[Federal Register Volume 75, Number 153 (Tuesday, August 10, 2010)]
[Notices]
[Pages 48370-48381]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2010-19678]


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NUCLEAR REGULATORY COMMISSION

[NRC-2010-0272]


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 July 15, 2010 to July 28, 2010. The last 
biweekly notice was published on July 27, 2010 (75 FR 44020).

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.
    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 ``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 O1F-21, 
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

[[Page 48371]]

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 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 (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

[[Page 48372]]

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://ehd.nrc.gov/EHD_Proceeding/home.asp, 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].

Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of amendment requests: April 29, 2010, as supplemented by 
letter dated July 22, 2010.
    Description of amendment requests: The proposed change will add to 
Technical Specification 5.6.5.b an additional topical report describing 
an NRC reviewed and approved analytical method for determining core 
operating limits. The new analytical method, which is described in 
AREVA Topical Report ANP-10298PA, ACE/ATRIUM 10XM Critical Power 
Correlation, Revision 0, March 2010, provides a new correlation for 
predicting the critical power for boiling water reactors containing 
ATRIUM 10XM fuel.
    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 probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. The 
proposed amendments add an additional analytical methodology to the 
list of NRC-approved analytical methods identified in Technical 
Specification 5.6.5.b that can be used to establish core operating 
limits. The proposed amendments support the use of the AREVA ATRIUM 
10XM fuel design at BSEP [Brunswick Steam Electric Plant]. The 
addition of an approved analytical methodology in Technical 
Specification Section 5.6.5 has no effect on any accident initiator 
or precursor previously evaluated and does not change the manner in 
which the core is operated. The NRC-approved methodology ensures 
that the output accurately models core behavior. Since no individual 
precursors of an accident are affected, the proposed amendments do 
not increase the probability of a previously analyzed event.
    The consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences. The proposed amendments add an additional analytical 
methodology to the list of NRC-approved analytical methods used to 
establish core operating limits. The addition of the topical report 
to Technical Specification 5.6.5.b will allow a new analytical 
methodology to be used to determine critical power ratio limits. 
Minimum Critical Power Ratio (MCPR) Safety Limit values, which are 
defined in Technical Specification 2.1.1.2, are calculated to ensure 
that greater than 99.9 percent of the fuel rods in the reactor core 
avoid transition boiling during plant operation, if the safety limit 
is not exceeded. The derivation of MCPR Safety Limit values in the 
Technical Specifications, using these NRC-accepted methods, will 
continue to ensure the MCPR Safety Limit is not exceeded during all 
modes of plant operation and anticipated operational occurrences. 
The addition of the analytical methodology described in Topical 
Report ANP-10298PA to Technical Specification 5.6.5.b does not alter 
the assumptions of accident analyses or the Technical Specification 
Bases. Based on the above, the proposed amendments do not increase 
the consequences of a previously analyzed accident.
    Therefore, the proposed amendments do 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.
    Creation of the possibility of a new or different kind of 
accident requires creating one or more new accident precursors. New 
accident precursors may be created by modifications of plant 
configuration, including changes in allowable modes of operation. 
The proposed amendments do not involve any plant configuration 
modifications, do not involve any changes to allowable modes of 
operation, and do not introduce any new failure mechanisms. The 
proposed topical report addition to Technical Specification 5.6.5.b 
provides an analytical methodology for determining core critical 
power limits that ensures no new accident precursors are created.
    Therefore, the proposed amendments do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed amendments add an additional analytical methodology 
to the list of NRC-approved analytical methods identified in 
Technical Specification 5.6.5.b that can be used to establish core 
operating limits. This addition to Technical Specification 5.6.5.b 
will allow a new NRC-accepted analytical methodology to be used to 
determine critical power ratio limits. The MCPR Safety Limit 
provides a margin of safety by ensuring that at least 99.9 percent 
of the fuel rods do not experience transition boiling during normal 
operation and anticipated operational occurrences if the MCPR Safety 
Limit is not exceeded. The proposed change will ensure the current 
level of fuel protection is maintained by continuing to ensure that 
the fuel design safety criterion (i.e., that no more than 0.1 
percent of the rods are expected to be in boiling transition if the 
MCPR Safety Limit is not exceeded) is met.
    Therefore, the proposed amendments do 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: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, NC 27602.
    NRC Branch Chief: Douglas A. Broaddus.

[[Page 48373]]

Carolina Power and Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of amendment requests: April 29, 2010, as supplemented by 
letter dated July 22, 2010.
    Description of amendment requests: The proposed change would add, 
to Technical Specification 5.6.5.b, an additional topical report 
describing an NRC reviewed and approved analytical method for 
determining core operating limits. The new analytical method, which is 
described in AREVA Topical Report BAW-10247PA, Realistic Thermal-
Mechanical Fuel Rod Methodology for Boiling Water Reactors, Revision 0, 
April 2008, provides a new statistical thermal-mechanical evaluation 
methodology for determining reactor core linear heat generation limits 
in boiling water reactors.
    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 probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. The 
proposed amendments add an additional analytical methodology to the 
list of NRC-approved analytical methods identified in Technical 
Specification 5.6.5.b that can be used to establish core operating 
limits. The proposed amendments support the use the AREVA ATRIUM 
10XM fuel design at BSEP [Brunswick Steam Electric Plant]. The 
addition of an approved analytical methodology in Technical 
Specification Section 5.6.5 has no effect on any accident initiator 
or precursor previously evaluated and does not change the manner in 
which the core is operated. The NRC-approved methodology ensures 
that the output accurately models core behavior. Since no individual 
precursors of an accident are affected, the proposed amendments do 
not increase the probability of a previously analyzed event.
    The consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences. The proposed amendments add an additional analytical 
methodology to the list of NRC-approved analytical methods used to 
establish core operating limits. The addition of the topical report 
to Technical Specification 5.6.5.b will allow a new thermal-
mechanical methodology, based on the RODEX4 fuel performance code, 
to be used to determine reactor core linear heat generation rate 
limits monitored as specified by Technical Specification 3.2.3. The 
addition of the analytical methodology described in Topical Report 
BAW-10247PA to Technical Specification 5.6.5.b does not alter the 
assumptions of accident analyses or the Technical Specification 
Bases. Based on the above, the proposed amendments do not increase 
the consequences of a previously analyzed accident.
    Therefore, the proposed amendments do 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.
    Creation of the possibility of a new or different kind of 
accident requires creating one or more new accident precursors. New 
accident precursors may be created by modifications of plant 
configuration, including changes in allowable modes of operation. 
The proposed amendments do not involve any plant configuration 
modifications, do not involve any changes to allowable modes of 
operation, and do not introduce any new failure mechanisms. The 
proposed topical report addition to Technical Specification 5.6.5.b 
provides an analytical methodology for determining reactor core 
linear heat generation rate limits that ensures no new accident 
precursors are created.
    Therefore, the proposed amendments do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed amendments add an additional analytical methodology 
to the list of NRC-approved analytical methods identified in 
Technical Specification 5.6.5.b that can be used to establish core 
operating limits. This addition to Technical Specification 5.6.5.b 
will allow a new NRC-accepted analytical methodology to be used to 
determine reactor core linear heat generation rate limits.
    Limits on the linear heat generation rate are specified to 
ensure that fuel design limits are not exceeded anywhere in the core 
during normal operation, including anticipated operational 
occurrences. Exceeding the linear heat generation rate limit could 
potentially result in fuel damage and subsequent release of 
radioactive materials. The mechanisms that could cause fuel damage 
during normal operations and operational transients and that are 
considered in fuel evaluations are rupture of the fuel rod cladding 
caused by strain and overheating of the fuel. The proposed change 
will ensure the current level of fuel protection is maintained 
(i.e., that the fuel design safety criteria of less than one percent 
plastic strain of the fuel cladding is met and incipient centerline 
melting of the fuel does not occur) and thus assure that rupture of 
the fuel rod cladding caused by strain and overheating of the fuel 
does not occur.
    Therefore, the proposed amendments do 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: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, NC 27602.
    NRC Branch Chief: Douglas A. Broaddus.

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

    Date of amendment request: June 10, 2009, supplemented by letters 
dated September 16, 2009, and July 23, 2010.
    Description of amendment request: The proposed amendment would 
revise Fermi 2 Plant Operating License, Appendix A, Technical 
Specification (TS) Table 3.3.8.1-1, Function 2 (Degraded Voltage) to 
identify an additional time delay logic for Loss-of-Coolant Accident 
(LOCA) concurrent with degraded voltage conditions. Specifically, this 
proposed amendment adds a new time delay logic associated with Function 
2 for a degraded voltage concurrent with a LOCA. This will bring Fermi 
2 into compliance with 10 CFR Part 50, Appendix A, General Design 
Criterion (GDC)--17, ``Electric Power Systems.'' In addition, it would 
revise the TS maximum and minimum allowable values for the 4.16kV 
Emergency Bus Undervoltage (Degraded Voltage) and revise the minimum 
Emergency Diesel Generator (EDG) output voltage acceptance criterion in 
Surveillance Requirements (SRs) 3.8.1.2, 3.8.1.7, 3.8.1.10, 3.8.1.11, 
3.8.1.14, and 3.8.1.17. The additional changes resulted from a 
reconstitution effort of the electrical design bases calculations to 
support the backfit modifications, necessary to address issues 
identified in the Component Design Bases Inspection (CDBI) at Fermi 2. 
This notice supersedes the notice published in the Federal Register on 
August 11, 2009, (74 FR 40235), in its entirety.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below

    .1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.

[[Page 48374]]

    Providing the additional logic ensures the timely transfer of 
plant safety system loads to the Emergency Diesel Generators in the 
event a sustained degraded bus voltage is present with a Loss of 
Coolant Accident (LOCA) signal. This ensures that under these 
degraded bus voltage conditions, Emergency Core Cooling System 
(ECCS) equipment is powered from the emergency diesel generators in 
a timely manner. This change is needed to bring Fermi 2 into full 
compliance with 10 CFR Part 50, Appendix A, General Design 
Criterion-17, ``Electric Power Systems,'' and to meet the 
requirements of NUREG-0800 Rev. 2, Branch Technical Position (BTP) 
Power Systems Branch (PSB)-1. The time delay supports the time 
assumed in the accident analysis for water injection into the 
reactor vessel under LOCA conditions.
    The proposed TS change to the maximum and minimum allowable 
voltages for the 4160 volt Emergency Bus Undervoltage (Degraded 
Voltage) affects the separation of an Emergency Bus that is 
experiencing degraded voltage from the offsite power system and the 
transfer to an emergency diesel generator. While the allowed voltage 
range is narrower, the function remains the same. The narrower 
voltage range has been analyzed and is needed to ensure spurious 
trips are avoided. The proposed change does not affect any accident 
initiators or precursors. As a result, the probability of any 
accident previously evaluated is not significantly increased.
    The consequences of any accident previously evaluated are not 
increased since the 4160 volt Emergency Bus Undervoltage (Degraded 
Voltage) relays will continue to meet their required function to 
transfer the 4160 volt Emergency Buses to the emergency diesel 
generators in the event of a degraded voltage condition on the 
offsite power supply. This transfer ensures that the electrical 
equipment is capable of performing its intended function to meet the 
requirements of the accident analyses.
    The increase in the minimum EDG output voltage acceptance 
criterion value in TS 3.8.1 surveillance requirements does not 
adversely affect any of the parameters in the accident analyses. The 
change increases the minimum allowed EDG output voltage acceptance 
criterion to ensure that sufficient voltage is available to operate 
the required Emergency Safety Feature (ESF) equipment under accident 
conditions. The increase in the minimum allowed EDG output voltage 
in the TS surveillance requirements ensures that adequate voltage is 
available to support the assumptions made in the Design Bases 
Accident (DBA) analyses. DBA analyses assume that onsite standby 
emergency power will provide an adequate power source to operate 
safe shutdown equipment and to mitigate consequences of design bases 
accidents. This conservative change of the acceptance criterion 
enhances the testing requirements of the onsite emergency diesel 
generators and ensures the reliability of this power source. 
Changing the acceptance criterion does not affect the probability of 
evaluated accidents and it provides better assurance of EDG 
reliability in mitigating consequences of accidents.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not affect any of the current degraded 
voltage logic schemes or any other equipment provided to mitigate 
accidents. It utilizes existing logic systems to isolate safety 
buses from the grid and re-power those safety buses using the onsite 
emergency power system. The change utilizes a narrower voltage range 
and a shorter time delay to ensure that in the case of a sustained 
degraded voltage condition concurrent with a LOCA signal, the safety 
electrical power buses will be transferred from the offsite power 
system to the onsite power system in a timely manner to ensure water 
is injected into the reactor vessel in the time assumed and 
evaluated in the accident analysis.
    No new or different accidents result from the proposed change. 
The proposed TS change to the maximum and minimum allowable voltages 
for the 4160 volt Emergency Bus Undervoltage (Degraded Voltage) does 
not affect existing accident precursors or modes of operation nor 
does it introduce new ones. The relays will continue to detect 
degraded voltage conditions and transfer the Emergency Buses to 
their respective emergency diesel generators in time to ensure 
adequate voltage is available for proper safety equipment 
performance, and to prevent equipment damage. The function of the 
relays remains the same.
    The change in the value of the minimum EDG output voltage 
acceptance criterion supports the assumptions in the accident 
analyses that sufficient voltage will be available to operate ESF 
equipment on the Class 1E buses when these buses are powered from 
the onsite emergency diesel generators. The maximum EDG output 
voltage of 4580 volts is not affected by this change. The change in 
the minimum EDG output voltage from 3873 to 3950 volts ensures the 
reliability of the onsite emergency power source.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The proposed change implements a new design for a reduced time 
delay to isolate safety buses from offsite power if a Loss of 
Coolant Accident were to occur concurrent with a sustained degraded 
voltage condition and uses a narrower voltage range for degraded bus 
undervoltage. This ensures that emergency core cooling system pumps 
inject water into the reactor vessel within the time assumed and 
evaluated in the accident analysis, consistent with the requirements 
of BTP PSB-1 Section B.1.b. and 10 CFR Part 50, Appendix A, General 
Design Criterion-17, ``Electric Power Systems.''
    The proposed TS change to the maximum and minimum allowable 
voltages for the 4160 volt Emergency Bus Undervoltage (Degraded 
Voltage) will allow all safety loads to have sufficient voltage to 
perform their intended safety functions while ensuring spurious 
trips are avoided. Thus, the results of the accident analyses will 
not be affected as the input assumptions are protected.
    The proposed TS change for the maximum allowable values for the 
4160 volt Emergency Bus Undervoltage (Degraded Voltage) provides a 
greater margin between the predicted worst case transient voltages 
and the maximum reset value of the degraded voltage relays. This 
change increases the probability that the offsite power source 
remains available and connected to the auxiliary power system during 
postulated transients. The analytical limit voltage for the safety 
related 4160 volt buses is unchanged and the proposed TS changes for 
the minimum allowable values for the 4160 volt Emergency Bus 
Undervoltage (Degraded Voltage) still ensures that this limit is 
protected. This is consistent with the requirements of 10 CFR Part 
50, Appendix A, General Design Criterion-17, ``Electric Power 
Systems.''
    The proposed change in the minimum EDG output voltage acceptance 
criterion in TS 3.8.1 surveillance requirements does not affect the 
surveillance frequency or different testing requirements, only the 
acceptance criterion. The change provides a better assurance that 
the onsite power source is able to satisfy the design requirements 
assumed in the accident analyses to mitigate the consequences of 
design bases accidents.
    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 G. Pettinari, Legal Department, 688 
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279.
    NRC Branch Chief: Robert J. Pascarelli.

Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power 
Station, Kewaunee County, Wisconsin

    Date of amendment request: June 1, 2010.
    Description of amendment request: The licensee proposed to revise 
the Kewaunee licensing basis, approving the licensee to operate the 
load tap changers (LTCs) on two new transformers to operate in the 
automatic mode. The LTCs are subcomponents of the two new transformers, 
one has already been installed and one to be installed. The LTCs are 
designed to compensate for potential offsite power voltage variations 
and will provide

[[Page 48375]]

added assurance that acceptable voltage is maintained for safety-
related equipment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration (NSHC). The NRC staff reviewed the licensee's NSHC 
analysis and has prepared its own as follows:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The function of the LTCs is to ensure that acceptable voltage is 
maintained for safety-related equipment. The only postulated 
accident previously evaluated where the probability of occurrence 
may be potentially affected by operating the LTCs in automatic mode 
is the loss of offsite power (LOOP) accident. However, the 
licensee's analysis shows that, as a result of availability of 
backup equipment and systems, the probability of a LOOP would not be 
increased by operation of the LTCs in the automatic mode. 
Furthermore, operation of the LTCs in the automatic mode is not 
likely to degrade the Kewaunee electrical system; thus, the 
electrical system will continue to fulfill its design functions 
during normal and accident conditions. As a result, operating the 
LTCs in automatic mode will not be a factor to increase the 
consequences of previously evaluated accidents. In summary, the 
probability of occurrence and the consequences of the previously 
analyzed accidents would not be affected in any way by the proposed 
licensing basis change.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Other than the installation of the two new transformers (which 
is not the subject of the proposed amendment), the proposed change 
of licensing basis to allow the LTCs to be operated in the automatic 
mode does not involve any physical alteration of the plant, nor does 
it change methods and procedures governing plant operation. The 
proposed change will not impose any new or eliminate any old safety 
requirements on the plant electrical system.
    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 has no effect on any safety analysis 
methods, scenarios, or assumptions involving the electrical system.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    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 proposed amendment involves no significant hazards 
consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., Counsel for Dominion Energy Kewaunee, Inc., 
120 Tredegar Street, Richmond, VA 23219.
    NRC Branch Chief: Robert J. Pascarelli.

Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414, 
Catawba Nuclear Station, Units 1 and 2 (Catawba), York County, South 
Carolina; Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 
1 and 2 (McGuire), Mecklenburg County, North Carolina

    Date of amendment request: December 14, 2009.
    Description of amendment request: The amendments would revise the 
Technical Specifications Section 3.8.4 ``DC [Direct Current] Sources--
Operating'' Surveillance Requirements 3.8.4.2 and 3.8.4.5 for McGuire 
and 3.8.4.3 and 3.8.4.6 for Catawba.
    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.
    Performing the battery Surveillances is not an initiator to any 
accident sequence previously evaluated in the Updated Final Safety 
Analysis Report. The Batteries are still required to be operable, 
meet the Surveillance Requirements, and be capable of performing any 
mitigation function as designed. Revising the battery Surveillance 
resistance values and adding the total average resistance limit, as 
supported by calculations, will help ensure that the voltage and 
capacity of the Batteries remain within the design basis.
    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.
    This amendment does not involve a modification to the plant or a 
change in how the plant is operated. No new accident causal 
mechanisms are created as a result of this proposed amendment. No 
changes are being made to any structure, system, or component which 
will introduce any new accident causal mechanisms. This amendment 
request does not impact any plant systems that are accident 
initiators and does not impact any safety analysis.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    Response: No.
    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 the fuel cladding, reactor coolant and 
containment systems will not be impacted by the proposed change. The 
proposed McGuire and Catawba battery connection resistance limits 
ensure the continued availability and operability of the Batteries. 
As such, sufficient DC capacity to support operation of mitigation 
equipment remains within the design basis.
    Therefore, it is concluded that the proposed changes do 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: Lara S. Nichols, Associate General Counsel, 
Duke Energy Corporation, 526 South Church Street, EC07H, Charlotte, NC 
28202.
    NRC Branch Chief: Gloria Kulesa.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of amendment request: June 23, 2010.
    Description of amendment request: The current Arkansas Nuclear One, 
Unit No. 2 Technical Specification (TS) 6.5.8, ``Inservice Testing 
Program,'' contains references to the American Society of Mechanical 
Engineers (ASME) Boiler and Pressure Vessel Code, Section XI as the 
source of requirements for the inservice testing (IST) of ASME Code 
Class 1, 2, and 3 pumps and valves. The proposed amendment would delete 
the references to Section XI of the ASME Code and incorporate 
references to the ASME Code for Operation and Maintenance of Nuclear 
Power Plants (ASME OM Code). The proposed amendment would also correct 
some nonstandard frequencies utilized in the IST Program in which the 
provisions of Surveillance Requirement 3.0.2 are applicable. The 
proposed changes are consistent with Technical Specification Task Force 
(TSTF) Technical Change Travelers 479-A, ``Changes to Reflect Revision 
to 10 CFR 50.55a,'' and 497-A,

[[Page 48376]]

``Limit Inservice Testing Program SR 3.0.2 Application to Frequencies 
of 2 Years or Less.''
    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 TS 6.5.8, ``Inservice Testing 
Program,'' for consistency with the requirements of 10 CFR 
50.55a(f)(4) for pumps and valves which are classified as American 
Society of Mechanical Engineers (ASME) Code Class 1, Class 2 and 
Class 3. The proposed change incorporates revisions to the ASME Code 
which are consistent with the expectations of 10 CFR 50.55a.
    The proposed change does not impact any accident initiators or 
analyzed events or assumed mitigation of accident or transient 
events. The proposed change does not involve the addition or removal 
of any equipment, or any design changes to the facility.
    Therefore, this proposed change does not represent 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 a modification to the 
physical configuration of the plant (i.e., no new equipment will be 
installed) or change in the methods governing normal plant 
operation. The proposed change does not introduce a new accident 
initiator, accident precursor, or malfunction mechanism.
    Therefore, this proposed change does not create the possibility 
of an accident or a different kind than previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises TS 6.5.8, ``Inservice Testing 
Program,'' for consistency with the requirements of 10 CFR 
50.55a(f)(4) for pumps and valves which are classified as ASME Code 
Class 1, Class 2 and Class 3. The proposed change incorporates 
revisions to the ASME Code, which are consistent with the 
expectations of 10 CFR 50.55a. The safety function of the affected 
pumps and valves are maintained.
    Therefore, this proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Joseph A. Aluise, Associate General 
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: May 27, 2010.
    Brief description of amendments: The proposed amendments would 
revise the Comanche Peak Nuclear Power Plant (CPNPP), Units 1 and 2, 
Technical Specification (TS) 3.8.3, ``Diesel Fuel Oil, Lube Oil, and 
Starting Air,'' by relocating the current stored diesel fuel oil and 
lube oil numerical volume requirements from the TS to the TS Bases so 
that it may be modified under licensee control. The TS would be 
modified so that the stored diesel fuel oil and lube oil inventory will 
require that a 7-day supply be available for each diesel generator. 
Condition A and Condition B in the Action table and Surveillance 
Requirements (SRs) 3.8.3.1 and 3.8.3.2 would also be revised to reflect 
the above change. The proposed changes are consistent with U.S. Nuclear 
Regulatory Commission (NRC)-approved Revision 1 to Technical 
Specification Task Force (TSTF) Improved Standard Technical 
Specification Change Traveler 501, ``Relocate Stored Fuel Oil and Lube 
Oil Volume Values to Licensee Control.''
    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 relocates the volume of diesel fuel oil and 
lube oil required to support 7-day operation of the onsite diesel 
generators, and the volume equivalent to a 6-day supply for fuel oil 
and, for lube oil, a 2-day supply to licensee control. The specific 
volume of fuel oil equivalent to a 7- and 6-day supply is calculated 
using the NRC-approved methodology described in Regulatory Guide 
1.137, Revision 1, ``Fuel-Oil Systems for Standby Diesel 
Generators'' and ANSI [American National Standards Institute] N195 
1976, ``Fuel Oil Systems for Standby Diesel-Generators.'' The CPNPP 
specific volumetric requirements for lube oil were originally based 
on the manufacturer's consumption values; however, the volumetric 
requirements have been refined over time based on actual plant data 
and engine performance. As approved in CPNPP TS License Amendment 
75, the current lube oil volumetric requirements are based on the 
diesel generator lube oil consumption rate, avoidance of vortexing, 
static versus run lube oil level changes, and volume versus tank 
level data.
    Therefore, this proposed change is consistent with TSTF-501 as 
approved by the NRC. Because the requirement to maintain a 7-day 
supply of diesel fuel oil and lube oil is not changed and is 
consistent with the assumptions in the accident analyses, and the 
actions taken when the volume of fuel oil and lube oil are less than 
a 6-day and 2-day supply have not changed, neither the probability 
or the consequences of any accident previously evaluated will be 
affected.
    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 change does not involve a physical alteration of the plant 
(i.e., no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. The change 
does not alter assumptions made in the safety analysis but ensures 
that the diesel generator operates as assumed in the accident 
analysis. The proposed change is consistent with the safety analysis 
assumptions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change relocates the volume of diesel fuel oil and 
lube oil required to support 7-day operation of the onsite diesel 
generators, and the volume equivalent to a 6- and 2- (for fuel oil 
and lube oil, respectively) day supply to licensee control. As the 
bases for the existing limits on diesel fuel oil and lube oil are 
not changed, no change is made to the accident analysis assumptions 
and no margin of safety is reduced as part of this change.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Timothy P. Matthews, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.

[[Page 48377]]

    NRC Branch Chief: Michael T. Markley.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: November 23, 2009, as supplemented on 
December 11 and December 18, 2009, and July 23, 2010 (TS 09-06).
    Description of amendment request: On March 27, 2009, the Federal 
Register Notice 74 FR 13926 issued the final rule that amended Title 10 
of the Code of Federal Regulations (10 CFR), Part 73, ``Physical 
Protection of Plants and Materials.'' Specifically, the regulations in 
10 CFR 73.54 ``Protection of Digital Computer and Communication Systems 
and Networks'' establish the requirements for a cyber security program 
to protect digital computer and communication systems and networks 
against cyber attacks. The proposed amendment would include the 
proposed Cyber Security Plan, its implementation schedule, and a 
revised Physical Protection license condition for Sequoyah Nuclear 
Plant, Units 1 and 2 to fully implement and maintain in effect all 
provisions of the Nuclear Regulatory Commission approved Cyber Security 
Plan as required by 10 CFR 73.54.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Criterion 1: The proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Neither the proposed additional license condition nor the Cyber 
Security Plan directly impacts the physical configuration or 
function of plant structures, systems, or components (SSCs). 
Likewise, they do not change the manner in which SSCs are operated, 
maintained, modified, tested, or inspected. Neither the proposed 
additional license condition nor the Cyber Security Plan introduces 
any initiator of any accident previously evaluated. Any 
modifications to the physical configuration or function of SSCs or 
the manner in which SSCs are operated, maintained, modified, tested, 
or inspected that might result from the implementation of the Cyber 
Security Plan will be fully evaluated by existing regulatory 
processes (e.g., 10 CFR 50.59) prior to their implementation to 
ensure that they do not result in the probability or consequences of 
an accident previously evaluated.
    Therefore, it is concluded that this amendment does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    Criterion 2: The proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    This proposed amendment is intended to provide high assurance 
that safety-related SSCs are protected from cyber attacks. Inclusion 
of the additional condition in the Facility Operating License to 
implement the Cyber Security Plan does not directly alter the plant 
configuration, require new plant equipment to be installed, alter or 
create new accident analysis assumptions, add any initiators, or 
affect the function of plant systems or the manner in which systems 
are operated, maintained, modified, tested, or inspected.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    Criterion 3: The proposed amendment does not involve a 
significant reduction in a margin of safety.
    The proposed amendment does not involve any physical changes to 
plant or alter the manner in which plant systems are operated, 
maintained, modified, tested, or inspected. The proposed change does 
not alter the manner in which safety limits, limiting safety system 
settings or limiting conditions for operation are determined. The 
safety analysis acceptance criteria are not affected by this change. 
The proposed change will not result in plant operation in a 
configuration outside the design basis. The proposed change does not 
adversely affect systems that respond to safely shutdown the plant 
and to maintain the plant in a safe shutdown condition. Adding a 
license condition to require implementation of Cyber Security Plan 
will not reduce a margin of safety because the requirements of the 
Plan are designed to provide high assurance that safety-related SSCs 
are protected from cyber attacks.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Douglas A. Broaddus.

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

    Date of amendment request: March 30, 2010.
    Description of amendment request: This amendment request involves 
the adoption of approved changes to the Standard Technical 
Specifications (STSs) for Westinghouse Pressurized Water Reactors 
(NUREG-1431), to allow relocation of specific TS surveillance 
frequencies to a licensee-controlled program. The proposed changes are 
described in Technical Specification Task Force (TSTF) Traveler, TSTF-
425, Revision 3 (ADAMS Accession No. ML090850642) related to the 
``Relocation of Surveillance Frequencies to Licensee Control--Risk 
Informed Technical Specification Task Force (RITSTF) Initiative 5b,'' 
and are described in the Notice of Availability published in the 
Federal Register on July 6, 2009 (74 FR 31996). The proposed changes 
are consistent with NRC-approved Industry/TSTF Traveler, TSTF-425, 
Revision 3, ``Relocate Surveillance Frequencies to Licensee Control-
[RITSTF] Initiative 5b.'' The proposed changes relocate surveillance 
frequencies to a licensee-controlled program, the Surveillance 
Frequency Control Program (SFCP). The changes are applicable to 
licensees using probabilistic risk guidelines contained in NRC-approved 
NEI 04-10, ``Risk-Informed Technical Specifications Initiative 5b, Risk 
Informed Method for Control of Surveillance Frequencies,'' (ADAMS 
Accession No. 071360456). In addition, administrative/editorial 
deviations of the TSTF-425 inserts and the existing TS wording are 
being proposed to fit the custom TS format.
    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 any accident previously evaluated?
    Response: No.
    The proposed changes relocate the specified frequencies for 
periodic surveillance requirements to licensee control under a new 
Surveillance Frequency Control Program. Surveillance frequencies are 
not an initiator to any accident previously evaluated. As a result, 
the probability of any accident previously evaluated is not 
significantly increased. The systems and components required by the 
technical specifications for which the surveillance frequencies are 
relocated are still required to be operable, meet the acceptance 
criteria for the surveillance requirements, and be capable of 
performing any mitigation function assumed in the accident analysis. 
As a result, the consequences of any accident previously evaluated 
are not significantly increased.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?

[[Page 48378]]

    Response: No.
    No new or different accidents result from utilizing the proposed 
changes. The changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements. The changes do not alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumptions and current plant operating practice.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in 
[a] margin of safety?
    Response: No.
    The design, operation, testing methods, and acceptance criteria 
for systems, structures, and components (SSCs), specified in 
applicable codes and standards (or alternatives approved for use by 
the NRC) will continue to be met as described in the plant licensing 
basis (including the final safety analysis report and bases to TS), 
since these are not affected by changes to the surveillance 
frequencies. Similarly, there is no impact to safety analysis 
acceptance criteria as described in the plant licensing basis. To 
evaluate a change in the relocated surveillance frequency, Dominion 
will perform a probabilistic risk evaluation using the guidance 
contained in NRC approved NEI 04-10, Rev. 1 in accordance with the 
TS SFCP. NEI 04-10, Rev. 1, methodology provides reasonable 
acceptance guidelines and methods for evaluating the risk increase 
of proposed changes to surveillance frequencies consistent with 
Regulatory Guide 1.177.
    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 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 St., RS-2, Richmond, VA 23219.
    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. 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].

Carolina Power and Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of application for amendments: October 27, 2009.
    Brief Description of amendments: The proposed amendments modified 
technical specifications (TSs) requirements related to primary 
containment isolation instrumentation in accordance with the Nuclear 
Regulatory Commission-approved Technical Specification Task Force 
(TSTF), Standard Technical Specifications Change Traveler, TSTF-306, 
Revision 2, ``Add action to LCO [Limiting Condition for Operation] 
3.3.6.1 to give option to isolate the penetration.'' The proposed 
amendment would revise TS Section 3.3.6.1, ``Primary Containment 
Isolation Instrumentation,'' by adding an ACTIONS note allowing 
intermittent opening, under administrative control, of penetration flow 
paths that are isolated. Additionally, the traversing in-core probe 
system would be added as a separate isolation function with an 
associated Required Action to isolate the penetration within 24 hours 
rather than immediately initiating a unit shutdown.
    Date of issuance: July 23, 2010.
    Effective date: Date of issuance, to be implemented within 60 days.
    Amendment Nos.: 255 and 283.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
changed the Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2010 (75 FR 
4114).
    The Commission's related evaluation of the amendments is contained 
in the Safety Evaluation dated July 23, 2010.
    No significant hazards consideration comments received: No.

Carolina Power and Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of application for amendment: January 27, 2010, as 
supplemented by letter dated March 22, 2010.
    Brief description of amendment: The amendment revises a Limiting 
Condition for Operation (LCO) in Technical Specifications (TS) Section 
3.6.2.2.a to incorporate an expanded range of eductor flow rates for 
the Containment Spray Additive System as a result of the use of a new 
chemical model and new boric acid equilibrium data, revised sump pH 
limits, and changes to the Containment Spray Additive Tank 
concentration and volume limits.
    Date of issuance: July 16, 2010.
    Effective date: Effective as of the date of issuance and shall be 
implemented within 30 days.
    Amendment No. 134.
    Renewed Facility Operating License No. NPF-63: The amendment 
revises the technical specifications and facility operating license.
    Date of initial notice in Federal Register: March 23, 2010 (75 FR 
13788). The supplement dated March 22, 2010, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed,

[[Page 48379]]

and did not change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register on 
March 23, 2010 (75 FR 13788).
    The Commission's related evaluation of the amendment is contained 
in a safety evaluation dated July 16, 2010.
    No significant hazards consideration comments received: No.

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: August 6, 2009, as supplemented 
by letter dated February 23, 2010.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) by changing the surveillance requirement 
frequency for TS 3.4.12, ``Low Temperature Overpressure Protection 
System,'' from 6 months to 18 months.
    Date of Issuance: July 21, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 368, 370, and 369.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the licenses and the TSs.
    Date of initial notice in Federal Register: March 9, 2010 (75 FR 
10827). The supplement dated February 23, 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 the amendments is contained 
in a Safety Evaluation dated July 21, 2010.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 23, 2009, as superseded 
on March 18, 2010, as supplemented on May 11 and June 3, 2010.
    Brief description of amendment: The amendment revises TS 
Surveillance Requirements (SRs) 3.4.3.2 and 3.5.1.13 by deleting the 
current requirement to manually actuate each main steam safety/relief 
valve (SRV) during plant startup. SRs 3.4.3.2 and 3.5.1.13 have been 
modified to require that the SRVs be tested in accordance with the 
inservice test program that meets the requirements of American Society 
of Mechanical Engineers Code for Operation and Maintenance of Nuclear 
Power Plants.
    Date of issuance: July 21, 2010.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 297.
    Renewed Facility Operating License No. DPR-59: The amendment 
revised the License and the Technical Specifications.
    Date of initial notice in Federal Register: April 20, 2010 (75 FR 
20631). The May 11 and June 3, 2010, supplements provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the NRC 
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 July 21, 2010.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 31, 2009, as supplemented 
by letters dated March 5 and June 17, 2010.
    Brief description of amendment: The change revised the JAFNPP 
Technical Specifications (TSs) Surveillance Requirements (SRs) for 
testing of the Residual Heat Removal System Shutdown Cooling (SDC) mode 
Containment Isolation, Reactor Pressure--High Function by replacing the 
current requirement to perform TS SR 3.3.6.1.3, Perform Channel 
Calibration, with TS SR 3.3.6.1.1 Perform Channel Check, SR 3.3.6.1.2, 
Perform Channel Functional Test, SR 3.3.6.1.4, Calibrate the Trip 
Units, and SR 3.3.6.1.5, Perform Channel Calibration. These changes are 
to support a proposed plant modification to increase the reliability of 
SDC isolation logic by changing the source of the reactor high pressure 
input signal.
    Date of issuance: July 21, 2010.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 298.
    Renewed Facility Operating License No. DPR-59: The amendment 
revised the License and the Technical Specifications.
    Date of initial notice in Federal Register: October 6, 2009 (74 FR 
51239).
    The supplements dated March 5 and June 17, 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 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 July 21, 2010.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, and PSEG Nuclear, LLC, Docket Nos. 50-
277, Peach Bottom Atomic Power Station (PBAPS), Units 2, York and 
Lancaster Counties, Pennsylvania

    Date of application for amendments: August 28, 2009, as 
supplemented on February 25, 2010, and May 24, 2010.
    Brief description of amendments: The amendment modifies the PBAPS 
Unit 2 Technical Specification (TS) Section 5.5.12 to reflect a one-
time extension of the Type A containment Integrated Leak Rate Test 
(ILRT) to no later than October 2015. The TS revision allows a one-time 
extension of 5 years to the 10-year frequency of the performance-based 
leakage rate testing program for the PBAPS Unit 2 containment Type A 
ILRT test.
    Date of issuance: July 20, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 276.
    Renewed Facility Operating License Nos. DPR-44: Amendment revised 
the Technical Specifications.
    Date of Initial Notice in Federal Register: May 18, 2010 (75 FR 
27830).
    The supplements dated February 25, 2010, and May 24, 2010, 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 20, 2010.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 18, 2009, as 
supplemented on October 15, 2009, and April 14, 2010.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TSs) by modifying TS

[[Page 48380]]

Section 3.2.7.1 and 4.2.7.1, ``Primary Coolant System Pressure 
Isolation Valves,'' to incorporate requirements that are consistent 
with Section 3.4.5 of the Improved Standard Technical Specifications, 
NUREG-1433, Revision 3.0, ``Standard Technical Specifications General 
Electric Plants, BWR/4.''
    Date of issuance: July 26, 2010.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 206.
    Renewed Facility Operating License No. DPR-63: The amendment 
revises the License and TSs.
    Date of initial notice in Federal Register: October 14, 2009 (74 FR 
52824). The supplemental letters dated October 15, 2009, and April 14, 
2010, provided additional information that clarified the application, 
did not expand the scope of the application as originally noticed, and 
did not change the Nuclear Regulatory Commission (NRC) staff's original 
proposed no significant hazards consideration determination noticed in 
the Federal Register on October 14, 2009 (74 FR 52824).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 26, 2010.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power 
Plant, Unit 3, Humboldt County, California

    Date of application for amendment: April 9, 2010, and supplemented 
May 7, 2010.
    Brief description of amendment: The amendment Request deletes 
Technical Specification 3.1.3, ``Fuel Storage Pool Liner Water Level.'' 
Additional conforming and administrative changes are also made.
    Date of issuance: July 23, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 44.
    Facility Operating License No. DPR-7: This amendment revises the 
License.
    Date of initial notice in Federal Register: June 15, 2010 (75 FR 
33842).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 23, 2010.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of application for amendment: September 9, 2009.
    Brief description of amendment: The amendment changes the frequency 
of control rod notch testing, as specified in Technical Specification 
(TS) surveillance requirement 4.1.3.1.2.a, from at least once per 7 
days to at least once per 31 days. The amendment also adds the word 
``fully'' to the Action for TS Limiting Condition for Operation 3.9.2 
to clarify the requirement to fully insert all insertable control rods 
when the required source range monitor (SRM) instrumentation is 
inoperable. The proposed amendment is based on TS Task Force (TSTF) 
change, TSTF-475, Revision 1, ``Control Rod Notch Testing Frequency and 
SRM Insert Control Rod Action.''
    Date of issuance: July 21, 2010.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 182.
    Facility Operating License No. NPF-57: The amendment revised the 
TSs and the License.
    Date of initial notice in Federal Register: December 1, 2009 (74 FR 
62836).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 21, 2010.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company et al., Docket No. 52-011, Vogtle 
Electric Generating Plant ESP Site, Burke County, Georgia

    Date of amendment request: May 24, 2010, as supplemented June 2 and 
22, 2010.
    Description of amendment request: This amendment revises the Vogtle 
Electric Generating Plant ESP Site Safety Analysis Report (SSAR) to 
change the classification of backfill over the slopes of the Units 3 
and 4 excavations from Category 1 and 2 backfill to engineered granular 
backfill (EGB).
    Date of issuance: July 9, 2010.
    Effective date: As of date of issuance and shall be implemented 
within 15 days from the date of issuance.
    Amendment No.: 3.
    Early Site Permit No. ESP-004: Amendment revised the VEGP ESP SSAR.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes. June 3, 2010 (75 FR 31477). The supplements 
dated June 2 and 22, 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 staff's original 
proposed no significant hazards consideration determination. The June 
3, 2010 notice provided an opportunity to submit comments on the 
Commission's proposed NSHC determination. No comments have been 
received. The June 3, 2010 notice also provided an opportunity to 
request a hearing by August 2, 2010, but indicated that if the 
Commission makes a final NSHC determination, any such hearing would 
take place after issuance of the amendment.
    The Commission's related evaluation of the requested amendment, 
state consultation, and final NSHC determination are contained in a 
safety evaluation dated July 9, 2010. The NRC staff prepared an 
environmental assessment (75 FR 39284) and determined that the 
requested amendment will not have a significant effect on the quality 
of the human environment.
    Attorney for licensee: M. Stanford Blanton, Balch & Bingham, LLP.
    NRC Branch Chief: Jeffrey Cruz.

Southern Nuclear Operating Company et al., Docket No. 52-011, Vogtle 
Electric Generating Plant ESP Site, Burke County, Georgia

    Date of amendment request: April 20, 2010, as supplemented April 23 
and 28, May 5, 10, 13, 20, and 24, 2010.
    Description of amendment request: The amendment revised the Vogtle 
Electric Plant (VEGP) ESP Site Safety Analysis Report (SSAR) to allow 
the use of Category 1 and 2 backfill material from additional onsite 
areas that were not specifically identified in the VEGP ESP SSAR as 
backfill sources for the activities approved under the ESP and Limited 
Work Authorization.
    Date of issuance: June 25, 2010.
    Effective date: As of date of issuance and shall be implemented 
within 15 days from the date of issuance.
    Amendment No.: 2.
    Early Site Permit No. ESP-004: Amendment revised the VEGP ESP SSAR.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes. May 6, 2010 (75 FR 24993). The supplements 
dated May 5, 10, 13, 20, and 24, 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 staff's 
original proposed no significant hazards consideration determination. 
The May 6, 2010 notice provided an opportunity to submit comments on 
the Commission's proposed NSHC determination. No comments have been 
received. The May 6, 2010 notice also provided an opportunity to 
request a hearing by July 6, 2010, but indicated that if the Commission 
makes a final

[[Page 48381]]

NSHC determination, any such hearing would take place after issuance of 
the amendment.
    The Commission's related evaluation of the requested amendment, 
state consultation, and final NSHC determination are contained in a 
safety evaluation dated June 25, 2010. The NRC staff prepared an 
environmental assessment (75 FR 36446) and determined that the 
requested amendment will not have a significant effect on the quality 
of the human environment.
    Attorney for licensee: M. Stanford Blanton, Balch & Bingham, LLP.
    NRC Branch Chief: Jeffrey Cruz.

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

    Date of amendment request: October 10, 2010, as supplemented by 
letter dated March 8, 2010.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.1.7, ``Rod Position Indication,'' TS 3.2.1, ``Heat 
Flux Hot Channel Factor (FQ(Z)) (FQ 
Methodology),'' TS 3.2.2, ``Nuclear Enthalpy Rise Hot Channel Factor 
(F\N\[Delta]H),'' TS 3.2.4, ``Quadrant Power Tilt Ratio (QPTR),'' and 
TS 3.3.1, ``Reactor Trip System (RTS) Instrumentation,'' for use of the 
Best Estimate Analyzer for Core Operations--Nuclear (BEACON) Power 
Distribution Monitoring System (PDMS), as described in WCAP-12472-P-A, 
``BEACON Core Monitoring and Operations Support System,'' to perform 
power distribution surveillances.
    Date of issuance: July 23, 2010.
    Effective date: As the date of issuance and shall be implemented by 
December 29, 2010.
    Amendment No.: 188.
    Renewed Facility Operating License No. NPF-42. The amendment 
revised the Operating License and Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2010 (75 FR 
4120). The supplemental letter dated March 8, 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 the amendment is contained 
in a Safety Evaluation dated July 23, 2010.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 30th day of July, 2010.

    For The Nuclear Regulatory Commission.
Robert A. Nelson,
Acting Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2010-19678 Filed 8-9-10; 8:45 am]
BILLING CODE 7590-01-P