[Federal Register Volume 76, Number 35 (Tuesday, February 22, 2011)]
[Notices]
[Pages 9821-9832]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2011-3721]


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NUCLEAR REGULATORY COMMISSION

[NRC-2011-0040]


Biweekly Notice Applications and Amendments to Facility Operating 
Licenses Involving No Significant Hazards Considerations

I. 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 
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 January 27, 2011, to February 10, 2011. The 
last biweekly notice was published on February 8, 2011 (76 FR 6830).

Notice of Consideration of Issuance of Amendments to Facility Operating 
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), Section 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.
    Written comments may be submitted by mail to the Chief, Rules, 
Announcements and Directives Branch (RADB), TWB-05-B01M, Division of 
Administrative Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be faxed to the RADB at 301-492-3446. 
Documents may be examined, and/or copied for a fee, 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.
    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. 
Requests for a hearing and a petition for leave to intervene shall be 
filed in accordance with the Commission's

[[Page 9822]]

``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 Commission's PDR, located at One White Flint 
North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. Publicly available records will be accessible from the 
Agencywide Documents Access and Management System's (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, 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, 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 
ten (10) days prior to the filing deadline, the participant should 
contact the Office of the Secretary by e-mail at 
[email protected], or by telephone at (301) 415-1677, to request 
(1) a digital ID certificate, which allows the participant (or its 
counsel or representative) to digitally sign documents and access the 
E-Submittal server for any proceeding in which it is participating; and 
(2) advise the Secretary that the participant will be submitting a 
request or petition for hearing (even in instances in which the 
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the 
Secretary will establish an electronic docket for the hearing in this 
proceeding if the Secretary has not already established an electronic 
docket.
    Information about applying for a digital ID certificate is 
available on NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing 
the E-Submittal server are detailed in NRC's ``Guidance for Electronic 
Submission,'' which is available on the agency's public Web site at 
http://www.nrc.gov/site-help/e-submittals.html. Participants may 
attempt to use other software not listed on the Web site, but should 
note that the NRC's E-Filing system does not 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 EIE, users will be required to install a Web 
browser plug-in from the NRC Web site. Further information on the Web-
based submission form, including the installation of the Web browser 
plug-in, is available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with NRC guidance 
available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an e-mail notice confirming receipt of the document. The 
E-Filing system also distributes an e-mail notice that provides access 
to the

[[Page 9823]]

document to the NRC 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 e-mail 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 
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 Commission's PDR, located at One White Flint 
North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. Publicly available records will be accessible from the ADAMS 
Public Electronic Reading Room on the Internet at the NRC Web site, 
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 PDR Reference staff at 1-800-
397-4209, 301-415-4737, or by e-mail to [email protected].

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: January 13, 2011.
    Description of amendment request: The proposed amendment would 
modify the Facility Operating License (FOL) by deleting references to 
specific Safety Evaluation Reports (SER), Technical Specification (TS) 
Amendments, and Exemptions from License Condition 2.C(3), Fire 
Protection, and replacing them with the words ``as supplemented.'' This 
is an administrative amendment to the FOL.
    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. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The proposed FOL change is administrative and does not involve a 
plant or design function change. It has no effect on reactor 
operation or accident analyses, and thus, the proposed FOL change 
does not increase the probability or consequence of an accident 
previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: No.
    The proposed FOL change is administrative and does not involve a 
plant or design function change. Because the proposed amendment 
would not change the design, configuration, or method of operation 
of the plant, it would not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    The proposed FOL change is administrative and does not involve a 
plant or design function change. No design or safety margin is 
involved. Therefore, the proposed change does not involve a 
reduction in any margin of safety.

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

Entergy Nuclear Vermont Yankee (VY), LLC and Entergy Nuclear 
Operations, Inc., Docket No. 50-271, Vermont Yankee Nuclear Power 
Station, Vernon, Vermont

    Date of amendment request: December 21, 2010.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TS) Section 3.6.A ``Pressure and 
Temperature Limitation'' to reflect the pressure and temperature (P-T) 
limits for the reactor coolant system through, approximately the end of 
the prospective 20-year

[[Page 9824]]

renewed license period, depending on the plant capacity factor.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the period of applicability of the 
P-T limits. The technical bases for the new period of applicability 
have been previously reviewed and approved by the NRC as discussed 
in the submittal. Because the applicable regulatory requirements 
continue to be met, the change does not significantly increase the 
probability of any accident previously evaluated. The proposed 
change provides the same assurance of RPV integrity as previously 
provided.
    The change will require that the reactor pressure vessel and 
interfacing coolant system continue to be operated within their 
design, operational or testing limits. Also, the change will not 
alter any assumptions previously made in evaluating the radiological 
consequences of accidents.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve a modification of the 
design of plant structures, systems, or components. The change will 
not impact the manner in which the plant is operated and will not 
degrade the reliability of structures, systems, or components 
important to safety as equipment protection features will not be 
deleted or modified, equipment redundancy or independence will not 
be reduced, supporting system performance will not be affected and 
no severe testing of equipment will be imposed. No new failure modes 
or mechanisms will be introduced as a result of this proposed 
change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Appendix G to 10 CFR 50 describes the conditions that require 
pressure-temperature (P-T) limits and provides the general bases for 
these limits. Operating limits based on the criteria of Appendix G, 
as defined by applicable regulations, codes and standards, provide 
reasonable assurance that non-ductile or rapidly propagating failure 
will not occur. The P-T limits are prescribed for all plant modes to 
avoid encountering pressure, temperature and temperature rate of 
change conditions that might cause undetected flaws to propagate and 
cause non-ductile failure of the reactor coolant pressure boundary. 
Calculation of P-T limits in accordance with the criteria of 
Appendix G to 10 CFR 50 and applicable regulatory requirements 
ensures that adequate margins of safety are maintained and there is 
no significant reduction in a margin of safety.
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. There is no change or impact on any safety 
analysis assumption or in any other parameter affecting the course 
of an accident analysis supporting the basis of any Technical 
Specification. The proposed change does not involve any increase in 
calculated off-site dose consequences.
    Therefore, operation of VY in accordance with the proposed 
amendment will not involve a significant reduction in a margin to 
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., 400 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Nancy Salgado.

FirstEnergy Nuclear Operating Company (FENOC), et al., Docket No. 50-
440, Perry Nuclear Power Plant, Unit 1 (PNPP), Lake County, Ohio

    Date of amendment request: December 15, 2010.
    Description of amendment request: The proposed amendment would 
modify the requirements for testing control rod scram times following 
fuel movement within the reactor pressure vessel by incorporating 
Nuclear Regulatory Commission (NRC) approved Technical Specification 
Task Force (TSTF) change traveler TSTF-222-A, Revision 1.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (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 control rod drive system is not an initiator to any accident 
sequence analyzed in the PNPP Updated Final Safety Analysis Report 
(USAR), including Appendix 15C, ``Anticipated Transients Without 
Scram (ATWS).'' The proposed TS changes improve existing 
surveillance requirements by eliminating unnecessary control rob 
scram time testing, while continuing to provide adequate assurance 
of control rod performance for those control rods in core cells in 
which fuel is moved or replaced, or control rod maintenance was 
performed.
    Historically, testing results indicate the control rod drive 
system is highly reliable. Since the fall 1996 implementation of 
Improved Technical Specifications, during 6036 control rod tests 
covering all 177 control rods, only 7 control rod tests (0.12 
percent) yielded results slower than the required insertion time 
limit, and no control rods were inoperable as a result of scream 
time testing. All seven slow insertion time test results have been 
attributed to control rod scream solenoid pilot valves (SSPVs). 
These seven slow tests occurred prior to May 1999, and prior to a 
control rod SSPV upgrade program during which all 177 SSPV's were 
replaced.
    As such, this type of change does not affect initiators of 
analyzed events and does not affect the mitigation of any accidents 
or transients.
    Therefore, the proposed TS changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed TS changes do not involve a physical alteration of 
the plant. No new equipment is being introduced, and installed 
equipment is not being operated in a new or different manner. There 
are no setpoints affected by the changes at which protective or 
mitigative actions are initiated. The changes will not alter the 
manner in which equipment operation is initiated, nor will the 
functional demands on credited equipment be changed. No alterations 
in the procedures that ensure the plant remains within analyzed 
limits are being proposed, and no changes are being made to the 
procedures relied upon to respond to an off-normal event as 
described in the USAR. This change does not alter assumptions made 
in the safety analysis and licensing basis. As such, no new failures 
modes are being introduced. Accordingly, the proposed changes do not 
create any new credible failure mechanisms, malfunction, or accident 
initiators not previously considered in PNPP design and licensing 
basis.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Margin of safety is related to the ability of the fission 
product barriers to perform their design functions during and 
following accident conditions. These barriers include the fuel 
cladding, the reactor coolant system, and the containment. This 
request does not involve a change to the fuel cladding, the reactor 
coolant system, or the containment.
    The proposed TS changes associated with TSTF-222-1 modify 
current frequency requirements for scram time testing control rods 
following refueling outages and for control rod requiring testing 
due to work

[[Page 9825]]

activities. Scram times for control rods not affected by fuel 
movement or control rod maintenance remain unaffected.
    The proposed TS changes have no affect on any safety analysis 
assumptions or methods of performing safety analyses. The changes do 
not adversely affect system design or operational requirements, and 
the equipment continues to be tested in a manner and at a frequency 
necessary to provide confidence that the equipment can perform its 
intended safety functions.
    Therefore, the proposed TS 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: David W. Jenkins, Attorney, FirstEnergy 
Corporation, Mail Stop. A-GO-15, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Robert D. Carlson.

FirstEnergy Nuclear Operating Company (FENOC), et al., Docket No. 50-
440, Perry Nuclear Power Plant, Unit 1 (PNPP), Lake County, Ohio

    Date of amendment request: December 15, 2010
    Description of amendment request: The proposed amendment would 
revise the required testing frequency of Surveillance Requirement (SR) 
3.1.4.2 from ``120 days cumulative operation in MODE 1'' to ``200 days 
cumulative operation in MODE 1'' by incorporating Nuclear Regulatory 
Commission (NRC) approved Technical Specification Task Force (TSTF) 
change traveler TSTF-460, Revision 0.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The frequency 
of surveillance testing is not an initiator of any accident 
previously evaluated. The frequency of surveillance testing does not 
affect the ability to mitigate any accident previously evaluated, as 
the tested component is still required to be operable.
    Therefore, the proposed TS changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The proposed 
change does not result in any new or different modes of plant 
operation.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The proposed 
change continues to test the control rod scram time to ensure the 
assumptions in the safety analysis are protected.
    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: David W. Jenkins, Attorney, FirstEnergy 
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Robert D. Carlson.

FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of amendment request: October 15, 2010.
    Description of amendment request: The proposed amendment would 
revise Operating License No. DPR-49 by modifying the License to delete 
the parent guarantee License Condition.
    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 is an administrative change deleting the 
parent guarantee License Condition, as well as other minor editorial 
changes in format. Deletion of this License Condition does not 
involve any modifications to the safety-related structures, systems 
or components (SSCs). Deletion of this License Condition will not 
alter previously evaluated Final Safety Analysis Report (FSAR) 
design basis accident analysis assumptions, add any accident 
initiators, or affect the function of the plant safety-related SSCs 
as to how they are operated, maintained, modified, tested, or 
inspected. Therefore, the proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment only deletes the parent guarantee License 
Condition and makes other minor editorial changes. Deletion of this 
License Condition does not result in the need for any new or 
different FSAR design basis accident analysis. It does not introduce 
new equipment that could create a new or different kind of accident, 
and no new equipment failure modes are created. As a result, no new 
accident scenarios, failure mechanisms, or limiting single failures 
are introduced as a result of this proposed amendment. Therefore, 
the proposed amendment does not create a possibility for an accident 
of a new or different type than those previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The margin of safety is associated with the confidence in the 
ability of the fission product barriers (i.e., fuel cladding, 
reactor coolant pressure boundary, and containment structure) to 
limit the level of radiation to the public. The proposed amendment 
would not alter the way any safety-related SSC functions and would 
not alter the way the plant is operated. The amendment only involves 
deletion of the parent guarantee License Condition and minor 
editorial changes. The proposed amendment would not introduce any 
new uncertainties or change any existing uncertainties associated 
with any safety limit. The proposed amendment would have no impact 
on the structural integrity of the fuel cladding, reactor coolant 
pressure boundary, or containment structure. Based on the above 
considerations, the proposed amendment would not degrade the 
confidence in the ability of the fission product barriers to limit 
the level of radiation to the public. 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: Ms. Marjan Mashhadi, Florida Power & Light 
Company, 801 Pennsylvania Avenue, NW., Suite 220, Washington, DC 20004.

[[Page 9826]]

    NRC Branch Chief: Robert J. Pascarelli.

Indiana Michigan Power Company (the licensee), Docket No. 50-315, 
Donald C. Cook Nuclear Plant, Unit 1 (DCCNP-1), Berrien County, 
Michigan

    Date of amendment request: December 16, 2010.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 4.2.1, adding Optimized ZIRLO \TM\ 
fuel rods to the fuel matrix in addition to Zircaloy or ZIRLO fuel rods 
that are currently in use. The proposed amendment would also add a 
Westinghouse topical report regarding Optimized ZIRLO \TM\ as reference 
8 in TS 5.6.5.b, which lists the analytical methods used to determine 
the core operating limits.
    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 would allow the use of Optimized ZIRLO \TM\ 
clad nuclear fuel in the reactors. The NRC approved topical report 
WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A ``Optimized ZIRLO 
\TM\,'' prepared by Westinghouse Electric Company LLC 
(Westinghouse), addresses Optimized ZIRLO \TM\ and demonstrates that 
Optimized ZIRLO \TM\ has essentially the same properties as 
currently licensed ZIRLO \TM\. The fuel cladding itself is not an 
accident initiator and does not affect accident probability. Use of 
Optimized ZIRLO \TM\ fuel cladding has been shown to meet all 10 CFR 
50.46 acceptance criteria and, therefore, will not increase the 
consequences of an accident.
    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.
    Use of Optimized ZIRLO \TM\ clad fuel will not result in changes 
in the operation or configuration of the facility. Topical Report 
WCAP-12610-P-A and CENPD-404-P-A demonstrated that the material 
properties of Optimized ZIRLO \TM\ are similar to those of standard 
ZIRLO \TM\. Therefore, Optimized ZIRLO \TM\ fuel rod cladding will 
perform similarly to those fabricated from standard ZIRLO \TM\, thus 
precluding the possibility of the fuel becoming an accident 
initiator and causing a new or different type of accident.
    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 not involve a significant reduction in 
the margin of safety because it has been demonstrated that the 
material properties of the Optimized ZIRLO \TM\ are not 
significantly different from those of standard ZIRLO \TM\. Optimized 
ZIRLO \TM\ is expected to perform similarly to standard ZIRLO \TM\ 
for all normal operating and accident scenarios, including both loss 
of coolant accident (LOCA) and non-LOCA scenarios. For LOCA 
scenarios, where the slight difference in Optimized ZIRLO \TM\ 
material properties relative to standard ZIRLO \TM\ could have some 
impact on the overall accident scenario, plant-specific LOCA 
analyses using Optimized ZIRLO \TM\ properties will be performed 
prior to the use of fuel assemblies with fuel rods containing 
Optimized ZIRLO \TM\. These LOCA analyses will demonstrate that the 
acceptance criteria of 10 CFR 50.46 will be satisfied when Optimized 
ZIRLO \TM\ fuel rod cladding is implemented.
    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: James M. Petro, Jr., Senior Nuclear Counsel, 
Indiana Michigan Power Company, One Cook Place, Bridgman, MI 49106.
    NRC Branch Chief: Robert J. Pascarelli.

Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket No. 50-220, Nine 
Mile Point Nuclear Station Unit 1 (NMP1), Oswego County, New York

    Date of amendment request: September 29, 2010.
    Description of amendment request: The proposed amendment would 
revise the NMP1 Technical Specifications (TSs) Section 3/4.1.5, 
``Solenoid-Actuated Pressure Relief Valves (Automatic Depressurization 
System),'' and 3/4.2.9, ``Pressure Relief Systems--Solenoid-Actuated 
Pressure Relief Valves (Overpressurization),'' to provide for an 
alternative means of testing the main steam electromatic relief valves 
(ERVs). Specifically, the proposed amendment would revise TS 
Surveillance Requirements (SRs) 4.1.5.a and 4.2.9.b to verify each ERV 
actuator strokes when manually actuated at least once each operating 
cycle. The functional testing requirements for the ERVs would be 
described in the Inservice Testing (IST) Program and controlled 
pursuant to TS Administrative Controls Section 6.5.4, ``Inservice 
Testing Program.'' The proposed change would allow demonstration of the 
capability of the valves to perform their safety function without 
requiring the ERVs to be cycled with reactor steam pressure while 
installed in the plant.
    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 revises the TS Surveillance Requirements 
(SRs) to provide for an alternative means of testing the main steam 
ERVs. The ERVs perform automatic depressurization system (ADS) and 
overpressurization relief mode safety functions to mitigate the 
consequences of a small break loss of coolant accident (SBLOCA) and 
other accidents and transients. The ERVs are not considered an 
initiator for any accident previously evaluated except for the 
stuck-open ERV event, which is evaluated in Section XV-B.3.11 of the 
NMP1 Updated Final Safety Analysis Report (UFSAR). The proposed 
amendment would allow demonstration of the capability of the valves 
to perform their safety function through a series of tests, 
inspections, and maintenance activities without requiring the ERVs 
to be cycled with reactor steam pressure while installed in the 
plant, thereby eliminating the possibility of a stuck-open ERV event 
due to testing. Thus, the proposed amendment does not increase the 
probability of a stuck-open ERV event. The testing methodology, 
comprehensive inspections and preventive maintenance, and associated 
programmatic controls will provide an equivalent level of assurance 
that the ERVs are capable of performing their intended accident 
mitigation safety functions and, as such, will have no effect on the 
types or amounts of radiation released or the predicted offsite 
doses in the event of an accident. Accordingly, the proposed 
amendment does not alter the initial conditions, assumptions, or 
conclusions of any accident analysis.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment does not affect the assumed accident 
performance of the ERVs, or of any plant structure, system, or 
component previously evaluated. The proposed amendment does not 
involve the

[[Page 9827]]

installation of new equipment, and installed equipment is not being 
operated in a new or different manner. The proposed amendment 
provides for an alternative means of testing the ERVs that does not 
involve 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 ERVs are 
capable of performing their accident mitigation safety functions. No 
setpoints are being changed that would alter the dynamic response of 
plant equipment. As such, the proposed amendment will not introduce 
any new failure modes.
    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 provides for an alternative means of 
testing the ERVs, in that the testing requirements will be satisfied 
by a combination of required testing in accordance with the 
Inservice Testing Program (controlled in accordance with TS 
administrative controls) and the revised TS SRs. The proposed 
changes will provide a complete verification of the functional 
capability of the ERVs 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 ERVs 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 ERV 
leakage resulting from testing the ERVs 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: Carey W. Fleming, Senior Counsel, 
Constellation Energy Nuclear Group, LLC, 100 Constellation Way, Suite 
200C, Baltimore, MD 21202.
    NRC Branch Chief: Nancy L. Salgado.

Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue 
County, Minnesota

    Date of amendment request: February 3, 2011.
    Description of amendment request: The proposed amendments would 
revise the Technical Specification (TS) 3.8.1, ``AC Sources--
Operating'', Surveillance Requirement 3.8.1.10 footnote, which concerns 
battery charger modifications to be installed during or prior to the 
Unit 1 2011 refueling outage. The proposed change will allow use of 
different battery charger modifications to those considered when the 
footnote was added to the TS.
    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.
    This license amendment request proposes to revise the footnote 
to the emergency diesel generator Technical Specification 
surveillance requirement for loss of offsite power with safety 
injection actuation. The proposed footnote revision removes some 
specific requirements for battery charger modifications but will 
continue to assure that the applicable emergency diesel generator 
and its associated battery charger perform their required safety 
functions.
    The emergency diesel generators and their associated battery 
chargers are not accident initiators and therefore, these changes do 
not involve a significant increase [in] the probability of an 
accident.
    The proposed changes to the Technical Specification footnote 
will assure that the emergency diesel generator and the associated 
battery charger continue to perform their required safety function. 
Since the emergency diesel generator and the associated battery 
charger will provide required electrical power as assumed in the 
accident analyses, the results of the previous accident analyses are 
not changed and the changes proposed in this license amendment 
request do not involve a significant increase in the consequences of 
an accident.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    This license amendment request proposes to revise the footnote 
to the emergency diesel generator Technical Specification 
surveillance requirement for loss of offsite power with safety 
injection actuation. The proposed footnote revision removes some 
specific requirements for battery charger modifications but will 
continue to assure that the applicable emergency diesel generator 
and its associated battery charger perform their required safety 
functions.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed change. 
The proposed change does not challenge the performance or integrity 
of any safety-related system. The proposed change does involve 
modification of plant battery chargers, however, failures of battery 
chargers has been previously considered and bounded by assuming one 
safety related train of equipment fails. The modified battery 
chargers do not create new failure modes or mechanisms and no new 
accident precursors are generated. Surveillance testing requirements 
for the emergency diesel generator and battery charger will continue 
to demonstrate that the Limiting Conditions for Operation are met 
and the system components are functional.
    Therefore, the proposed changes do 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.
    This license amendment request proposes to revise the footnote 
to the emergency diesel generator Technical Specification 
surveillance requirement for loss of offsite power with safety 
injection actuation. The proposed footnote revision removes some 
specific requirements for battery charger modifications but will 
continue to assure that the applicable emergency diesel generator 
and its associated battery charger perform their required safety 
functions.
    The proposed Technical Specification footnote change does not 
affect the availability, operability, or performance of safety-
related systems and components: The affected emergency diesel 
generator and its associated battery will continue to perform their 
safety functions. The ability of operable structures, systems, and 
components to perform their designated safety function is unaffected 
by this proposed change. The proposed change does not involve a 
significant reduction in a margin of safety because the proposed 
footnote changes do not reduce the margin of safety that exists in 
the present Technical Specifications or Updated Safety Analysis 
Report. The operability requirements of the Technical Specifications 
are consistent with the initial condition assumptions of the safety 
analyses and the surveillance testing requirements will continue to 
demonstrate the operability of the emergency diesel generator.
    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

[[Page 9828]]

satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401
    NRC Branch Chief: Robert J. Pascarell.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

    Date of amendment request: November 10, 2010.
    Description of amendment request: The change to the PPL 
Susquehanna, LLC (PPL) Unit 1 and Unit 2 Technical Specification (TS) 
Surveillance Requirement (SR) 3.4.3.1 ``Safety/Relief Valves (S/RVs)'' 
proposes a new safety function lift setpoint lower tolerance for the S/
RVs. The proposed change will revise the lower tolerances from -3% to -
5%. This change would be limited to the lower tolerances and does not 
affect the upper tolerances. This change only applies to the lower as-
found tolerance and not to the as-left tolerance, which will remain 
unchanged at 1% of the safety lift setpoint. The as-found 
tolerances are used for determining past operability and to increase 
sample sizes for S/RV testing should the upper tolerance be exceeded. 
There will be no revision to the actual setpoints of the valves 
installed in the plant due to this change.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    This change has no influence on the probability or consequences 
of any accident previously evaluated. The lower setpoint tolerance 
change does not affect the operation of the valves and it does not 
change the as-left setpoint tolerance. The change only affects the 
lower tolerance for opening the valve and does not change the upper 
tolerance, which is the limit that protects from overpressurization.
    The proposed action does not involve physical changes to the 
valves, nor does it change the safety function of the valves. The 
proposed TS revision involves no significant changes to the 
operation of any systems or components in normal or accident 
operating conditions and no changes to existing structures, systems, 
or components.
    The proposed action does not change any other behavior or 
operation of any S/RVs, and, therefore, has no significant impact on 
reactor operation. It also has no significant impact on response to 
any perturbation of reactor operation including transients and 
accidents previously analyzed in the Final Safety Analysis Report 
(FSAR).
    Therefore, the proposed amendment does not result in a 
significant increase in the probability or consequences of any 
previously evaluated accident.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed lower setpoint tolerance change only affects the 
criteria to determine when an as-found S/RV test is considered to be 
acceptable. This change does not affect the criteria for the upper 
setpoint tolerance.
    The proposed lower setpoint tolerance change does not adversely 
affect the operation of any safety-related components or equipment. 
Since the proposed action does not involve hardware changes, 
significant changes to the operation of any systems or components, 
nor change to existing structures, systems, or components, there is 
no possibility that a new or different kind of accident is created.
    The proposed change does not involve physical changes to the S/
RVs, nor does it change the safety function of the S/RVs. The 
proposed change does not require any physical change or alteration 
of any existing plant equipment. No new or different equipment is 
being installed, and installed equipment is not being operated in a 
new or different manner. There is no alteration to the parameters 
within which the plant is normally operated. This change does not 
alter the manner in which equipment operation is initiated, nor will 
the functional demands on credited equipment be changed. No 
alterations in the procedures that ensure the plant remains within 
analyzed limits are being proposed, and no changes are being made to 
the procedures relied upon to respond to an off-normal event as 
described in the FSAR. As such, no new failure modes are being 
introduced. The change does not alter assumptions made in the safety 
analysis and licensing basis.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed lower setpoint tolerance change only affects the 
criteria to determine when an as-found S/RV test is considered to be 
acceptable. This change does not affect the criteria for the upper 
setpoint tolerance. The TS setpoints for the S/RVs are not changed. 
The as-left setpoint tolerances are not changed by this proposed 
change.
    The margin of safety is established through the design of the 
plant structures, systems, and components, the parameters within 
which the plant is operated, and the establishment of the setpoints 
for the actuation of equipment relied upon to respond to an event. 
The proposed change does not significantly impact the condition or 
performance of structures, systems, and components relied upon for 
accident mitigation.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Branch Chief : Nancy L. Salgado.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, 
Georgia, and Southern Nuclear Operating Company, Inc., Georgia Power 
Company, Oglethorpe Power Corporation, Municipal Electric Authority of 
Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin 
I. Hatch Nuclear Plant, Unit 1 and 2, Appling County, Georgia

    Date of amendment request: December 16, 2010.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) Section 2.0 ``Safety Limits.'' 
Specifically, the removal of the requirement to report a Safety Limit 
Violation, that is redundant to existing regulations, Title 10 of the 
Code of Federal Regulations (10 CFR), Part 50.36(c)(8) ``Written 
Reports.'' The proposed change is described in Technical Specification 
Task Force Traveler TSTF-5-A, Revision 1, ``Delete Safety Limit 
Violation Notification Requirements,'' (Agencywide Documents Access and 
Management System (ADAMS) Accession No. ML052010227), and was described 
in the Notice of Availability published in the Federal Register (FR) on 
November 4, 2005 (70 FR 67202). The proposed changes are consistent 
with the NRC-approved TSTF-5-A, Revision 1.
    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.

[[Page 9829]]

    The proposed change to remove the duplicative safety limit 
reporting, notification, and restart constraint requirements from 
the TS does not affect the plant or operation of the plant. The 
change simply removes duplicative information from the TS that is 
covered in the NRC regulations. 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 previously evaluated?
    Response: No.
    The proposed change to remove the duplicative safety limit 
reporting, notification, and restart constraint requirements from 
the TS does not introduce any new accident scenarios, failure 
mechanisms, or limiting single failures. All systems, structures, 
and components previously required for the mitigation of a transient 
remain capable of fulfilling their intended design functions. The 
proposed change has no adverse effect on any safety-related system 
or component and does not challenge the performance or integrity of 
any safety related system. This change is considered an 
administrative action to remove duplicative reporting, notification, 
and restart constraint requirements. Therefore, this 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 changes are administrative and do not involve any 
reduction in a margin of safety. All systems, structures, and 
components previously required for the mitigation of a transient 
remain capable of fulfilling their intended design functions. The 
proposed change has no adverse effect on any safety-related system 
or component and does not [involve a significant reduction in a 
margin of safety.]

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Branch Chief: Gloria Kulesa.

Notice of Issuance of Amendments to Facility Operating 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.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, 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 Commission's Public Document Room (PDR), located at One White Flint 
North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland 20852. Publicly available records will be accessible from the 
Agencywide Documents Access and Management System (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, 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 Reference staff at 1-800-397-4209, 301-415-4737 or by 
e-mail to [email protected].

Dairyland Power Cooperative, Docket No. 50-409, La Crosse Boiling Water 
Reactor, Vernon County, Wisconsin

    Date of application for amendment: July 28, 2009, and supplemented 
August 7, 2009, May 19, 2010, and August 12, 2010.
    Brief description of amendment: The amendment revises the La Crosse 
Boiling Water Rector (LACBWR) Technical Specifications, in support of 
the dry cask storage project at LACBWR.
    Date of issuance: January 25, 2011.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 71.
    Facility Operating License No. DPR-7: This amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: October 6, 2009 (74 FR 
51326).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 25, 2011.
    No significant hazards consideration comments received: No.

Dominion Electric Kewaunee, Inc. Docket No. 50-305, Kewaunee Power 
Station (KPS), Kewaunee County, Wisconsin

    Date of application for amendment: August 24, 2009 (Agencywide 
Documents and Management System (ADAMS) Accession No. ML092440398), as 
supplemented by letters dated October 22, 2009 (ADAMS Accession No. 
ML093070092), April 13, 2010 (ADAMS Accession Nos. ML101060517 and 
ML101040090), May 12, 2010 (ADAMS Accession No. ML101380399), July 1, 
2010 (ADAMS Accession No. ML101890404), July 16, 2010 (ADAMS Accession 
No. ML102370370), August 18, 2010 (ADAMS Accession No. ML102371064), 
September 7, 2010 (ADAMS Accession No. ML102730383), September 8, 2010 
(ADAMS Accession No. ML102580700), October 15, 2010 (ADAMS Accession 
No. ML102920037), and December 2, 2010 (ADAMS Accession No. 
ML103400328).
    Brief description of amendment: This amendment converts the current 
technical specifications (CTSs) to the improved TSs (ITSs) and 
relocates certain requirements to other licensee-controlled documents. 
The ITSs are based on NUREG-1431, Rev. 3.0, ``Standard Technical 
Specifications, Westinghouse Plants,'' Revision 3.0; ``NRC Final Policy 
Statement on Technical Specification Improvements for Nuclear Power 
Reactors,'' dated July 22, 1993 (58 FR 39132); and 10 CFR 50.36, 
``Technical Specifications.'' Technical Specification Task Force 
changes were also incorporated. The purpose of the conversion is to 
provide clearer and more readily understandable requirements in the TSs 
for KPS to ensure safe operation. In addition, the amendment includes a 
number of issues that were considered beyond the scope of NUREG-1431.
    Date of issuance: February 2, 2011.
    Effective date: As of the date of issuance and shall be implemented 
on or before February 23, 2011.

[[Page 9830]]

    Amendment No.: 207.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: December 15, 2009 (74 
FR 66384). The supplements provided, contained clarifying information 
and did not expand the scope of the application as originally noticed.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 2, 2011.
    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: January 27, 2010.
    Brief description of amendment:
    The amendment revises Section 2.E. of the Palisades Nuclear Plant 
(PNP) Renewed Facility Operating License to remove the name of the 
former operator of the plant in the title of the PNP physical security 
plan and replace it with Entergy Nuclear. The change also removes the 
security plan revision number and the date the plan was submitted to 
the Nuclear Regulatory Commission.
    Date of issuance: January 25, 2011.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 241.
    Facility Operating License No. DPR-20: Amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
considerations (NSHC): The notice provided an opportunity to submit 
comments on the Commission's proposed NSHC determination. No comments 
have been received.
    Date of initial notice in Federal Register: November 18, 2010 (75 
FR 70708), followed by the repeat biweekly notice in the Federal 
Register on January 25, 2011 (76 FR 4389).
    The Commission's related evaluation of the amendment, state 
consultation, and final NSHC determination are contained in a Safety 
Evaluation dated January 25, 2011.
    Attorney for licensee: Mr. William Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White 
Plains, NY 10601.
    NRC Branch Chief: Robert J. Pascarelli.

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

    Date of application for amendment: January 24, 2010, as 
supplemented by letters dated September 7 and November 4, 2010.
    Brief description of amendment: This amendment request would revise 
the Technical Specifications (TSs) Section 1.0, Definitions, TS Section 
3.6, Primary System Boundary Specifications 3.6.A, and TS Programs and 
Manuals Section 5.5, to include reference to the Pressure and 
Temperature Limits Report (PTLR). The proposed PTLR would include 
revised 43 effective full-power years pressure-temperature curves, 
neutron fluence, and adjusted reference temperature values.
    Date of issuance: January 26, 2011.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 234.
    Facility Operating License No. DPR-35: The amendment revised the 
License and Technical Specifications.
    Date of initial notice in Federal Register: April 6, 2010 (75 FR 
17443). The supplemental letters dated September 7 and November 4, 
2010, provided additional information that clarified the application, 
did not expand the scope of the application as originally noticed, and 
did not change the staff's original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated January 26, 2011.
    No significant hazards consideration comments received: No.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

    Date of amendment request: April 13, 2010 as supplemented by letter 
dated. February 2, 2011.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) to update the Table of Contents and the 
Applicability and Objective portions of TS 4.12 as a result of changes 
made by License Amendment Nos. 230 and 239 and to revise wording in TS 
3.7.A.8. The changes are considered administrative in nature and do not 
materially change any technical requirement.
    Date of Issuance: February 9, 2011.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 245.
    Facility Operating License No. DPR-28: Amendment revised the 
License and Technical Specifications.
    Date of initial notice in Federal Register: June 29, 2010 (75 FR 
37474). The supplement letter dated February 2, 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 this amendment is contained 
in a Safety Evaluation dated February 9, 2011.
    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: July 20, 2010.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.7.1.2, ``Emergency Feedwater System,'' Limiting 
Condition for Operation (LCO) 3/4.7.1.2, ``Emergency Feedwater,'' to 
clarify the acceptability of transitioning from Mode 4, Hot Shutdown, 
to Mode 3, Hot Standby, with the turbine-driven emergency feedwater 
(EFW) pump inoperable but available. The amendment granted an exception 
to TS LCO 3.0.4 and Surveillance Requirement 4.0.4 allowing entry into 
operational Mode 3 with TS LCO equipment, the turbine-driven EFW pump, 
associated with a shutdown action inoperable.
    Date of issuance: January 31, 2011.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 230.
    Facility Operating License No. NPF-38: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: September 21, 2010 (75 
FR 57523).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 31, 2011.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of amendment request: February 4, 2010 as supplemented by 
letters

[[Page 9831]]

dated September 15, 2010, October 6, 2010, and December 13, 2010.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3.3.6.1, ``Primary Containment 
Isolation Instrumentation,'' ``Table 3.3.6.1-1, ``Primary Containment 
Isolation Instrumentation,'' Function 6.a ``Shutdown Cooling System 
Isolation, Recirculation Line Water Temperature--High,'' to enable 
implementation with reactor pressure-based isolation instrumentation, 
for the Dresden Nuclear Power Station, Units 2 and 3.
    Date of issuance: February 7, 2011.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 236/229.
    Facility Operating License Nos. DPR-19 and DPR-25: The amendment 
revised the Technical Specifications and License.
    Date of initial notice in Federal Register: April 20, 2010 (75 FR 
20635).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 7, 2011.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: Dated October 5, 2009 as 
supplemented by letters dated June 10, November 23, December 14, and 
December 22, 2010, and January 11, 24, and 28, 2011.
    Brief description of amendments: The proposed amendment would 
revise Technical Specification (TS) 4.3.1, ``Criticality,'' to address 
a non-conservative TS. The proposed change addresses the Boraflex 
degradation issue in the LSCS Unit 2 spent fuel storage racks by 
revising TS Section 4.3.1 to allow the use of NETCO-SNAP-IN[supreg] 
rack inserts in LSCS Unit 2 spent fuel storage rack cells as a 
replacement for the neutron absorbing properties of the existing 
Boraflex panels.
    Date of issuance: January 28, 2011.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days after the end of Unit 2 refueling outage 13.
    Amendment Nos.: 199 and 186.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications and License.
    Date of initial notice in Federal Register: January 5, 2010 (75 FR 
463). The June 10, November 23, December 14, and December 22, 2010, and 
January 11, 24, and 28, 2011, submittals contained clarifying 
information and did not change the NRC 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 January 28, 2011.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: April 15, 2009, as supplemented by 
letters dated December 18, 2009, October 8, 2010 and January 10, 2011.
    Brief description of amendment request: The amendment request and 
proposed exemption request were to incorporate a new methodology for 
the development of Reactor Coolant System (RCS) pressure-temperature 
limits into Technical Specification (TS) 5.6.4, ``Reactor Coolant 
System (RCS) Pressure and Temperature Limits Report (PTLR).'' The 
amendment also requested a revision to the period of validity of the 
analysis for the low temperature overpressure protection (LTOP) system 
contained in Operating License Condition 2.C(3)(d). An associated 
revision to the Technical Specification Basis 3.4.12 ``Low Temperature 
Overpressure Protection (LTOP)'' supports the change to the operating 
license condition.
    Date of issuance: January 28, 2011.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 282.
    Facility Operating License No. NPF-3: The amendment revised the TS 
and license.
    Date of initial notice in Federal Register: June 16, 2009 (72 FR 
28577). The supplemental letters contained clarifying information, 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 January 28, 2011.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 9, 2009.
    Brief description of amendment: The amendment changes the NMP2 
Technical Specification (TS) 3.8.4, ``DC Sources--Operating,'' to 
remove the Mode restrictions for performance of TS Surveillance 
Requirements (SRs) 3.8.4.7 and 3.8.4.8 for the Division 3 direct 
current (DC) electrical power subsystem battery. The Division 3 DC 
electrical power subsystem feeds emergency DC loads associated with the 
high-pressure core spray (HPCS) system. These surveillances verify that 
the battery capacity is adequate for the battery to perform its 
required functions. The amendment removes these Mode restrictions for 
the Division 3 battery, thereby allowing performance of the SRs during 
Mode 1, 2, or 3 in conjunction with scheduled HPCS system outages.
    Date of issuance: January 31, 2011.
    Effective date: As of the date of issuance to be implemented within 
90 days.
    Amendment No.: 136.
    Renewed Facility Operating License No. NPF-069: The amendment 
revises the License and TSs.
    Date of initial notice in Federal Register: April 6, 2010 (75 FR 
17444).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 31, 2011.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, 
Georgia

    Date of application for amendments: February 2, 2010.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) Table 3.3.1-1 ``Reactor Trip System 
Instrumentation [RTS],'' Function 3, ``Power Range Neutron Flux High 
Positive Rate.'' Specifically, the revision added surveillance 
requirement 3.3.1.15 to verify the RTS response time is within limits.
    Date of issuance: February 7, 2011.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 159 and 141.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the licenses and the TSs.
    Date of initial notice in Federal Register: May 4, 2010 (75 FR 
23817). The supplement dated October 29, 2010, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the NRC

[[Page 9832]]

staff's original proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 7, 2011.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 30, 2010.
    Brief description of amendment: The amendments revised the North 
Anna Technical Specifications (TSs) by relocating specific surveillance 
frequencies to a licensee-controlled program with the implementation of 
Nuclear Energy Institute (NEI) 04-10, ``Risk-Informed Technical 
Specifications Initiative 5b, Risk-Informed Method for Control of 
Surveillance Frequencies.''
    Date of issuance: January 31, 2011.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days from the date of issuance.
    Amendment Nos.: 262 and 243.
    Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments 
changed the licenses and the technical specifications.
    Date of initial notice in Federal Register: May 18, 2010 (75 FR 
27833). The supplements dated August 30, 2010, and January 18, 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 January 31, 2011.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 10th day of February 2011.

    For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2011-3721 Filed 2-18-11; 8:45 am]
BILLING CODE 7590-01-P