[Federal Register Volume 75, Number 85 (Tuesday, May 4, 2010)]
[Notices]
[Pages 23808-23819]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2010-10105]


=======================================================================
-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION

[NRC-2010-0169]


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 April 8, 2010 to April 21, 2010. The last 
biweekly notice was published on April 20, 2010 (75 FR 20627).

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.

[[Page 23809]]

    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, Public File Area 
O1F21, 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, Public File 
Area O1F21, 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

[[Page 23810]]

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 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, Public File Area O1F21, 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].

Calvert Cliffs Nuclear Power Plant, LLC, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, 
Maryland

    Date of amendments request: November 23, 2009.
    Description of amendments request: The amendment would modify the 
licensing basis and the Technical Specifications by allowing for the 
transition from Westinghouse Turbo fuel to AREVA Advanced CE-14 High 
Thermal Performance (HTP) fuel in the Calvert Cliffs reactors. The 
licensee plans to refuel and operate with AREVA fuel beginning with the 
refueling outage in 2011 for Unit No. 2 and 2012 for Unit No. 1. The 
transition is planned to occur over three refueling cycles on each 
unit.
    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?

[[Page 23811]]

    No.
    The reactor fuel and the analyses associated with it are not 
accident initiators. The response of the fuel to an accident is 
analyzed using conservative techniques and the results are compared 
to approved acceptance criteria. These evaluation results will show 
that the fuel response to an accident is within approved acceptance 
criteria for both cores loaded with the new AREVA Advanced CE-14 HTP 
fuel and cores loaded with both AREVA and Westinghouse Turbo fuel. 
Therefore, the change in fuel design does not affect accident or 
transient initiation or consequences.
    The proposed change to the Safety Limit Technical Specification 
(2.1.1.2) does not require any physical change to any plant system, 
structure, or component. The change to establish the peak fuel 
centerline temperature as the safety limit is consistent with the 
Standard Review Plan (SRP) for ensuring that the fuel design limits 
are met. Operations and analysis will continue to be in compliance 
with Nuclear Regulatory Commission (NRC) regulations. The peak fuel 
centerline temperature is the basis for protecting the fuel and is 
consistent with the analogous wording for other pressurized water 
reactor (PWR) plants. Providing the peak fuel centerline melt 
temperature as the safety limit does not impact the initiation or 
the mitigation of an accident.
    The proposed change to remove the total planar radial peaking 
factor (F\T\XY, Technical Specification 3.2.2) is based 
on a methodology change. During and after the transition to AREVA 
Advanced CE-14 HTP fuel, the core analyses are performed using AREVA 
methodologies. These methodologies do not use the total planar 
radial peaking factor (F\T\XY) as an initial value in the 
accident analyses. The linear heat rate algorithm limits are 
provided by the total integrated radial peaking factor, azimuthal 
power tilt, and axial shape index. The linear heat rate is evaluated 
in accordance with NRC-approved methodology and meets acceptance 
criteria. The total planar radial peaking factor is not an accident 
initiator and does not play a role in accident mitigation. A number 
of other changes are also made to remove references to Technical 
Specification 3.2.2 throughout the Technical Specifications.
    Topical reports have been reviewed and approved by the NRC for 
use in determining core operating limits. The core operating limits 
to be developed using the new methodologies will be established in 
accordance with the applicable limitations as documented in the 
appropriate NRC Safety Evaluation reports. The proposed change to 
add and remove various topical reports to Technical Specification 
5.6.5 enables the use of appropriate methodologies to re-analyze 
certain events. The proposed methodologies will ensure that the 
plant continues to meet applicable design criteria and safety 
analysis acceptance criteria.
    The proposed change to the list of NRC-approved methodologies 
listed in Technical Specification 5.6.5 is administrative in nature 
and has no impact on any plant configuration or system performance 
relied upon to mitigate the consequences of an accident. The 
proposed change will update the listing of NRC-approved 
methodologies to remove methods no longer used and add new methods 
consistent with the transition to AREVA Advanced CE-14 HTP fuel. 
Changes to the calculated core operating limits may only be made 
using NRC-approved methods, must be consistent with all applicable 
safety analysis limits and are controlled by the 10 CFR 50.59 
process. The list of methodologies in the Technical Specifications 
does not impact either the initiation of an accident or the 
mitigation of its consequences.
    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 type of accident from any accident previously 
evaluated?
    No.
    Use of AREVA Advanced CE-14 HTP fuel in the Calvert Cliffs 
reactor cores is consistent with the current plant design bases and 
does not adversely affect any fission product barrier, nor does it 
alter the safety function of safety systems, structures, or 
components, or their roles in accident prevention or mitigation. The 
operational characteristics of AREVA Advanced CE-14 HTP fuel are 
bounded by the safety analyses. The AREVA Advanced CE-14 HTP fuel 
design performs within fuel design limits and does not create the 
possibility of a new or different type of accident.
    The proposed change to the Safety Limit Technical Specification 
(2.1.1.2) does not require any physical change to any plant system, 
structure, or component, nor does it require any change in safety 
analysis methods or results. The existing analyses remain unchanged 
and do not affect any accident initiators that would create a new 
accident.
    The proposed change to remove the total planar radial peaking 
factor (F\T\XY, Technical Specification 3.2.2) is based 
on a change in analytical methods needed to support the physical 
fuel change. These methodologies do not use the total planar radial 
peaking factor (F\T\XY) as an initial value in the 
accident analysis. The total planar radial peaking factor does not 
play a role in accident mitigation and cannot create the possibility 
of a new or different kind of accident. A number of other changes 
are made to remove references to Technical Specification 3.2.2 
throughout the Technical Specifications.
    The proposed change to the list of topical reports used to 
determine the core operating limits is administrative in nature and 
has no impact on any plant configuration or on system performance. 
It updates the list of NRC-approved topical reports used to develop 
the core operating limits. There is no change to the parameters 
within which the plant is normally operated. The possibility of a 
new or different accident is not created.
    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?
    No.
    Use of AREVA Advanced CE-14 HTP fuel is consistent with the 
current plant design bases and does not adversely affect any fission 
product barrier, nor does it alter the safety function of safety 
systems, structures, or components, or their roles in accident 
prevention or mitigation. The operational characteristics of AREVA 
Advanced CE-14 HTP fuel are bounded by the safety analyses. The 
AREVA Advanced CE-14 HTP fuel design performs within fuel design 
limits. The proposed changes do not result in exceeding design basis 
limits. Therefore, all licensed safety margins are maintained.
    The proposed change to the Safety Limit Technical Specification 
(2.1.1.2) does not require any physical change to any plant system, 
structure, or component, nor does it require any change in safety 
analysis methods or results. Therefore, by changing the safety limit 
from peak linear heat rate to peak fuel centerline temperature, the 
margin as established in the current licensing basis remains 
unchanged.
    The proposed change to remove the total planar radial peaking 
factor (F\T\XY,Technical Specification 3.2.2) is based on 
a methodology change. The linear heat rate algorithm limits are 
provided by the total integrated radial peaking factor, azimuthal 
power tilt, and axial shape index. The linear heat rate is evaluated 
in accordance with NRC-approved methodology and meets acceptance 
criteria. Therefore, the margin as established for the linear heat 
rate remains unchanged. A number of other changes are made to remove 
references to Technical Specification 3.2.2 throughout the Technical 
Specifications.
    The proposed change to the list of topical reports does not 
amend the cycle specific parameters presently required by the 
Technical Specifications. The individual Technical Specifications 
continue to require operation of the plant within the bounds of the 
limits specified in the COLR [Core Operating Limits Report]. The 
proposed change to the list of analytical methods referenced in the 
COLR is administrative in nature and does not impact the margin of 
safety.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear 
Generation, Constellation Generation Group, LLC, 750 East Pratt Street, 
17th floor, Baltimore, MD 21202.
    NRC Branch Chief: Nancy L. Salgado.

[[Page 23812]]

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

    Date of amendment request: March 15, 2010.
    Description of amendment request: The proposed amendment would 
revise a Technical Specification (TS) to address the increased 
setpoints and setpoint tolerances for Safety Relief Valves (SRVs) and 
Spring Safety Valves (SSVs) and changes related to the replacement of 
four Target Rock two-stage SRVs with more reliable three-stage SRVs and 
two existing Dresser 3.749 inch throat diameter SSVs with Dresser 4.956 
inch diameter SSVs.
    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 increases the allowable as-found SRV and SSV 
setpoint tolerance, determined by test after the valves have been 
removed from service, from  1% to  3%. The 
proposed change also increases the SRV and SSV setpoints. Analysis 
of these changes demonstrates that reactor pressure will be 
maintained below the applicable code overpressure limits. The 
proposed change increases the SSV discharge capacity due to its 
increased throat diameter. The proposed change does not alter the TS 
requirements for the number of SRVs and SSVs required to be 
operable, the allowable as-left lift setpoint tolerance, the testing 
frequency, or the manner in which the valves are operated. 
Consistent with current TS requirements, the proposed change 
continues to require that the safety valves be adjusted to within 
 1% of their nominal lift setpoints following testing. 
The proposed increase in the SRV and SSV setpoint complies with the 
ASME Boiler and Pressure Vessel (B&PV) Code (1965 Edition, including 
January 1966 Addendum) for the pressure vessel, USAS Piping Code 
Section B31.1 for the steam space piping, and ASME Section III for 
the reactor coolant system recirculation piping. Since the proposed 
change does not alter the manner in which the valves are operated, 
there is no significant impact on the reactor operation.
    The proposed change does not involve a change to 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. Therefore, these changes 
will not increase the probability of an accident previously 
evaluated.
    Since an SSV setpoint increase and setpoint tolerance will 
increase the SSV safety valve opening pressure and an increase in 
the SSV throat size will increase the SSV flow capacity, the SSV 
dynamic loads are expected to increase. Entergy has evaluated the 
SSV dynamic loads for the associated piping. All piping and 
structures were found to meet Code requirements.
    Since an SRV setpoint and the setpoint tolerance increase will 
increase the SRV valve opening pressure, the SRV discharge dynamic 
loads will increase. Entergy has evaluated the SRV dynamic load 
increases for the associated piping and torus submerged structures 
and the evaluation concluded that all piping and structures were 
found to meet Code requirements.
    The proposed revision to the HPCI [high-pressure coolant 
injection] and RCIC [Reactor Core Isolation Cooling] pump 
operability determination surveillance follows the format of BWR 
Standard Technical Specification surveillance, and complies with in-
service testing for pump operability determination in accordance 
with ASME OM Code requirement.
    Generic considerations related to the change in setpoints and 
setpoint tolerance were addressed in NEDC-31753P, ``BWROG In-Service 
Pressure Relief Technical Specification Revision Licensing Topical 
Report,'' and were reviewed and approved by the NRC in a safety 
evaluation dated March 8, 1993. General Electric Hitachi Company 
(GEH) completed plant-specific analyses to assess the impact of 
increase in SRV and SSV setpoints and increase in the setpoint 
tolerance from  1% to  3%. The impact of the 
increases in the SRV and SSV setpoints and increases in the setpoint 
tolerances, as addressed in this analysis, included vessel 
overpressure, Updated Final Safety Analysis Report (UFSAR) Chapter 
14 events, ATWS [Anticipated Transient Without Scram], Loss of 
Coolant Accident (LOCA), containment response and dynamic loads, 
high-pressure systems performance, operating mode and equipment out 
of service. The proposed change is supported by GEH analysis of 
events that credit the SRVs and SSVs.
    The plant specific evaluations, required by the NRC's safety 
evaluation and performed to support this proposed change, 
demonstrate that there is no change to the design core thermal 
limits and adequate margin to the reactor coolant system pressure 
limits exists. These analyses also demonstrate that operation of 
Core Standby Cooling Systems (CSCS) is not adversely affected and 
the containment response following a LOCA is acceptable. The plant 
systems associated with these proposed changes are capable of 
meeting applicable design basis requirements and retain the 
capability to mitigate the consequences of accidents described in 
the UFSAR. Therefore, these changes do not involve an increase in 
the consequences of an accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change increases the allowable as-found lift 
setpoint tolerance for the Pilgrim SRV and SSV valves. The proposed 
change to increase the tolerance was developed in accordance with 
the provisions contained in the NRC safety evaluation for NEDC-
31753P. SRVs and SSVs installed in the plant following testing will 
continue to meet the current tolerance acceptance criteria of  1% of the nominal setpoint. The proposed change does not 
affect the manner in which the overpressure protection system is 
operated; therefore, there are no new failure mechanisms for the 
overpressure protection system.
    The proposed changes do not change the safety function of the 
SRVs and SSVs, or HPCI and RCIC systems. There is no alteration to 
the parameters within which the plant is normally operated. The 
increase in SRV and SSV setpoints, setpoint tolerance, and increased 
SSV discharge capacity are not precursors to new or different kinds 
of accidents and do not initiate new or different kinds of 
accidents. The impact of these changes have been analyzed and found 
to be acceptable within the design limits and plant operating 
procedures.
    As a result, no new failure modes are being introduced. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    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 modifies the setpoints at which protective 
actions are initiated, and [* * *] does not change the requirements 
governing operation or availability of safety equipment assumed to 
operate to preserve the margin of safety.
    Establishment of the  3% SRV and SSV setpoint 
tolerance limit does not adversely affect the operation of any 
safety-related component or equipment. Evaluations performed in 
accordance with the NRC safety evaluation for NEDC-31753P have 
concluded that all design limits will continue to be met.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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.

[[Page 23813]]

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: February 22, 2010.
    Description of amendment request: The proposed amendment will 
modify Technical Specification (TS) 3/4.9.4, ``Containment Building 
Penetrations,'' to allow alternative means of penetration closure 
during Core Alterations or irradiated fuel movement while in refueling 
operations. Additional improvements to the TS are also being proposed, 
as well as the elimination of TS 3/4.9.9, ``Containment Purge Valve 
Isolation System.'' The proposed changes are consistent with Revision 3 
of NUREG-1432, ``Standard Technical Specifications Combustion 
Engineering Plants.''
    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.
    TS 3/4.9.4 currently allows containment penetration flow paths 
to be open during Core Alterations or movement of irradiated fuel 
within containment under specific administrative controls. The 
proposed change would allow additional approved methods for ensuring 
positive penetration closure. The fuel handling accident (FHA) 
radiological analysis does not take credit for containment isolation 
or filtration. Therefore, the time required to close any open 
penetrations does not affect the radiological analysis dose 
calculations and the proposed change does not involve a significant 
increase in the consequences of an accident previously evaluated. 
The administrative controls for containment penetration closure are 
conservative even though not required by the accident analysis.
    The proposed revision only provides alternate methods of 
penetration closure and does not alter any plant equipment where the 
probability of an accident would be increased. The incorporation of 
purge valve isolation surveillance requirements for assuring purge 
valve Operability has no effect on the probability or consequences 
of the analyzed accidents.
    Therefore, this 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.
    Alternative methods of providing penetration closure do not 
create accident initiators and do not represent a significant change 
in the configuration of the plant. The proposed allowance to secure 
containment penetrations during refueling operations will not 
adversely effect plant safety functions or equipment operating 
practices such that a new or different accident could be created.
    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.
    TS Limiting Condition for Operation (LCO) 3.9.4 closure 
requirements for containment penetrations ensure that the 
consequences of a postulated FHA inside containment during Core 
Alterations or fuel handling activities are minimized. The LCO 
establishes containment closure requirements, which limit the 
potential escape paths for fission products by ensuring that there 
is at least one barrier to the release of radioactive material. The 
proposed change to allow alternate methods of reaching containment 
penetration closure during Core Alterations or fuel movement does 
not affect the expected dose consequences of a FHA since it does not 
credit containment building closure. The proposed administrative 
controls provide assurance that prompt closure of the penetration 
flow paths will be accomplished in the event of a FHA inside 
containment thus minimizing the transmission of radioactive material 
from the containment to the outside environment. The incorporation 
of purge valve isolation surveillance requirements does not reduce 
any margins of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: February 24, 2010.
    Description of amendment request: The proposed amendment deletes 
Operating License Condition 2.C.14 (Fuel Movement in the Fuel Handling 
Building) due to electing to comply with Section 50.68, ``Criticality 
accident requirements,'' of Title 10 of the Code of Federal Regulations 
(10 CFR). The Operating License Condition 2.C.14, ``no more than one 
fuel assembly shall be out of its shipping container or storage 
location at a given time,'' was one basis for the exemption from the 
criticality alarm system requirements of 10 CFR 70.24. The criticality 
accident requirements can be met either by complying with 10 CFR 70.24 
or 10 CFR 50.68 requirements. The 10 CFR 50.68 criteria are now being 
used; therefore, Operating License Condition 2.C.14 is no longer 
applicable.
    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 amendment deletes Operating License Condition 
2.C.14 (Fuel Movement in the Fuel Handling Building) due to electing 
to comply with 10 CFR 50.68 requirements.
    The proposed changes will not alter the configuration of the 
storage racks or their environment. The fuel racks will not be 
operated outside of their design limits, and no additional loads 
will be imposed on them. Therefore, these changes will not affect 
fuel storage rack performance or reliability. No new equipment will 
be introduced into the plant. The accuracies and response 
characteristics of existing instrumentation will not be modified. 
The proposed changes will not require, or result in, a change in 
safety system operation, and will not affect any system interface 
with the fuel storage racks. Fuel assembly placement will continue 
to be controlled in accordance with approved fuel handling 
procedures. All the requirements of 10 CFR 50.68 continue to be met 
which ensures no significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes will not affect any barrier that mitigates 
dose to the public, and will not result in a new release pathway 
being created. The functions of equipment designed to control the 
release of radioactive material will not be impacted, and no 
mitigating actions described or assumed for an accident in the UFSAR 
[Updated Final Safety Analysis Report] will be altered or prevented. 
No assumptions previously made in evaluating the consequences of an 
accident will need to be modified. Onsite dose will not be 
increased, so the access of plant personnel to vital areas of the 
plant will not be restricted, and mitigating actions will not be 
impeded.
    Therefore, it is concluded that the proposed changes do not 
significantly increase either the probability or consequences of any 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of

[[Page 23814]]

accident from any accident previously evaluated?
    Response: No.
    The proposed amendment deletes Operating License Condition 
2.C.14 (Fuel Movement in the Fuel Handling Building) due to electing 
to comply with 10 CFR 50.68 requirements.
    10 CFR 50.68(b)(1) provides the requirements to ensure that 
plant procedures shall prohibit the handling and storage at any one 
time of more fuel assemblies than have been determined to be safely 
subcritical under the most adverse moderation conditions feasible by 
unborated water. By meeting this criteria, the removal of Operating 
License Condition 2.C.14 will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Therefore, it is concluded that the proposed changes do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed amendment deletes Operating License Condition 
2.C.14 (Fuel Movement in the Fuel Handling Building) due to electing 
to comply with 10 CFR 50.68 requirements.
    10 CFR 50.68(b)(1) provides similar requirements as that 
contained in Operating License Condition 2.C.14. The NRC has 
approved the [Waterford Steam Electric Station, Unit 3] use of 10 
CFR 50.68 criteria. By meeting the 10 CFR 50.68(b)(1) requirements, 
there will not be a significant reduction in a margin of safety.
    Therefore, it is concluded that the proposed changes do not 
involve a significant reduction in a margin of safety.

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

Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power 
Station, Unit No. 1, DeWitt County, Illinois

    Date of amendment request: February 15, 2010.
    Description of amendment request: The proposed amendment would 
relocate selected Surveillance Requirement frequencies from the Clinton 
Power Station, Unit No. 1 (Clinton) Technical Specifications (TSs) to a 
licensee-controlled program. This change is based on the NRC-approved 
Industry Technical Specifications Task Force (TSTF) change TSTF-425, 
``Relocate Surveillance Frequencies to Licensee Control--Risk Informed 
Technical Specification Task Force (RITSTF) Initiative 5b,'' Revision 
3, (Agencywide Documents Access and Management System (ADAMS) Accession 
Package No. ML090850642). Plant-specific deviations from TSTF-425 are 
proposed to accommodate differences between the Clinton TSs and the 
model TSs originally used to develop TSTF-425.
    The Nuclear Regulatory Commission (NRC) staff issued a Notice of 
Availability for TSTF-425 in the Federal Register on July 6, 2009 (74 
FR 31996). The notice included a model safety evaluation (SE) and a 
model no significant hazards consideration (NSHC) determination. In its 
application dated February 15, 2010 (ADAMS Accession No. ML100470787), 
the licensee affirmed the applicability of the model NSHC determination 
which is presented below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of any accident previously 
evaluated?
    Response: No. The proposed change relocates 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 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. No new or different accidents result from 
utilizing the proposed change. The changes do not involve a physical 
alteration of the plant (i.e., no new or different type of equipment 
will be installed) or 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. Does the proposed change involve a significant reduction in 
the 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, Exelon will perform a probabilistic risk evaluation using 
the guidance contained in NRC approved NEI 04-01, Rev. 1. The 
methodology provides reasonable acceptance guidelines and methods 
for evaluating the risk increase of proposed changes to surveillance 
frequencies consistent with Regulatory Guide 1.177 [An Approach for 
Plant-Specific, Risk-Informed Decision-making: Technical 
Specifications].
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Stephen J. Campbell.

Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power 
Station, Unit No. 1, DeWitt County, Illinois

    Date of amendment request: March 3, 2010.
    Description of amendment request: The proposed amendment revises 
Technical Specification (TS) 3.1.7, ``Standby Liquid Control (SLC) 
System,'' to extend the completion time (CT) for Condition B (i.e., 
``Two SLC subsystems inoperable'') from 8 hours to 72 hours.
    Basis for proposed no significant hazards consideration: As 
required by 10 CFR 50.91(a), an analysis of the issue of no significant 
hazards consideration is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or

[[Page 23815]]

consequences of any accident previously evaluated?
    Response: No.
    The proposed amendment revises Technical Specification (TS) 
3.1.7, ``Standby Liquid Control (SLC) System,'' to extend the 
completion time (CT) for Condition B (i.e., ``Two SLC subsystems 
inoperable.'') from eight hours to 72 hours.
    The proposed change is based on a risk-informed evaluation 
performed in accordance with Regulatory Guides (RG) 1.174, ``An 
Approach for Using Probabilistic Risk Assessment in Risk-Informed 
Decisions On Plant-Specific Changes to the Licensing Basis,'' and RG 
1.I77, ``An Approach for Plant-Specific, Risk-Informed Decision-
making: Technical Specifications.''
    The proposed amendment modifies an existing CT for a dual-train 
SLC system inoperability. The condition evaluated, the action 
requirements, and the associated CT do not impact any initiating 
conditions for any accident previously evaluated.
    The proposed amendment does not increase postulated frequencies 
or the analyzed consequences of an Anticipated Transient Without 
Scram (ATWS). Requirements associated with 10 CFR 50.62 will 
continue to be met. In addition, the proposed amendment does not 
increase postulated frequencies or the analyzed consequences or a 
large-break loss-of-coolant accident for which the SLC system will 
be used for pH control. The extended CT provides additional time to 
implement actions in response to a dual-train SLC system 
inoperability, while also minimizing the risk associated with 
continued operation. 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 previously evaluated?
    Response: No.
    The proposed amendment revises TS 3.1.7 to extend the CT for 
Condition B from eight hours to 72 hours. The proposed amendment 
does not involve any change to plant equipment or system design 
functions. This proposed TS amendment does not change the design 
function of the SLC system and does not affect the system's ability 
to perform its design function. The SLC system provides a method to 
bring the reactor, at any time in a fuel cycle, from full power and 
minimum control rod inventory to a subcritical condition with the 
reactor in the most reactive xenon free state without taking credit 
for control rod movement. Required actions and surveillance 
requirements are sufficient to ensure that the SLC system functions 
are maintained. No new accident initiators are introduced by this 
amendment. Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment revises TS 3.1.7 to extend the CT for 
Condition B from eight hours to 72 hours. The proposed amendment 
does not involve any change to plant equipment or system design 
functions. 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 setpoints for the actuation of 
equipment relied upon to respond to an event.
    The proposed amendment does not modify the condition or point at 
which SLC is initiated, nor does it affect the system's ability to 
perform its design function. In addition, the proposed change 
complies with the intent of the defense-in-depth philosophy and the 
principle that sufficient safety margins are maintained, consistent 
with RG 1.177 requirements (i.e., Section C, ``Regulatory 
Position,'' paragraph 2.2 ``Traditional Engineering 
considerations'').
    Based on the above analysis, EGC concludes that the proposed 
amendment presents no significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

    The NRC staff has reviewed the analysis adopted by the licensee 
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. Bradley J. Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Stephen J. Campbell.

Exelon Generation Company, LLC, and PSEG Nuclear, LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station (PBAPS), Units 2 and 
3, York and Lancaster Counties, Pennsylvania

    Date of amendment request: August 31, 2009.
    Description of amendment request: The proposed amendment would 
modify the PBAPS Technical Specifications (TS) 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.'' Additionally, the change would 
add a new program, the Surveillance Frequency Control Program, to TS 
Section 5, Administrative Controls. The changes are based on NRC-
approved Industry Technical Specifications Task Force (TSTF) Traveler 
425, Revision 3, ``Relocate Surveillance Frequencies to Licensee 
Control--Risk Informed Technical Specification Task Force Initiative 
5b,'' with optional changes and variations as described in Attachment 
1, Section 2.2 of the licensee's submittal dated August 31, 2009.
    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 [SFCP]. 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 previously evaluated?
    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 
the margin of safety?
    Response: No.
    [* * * T]here is no impact to safety analysis acceptance 
criteria as described in the plant licensing basis. To evaluate a 
change in the relocated surveillance frequency, Exelon 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

[[Page 23816]]

not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, and with the changes noted above, it appears that the 
three standards of 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves NSHC.
    Attorney for licensee: Mr. J. Bradley Fewell, Associate General 
Counsel, Exelon Generation Company LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Harold K. Chernoff.

FPL Energy Seabrook, LLC Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire

    Date of amendment request: March 16, 2010.
    Description of amendment request: The proposed changes would revise 
the Seabrook Technical Specifications requirement that the Operations 
Manager shall have held a senior reactor operator license for the 
Seabrook Station prior to assuming the Operations Manager position. 
Specifically, the proposed change would require the Operations Manager 
to meet one of the following: (1) Hold a senior operator license; (2) 
have held a senior operator license for a similar unit; or (3) have 
been certified for equivalent senior operator knowledge. In its 
application dated March 16, 2010, the licensee concluded that the no 
significant hazards consideration (NSHC) determination presented in the 
notice is applicable to Seabrook Station.
    Basis for proposed NSHC determination: As required by 10 CFR 
50.91(a), the licensee has provided its analysis of the issue of NSHC, 
which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    [The requested change would only affect the qualification 
requirements for the Operations Manager Position]. The proposed 
change does not impact the configuration or function of plant 
structures, systems, or components (SSCs) or the manner in which 
SSCs are operated, maintained, modified, tested, or inspected. No 
actual facility equipment or accident analyses will be affected by 
the proposed changes. Therefore, this request has no [significant] 
impact on 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 requested change would only affect the qualification 
requirements for the Operations Manager Position]. The proposed 
change does not alter the plant configuration, require new plant 
equipment to be installed, alter 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, this request 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 a margin of safety.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel cladding, reactor coolant 
system pressure boundary, and containment structure) to limit the 
level of radiation dose to the public. [The requested change would 
only affect the qualification requirements for the Operations 
Manager Position]. No actual plant equipment or accident analyses 
will be affected by the proposed changes. Additionally, the proposed 
changes will not relax any criteria used to establish safety limits, 
will not relax any safety system settings, and will not relax the 
bases for any limiting conditions for operation. 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. Therefore, these 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, and with the changes noted above, it appears that the 
three standards of 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves NSHC.
    Attorney for licensee: M.S. Ross, Florida Power & Light Company, 
P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Harold K. Chernoff.

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

    Date of amendment request: November 24, 2009.
    Description of amendment request: The proposed amendments would 
make changes to Technical Specification (TS) Section 4.2.1, Fuel 
Assemblies, and TS Section 5.6.5, Core Operating Limit Report, by 
revising the TS to allow the use of Optimized ZIRLO\TM\ fuel rod 
cladding material.
    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.
    Westinghouse Electric Company, LLC (Westinghouse) topical report 
WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A ``Optimized 
ZIRLO\TM\'', July 2006, provides the details and results of material 
testing of Optimized ZIRLO\TM\ compared to standard ZIRLO\TM\ as 
well as the material properties to be used in various models and 
methodologies when analyzing Optimized ZIRLO\TM\. The Nuclear 
Regulatory Commission (NRC) has allowed use of Optimized ZIRLO\TM\ 
fuel cladding material in Westinghouse fueled reactors provided that 
licensees ensure compliance with the conditions and limitations set 
forth in the NRC Safety Evaluation (SE) for the topical report. By 
satisfying the conditions and limitations of the NRC SE through 
completed actions and its approved reload safety evaluation process, 
the licensee ensures that the effects of Optimized ZIRLO\TM\ on 
PINGP core performance are evaluated and that the probability or 
consequences of previously-evaluated accidents are not increased.
    Therefore, the proposed change of adding a cladding material 
does not result in an increase to 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.
    Material properties of this fuel design have been evaluated in 
Westinghouse topical report WCAP-12610-P-A and CENPD-404-P-A, 
Addendum 1-A ``Optimized ZIRLO\TM\'' July 2006. That report provides 
the details and results of material testing of Optimized ZIRLO\TM\ 
compared to standard ZIRLO\TM\ as well as the material properties to 
be used in various models and methodologies when analyzing Optimized 
ZIRLO\TM\. Neither that topical report nor the associated NRC SE 
identifies the possibility of a new or different kind of accident 
resulting from this change for generic application in Westinghouse 
reactors. As demonstrated in that topical report and stated in the 
NRC SE, there is reasonable assurance that under both normal and 
accident conditions, the Optimized ZIRLO\TM\ fuel cladding will be 
able to safely operate and comply with NRC regulations. By 
satisfying the conditions and limitations of the NRC SE by virtue of 
its completed actions and its approved reload safety evaluation 
process, the licensee ensures that the effects of Optimized 
ZIRLO\TM\ are evaluated and will not create the possibility of a new 
or different kind of accident. Assurance that the possibility of new 
or different type of accidents will not be created on a site-
specific basis is inherent to the reload safety evaluation process 
approved for use at the PINGP. Site specific evaluation of the PINGP 
core designs with Optimized ZIRLO\TM\ will be performed 
programmatically and necessarily by the approved reload safety 
evaluation process.

[[Page 23817]]

    Therefore, the proposed change of adding a cladding material 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The cladding material used in the fuel rods is designed and 
tested to prevent excessive fuel temperatures, excessive internal 
rod gas pressure due to fission gas releases, and excessive cladding 
stresses and strains. Optimized ZIRLO\TM\ was developed to meet 
these needs and provides a reduced corrosion rate while maintaining 
the benefits of mechanical strength and resistance to accelerated 
corrosion from abnormal chemistry conditions. Westinghouse topical 
report WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A ``Optimized 
ZIRLO\TM\, July 2006, provides the details and results of material 
testing of Optimized ZIRLO\TM\ compared to standard ZIRLO\TM\ as 
well as the material properties to be used in various models and 
methodologies when analyzing Optimized ZIRLO\TM\. The NRC has 
allowed use of Optimized ZIRLO\TM\ fuel cladding material detailed 
within this topical report as detailed within their SE. Therefore, 
the change in material does not result in a significant reduction in 
a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: Robert J. Pascarelli.

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: January 27, 2010.
    Description of amendment request: The proposed amendments would 
make changes to the Technical Specifications (TS) to revise TS 3.8.3, 
``Diesel Fuel Oil''. The amendments would revise the diesel fuel oil 
(DFO) storage volumes applicable to Unit 1 in TS 3.8.3 Condition 
statements A and D, and increase the Unit 1 DFO supply required by 
surveillance requirement 3.8.3.1. The amendments would clarify wording 
in TS 3.8.3 Condition B statement which applies to both units.
    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 increase the 
emergency diesel generator fuel oil storage volumes specified in the 
Technical Specification Condition statements and Surveillance 
Requirements. Also a word was added to a Condition statement to 
clarify its meaning.
    The emergency diesel generators and their supporting diesel fuel 
oil storage systems are not accident initiators and therefore the 
proposed fuel oil storage volume increases do not involve an 
increase in the probability of an accident.
    The proposed increased diesel fuel oil storage volumes provide 
sufficient volumes to maintain the current licensing basis for 
emergency diesel generator operation. Thus the proposed fuel oil 
storage volume increases do not involve a significant increase in 
the consequences of an accident.
    The proposed Technical Specification Condition statement wording 
clarification is administrative and thus does not involve an 
increase in the probability of an accident or an increase in the 
consequences of an accident.
    Therefore, the proposed Technical Specification 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 increase the 
emergency diesel generator fuel oil storage volumes specified in the 
Technical Specification Condition statements and Surveillance 
Requirements. Also a word was added to a Condition statement to 
clarify its meaning.
    The proposed Technical Specification changes which increase 
emergency diesel generator fuel oil storage volumes do not change 
any system operations or maintenance activities. The changes do not 
involve physical alteration of the plant, that is, no new or 
different type of equipment will be installed. The changes do not 
alter assumptions made in the safety analyses but ensures that the 
diesel generators operate as assumed in the accident analyses. These 
changes do not create new failure modes or mechanisms which are not 
identifiable during testing and no new accident precursors are 
generated.
    The proposed Technical Specification Condition statement wording 
clarification is administrative and thus does not create the 
possibility of a new or different kind of accident.
    Therefore, the proposed Technical Specification 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 increase the 
emergency diesel generator fuel oil storage volumes specified in the 
Technical Specification Condition statements and Surveillance 
Requirements. Also a word was added to a Condition statement to 
clarify its meaning.
    Since this license amendment proposes Technical Specification 
changes which increase the required fuel oil storage volumes, 
margins of safety are increased and thus no margin of safety is 
reduced as part of this change.
    The proposed Technical Specification Condition statement wording 
clarification is administrative and thus does not involve a 
significant reduction in a margin of safety.
    Therefore, the proposed Technical Specification 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 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. Pascarelli.

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 amendment request: February 2, 2010.
    Description of amendment request: The proposed amendments would 
revise the verification requirements for the Reactor Trip System 
Instrumentation. Specifically, the amendment proposes the addition to 
Table 3.3.1-1 of a response time measurement for the verification of 
the Power Range Neutron High Positive Rate Trip (PFRT) function as 
recommended by Westinghouse Nuclear Safety Advisory Letter (NSAL-09-01) 
``Rod Withdrawal at Power Analysis for Reactor Coolant System 
Overpressure.''
    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 to Vogtle Electric Generating Plant (VEGP) 
Technical Specification (TS) 3.3.1, ``Reactor Trip

[[Page 23818]]

System (RTS) Instrumentation,'' Table 3.3.1-1, ``Reactor Trip System 
Instrumentation'' does not significantly increase the probability or 
consequences of an accident previously evaluated in the Updated 
Final Safety Analysis Report (UFSAR). The overall protection system 
performance will remain within the bounds of the accident analysis 
since there are no hardware changes. The design of the Reactor Trip 
System (RTS) instrumentation, specifically the positive range 
neutron flux high positive rate trip (PFRT) function, will be 
unaffected. The reactor protection system will continue to function 
in a manner consistent with the plant design basis. All design, 
material, and construction standards that were applicable prior to 
the request are maintained.
    The proposed change adds an additional surveillance requirement 
to assure that the PFRT is verified to be consistent with the safety 
analysis and licensing basis. In this specific case, a response time 
verification requirement will be added to the PFRT function.
    The proposed changes will not modify any system interface. The 
proposed changes will not affect the probability of any event 
initiators. There will be no degradation in the performance of or an 
increase in the number of challenges imposed on safety-related 
equipment assumed to function during an accident situation. There 
will be no change to normal plant operating parameters or accident 
mitigation performance. The proposed change will not alter any 
assumptions nor change any mitigation actions in the radiological 
consequences evaluations in the UFSAR.
    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility or the manner in which 
the plant is operated and maintained. The proposed changes do not 
alter nor prevent the ability of SSCs from performing their intended 
function to mitigate the consequences of an initiating event within 
the assumed acceptance limits. The proposed change is consistent 
with the safety analyses assumptions and resultant consequences. The 
RCS overpressure limit listed in Specification 2.1.2 of the VEGP 
Technical Specifications (i.e., 2735 psig) is not violated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any previously evaluated?
    Response: No.
    There are no hardware changes nor are there any changes in the 
method by which any safety related plant system performs its safety 
function. This change will not affect the normal method of plant 
operation nor change any operating parameters.
    No performance requirements will be affected; however, the 
proposed change adds an additional surveillance requirement. The 
additional surveillance requirement is consistent with assumptions 
made in the safety analyses and licensing basis.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of this change. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of this change.
    Therefore, the proposed change does not create the possibility 
of a new or different accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change does not affect the acceptance criteria for 
any analyzed event nor is there a change to any Safety Limits. There 
will be no effect on the manner in which Safety Limits or Limiting 
Conditions of Operations are determined, nor will there be any 
effect on those plant systems necessary to assure the accomplishment 
of protection functions.
    This change is consistent with the assumptions made in the 
safety analyses. The addition of a surveillance requirement 
increases the margin of safety by assuring that the associated 
safety analysis assumption on the PFRT response time is verified.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standard set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

Previously Published Notices of Consideration of Issuance of Amendments 
to Facility Operating Licenses, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear 
Plant, Van Buren County, Michigan

    Date of amendment request: March 31, 2010.
    Brief description of amendment request: The proposed amendment 
would add new license condition 2.C(4) stating that performance of 
Technical Specification surveillance requirement 3.1.4.3, which 
verifies control rod freedom of movement, is not required for control 
rod drive 22 during cycle 21 until the next entry into Mode 3 in a 
maintenance or refueling outage, whichever is earlier.
    Date of publication of individual notice in Federal Register: April 
14, 2010 (75 FR 19428).
    Expiration date of individual notice: June 13, 2010.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of amendment request: March 29, 2010, as supplemented by 
letter dated March 29, 2010.
    Brief description of amendment request: The proposed amendment 
would revise the Technical Specification (TS) 3.3.2, ``Engineered 
Safety Feature Actuation System (ESFAS) Instrumentation,'' regarding 
function 6.g in TS Table 3.3.2-1. Function 6.g provides an auxiliary 
feedwater (AFW) start signal that is provided to the motor-driven AFW 
pumps in the event of a trip of both turbine-driven main feedwater 
pumps. The changes would revise Condition J for ESFAS instrumentation 
function 6.g to read, ``One or more Main Feedwater Pumps trip 
channel(s) inoperable.'' The licensee will make corresponding changes 
to Required Action J.1 and the Note above Required Actions J.1 and J.2 
for consistency with the revised Condition.
    Date of publication of individual notice in Federal Register: April 
14, 2010 (75 FR 19431).
    Expiration date of individual notice: April 28, 2010, for public 
comments; June 14, 2010, for hearing requests.

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

[[Page 23819]]

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, Public File Area 01F21, 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].

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 
supplemented by letter dated February 5, 2010.
    Brief description of amendment: The amendment modified the 
Technical Specification (TS) 5.5.7, ``Inservice Testing Program,'' by 
replacing the references from the American Society of Mechanical 
Engineers (ASME) Boiler and Pressure Vessel Code to the current Code of 
Record, the ASME Operation and Maintenance Nuclear Power Plants Code 
(ASME OM Code), the Code of Record for the James A. FitzPatrick Nuclear 
Power Plant (JAFNPP) Inservice Testing (IST) Program. This is an 
administrative amendment to maintain the TS current with the NRC 
accepted Code of Record for JAFNPP IST Program.
    Date of issuance: April 12, 2010.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 296.
    Renewed Facility Operating License No. DPR-59: The amendment 
revised the License and the Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2010 (75 FR 
4117).
    The February 5, 2010, supplement 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 April 12, 2010.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station (Byron), Unit Nos. 1 and 2, Ogle County, Illinois

    Date of application for amendment: September 24, 2009, as 
supplemented by letters dated November 13, 2009; January 19, 2010; 
March 1, 2010; March 9, 2010 (two letters); and March 19, 2010.
    Brief description of amendment: The amendments adds a new 
Completion Time (CT) of 144 hours to restore a unit-specific essential 
service water train to operable status associated with the Limiting 
Condition for Operation for Technical Specification (TS) 3.7.8, 
``Essential Service Water (SX) System.'' The new CT will be used for 
maintenance during the Byron, Unit No. 2, spring 2010, refueling 
outage. The licensee requested the new CT to replace two of the four SX 
pump suction isolation valves without having to shutdown Byron, Unit 
No. 1; maintenance history has shown that replacement of the SX pump 
suction isolation valves cannot be assured within the existing 72 hour 
CT window.
    Date of issuance: April 9, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: Unit No. 1--168; Unit No. 2--168.
    Facility Operating License Nos. NPF-37 and NPF-66: The amendments 
revise the TSs and Licenses.
    Date of initial notice in Federal Register: December 1, 2009 (74 FR 
62835).
    The supplemental letters 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 Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 9, 2010.
    No significant hazards consideration comments received: No.

R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of application for amendment: September 18, 2009.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 5.5.7, ``Inservice Testing Program,'' by 
incorporating TS Task Force Traveler (TSTF)-479, ``Changes to Reflect 
Revision of 10 CFR 50.55a,'' and TSTF-497, ``Limit Inservice Testing 
Program SR [Surveillance Requirement] 3.0.2 Application to Frequencies 
of 2 Years or Less.'' Specifically, the amendments (1) replace 
references to the American Society of Mechanical Engineers (ASME) 
Boiler and Pressure Vessel Code, Section XI with the ASME Code for 
Operation and Maintenance of Nuclear Power Plants for inservice testing 
activities, and (2) applies the extension allowance of SR 3.0.2 to 
other normal and accelerated inservice testing frequencies of 2 years 
or less that were not included in the frequencies listed in TS 5.5.7.a.
    Date of issuance: April 8, 2010.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 110.
    Renewed Facility Operating License No. DPR-18: Amendment revised 
the License and Technical Specifications.
    Date of initial notice in Federal Register: November 3, 2009 (74 FR 
56887).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 8, 2010.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 22nd day of April 2010.

    For the Nuclear Regulatory Commission.
Robert A. Nelson,
Deputy Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2010-10105 Filed 5-3-10; 8:45 am]
BILLING CODE 7590-01-P