[Federal Register Volume 77, Number 94 (Tuesday, May 15, 2012)]
[Notices]
[Pages 28626-28637]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2012-11599]


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NUCLEAR REGULATORY COMMISSION

[NRC-2012-0107]


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 19, 2012 to May 2, 2012. The last 
biweekly notice was published on May 1, 2012 (77 FR 25753).

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-2012-0107. 
You may submit comments by the following methods:
     Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0107. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; email: [email protected].
     Mail comments to: Cindy Bladey, Chief, Rules, 
Announcements, and Directives Branch (RADB), Office of Administration, 
Mail Stop: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
     Fax comments to: RADB at 301-492-3446.
    For additional direction on accessing information and submitting 
comments, see ``Accessing Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

SUPPLEMENTARY INFORMATION:

I. Accessing Information and Submitting Comments

A. Accessing Information

    Please refer to Docket ID NRC-2012-0107 when contacting the NRC 
about the availability of information regarding this document. You may 
access information related to this document, which the NRC possesses 
and is publicly available, by the following methods:
     Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0107.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may access publicly available documents online in the NRC

[[Page 28627]]

Library at http://www.nrc.gov/reading-rm/adams.html. To begin the 
search, select ``ADAMS Public Documents'' and then select ``Begin Web-
based ADAMS Search.'' For problems with ADAMS, please contact the NRC's 
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to [email protected]. Documents may be viewed in 
ADAMS by performing a search on the document date and docket number.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2012-0107 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; or (2) create the possibility of a new or 
different kind of accident from any accident previously evaluated; or 
(3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination; any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license or 
combined license. Requests for a hearing and a petition for leave to 
intervene shall be filed in accordance with the Commission's ``Rules of 
Practice for Domestic Licensing Proceedings'' in 10 CFR part 2. 
Interested person(s) should consult a current copy of 10 CFR 2.309, 
which is available at the NRC's PDR, located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. 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

[[Page 28628]]

determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held. If 
the final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards consideration, then any hearing 
held would take place before the issuance of any amendment.
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC E-Filing rule (72 FR 49139, 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the Internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at [email protected], or by 
telephone at 301-415-1677, to request (1) a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a request or 
petition for hearing (even in instances in which the participant, or 
its counsel or representative, already holds an NRC-issued digital ID 
certificate). Based upon this information, the Secretary will establish 
an electronic docket for the hearing in this proceeding if the 
Secretary has not already established an electronic docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the agency's public Web 
site at http://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with the NRC guidance 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC Web site 
at http://www.nrc.gov/site-help/e-submittals.html, by email at 
[email protected], or by a toll-free call at 1-866 672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, 
Maryland, 20852, Attention: Rulemaking and Adjudications Staff. 
Participants filing a document in this manner are responsible for 
serving the document on all other participants. Filing is considered 
complete by first-class mail as of the time of deposit in the mail, or 
by courier, express mail, or expedited delivery service upon depositing 
the document with the provider of the service. A presiding officer, 
having granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. With 
respect to copyrighted works, except for limited excerpts that serve 
the purpose of the adjudicatory filings and would constitute a Fair Use 
application, participants are requested not to include copyrighted 
materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Non-

[[Page 28629]]

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 [email protected].
Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power 
Station, Unit 3, New London County, Connecticut
    Date of amendment request: November 17, 2011.
    Description of amendment request: The proposed amendment would add 
Optimized ZIRLOTM as an allowable fuel rod cladding material 
and add the Westinghouse topical report on Optimized ZIRLOTM 
to the Millstone Power Station, Unit 3 (MPS3) Technical Specifications. 
In addition, a typographical error would be corrected.
    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 Technical Specifications changes are: (1) Adding 
Optimized ZIRLOTM to the allowable or approved cladding 
materials to be used at MPS3, and (2) correcting a typographical 
error in the title of Reference 8 in Technical Specification (TS) 
6.9.1.6.b. The proposed change of adding a cladding material does 
not result in an increase to the probability or consequences of an 
accident previously evaluated. Technical Specification 5.3.1 
addresses the fuel assembly design, and currently specifies that 
``Each assembly shall consist of a matrix of Zircaloy or 
ZIRLO[supreg] fuel rods * * *''. The proposed change will add 
Optimized ZIRLOTM to the approved fuel rod cladding 
materials listed in this technical specification. In addition, a 
reference to the topical report for Optimized ZIRLOTM, 
WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A, will be added to the 
listing of approved methods used to determine the core operating 
limits for MPS3 provided in Technical Specification 6.9.1.6.b.
    Westinghouse topical report WCAP-12610-P-A & CENPD-404-P-A, 
Addendum 1-A, ``Optimized ZIRLOTM,'' provides the details and 
results of material testing of Optimized ZIRLOTM compared 
to standard ZIRLO[supreg], as well as the material properties to be 
used in various models and methodologies when analyzing Optimized 
ZIRLOTM. As the nuclear industry pursues longer operating 
cycles with increased fuel discharge burnup and fuel duty, the 
corrosion performance requirements for the nuclear fuel cladding 
become more demanding. Optimized ZIRLOTM was developed to 
meet these industry needs by providing a reduced corrosion rate 
while maintaining the composition and physical properties, such as 
mechanical strength, similar to standard ZIRLO[supreg]. In addition, 
margin to the fuel rod design criterion on fuel rod internal 
pressure has been impacted by increased fuel duty, use of integral 
fuel burnable absorbers, and corrosion/temperature feedback effects. 
Reducing the associated corrosion buildup reduces temperature 
feedback effects, providing additional margin to the fuel rod 
internal pressure design criterion. The fuel will continue to 
satisfy the pertinent design basis operating limits, so cladding 
integrity is maintained. There are no changes that will adversely 
affect the ability of existing components and systems to mitigate 
the consequences of any accident. Addition of Optimized 
ZIRLOTM to the allowable cladding materials for MPS3 
therefore does not result in a significant increase in the 
probability or consequences of an accident previously evaluated.
    The NRC has previously approved use of Optimized 
ZIRLOTM fuel cladding material in Westinghouse fueled 
reactors provided that licensees ensure compliance with the 
Conditions and Limitations set forth in the NRC Safety Evaluation 
for the topical report. Confirmation that these Conditions are 
satisfied is performed under 10 CFR 50.59 as part of the normal core 
reload process.
    The change to the title of Reference 8 in Technical 
Specification 6.9.1.6.b is administrative in nature and does not 
alter any of the requirements of the affected TS. The proposed 
change does not modify any plant equipment and does not impact any 
failure modes that could lead to an accident. Additionally, the 
proposed change has no effect on the consequence of any analyzed 
accident since the change does not affect any equipment related to 
accident mitigation.
    Based on this discussion, the proposed change does not 
significantly 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 Specifications change adds Optimized 
ZIRLOTM to the approved fuel rod cladding materials that 
may be used at MPS3. Optimized ZIRLOTM was developed to 
provide a reduced cladding corrosion rate while maintaining the 
benefits of mechanical strength and resistance to accelerated 
corrosion from potential abnormal chemistry conditions. The fuel rod 
design bases are established to satisfy the general and specific 
safety criteria addressed in the MPS3 Final Safety Analysis Report 
(FSAR), Chapter 15 (Accident Analyses). The fuel rods are designed 
to prevent excessive fuel temperatures, excessive fuel rod internal 
gas pressures due to fission gas releases, and excessive cladding 
stresses and strains. Westinghouse topical report WCAP-12610-P-A & 
CENPD-404-P-A, Addendum 1-A, ``Optimized ZIRLOTM,'' 
provides the details and results of material testing of Optimized 
ZIRLOTM compared to standard ZIRLO[supreg], as well as 
the material properties to be used in various models and 
methodologies when analyzing Optimized ZIRLOTM. The 
original fuel design basis requirements have been maintained. No new 
single failure mechanisms will be created, and there are no 
alterations to plant equipment or procedures that would introduce 
any new or unique operational modes or accident precursors.
    Therefore, the proposed changes to the MPS3 TSs related to the 
use of Optimized ZIRLOTM do not create the possibility of 
a new or different kind of accident or malfunction from those 
previously evaluated within the FSAR.
    The change to the title of Reference 8 in Technical 
Specification 6.9.1.6.b is administrative in nature. It does not 
modify any plant equipment and there is no impact on the capability 
of the existing equipment to perform their intended functions. No 
system setpoints are being modified and no changes are being made to 
the method in which plant operations are conducted. No new failure 
modes are introduced by the proposed changes. The proposed change 
does not introduce accident initiators or malfunctions that would 
cause a new or different kind of accident.
    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 change involve a significant reduction in a 
margin of safety?
    Response: No.
    The cladding materials used for fuel rods are designed and 
tested to prevent excessive fuel temperatures, excessive fuel rod 
internal gas pressures due to fission .as releases, and excessive 
cladding stresses and strains. Optimized ZIRLOTM was 
developed to meet these needs while providing a reduced cladding 
corrosion rate and maintaining the benefits of mechanical strength 
and resistance to accelerated corrosion from potential abnormal 
chemistry conditions. Reducing the associated corrosion buildup 
reduces temperature feedback effects, providing additional margin to 
the fuel rod internal pressure design criterion. Westinghouse 
topical report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, 
``Optimized ZIRLOTM'' provides the details and results of 
material testing of Optimized ZIRLOTM compared to 
standard ZIRLO[supreg], as

[[Page 28630]]

well as the material properties to be used in various models and 
methodologies when analyzing Optimized ZIRLOTM. The NRC 
has previously approved use of the Optimized ZIRLOTM fuel 
cladding material as detailed in their Safety Evaluation for this 
topical report. The original fuel design basis requirements have 
been maintained, and evaluations will be performed under 10 CFR 
50.59 for each reload cycle that incorporates Optimized 
ZIRLOTM cladding to confirm that design and safety limits 
are satisfied. Therefore, inclusion of Optimized ZIRLOTM 
as an additional fuel rod cladding material for MPS3 does not result 
in a significant reduction in margin required to preclude or reduce 
the effects of an accident or malfunction previously evaluated in 
the FSAR.
    The change to the title of Reference 8 in Technical 
Specification 6.9.1.6.b is administrative in nature. This change 
does not alter any of the requirements of the affected TS. The 
proposed change does not affect any of the assumptions used in the 
accident analysis, nor does it affect any operability requirements 
for equipment important to plant safety.
    Therefore, the proposed change will not result in a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 
23219.
    NRC Branch Chief: George A. Wilson.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
    Date of amendment request: March 5, 2012.
    Description of amendment request: The proposed amendments would 
implement a measurement uncertainty recapture power uprate at the 
McGuire Nuclear Station, Units 1 and 2.
    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 3411 
megawatts thermal (MWt) to 3469 MWt; an increase of approximately 
1.7% 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 3469 MWt. A 
review of station accident analyses found that all acceptance 
criteria are still met at the uprated power of 3469 MWt.
    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 safety and support 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 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 structures, systems or components 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 design basis continues to be met after the measurement 
uncertainty recapture (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 McGuire safety analyses including the revised design 
basis radiological accident dose calculations, bound the power 
uprate.
    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.
Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414, 
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina
    Date of amendment request: November 22, 2011.
    Description of amendment request: The amendments would revise the 
Technical Specifications (TSs) to allow single discharge header 
operation of the nuclear service water system (NSWS) for a time period 
of 14 days.
    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:


[[Page 28631]]



First Standard

    Does operation of the facility in accordance with the proposed 
amendment involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The proposed single discharge header operation configuration for 
NSWS operation and the associated proposed TS and Bases changes have 
been evaluated to assess their impact on plant operation and to 
ensure that the design basis safety functions of safety related 
systems are not adversely impacted. During single discharge header 
operation, the operating NSWS header will be able to discharge all 
required NSWS flow from safety related components. [Probabilistic 
risk assessment] PRA has demonstrated that due to the limited 
proposed time in the single discharge header configuration, the 
resultant plant risk remains acceptable.
    The purpose of this amendment request is to ultimately 
facilitate inspection and maintenance of the Unit 2 NSWS discharge 
headers within the Auxiliary Building. Therefore, NRC approval of 
this request will ultimately help to enhance the long-term 
structural integrity of the NSWS and will help to ensure the 
system's reliability for many years.
    In general, the NSWS serves as an accident mitigation system and 
cannot by itself initiate an accident or transient situation. The 
only exception is that the NSWS piping can serve as a source of 
floodwater to safety related equipment in the Auxiliary Building or 
in the diesel generator buildings in the event of a leak or a break 
in the system piping. The probability of such an event is not 
significantly increased as a result of this proposed request. Safety 
related NSWS piping is tested and inspected in accordance with all 
applicable inservice testing and inservice inspection requirements. 
Given the negligible influence of flooding events on the NSWS for 
the submittal configuration (i.e., no dominant contribution from 
floods), it is judged that the analyses assessing the influence of 
these events provide an acceptable evaluation of the contribution of 
the flood risk for the requested Completion Time of 14 days.
    The proposed 14 day TS Required Action Completion Time has been 
evaluated for risk significance and the results of this evaluation 
have been found acceptable. The probabilities of occurrence of 
accidents presented in the [updated final safety analysis report] 
UFSAR will not increase as a result of implementation of this 
change. Because the PRA analysis supporting the proposed change 
yielded acceptable results, the NSWS will maintain its required 
availability in response to accident situations. Since NSWS 
availability is maintained, the response of the plant to accident 
situations will remain acceptable and the consequences of accidents 
presented in the UFSAR will not increase.

Second Standard

    Does operation of the facility in accordance with the proposed 
amendment create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: No.
    Implementation of this amendment will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated. The proposed request does not affect the basic operation 
of the NSWS or any of the systems that it supports. These include 
the Emergency Core Cooling System, the Containment Spray System, the 
Containment Valve Injection Water System, the Auxiliary Feedwater 
System, the Component Cooling Water System, the Control Room Area 
Ventilation System, the Control Room Area Chilled Water System, the 
Auxiliary Building Filtered Ventilation Exhaust System, or the 
Diesel Generators. During proposed single discharge header 
operation, the NSWS will remain capable of fulfilling all of its 
design basis requirements. No new accident causal mechanisms are 
created as a result of NRC approval of this amendment request. No 
changes are being made to the plant which will introduce any new 
type of accident outside those assumed in the UFSAR.

Third Standard

    Does operation of the facility in accordance with the proposed 
amendment involve a significant reduction in the margin of safety?
    Response: No.
    Implementation of this amendment will not involve a significant 
reduction in any margin of safety. Margin of safety is related to 
the confidence in the ability of the fission product barriers to 
perform their design functions during and following an accident 
situation. These barriers include the fuel cladding, the reactor 
coolant system, and the containment system. The performance of these 
fission product barriers will not be impacted by implementation of 
this proposed TS amendment. During single discharge header 
operation, the NSWS and its supported systems will remain capable of 
performing their required functions. No safety margins will be 
impacted.
    The PRA conducted for this proposed amendment demonstrated that 
the impact on overall plant risk remains acceptable during single 
discharge header operation. Therefore, there is not a significant 
reduction in the margin of safety.
    Based upon the preceding discussion, Duke Energy has concluded 
that the proposed amendment does not involve a significant hazards 
consideration.

    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.
Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1, Dauphin County, Pennsylvania
    Date of amendment request: March 26, 2012, as supplemented by 
letter dated April 2, 2012.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) Limiting Condition for 
Operation 3.1.1.2, TS Surveillance Requirement 4.19.2, TS 6.9.6 ``Steam 
Generator Tube Inspection Report,'' and TS 6.19 ``Steam Generator (SG) 
Program,'' changing certain inspection periods and making other 
administrative changes and clarifications. These proposed changes are 
consistent with Technical Specification Task Force (TSTF) Traveler, 
TSTF-510, Revision 2, ``Revision to Steam Generator Program Inspection 
Frequencies and Tube Sample Selection.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No
    The proposed change revises the Steam Generator (SG) Program to 
modify the frequency of verification of SG tube integrity and SG 
tube sample selection. A steam generator tube rupture (SGTR) event 
is one of the design basis accidents that are analyzed as part of a 
plant's licensing basis. The proposed SG tube inspection frequency 
and sample selection criteria will continue to ensure that the SG 
tubes are inspected such that the probability of a SGTR is not 
increased. The consequences of a SGTR are bounded by the 
conservative assumptions in the design basis accident analysis. The 
proposed change will not cause the consequences of a SGTR to exceed 
those assumptions. The proposed change to reporting requirements and 
clarifications of the existing requirements have no affect on the 
probability or consequences of SGTR.
    Therefore, it is concluded that 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 changes to the Steam Generator Program will not 
introduce any adverse changes to the plant design basis or 
postulated accidents resulting from potential tube degradation. The 
proposed change does not affect the design of the SGs or their

[[Page 28632]]

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

    The NRC staff has reviewed the licensee's analysis and based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: J. Bradley Fewell, Esquire, Associate 
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Meena Khanna.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo County, 
California
    Date of amendment request: October 24, 2011.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.3.5, ``Loss of Power (LOP) Diesel 
Generator (DG) Start Instrumentation,'' to correct the nonconservative 
first level undervoltage relays (FLUR) limits contained in Surveillance 
Requirement (SR) 3.3.5.3; revise the Final Safety Analysis Report 
Update (FSARU) Appendix 6.2D and Sections 6.3, 15.3, and 15.4; revise 
the loss-of-coolant accident (LOCA) control room operator and offsite 
dose analysis of record described in the FSARU; and provide a new 
process for revising input values to this analysis.
    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 change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The diesel generators (DGs) provide a source of emergency power 
when offsite power is either unavailable, or is degraded below a 
point that would allow safe unit operation. Undervoltage protection 
will generate a loss of power (LOP) DG start if a loss of voltage or 
degraded voltage condition occurs on the 4.16 kV [kilovolt] vital 
bus. The proposed technical specification (TS) change affects the 
voltage at which an emergency bus that is experiencing sustained 
degraded voltage will disconnect from offsite power and transfer to 
the DGs. While the TS limits are revised, the function remains the 
same and will continue to be performed. The first level undervoltage 
relays (FLUR) and second level undervoltage relays (SLUR) will 
continue to meet their required function to transfer 4.16 kV buses 
to the DGs in the event of insufficient offsite power voltage. This 
transfer will ensure that the class 1E equipment is capable of 
performing its function to meet the requirements of the accident 
analysis. The revised TS surveillance requirement (SR) 3.3.5.3 
setpoints will not cause unnecessary separation of engineered safety 
[feature] (ESF) loads from the 230 kV System. The proposed change 
does not affect any accident initiators or precursors.
    The ESF function delay times are bounding input parameters that 
represent actual plant performance for when these ESF functions can 
be credited to begin operating after an accident has already 
occurred. The increased ESF delay times are not physically related 
to the cause of any accident. Therefore, the increase in ESF delay 
times do not introduce the possibility of a change in the frequency 
of an accident previously evaluated. The revised LOCA control room 
operator and offsite dose analysis results remain within the 
applicable [General Design Criterion (GDC)] 19-1971 and 10 CFR 100 
limits. Therefore, the proposed activity does not result in an 
increase in the consequence of an accident previously evaluated in 
the FSARU.
    Therefore, the probability or consequences of any accident 
previously evaluated will not be significantly increased as a result 
of the proposed change.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of the proposed change. The revised surveillance requirements will 
continue to assure equipment reliability such that plant safety is 
maintained or will be enhanced. An increased ESF delay time is not 
an initiator of any accident and does not create any new system 
interactions or failure modes of any structures, systems or 
components (SSC).
    Equipment important to safety will continue to operate as 
designed. The changes do not result in adverse conditions or result 
in any increase in the challenges to safety systems. Therefore, 
operation of the Diablo Canyon Power Plant in accordance with the 
proposed amendment will not create the possibility of a new or 
different type of accident from any accident previously evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The DGs provide emergency electrical power to the safeguard 
buses in support of equipment required to mitigate the consequences 
of design basis accidents and anticipated operational occurrences, 
including an assumed loss of all offsite power. SR 3.3.5.3 verifies 
that the LOP DG start instrumentation channels respond to measured 
parameters within the necessary range and accuracy. The proposed 
amendment corrects nonconservative values in the TS limits for the 
degraded voltage protection function. The proposed change to this SR 
assures that design requirements of the emergency electrical power 
system continue to be met.
    There are no new or significant changes to the initial 
conditions contributing to accident severity or consequences. The 
proposed increase in ESF delay times is considered an analysis input 
change. However, the safety analyses continue to meet all applicable 
acceptance criteria. The proposed amendment will not otherwise 
affect the plant protective boundaries, will not cause a release of 
fission products to the public, nor will it degrade the performance 
of any other SSCs important to safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Jennifer Post, Esq., Pacific Gas and 
Electric Company, 77 Beale Street, Room 2496, Mail Code B30A, San 
Francisco, CA 94105.
    NRC Branch Chief: Michael T. Markley.

[[Page 28633]]

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo County, 
California
    Date of amendment request: January 25, 2012.
    Description of amendment request: The proposed amendment would 
revise the Diablo Canyon Power Plant, Units 1 and 2, Updated Final 
Safety Analysis Report Update (UFSAR) Section 4.3.2.2, ``Power 
Distribution,'' to allow use of the BEACON Power Distribution 
Monitoring System methodology described in Westinghouse Electric 
Company LLC (Westinghouse) WCAP-12472-P-A, Addendum 1-A, ``BEACON Core 
Monitoring and Operations Support System,'' dated January 2000.
    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 is to revise the Updated Final Safety 
Analysis Report to allow the use of the BEACON code methodology 
contained in WCAP-12472-P-A, Addendum 1-A. The BEACON code will be 
used to perform core flux mapping to support the performance of 
Technical Specification surveillances for power distribution limits 
and the use of the BEACON code will not cause an accident.
    No physical changes are being made to the plant. With the 
change, Diablo Canyon Power Plant will continue to operate within 
the power distribution limits contained in the plant Technical 
Specifications.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve any physical changes to the 
plant. The BEACON code performs flux mapping of the core and is not 
used to control the performance of any plant equipment. Therefore, 
use of the BEACON code cannot cause an accident. If it is determined 
that the plant is not operating within the power distribution limits 
during the performance of a Technical Specification Surveillance 
using BEACON, then the applicable Technical Specification Condition 
and Required Action(s) will be entered.
    Therefore, the proposed change does not create the possibility 
of a new or different accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    With the use of the BEACON code methodology contained in WCAP-
12472-P-A, Addendum 1-A, the plant will continue to operate within 
the power distribution limits contained in the plant Technical 
Specifications. The use of the BEACON code does not involve any 
changes to the fuel, reactor vessel, or containment fission product 
barriers. The use of the BEACON code methodology includes 
requirements for control of uncertainties associated with use of the 
methodology and therefore there will be no impact on the accident 
analyses that are contained in the Updated Final Safety Analysis 
Report.
    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 requests involve no significant hazards consideration.
    Attorney for licensee: Jennifer Post, Esq., Pacific Gas and 
Electric Company, 77 Beale Street, Room 2496, Mail Code B30A, San 
Francisco, CA 94105.
    NRC Branch Chief: Michael T. Markley.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant 
(WBN), Unit 1, Rhea County, Tennessee
    Date of amendment request: March 8, 2012.
    Description of amendment request: The proposed amendment would 
revise (1) Technical Specification (TS) 3.3.7, ``Control Room Emergency 
Ventilation System (CREVS) Actuation Instrumentation,'' by changing the 
Allowable Value for the main control room air intake radiation 
monitoring instrumentation in Table 3.3.7-1 from <= 9.45E-05 micro-
Curie/cubic centimeter ([micro]Ci/cc) (3,308 counts per minute (cpm)) 
to <= 1.647E-04 [micro]Ci/cc (3,308 cpm); and (2) TS 3.4.16, ``RCS 
[reactor coolant system] Specific Activity,'' by lowering the DOSE 
EQUIVALENT 1-131 spike limit from 21 micro-Curie/gram ([micro]Ci/gm) to 
14 [micro]Ci/gm in Required Action A.1 and Condition C.
    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 consequence of an accident previously evaluated?
    Response: No.
    The proposed TS changes do not adversely affect any fission 
product barrier nor do they alter the safety function of safety 
systems, structures, or components, or their roles in accident 
prevention or mitigation. They do not change any credited operator 
actions nor do they physically change any plant system, structure, 
or component. The amount of iodine in the primary coolant and the 
Allowable Value for the main control room radiation monitors do not 
affect the initiation of any accident or transient. Therefore, the 
proposed amendment does not result in a significant increase in the 
probability of an accident previously evaluated. The changes do not 
adversely affect the protective and mitigative capabilities of the 
plant. The SSCs [structures, systems, and components] will continue 
to perform their intended safety functions. The proposed reduction 
in the amount of DOSE EQUIVALENT 1-131 (DEI-131) in the reactor 
coolant following a load transient has no impact on any plant 
configuration or system performance relied upon to mitigate the 
consequences of an accident.
    The calculated radiological doses remain within the limits 
prescribed in 10 CFR Part 100 and GDC-19 [General Design Criterion 
19 of Appendix A to 10 CFR Part 50] and are consistent with the 
methodology and acceptance criteria of Section 15.6.3 of NUREG-0800 
and Appendix A of Section 15.1.5 of NUREG-0800.
    The change to the Allowable Value for the main control room 
radiation monitors continues to ensure that the monitors are capable 
of performing their intended design function of isolating the main 
control room subsequent to an accident.
    Therefore, the proposed amendment does not involve a significant 
increase in the consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed TS changes do not alter the configuration of the 
plant nor do they directly affect plant operation. The proposed TS 
changes do not result in the installation of any new equipment or 
system or the modification of any existing equipment or systems. No 
new operation procedures, conditions, or modes are created. As a 
result, the proposed TS changes do not introduce any new failure 
mechanisms, malfunctions, or accident initiators not already 
considered in the design and licensing basis. There will be no 
adverse effects or challenges imposed on any safety-related system 
as a result of these changes.
    Therefore, the proposed amendment 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 calculated radiological doses remain within the limits 
prescribed in 10 CFR Part 100 and GDC-19, and are consistent with 
the methodology and acceptance criteria of

[[Page 28634]]

Section 15.6.3 of NUREG-0800 and Appendix A of Section 15.1.5 of 
NUREG-0800. The Allowable Value for the main control room radiation 
monitor continues to ensure that the monitors are capable of 
performing their intended design function of isolating the main 
control room subsequent to an accident.
    Therefore, the proposed amendment does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Stephen J. Campbell.
Notice of Issuance of Amendments to Facility Operating Licenses and 
Combined Licenses
    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    A notice of consideration of issuance of amendment to facility 
operating license or combined license, as applicable, proposed no 
significant hazards consideration determination, and opportunity for a 
hearing in connection with these actions, was published in the Federal 
Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the NRC's Public Document Room (PDR), located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. Publicly available documents created or received at the NRC are 
accessible electronically through the Agencywide Documents Access and 
Management System (ADAMS) in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
PDR's Reference staff at 1-800-397-4209, 301-415-4737 or by email to 
[email protected].
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 application of amendments: May 6, 2010, as supplemented by 
letters dated February 11, 2011, April 28, 2011, July 19, 2011, and 
September 16, 2011.
    Brief description of amendments: The amendments revised the 
Technical Specifications related to supporting operation with 24-month 
fuel cycles. Specifically, the change would revise the frequency of 
certain TS Surveillance Requirements (SRs) from ``18 months'' to ``24 
months,'' in accordance with the guidance of Generic Letter (GL) 91-04, 
``Changes in Technical Specification Surveillance Intervals to 
Accommodate a 24-Month Fuel Cycle.''
    Date of Issuance: April 20, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1--379, Unit 2--381, and Unit 3--380.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: September 7, 2010 (75 
FR 54394). The supplements dated February 11, 2011, April 28, 2011, 
July 19, 2011, and September 16, 2011, 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 20, 2012.
    No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington
    Date of application for amendment: April 11, 2011.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.7.4, ``RCS Leakage Detection Instrumentation,'' to 
define a new time limit for restoring inoperable Reactor Coolant System 
(RCS) leakage detection instrumentation to operable status; to 
establish alternate methods of monitoring RCS leakage when one or more 
required monitors are inoperable; and to make TS Bases changes which 
reflect the proposed changes and more accurately reflect the contents 
of the facility design basis related to operability of the RCS leakage 
detection instrumentation. These changes are consistent with the 
guidance contained in NRC-approved Technical Specifications Task Force 
(TSTF) change traveler TSTF-514, Revision 3, ``Revise BWR [Boiling-
Water Reactor] Operability Requirements and Actions for RCS Leakage 
Instrumentation.'' The NRC announced the availability of this TS 
improvement in the Federal Register on December 17, 2010 (75 FR 79048), 
as part of the consolidated line item improvement process.
    Date of issuance: April 23, 2012.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 224.
    Facility Operating License No. NPF-21: The amendment revised the 
Facility Operating License.
    Date of initial notice in Federal Register: May 31, 2011 (76 FR 
31373).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 23, 2012.
    No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian 
Point Nuclear Generating Units 2 and 3 (IP2 and IP3), Westchester 
County, New York
    Date of application for amendment: September 16, 2011.
    Brief description of amendment: The amendment revises the Inservice 
Testing Program, Technical Specification (TS) 5.5.6 for IP2 and TS 
5.5.7 for IP3.
    Date of issuance: May 2, 2012.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 267 and 245.

[[Page 28635]]

    Facility Operating License Nos. DPR-26 and DPR-64: The amendment 
revised the License and the TSs.
    Date of initial notice in Federal Register: December 27, 2011 (76 
FR 80976).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 2, 2012.
    No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear 
Plant, Van Buren County, Michigan
    Date of application for amendment: April 6, 2011, supplemented by 
letter dated October 28, 2011.
    Brief description of amendment: The amendment revised Technical 
Specification 5.5.14, ``Containment Leak Rate Testing Program,'' by 
replacing the reference to RG 1.163, ``Performance-Based Containment 
Leak-Test Program,'' with a reference to Topical Report NEI 94-01, 
Revision 2-A, ``Industry Guideline for Implementing Performance-Based 
Option of 10 CFR Part 50, Appendix J,'' as the implementation document 
for the 10 CFR Part 50, Appendix J, Option B, performance-based 
containment leak rate testing program at the Palisades Nuclear Plant 
(PNP). This amendment allows PNP to extend its performance-based 
containment integrated leakage rate test (ILRT, or Type A test) 
interval up to 15 years. Accordingly, the licensee has also requested 
to extend its current Type A test interval from the current one-time 
approved 11.25 years to 15 years so that the next Type A test can be 
conducted by May 3, 2016, instead of the current due date of August 3, 
2012.
    Date of issuance: April 23, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 247.
    Facility Operating License No. DPR-20: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 14, 2011, (76 FR 
34766).
    The supplemental letter contained clarifying information and did 
not change the initial no significant hazards consideration 
determination, and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 23, 2012.
    No significant hazards consideration comments received: No.
Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi
    Date of application for amendment: October 28, 2011, as 
supplemented by letter dated January 26, 2012.
    Brief description of amendment: The amendment increased the numeric 
values of the Safety Limit Minimum Critical Power Ratio in Technical 
Specification Section 2.1.1.2 from 1.09 to 1.11 for two recirculation 
loop operation (TLO) and from 1.12 to 1.14 for single recirculation 
loop operation (SLO). The Minimum Critical Power Ratio Safety Limit 
values for both TLO and SLO are determined in accordance with the 
requirements set forth in NRC-approved General Electric Company (GE) 
licensing topical report NEDC-33173P, ``Applicability of GE Methods to 
Expanded Operating Domains,'' Revision 0, February 2006.
    Date of issuance: April 20, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No.: 189.
    Facility Operating License No. NPF-29: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: February 14, 2012 (77 
FR 8291).
    The supplemental letter dated January 26, 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 April 20, 2012.
    No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana
    Date of amendment request: April 13, 2011.
    Brief description of amendment: The amendment modified the 
Technical Specifications (TSs) as a result of a revised Fuel Handling 
Accident analysis. The new analysis determined that the current TSs may 
not be conservative for all scenarios. The amendment provides new 
applicability and/or action language in the TSs that includes load 
movements over irradiated fuel assemblies. Specifically, the amendment 
modified the following TSs: TS 3.3.3.1 (Radiation Monitoring 
Instrumentation); TS 3.7.6.1 (Control Room Emergency Air Filtration 
System); TS 3.7.6.3 (Control Room Air Temperature--Operating); TS 
3.7.6.4 (Control Room Air Temperature--Shutdown); TS 3.8.1.2 (A.C. 
[Alternating Current] Sources--Shutdown); TS 3.8.2.2 (D.C. [Direct 
Current] Sources--Shutdown); TS 3.8.3.2 (Onsite Power Distribution--
Shutdown); TS 3.9.3 (Decay Time); TS 3.9.4 (Containment Building 
Penetrations); and TS 3.9.7 (Crane Travel--Fuel Handling Building).
    Date of issuance: April 25, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 235.
    Facility Operating License No. NPF-38: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: August 23, 2011 (76 FR 
52701).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 25, 2012.
    No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois Docket Nos. STN 
50-454 and STN 50-455, Byron Station, Units 1 and 2, Ogle County, 
Illinois
    Date of application for amendment: March 14, 2011, as supplemented 
by letters dated September 2, 2011, and November 18, 2011.
    Brief description of amendment: The license amendment request 
changes the facility operating licenses and the Technical 
Specifications (TSs) 3.4.12-1, for the Braidwood Station, Units 1 and 2 
and Byron Station, Units 1 and 2. The proposed change will reflect 
standard wording incorporated in NUREG-1431, Revision 3, ``Standard 
Technical Specifications--Westinghouse Plants,'' for plants with 
installed bypass test capability. The proposed change is needed to 
support utilization of bypass test capability that is planned to be 
installed, which will reduce the potential for unnecessary reactor 
trips or safeguards actuation due to a failure or transient in a 
redundant channel.
    Date of issuance: March 30, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.

[[Page 28636]]

    Amendment Nos.: Braidwood Unit 1--169; Braidwood Unit 2--169; Byron 
Unit 1--176 and Byron Unit 2--176.
    Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66: 
The amendments revised the Technical Specifications and License.
    Date of initial notice in Federal Register: August 16, 2011 (76 FR 
50759).
    The September 2, 2011, and November 18, 2011, supplements contained 
clarifying information and did not change the staff's initial proposed 
finding of no significant hazards consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 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 (PBAPS), Units 2 and 
3, York and Lancaster Counties, Pennsylvania
    Date of application for amendments: April 6, 2011.
    Brief description of amendments: The amendment modifies the actions 
to be taken when the atmospheric gaseous radioactivity monitor is the 
only operable reactor coolant leakage detection instrument. The 
modified actions require additional, more frequent monitoring of other 
indications of Reactor Coolant System (RCS) leakage and provide 
appropriate time to restore another leakage detection instrument to 
operable status. This change is consistent with the U.S. Nuclear 
Regulatory Commission (NRC) approved safety evaluation on Technical 
Specification Task Force (TSTF) Traveler, TSTF-514-A, Revision 3, 
``Revised BWR [boiling-water reactor] Operability Requirements and 
Actions for RCS Leakage Instrumentation'' dated November 24, 2010.
    Date of issuance: April 23, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 283 and 286.
    Renewed Facility Operating License Nos. DPR-44 and DPR-56: 
Amendments revised the License and Technical Specifications.
    Date of initial notice in Federal Register: September 6, 2011, (76 
FR 55128).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 23, 2012.
    No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric 
Station, Unit 2, Luzerne County, Pennsylvania
    Date of application for amendment: March 8, 2012, as supplemented 
by letters dated March 23, March 29, and April 2, 2012.
    Brief description of amendment: The amendment allows an extension 
of 24 hours to the Completion Time for Condition C in the Susquehanna 
Steam Electric Station (SSES) Unit 2 Technical Specification (TS) 
3.8.7, ``Distribution Systems--Operating,'' to allow a Unit 1 4160 V 
subsystem to be de-energized and removed from service for 96 hours to 
perform modifications on the bus. It also allows an extension of 24 
hours to the Completion Time for Condition A in SSES Unit 2 TS 3.7.1, 
``Plant Systems--RHRSW [residual heat removal service water system] and 
UHS [ultimate heat sink],'' to allow the UHS spray array and spray 
array bypass valves associated with applicable division RHRSW, and in 
Condition B, the applicable division Unit 2 RHRSW subsystem, to be 
inoperable for 96 hours during the Unit 1 4160 V bus breaker control 
logic modifications.
    Date of issuance: April 19, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 258.
    Facility Operating License No. NPF-22: This amendment revised the 
License and Technical Specifications.
    Date of initial notice in Federal Register: March 16, 2012 (77 FR 
15814).
    The supplements dated March 23, March 29, and April 2, 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 April 19, 2012, which also contains its 
final no significant hazards consideration determination.
    No significant hazards consideration comments received: No.
South Carolina Electric and Gas Company, Docket No. 50-395, Virgil C. 
Summer Nuclear Station, Unit 1, Jenkinsville, South Carolina
    Date of application for amendment: October 12, 2011, as 
supplemented by letter dated April 5, 2012.
    Brief description of amendment: This amendment revised the Virgil 
C. Summer Nuclear Station (VCSNS) Technical Specification to allow a 
one-time extension of the 10-year interval for the containment 
integrated leakage rate test such that the existing test interval would 
be extended from 120 months to 130 months.
    Date of Issuance: May 1, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No: 189.
    Renewed Facility Operating License No. NPF-12: Amendment revises 
the License and Technical Specifications.
    Date of initial notice in Federal Register: December 13, 2011 (76 
FR 77571).
    The licensee's supplemental letter contained clarifying 
information, did not change the scope of the original license amendment 
request, did not change the NRC staff's initial proposed finding of no 
significant hazards consideration determination, and did not expand the 
scope of the original Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 1, 2012.
    No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear 
Plant, Unit 1, Limestone County, Alabama
    Date of application for amendment: April 16, 2010, as supplemented 
by letters dated February 23, May 12, October 7, 2011, and April 18, 
2012 (TS-473).
    Brief description of amendment: The licensee proposes to transition 
Unit 1 to AREVA fuel. To support the transition to AREVA fuel, the 
proposed amendment adds the AREVA NP analysis methodologies to the list 
of approved methods to be used in determining the core operating limits 
in the core operating limits report. Additional technical specification 
changes are requested to reflect the AREVA NP specific methods for 
monitoring and enforcing of the thermal limits. The licensee's request 
is for non-extended power uprate conditions (i.e., 105 percent of 
Original Licensed Thermal Power level) only.
    Date of issuance: April 27, 2012.
    Effective date: This license amendment is effective as of its date 
of issuance and shall be implemented within 30 days from the date of 
issuance.
    Amendment No.: 281.

[[Page 28637]]

    Renewed Facility Operating License No. DPR-33: Amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: January 10, 2011 (76 FR 
1467). The supplemental letters provided clarifying information that 
did not expand the scope of the original application or change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 27, 2012.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 3rd day of May 2012.

    For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2012-11599 Filed 5-14-12; 8:45 am]
BILLING CODE 7590-01-P