzed 
transients, including Main Steam Isolation Valve closure. SRVs 
associated with the Automatic Depressurization System are also 
required to function in the relief mode to reduce reactor pressure 
to permit injection by low pressure Emergency Core Cooling System 
(ECCS) pumps during certain reactor coolant pipe break accidents. 
The current testing method demonstrates the proper mechanical 
functioning of the SRVs in both modes through manual actuation of 
the SRVs. The proposed new testing method demonstrates both 
operability and proper mechanical functioning using a series of 
overlapping tests that demonstrate proper functioning of the SRV and 
supporting control components. This proposed testing method results 
in acceptable demonstration of the SRV functions in both the safety 
and relief modes, and therefore provides assurance that the 
probability of SRV failure will not increase. None of the accident 
safety analyses is affected by the requested TS changes. Therefore, 
the consequences of accidents mitigated by the SRVs will not 
increase.
    Certain SRV malfunctions are included in the UFSAR safety 
analyses. Specifically, the plant safety analyses include the 
inadvertent opening of an SRV and a stuck open SRV. By reducing or 
not actuating the SRVs during plant operation for testing and thus 
reducing the potential incidence of pilot stage leakage of the SRVs, 
the proposed testing reduces a contributor to these events.
    Based on these considerations, the proposed test method does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change modifies the method of testing of the SRVs, 
but does not alter the functions or functional capabilities of the 
SRVs. Testing under the proposed method is performed in offsite test 
facilities and in the plant during outage periods when the SRV 
functions are not required. Existing analyses address events 
involving an SRV inadvertently opening or failing to reclose. 
Analyses also address the failure of one or more SRVs to open. The 
proposed change does not introduce any new failure mode, and 
therefore, does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The proposed amendment provides for an alternative means of 
testing the SRVs. The proposed changes will provide a complete 
verification of the functional capability of the SRVs by performing 
a series of tests, inspections, and maintenance activities without 
opening the valves with reactor steam while installed in the plant. 
The alternative testing and associated programmatic controls will 
provide an equivalent level of assurance that the SRVs are capable 
of performing their intended accident mitigation safety functions. 
The proposed amendment does not affect the valve setpoints or 
adversely affect any other operational criteria assumed for accident 
mitigation. No changes are proposed that 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. Moreover, it is expected that the alternative 
testing methodology will increase the margin of safety by reducing 
the potential for SRV leakage resulting from testing the SRVs with 
reactor steam pressure while installed in the plant.
    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: Bruce R. Masters, DTE Energy, General 
Council--Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226-1279.
    NRC Acting Branch Chief: Shawn A. Williams.

[[Page 25757]]

Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, 
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South 
Carolina

    Date of amendment request: September 20, 2011, as supplemented by 
letter dated November 21, 2011.
    Description of amendment request: The proposed amendments would 
allow revisions to the current licensing basis to allow a measurement 
uncertainty recapture (MUR) power uprate.
    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 amendment changes the rated thermal power from 2568 
megawatts thermal (MWt) to 2610 MWt; an increase of approximately 
1.64% Rated Thermal Power. Duke Energy's evaluations have shown that 
all structures, systems and components (SSCs) are capable of 
performing their design function at the uprated power of 2610 MWt. A 
review of station accident analyses found that all but two analyses 
remain bounding at the uprated power of 2610 MWt. These two analyses 
(High Energy Line Break and Double Main Steam Line Break) were 
reanalyzed at the higher power level and found to be acceptable.
    The radiological consequences of operation at the uprated power 
conditions have been assessed. The proposed power uprate does not 
affect release paths, frequency of release, or the analyzed reactor 
core fission product inventory for any accidents previously 
evaluated in the Final Safety Analysis Report. Analyses performed to 
assess the effects of mass and energy releases remain valid. All 
acceptance criteria for radiological consequences continue to be met 
at the uprated power level.
    As summarized in Sections IV, V, and VI of Enclosure 2, the 
proposed change does not involve any change to the design or 
functional requirements of the associated systems. That is, the 
increased power level neither degrades the performance of, nor 
increases the challenges to any safety systems assumed to function 
in the plant safety analysis.
    While power level is an input to accident analyses, it is not an 
initiator of accidents. The proposed change does not affect any 
accident precursors and does not introduce any accident initiators. 
The proposed change does not impact the usefulness of the 
Surveillance Requirements (SRs) in evaluating the operability of 
required systems and components.
    In addition, evaluation of the proposed TS [Technical 
Specification] change demonstrates that the availability of 
equipment and systems required to prevent or mitigate the 
radiological consequences of an accident is not significantly 
affected. Since the impact on the systems is minimal, it is 
concluded that the overall impact on the plant safety analysis is 
negligible.
    Therefore, the proposed TS changes do not significantly increase 
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.
    A Failure Modes and Effects Analysis of the new system was 
performed, and the possible effects of failures of the new equipment 
and the increased power level on the overall plant systems were 
reviewed. This review found that no new or different accidents were 
created by the new equipment or the uprated power levels.
    No installed equipment is being operated in a different manner. 
The proposed changes have no significant adverse affect on any 
safety-related SSCs and do not significantly change the performance 
or integrity of any safety-related system.
    The proposed changes do not adversely affect any current system 
interfaces or create any new interfaces that could result in an 
accident or malfunction of a different kind than previously 
evaluated. The uprated power does not create any new accident 
initiators. Credible malfunctions are bounded by the current 
accident analyses of record or recent evaluations demonstrating that 
applicable criteria are still met with the proposed changes.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Although the proposed amendment increases the operating power 
level of the plants, it retains the margin of safety because it is 
only increasing power by the amount equal to the reduction in 
uncertainty in the heat balance calculation. The margins of safety 
associated with the power uprate are those pertaining to core 
thermal power. These include fuel cladding, reactor coolant system 
pressure boundary, and containment barriers. Analyses demonstrate 
that the current design basis continues to be met after the MUR 
power uprate. Components associated with the reactor coolant system 
pressure boundary structural integrity, including pressure-
temperature limits, vessel fluence, and pressurized thermal shock 
are bounded by the current analyses. Systems will continue to 
operate within their design parameters and remain capable of 
performing their intended safety functions.
    The current Oconee safety analyses, and the revised design basis 
radiological accident dose calculations, bound the power uprate and 
therefore do not significantly impact margins.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lara S. Nichols, Associate General Counsel, 
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC 
28202.
    NRC Branch Chief: Nancy L. Salgado.

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: December 8, 2011.
    Description of amendment request: The amendment would revise 
Technical Specifications (TS) 3.8.1; ``AC [Alternating Current] 
Sources--Operating.'' Specifically, the amendment would revise TS 3.8.1 
and the associated Bases, to expand its scope to include provisions for 
testing of the automatic transfer function from the station 22 kiloVolt 
(kV) bus to offsite power for Division III. A new Surveillance 
Requirement (SR) would be added to ensure availability of offsite power 
after loss of the station (onsite) 22 kV bus when offsite power remains 
available. The amendment would also add notes to the Limiting Condition 
for Operation (LCO) and SR to require this feature when Division III is 
powered by onsite power. In addition, new ACTIONS would be added to 
ensure this transfer from onsite to offsite is maintained when a 
required offsite power source is lost.
    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 Technical Specification 
Surveillance Requirements to allow power for emergency systems to be 
supplied from onsite power prior to event initiation. This power 
supply will be transferred to the current accepted offsite power 
source if the main generator is no longer available. The proposed 
Surveillance Requirement is to confirm the automatic transfer 
function.
    The proposed changes do not involve a change in the design 
requirements of the electrical power systems, including the 
emergency power systems. The plant will continue to operate within 
acceptable parameters (electrical loading, etc.) The

[[Page 25758]]

proposed changes do not change the function of plant equipment, or 
affect the response of emergency power systems.
    The proposed changes do not involve a change in the design basis 
initiators for loss of offsite power to the emergency power systems. 
The proposed change utilizes existing components and circuits. The 
change will add a new surveillance requirement to confirm the design 
function operation.
    The proposed change does not impact other design basis accident 
initiators or analyzed events or assumed mitigation of accident or 
transient events.
    The proposed change does not involve a change to the 
consequences of a design basis event as described in the Safety 
Analysis Report.
    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 revises the Technical Specification 
surveillance requirements to confirm operation of existing 
components and circuits. The proposed changes do not involve a 
change in the design basis initiators for loss of offsite power to 
the emergency power systems.
    The proposed changes do not involve a change in the design 
requirements of the electrical power systems, including the 
emergency power systems. The proposed changes do not change the 
function of plant equipment, nor do they affect the response of 
emergency power systems.
    The proposed changes do not involve a change in the operational 
limits or physical design of the electrical power systems, 
particularly the emergency power systems. The proposed changes do 
not change the design function or operation of plant equipment, nor 
do they introduce any new failure mechanisms. This change will 
implement surveillance requirements to confirm the design function 
operation.
    The transfer function components supporting the safety-related 
buses have been designed to applicable quality standards and design 
criteria. As such, no new failure modes are being introduced. The 
plant equipment will continue to respond in accordance with the 
design and analyses, and no malfunction of a new or different type 
is being introduced by the proposed changes.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises the Technical Specification 
surveillance requirements. The proposed changes do not involve a 
change in the operational limits or the response of that equipment 
if it is called upon to operate.
    The performance capability of the emergency diesel generators 
will not be affected. The plant will continue to operate within 
acceptable parameters (electrical loading, etc.)[.]
    In addition, administrative controls will ensure there are 
adequate administrative controls are in place to ensure the plant 
configuration remains as evaluated.
    The results of the PRA performed to quantitatively assess the 
risk impact of this change indicate there is a minimal risk impact.
    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: Joseph A. Aluise, Associate General 
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Michael T. Markley.

Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point 
Nuclear Generating Unit 2, Westchester County, New York

    Date of amendment request: January 11, 2012.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) Table 3.3.6-1, ``Containment Purge 
System and Pressure Relief Line Isolation Instrumentation.'' The 
proposed amendment would change the term ``ALLOWABLE VALUE'' to ``TRIP 
SETPOINT'' and revise the current setpoint used for the Containment 
Purge Systems and Pressure Relief Line isolation. The proposed revision 
to TS Table 3.3.6-1 will change ``<= 3 x background'' to allow the trip 
setpoint to be as specified in the Offsite Dose Calculation Manual 
(ODCM).
    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 will revise the term ``ALLOWABLE VALUE'' to 
``TRIP SETPOINT'' and change the setpoint requirements from ``<= 3 
[x] background'' to allow the allowable value to be as specified in 
the Offsite Dose Calculation Manual (ODCM). The change to trip 
setpoint is a correction of an administrative error and will only 
affect the instrument setting specified. Therefore it does not 
involve the initiation of an accident or the consequences. The 
values for the instrument setting are provided for isolating the 
Containment Purge and Pressure Relief Systems due to increased 
source terms and are redundant to containment isolation signals. 
They have no effect on the probability of an accident previously 
evaluated. The change in the setting will be negligible for purposes 
of an accident termination. The ODCM limits are based on 10 CFR 20 
[Title 10 of the Code of Federal Regulations, Part 20] limits which 
are substantially below accident analysis release rates. Therefore 
the change has a minimum effect on the consequences of such 
accidents.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of previously evaluated 
accidents.
    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 will revise the term ``ALLOWABLE VALUE'' to 
``TRIP SETPOINT'' and change the setpoint requirements. The changes 
do not affect the system operations, plant operating procedures or 
affect how the plant is operated. The change does not create the 
possibility of any equipment failure or effect on other equipment.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change will revise the term ``ALLOWABLE VALUE'' to 
``TRIP SETPOINT'' and change the setpoint requirements. The change 
to trip setpoint is correcting an administrative error and has no 
significant affect on the margin of safety. The proposed change 
involves changes to existing setpoints for automatic isolation of 
the Containment Purge and Pressure Relief Systems. However, the 
ability of the systems to isolate remains within current evaluations 
and therefore does not significantly reduce the safety margin.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue White 
Plains, NY 10601.
    NRC Branch Chief: George Wilson.

[[Page 25759]]

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

    Date of amendment request: April 5, 2012.
    Description of amendment request: The licensee proposed to revise 
the MNGP Technical Specifications (TS) 3.3.5.1, ``Emergency Core 
Cooling System (ECCS) Instrumentation.'' Specifically, it is proposed 
to revise the lower allowable value limit for Table 3.3.5.1, Functions 
1.e and 2.e, ``Reactor Steam Dome Pressure Permissive--Bypass Timer 
(Pump Permissive).'' The licensee has determined that the upper 
allowable value limit for the Automatic Depressurization System (ADS) 
bypass timer function provides the operator sufficient time to assess 
the situation and inhibit ADS actuation if the event does not require 
rapid reactor depressurization. The lower allowable value limit ADS 
bypass timer function pertains to providing adequate margin to unwanted 
pump starts during reactor water level transients and is not credited 
in the safety analyses.
    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 (NSHC). The NRC staff reviewed the licensee's NSHC 
analysis and prepared its own as follows:

    (1) Does the proposed amendment involve a significant increase 
in the probability or consequences of an accident previously 
evaluated?
    Response: No.
    The proposed change does not physically impact the plant nor 
does it impact any design or functional requirements of the 
Automatic Depressurization System (ADS). The proposed change does 
not degrade the performance or increase the challenges to any safety 
systems assumed to function in the accident analysis. There is no 
change to normal plant operating parameters or accident mitigation 
performance.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    (2) Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    There are no hardware changes nor are there any changes in the 
method by which plant systems perform a safety function. This 
request does not affect the normal method of plant operation. No new 
equipment is introduced which could create a new or different kind 
of accident. No new equipment failure modes are created. No new 
accident scenarios, transient precursors, failure mechanisms, or 
limiting single failures are introduced.
    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 amendment does not affect the assumptions of the 
safety analysis or the availability or operability of any plant 
equipment. There is no reduction in the margin of safety because the 
criteria for the performance of the ADS are not changed and there 
are no changes to those plant systems necessary to assure the 
accomplishment of protection functions.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
its own analysis, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the proposed amendment involves no significant hazards 
consideration.
    Attorney for the licensee: Peter M. Glass, Assistant General 
Counsel, Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 
55401.
    NRC Branch Chief: Istvan Frankl, Acting.

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 
<a href="/cdn-cgi/l/email-protection#8afaeef8a4f8eff9e5fff8e9efcae4f8e9a4ede5fc"><span class="__cf_email__" data-cfemail="58283c2a762a3d2b372d2a3b3d18362a3b763f372e">[email protected]</span></a>.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, 
Davis-Besse Nuclear Power Station (DBNPS), Unit 1, Ottawa County, Ohio

    Date of application for amendment: May 20, 2011 as supplemented by 
letter dated February 7, 2012.
    Brief description of amendment: This amendment revised Technical 
Specification (TS) 5.5.8.g to perform the special visual inspections 
based on a condition rather than a specific frequency. Specifically, TS 
5.5.8.g requires visual inspection of the secured internal auxiliary 
feedwater header (AFWH), header to shroud attachment welds, and 
external header thermal sleeves of the steam generators (SGs) at DBNPS 
to be performed during the third period of each 10-year inservice 
inspection interval (ISI). With the proposed change, if eddy current 
inspections (required by TS 5.5.8.d.5) identify any SG peripheral rube 
to secured internal AFWH gaps less than [frac14] inches or there is 
evidence that the header is degrading or has moved, then the TS 5.5.8.g 
visual inspections shall be performed on the affected SG.
    Date of issuance: April 18, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 285.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: September 20, 2011 (76 
FR 58306). The February 7, 2012

[[Page 25760]]

supplement 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 finding of no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 18, 2012.
    No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile 
Point Nuclear Station (NMP2), Unit 2, Oswego County, New York

    Date of application for amendment: December 30, 2011, as 
supplemented on March 20, 2012.
    Brief description of amendment: The proposed amendment changes the 
NMP2 Updated Safety Analysis Report allowing the use of Modified Alloy 
718 material for fabrication of the NMP2 reactor recirculation system 
jet pump holddown beams.
    Date of issuance: April 13, 2012.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 141.
    Renewed Facility Operating License No. NPF-069: The amendment 
revises the License and Updated Safety Analysis Report.
    Date of initial notice in Federal Register: February 8, 2012 (77 FR 
6601). The supplemental letter dated March 20, 2012, provided 
additional information that clarified the application and did not 
expand the scope of the application as originally noticed, and did not 
change the Nuclear Regulatory Commission staff's initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 13, 2012.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 28, 2011, as supplemented 
on February 14 and March 14, 2012.
    Brief Description of amendments: These amendments permanently 
revise the Technical Specifications (TS) 6.4.Q, ``Steam Generator (SG) 
Program,'' to exclude portions of the SG tube below the top of the SG 
tubesheet from periodic inspections. In addition, this amendment 
request proposes to revise TS 6.6.A.3, ``Steam Generator Tube 
Inspection Report,'' to remove references to the previous Unit 1 one-
time and Unit 2 temporary alternate repair criteria and provides 
reporting requirements specific to the permanent alternate repair 
criteria. This amendment also addressed minor administrative revisions 
to reinstate the superscript number 1 as the end of the TS 4.13.B.
    Date of issuance: April 17, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: Unit 1--277 and Unit 2--277.
    Renewed Facility Operating License Nos. DPR-32 and DPR-37: 
Amendments change the licenses and the technical specifications.
    Date of initial notice in Federal Register: October 25, 2011 (76 FR 
66090). The supplements dated February 14, 2012 and March 14, 2012, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated April 17, 2012.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 20th day of April 2012.

    For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2012-10195 Filed 4-30-12; 8:45 am]
BILLING CODE 7590-01-P


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