[Federal Register Volume 75, Number 65 (Tuesday, April 6, 2010)]
[Notices]
[Pages 17439-17452]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2010-7451]


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NUCLEAR REGULATORY COMMISSION

[NRC-2010-0145]


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 March 11, 2010, to March 24, 2010. The last 
biweekly notice was published on March 23, 2010 (75 FR 13786).

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, Rulemaking 
and Directives Branch (RDB), 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 RDB 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

[[Page 17440]]

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 17441]]

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].

Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414, 
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of amendment request: October 29, 2009.
    Description of amendment request: The amendments would delete a 
license condition located in each of the unit's Facility Operating 
Licenses (FOLs) which restricts the maximum fuel rod average burnup. 
Deletion of this condition would allow the maximum fuel rod average 
burnup to increase.
    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.
    Deletion of the MNS [McGuire Nuclear Station] and CNS [Catawba 
Nuclear Station] FOL Appendix B conditions currently limiting 
maximum rod average burnup to 60 GWd/MTU [Gigawatt-day per Metric 
Ton Uranium] does not add, delete, or modify any MNS or CNS systems, 
structures, or components (SSCs). The proposed amendment would 
effectively allow future increases in the MNS and CNS maximum rod 
average burnup limit up to and including 62 GWd/MTU using existing 
fuel management methods, analyses, and models that have been 
reviewed and approved by the NRC [Nuclear Regulatory Commission]. 
Maximum average rod burnup limits will continue to be maintained 
within safe and acceptable limits using these fuel management 
methods and models.
    Increasing the MNS and CNS maximum rod average burnup limit does 
not affect the thermal hydraulic response or the radiological 
consequences of any previously evaluated accident. The fuel rod 
design criteria will continue to be met at the maximum burnup limits 
allowed utilizing the current fuel management, analysis, and 
evaluation processes. An increase to the maximum rod average burnup 
limit will not increase the likelihood of a malfunction of nuclear 
fuel since the fuel currently used at MNS and CNS has been designed 
to support a maximum rod average burnup up to and including 62 GWd/
MTU. Therefore, the proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment would delete MNS and CNS FOL Appendix B 
conditions which currently limits maximum rod average burnup to 60 
GWd/MTU. The proposed amendment would effectively allow future 
increases in the MNS and CNS maximum rod average burnup limit up to 
and including 62 GWd/MTU using existing fuel management methods, 
analyses, and models that have been reviewed and approved by the 
NRC. The proposed amendment does not change the design function of 
the nuclear fuel or create any credible new failure mechanisms or 
malfunctions for the nuclear fuel. Fuel rod design criteria will 
continue to be met at the maximum burnup limits allowed under the 
fuel management methods and models that have been previously 
reviewed and approved by the NRC. Therefore, the proposed amendment 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment would delete a MNS and CNS FOL Appendix B 
conditions which currently limits maximum rod average burnup to 60 
GWd/MTU. The proposed amendment would effectively allow future 
increases in the MNS and CNS maximum rod average burnup limit up to 
and including 62 GWd/MTU using existing fuel management methods, 
analyses, and models that have been reviewed and approved by the 
NRC. The proposed amendment does not result in altering or exceeding 
a design basis or safety limit for the plant. All current fuel 
design criteria will continue to be satisfied, and the safety 
analysis of record, including evaluations of the radiological 
consequences of design bases accidents, will remain applicable. 
Radiological consequences have been evaluated consistent with 
methodologies approved by the NRC. [Therefore, the proposed 
amendment does not involve a significant reduction in a margin of 
safety.]

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Associate General 
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South 
Church Street, EC07H, Charlotte, NC 28202.
    NRC Branch Chief: Gloria Kulesa.

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: October 29, 2009.
    Description of amendment request: The amendments would delete a 
license condition located in each of the unit's Facility Operating 
Licenses (FOLs) which restricts the maximum fuel rod average burnup. 
Deletion of this condition would allow the maximum fuel rod average 
burnup to increase.
    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.
    Deletion of the MNS [McGuire Nuclear Station] and CNS [Catawba 
Nuclear Station]

[[Page 17442]]

FOL Appendix B conditions currently limiting maximum rod average 
burnup to 60 GWd/MTU [Gigawatt-day per Metric Ton Uranium] does not 
add, delete, or modify any MNS or CNS systems, structures, or 
components (SSCs). The proposed amendment would effectively allow 
future increases in the MNS and CNS maximum rod average burnup limit 
up to and including 62 GWd/MTU using existing fuel management 
methods, analyses, and models that have been reviewed and approved 
by the NRC [Nuclear Regulatory Commission]. Maximum average rod 
burnup limits will continue to be maintained within safe and 
acceptable limits using these fuel management methods and models.
    Increasing the MNS and CNS maximum rod average burnup limit does 
not affect the thermal hydraulic response or the radiological 
consequences of any previously evaluated accident. The fuel rod 
design criteria will continue to be met at the maximum burnup limits 
allowed utilizing the current fuel management, analysis, and 
evaluation processes. An increase to the maximum rod average burnup 
limit will not increase the likelihood of a malfunction of nuclear 
fuel since the fuel currently used at MNS and CNS has been designed 
to support a maximum rod average burnup up to and including 62 GWd/
MTU. Therefore, the proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment would delete MNS and CNS FOL Appendix B 
conditions which currently limits maximum rod average burnup to 60 
GWd/MTU. The proposed amendment would effectively allow future 
increases in the MNS and CNS maximum rod average burnup limit up to 
and including 62 GWd/MTU using existing fuel management methods, 
analyses, and models that have been reviewed and approved by the 
NRC. The proposed amendment does not change the design function of 
the nuclear fuel or create any credible new failure mechanisms or 
malfunctions for the nuclear fuel. Fuel rod design criteria will 
continue to be met at the maximum burnup limits allowed under the 
fuel management methods and models that have been previously 
reviewed and approved by the NRC. Therefore, the proposed amendment 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment would delete a MNS and CNS FOL Appendix B 
conditions which currently limits maximum rod average burnup to 60 
GWd/MTU. The proposed amendment would effectively allow future 
increases in the MNS and CNS maximum rod average burnup limit up to 
and including 62 GWd/MTU using existing fuel management methods, 
analyses, and models that have been reviewed and approved by the 
NRC. The proposed amendment does not result in altering or exceeding 
a design basis or safety limit for the plant. All current fuel 
design criteria will continue to be satisfied, and the safety 
analysis of record, including evaluations of the radiological 
consequences of design bases accidents, will remain applicable. 
Radiological consequences have been evaluated consistent with 
methodologies approved by the NRC. [Therefore, the proposed 
amendment does not involve a significant reduction in a margin of 
safety.]

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Associate General 
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South 
Church Street, EC07H, Charlotte, NC 28202.
    NRC Branch Chief: Gloria Kulesa.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of amendment request: February 8, 2010.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) requirements related to TS 3.1.3, 
``Control Rod Operability,'' and TS 3.1.5, ``Control Rod Scram 
Accumulators,'' to be consistent with NUREG-1433, ``Standard Technical 
Specifications General Electric Plants, BWR/4.'' The proposed amendment 
also corrects certain typographical errors.
    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 changes involve an administrative change to LCO 
[limiting condition for operation] 3.1.3, ``Control Rod 
OPERABILITY,'' and a simplification in the modeling methodology for 
scram time analysis in LCO 3.1.5, ``Control Rod Scram 
Accumulators,'' that continue to ensure that control rod operability 
requirements for the number and distribution of operable, slow and 
stuck control rods satisfy scram reactivity rate assumptions used in 
the plant safety analysis.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not involve any physical alteration of 
the plant (no new or different type of equipment is being installed) 
and do not involve a change in the design, normal configuration, or 
basic operation of the plant. The proposed changes do not introduce 
any new accident initiators. The proposed changes do not involve 
significant changes in the fundamental methods governing normal 
plant operation and do not require unusual or uncommon operator 
actions. The proposed changes provide assurance that the plant will 
not be operated in a mode or condition that violates the assumptions 
or initial conditions in the safety analyses and that the systems, 
structures, and components (SSCs) remain capable of performing their 
intended safety functions as assumed in the same analyses. 
Consequently, the response of the plant and the plant operator to 
postulated events will not be significantly different.
    Therefore, the proposed TS 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.
    Margin of safety is related to confidence in the ability of 
fission product barriers to perform their intended design functions 
during and following an accident. The proposed changes address 
control rod operability and continue to ensure control rod scram 
time acceptance criteria is satisfied. The scram time test 
acceptance criteria and control rod operability restrictions are 
based on industry approved methodology and will continue to ensure 
control rod scram design functions and reactivity insertion 
assumptions used in the safety analyses continue to be protected.
    Therefore, 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: William A. Horin, Esq., Winston & Strawn, 
1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Michael T. Markley.

Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc., 
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, 
Louisiana

    Date of amendment request: January 28, 2010.
    Description of amendment request: The proposed license amendment

[[Page 17443]]

request modifies the licensee's commitment to Table B-1, ``Minimum 
Staffing Requirements for NRC Licensees for Nuclear Power Plant 
Emergencies,'' of NUREG-0654/FEMA-REP-1, Revision 1, ``Criteria for 
Preparation and Evaluation of Radiological Emergency Response Plans and 
Preparedness in Support of Nuclear Power Plants,'' dated November 1980. 
Current Table 13.3-17, ``Repair and Corrective Actions,'' of the 
Emergency Plan only allows that Electrical or Instrumentation & Control 
technicians may fill these two positions. This change will allow these 
two maintenance positions on shift to be filled with any combination of 
the three maintenance craft disciplines.
    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 not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    No.
    The proposed change does not increase the probability or 
consequences of an accident. The change only impacts the 
implementation of the Emergency Plan by changing staffing of the 
Repair and Corrective action functions after an event. It has no 
impact on plant equipment or the operation of plant equipment and 
thus has no impact on the probability or consequences of an event. 
The number of personnel on shift has not been revised from the 
current Emergency Plan. The repair and corrective action function 
would continue to be performed by trained personnel because the 
process, personnel, and equipment involved in implementing the 
Emergency Plan would complete the same functions as those completed 
under the existing Emergency Plan, the Plan would continue to ensure 
adequate protection of public health and safety.
    (2) Does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    No.
    The change only impacts the implementation of the Emergency Plan 
by changing staffing of the Repair and Corrective action functions 
after an event. The change does not impact any plant equipment or 
systems needed to respond to an accident, nor does it involve any 
analysis of plant accidents. The proposed change does not create a 
new or different kind of accident from any previously evaluated 
because this change only impacts emergency response repair 
functions.
    (3) Does not involve a significant reduction in a margin of 
safety.
    No.
    The change to the Emergency Plan does not reduce the margin of 
safety currently provided by the Plan as it maintains the current 
number of personnel on shift to perform Repair and Corrective action 
functions. Repair and corrective actions will continue to be 
performed by trained personnel. 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: Joseph A. Aluise, Associate General 
Council--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: January 24, 2010.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 1.0, Definitions, TS 
Section 3.6, Primary System Boundary Specifications 3.6.A, and TS 
Administrative Controls Section 5.5, to include reference to the 
Pressure and Temperature Limits Report (PTLR). The PTLR includes 
revised 34 effective full-power years (EFPY) P-T Curves, neutron 
fluence, and Adjusted Reference Temperature (ART) values.
    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 modifies Technical Specifications (TS) 
Section 1.0
    (``Definitions''), Specification 3.6.A.2, and revises 5.0 
(``Administrative Controls''), to include section 5.5.9 to include 
reference to the Pressure and Temperature Limits Report (PTLR). This 
change adopts the methodology of SIR-05-044-A, ``Pressure-
Temperature Limits Report Methodology for Boiling Water Reactors,'' 
dated April-2007 for preparation of the pressure and temperature 
curves, and incorporates the guidance of TSTF [Technical 
Specification Task Force] -419-A (``Revised PTLR Definition and 
References in ISTS 5.6.6, RCS [reactor coolant system] PTLR''). In 
an NRC Safety Evaluation [safety evaluation] Report dated February 
6, 2007, ``the NRC staff has found that SIR-05-044 is acceptable for 
referencing in licensing applications for General Electric-designed 
boiling water reactors to the extent,'' specified and under, the 
limitations delineated in the TR and in the enclosed final SE.'' As 
part of this change, the Pilgrim Pressure and Temperature Limits 
Report (PTLR) based on the methodology and template provided in SIR-
05-044-A is being supplied for review. The pressure and temperature 
curves utilize the methodology of SIR-05-044-A.
    The NRC has established requirements in Appendix G to 10 CFR 
[Part] 50 in order to protect the integrity of the reactor coolant 
pressure boundary (RCPB) in nuclear power plants. Additionally, the 
regulation in 10 CFR Part 50, Appendix H, provides the NRC staff's 
criteria for the design and implementation of RPV material 
surveillance programs for operating light water reactors. 
Implementing this NRC approved methodology does not reduce the 
ability to protect the reactor coolant pressure boundary as 
specified in Appendix G, nor will this change increase the 
probability of malfunction of plant equipment, or the failure of 
plant structures, systems, or components. Incorporation of the new 
methodology for calculating P-T curves, and the relocation of the P-
T curves from the TS to the PTLR provides an equivalent level of 
assurance that the RCPB is capable of performing its intended safety 
functions. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not affect the assumed accident 
performance of the RCPB, nor any plant structure, system, or 
component previously evaluated. The proposed change does not involve 
the installation of new equipment, and installed equipment is not 
being operated in a new or different manner. The change in 
methodology ensures that the RCPB remains capable of performing its 
safety functions. No set points are being changed which would alter 
the dynamic response of plant equipment. Accordingly, no new failure 
modes are introduced which could introduce 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 does not affect the function of the RCPB or 
its response during plant transients. There are no changes proposed 
which alter the set points at which protective actions are 
initiated, and there is no change to the operability requirements 
for equipment assumed to operate for accident mitigation. This 
change adopts the methodology of SIR-05-044-A, ``Pressure-
Temperature Limits Report Methodology for Boiling Water Reactors,'' 
dated April 2007 for preparation of the pressure and temperature 
curves. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.
    This change adopts the methodology of SIR-05-044-A, ``Pressure-
Temperature

[[Page 17444]]

Limits Report Methodology for Boiling Water Reactors,'' dated April 
2007 for preparation of the pressure and temperature curves, and 
incorporates the guidance of TSTF-419-A (``Revise PTLR Definition 
and References in [Improved Standard Technical Specification] ISTS 
5.6.6, RCS PTLR''). In an NRC Safety Evaluation Report dated 
February 6, 2007, the NRC staff has found that SIR-05-044 is 
acceptable for referencing in licensing applications for General 
Electric-designed boiling water reactors.''

    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.

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

    Date of amendment request: December 3, 2009.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) to incorporate Standard Technical 
Specification 3.1.8 ``Scram Discharge Volume (SDV) Vent and Drain 
Valves'' and associated Bases of NUREG-1433, Revision 3, ``Standard 
Technical Specifications General Electric Plants, BWR/4,'' modified to 
account for plant specific design details.
    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 operation of Vermont Yankee Nuclear Power Station (VY) in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed amendment does not impact the operability of any 
structure, system or component that affects the probability of an 
accident or that supports mitigation of an accident previously 
evaluated. The proposed amendment does not affect reactor operations 
or accident analysis and has no radiological consequences. The 
operability requirements for accident mitigation systems remain 
consistent with the licensing and design basis. Therefore, the 
proposed amendment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The operation of VY in accordance with the proposed amendment 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing plant operation. Thus, this 
change does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. The operation of VY in accordance with the proposed amendment 
will not involve a significant reduction in a margin of safety.
    The proposed change ensures that the safety functions of the SDV 
vent and drain valves are fulfilled. The isolation function is 
maintained by valves in the vent and drain lines and by the required 
action to isolate the affected line. The ability to vent and drain 
the SDVs is maintained through administrative controls. In addition, 
the reactor protection system ensures that an SDV will not be filled 
to the point that it has insufficient volume to accept a full scram. 
Maintaining the safety functions related to isolation of the SDV and 
insertion of control rods ensures that the proposed change does not 
involve a significant reduction in the margin of safety. The 
proposed amendment does not change the design or function of any 
component or system. The proposed amendment does not impact any 
safety limits, safety settings or safety margins. Therefore, 
operation of VY in accordance with the proposed amendment will not 
involve a significant reduction in the margin to safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Nancy Salgado.

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

    Date of amendment request: December 9, 2009.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.8.4, ``DC Sources--Operating,'' 
by removing the Mode restrictions for performance of TS Surveillance 
Requirements (SRs) 3.8.4.7 and 3.8.4.8 for the Division 3 direct 
current (DC) electrical power subsystem battery. These surveillances 
verify that the battery capacity is adequate for the battery to perform 
its required functions. The proposed amendment would remove these Mode 
restrictions for the Division 3 battery, thereby allowing performance 
of SR 3.8.4.7 and SR 3.8.4.8 for the Division 3 battery during Mode 1, 
2, or 3 in conjunction with scheduled high pressure core spray (HPCS) 
system outages. Eliminating the requirement to perform SR 3.8.4.7 and 
SR 3.8.4.8 during Mode 4 or 5 (cold shutdown or refueling conditions) 
will provide greater flexibility in scheduling Division 3 battery 
testing activities by allowing the testing to be performed during non-
outage times.
    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 Division 3 (HPCS) DC electrical power subsystem and its 
associated emergency loads are accident mitigating features, not 
accident initiators. Therefore, the proposed TS changes to allow 
performance of Division 3 battery surveillance testing (service test 
and the battery performance discharge test) in any plant operating 
mode will not significantly impact the probability of any previously 
evaluated accident.
    The design and function of plant equipment is not being modified 
by the proposed amendment. Neither the battery test frequency nor 
the time that the TSs allow the HPCS system to be inoperable are 
being revised. Battery testing in accordance with the proposed TS 
changes will continue to verify that the Division 3 DC electrical 
power subsystem is capable of performing its required function of 
providing DC power to HPCS system equipment, consistent with the 
plant safety analyses. The battery testing period is within the 
period of time that the HPCS system will already be out of service 
for a planned system outage. The battery testing does not increase 
unavailability of the supported HPCS system or represent any change 
in risk above the current practice of planned system maintenance 
outages. Any risk associated with the testing of the Division 3 
battery will be enveloped by the risk management of the HPCS system 
outage. In addition, the HPCS system reliability and availability 
are monitored and evaluated in relationship to Maintenance Rule 
goals to ensure that total outage times do not degrade operational 
safety over time.
    Testing is limited to only one electrical division of equipment 
at a time to ensure that design basis requirements are met. Should a 
fault occur while testing the Division 3 battery, there would be no 
significant impact on any accident consequences since the other two 
divisional DC electrical power subsystems and their associated 
emergency loads would be available to provide the minimum safety 
functions necessary to shut

[[Page 17445]]

down the unit and maintain it in a safe shutdown condition.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    No changes are being made to the plant that would introduce any 
new accident causal mechanisms. Equipment will be operated in the 
same configuration with the exception of the plant operating mode in 
which the Division 3 battery surveillance testing is conducted. 
Performance of these surveillance tests while online will continue 
to verify operability of the Division 3 battery. The proposed 
license amendment does not impact any plant systems that are 
accident initiators and does not adversely impact any accident 
mitigating systems, since the HPCS system will already be out of 
service. The battery testing will not increase the out-of-service 
time for the HPCS system.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Margin of safety is related to confidence in the ability of the 
fission product barriers (fuel cladding, reactor coolant system, and 
primary containment) to perform their design functions during and 
following postulated accidents. The proposed changes to the TS 
surveillance testing requirements for the Division 3 battery do not 
affect the operability requirements for the battery, as verification 
of such operability will continue to be performed as required. 
Continued verification of operability supports the capability of the 
Division 3 DC electrical power subsystem to perform its required 
function of providing DC power to HPCS system equipment, consistent 
with the plant safety analyses. Consequently, the performance of the 
fission product barriers will not be adversely impacted by 
implementation of the proposed amendment. In addition, the proposed 
changes do not alter setpoints or limits established or assumed by 
the accident analysis.
    The battery testing will be performed when the HPCS system is 
already out of service for a planned system outage. The battery 
testing does not increase unavailability of the supported HPCS 
system or represent any change in risk above the current practice of 
planned system maintenance outages, as currently allowed by the TS.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
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: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006.
    NRC Branch Chief: Nancy L. Salgado.

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

    Date of amendment request: December 18, 2009.
    Description of amendment request: The proposed amendment would 
modify Technical Specifications (TS) requirements for unavailable 
barriers by adding limiting condition for operation (LCO) 3.0.9. The 
NRC staff issued a Notice of Opportunity to Comment in the Federal 
Register on June 2, 2006 (71 FR 32145), on possible amendments to 
revise the plant-specific TSs, including a model safety evaluation and 
model no significant hazards consideration determination using the 
consolidated line-item improvement process. The NRC staff subsequently 
issued a Notice of Availability of the models for referencing in 
license amendment applications in the Federal Register on October 3, 
2006 (71 FR 58444). The licensee affirmed the applicability of the 
model no significant hazards consideration determination in its 
application dated December 18, 2009.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated
    The proposed change allows a delay time for entering a supported 
system technical specification (TS) when the inoperability is due 
solely to an unavailable barrier if risk is assessed and managed. The 
postulated initiating events which may require a functional barrier are 
limited to those with low frequencies of occurrence, and the overall TS 
system safety function would still be available for the majority of 
anticipated challenges. Therefore, the probability of an accident 
previously evaluated is not significantly increased, if at all. The 
consequences of an accident while relying on the allowance provided by 
proposed LCO 3.0.9 are no different than the consequences of an 
accident while relying on the TS required actions in effect without the 
allowance provided by proposed LCO 3.0.9. Therefore, the consequences 
of an accident previously evaluated are not significantly affected by 
this change. The addition of a requirement to assess and manage the 
risk introduced by this change will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Previously Evaluated
    The proposed change does not involve a physical alteration of the 
plant (no new or different type of equipment will be installed). 
Allowing delay times for entering supported system TS when 
inoperability is due solely to an unavailable barrier, if risk is 
assessed and managed, will not introduce new failure modes or effects 
and will not, in the absence of other unrelated failures, lead to an 
accident whose consequences exceed the consequences of accidents 
previously evaluated. The addition of a requirement to assess and 
manage the risk introduced by this change will further minimize 
possible concerns. Thus, this change does not create the possibility of 
a new or different kind of accident from an accident previously 
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety
    The proposed change allows a delay time for entering a supported 
system TS when the inoperability is due solely to an unavailable 
barrier, if risk is assessed and managed. The postulated initiating 
events which may require a functional barrier are limited to those with 
low frequencies of occurrence, and the overall TS system safety 
function would still be available for the majority of anticipated 
challenges. The risk impact of the proposed TS changes was assessed 
following the three-tiered approach recommended in RG [Regulatory 
Guide] 1.177. A bounding risk assessment was performed to justify the 
proposed TS changes. This application of LCO 3.0.9 is predicated upon 
the licensee's performance of a risk assessment and the management of 
plant risk. The net change to the margin of safety is insignificant as 
indicated by the anticipated low levels of associated risk (ICCDP 
[Incremental Conditional Core Damage Probability] and ICLERP 
[Incremental Conditional Large Early Release Probability]) as shown in 
Table 1 of Section 3.1.1 in the Safety Evaluation published in the 
Federal Register on October 3, 2006. Therefore, this change does not 
involve a

[[Page 17446]]

significant reduction in a margin of safety.
    The NRC staff has reviewed the 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: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006.
    NRC Branch Chief: Nancy L. Salgado.

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

    Date of amendment request: October 27, 2009.
    Description of amendment request: The proposed amendment would 
adopt the Alternative Source Term (AST) methodology, in addition to 
Technical Specification (TS) changes supported by the AST design basis 
accident radiological consequences analyses. The proposed amendment 
would also incorporate Technical Specification Task Force (TSTF)-490, 
``Deletion of E-Bar Definition and Revision to RCS [reactor coolant 
system] Specific Activity Tech Spec,'' Revision 0.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    With this change, Prairie Island Nuclear Generating Plant 
(PINGP) proposes to implement 10 CFR 50.67, alternative source term 
methodologies, implement approved industry improved Standard 
Technical Specification traveler, TSTF-490, and revise TS 3.3.7, 
``Spent Fuel Pool Special Ventilation System Actuation 
Instrumentation,'' TS 3.7.12, ``Auxiliary Building Special 
Ventilation System,''
    TS 3.7.13, ``Spent Fuel Pool Special Ventilation System,'' TS 
3.9.4, ``Containment Penetrations,'' TS 5.5.9, ``Ventilation Filter 
Testing Program,'' TS 5.5.14, ``Containment Leakage Rate Testing 
Program,'' and TS 5.5.16, ``Control Room Habitability Program.''
    Alternative source term (AST) calculations have been performed 
for PINGP that demonstrate the dose consequences are consistent with 
the regulatory limits of 10 CFR 50.67 and the guidance of Regulatory 
Guide (RG) 1.183. The use of the AST methodology changes the 
regulatory assumptions regarding the analytical treatment of the 
design basis accidents and has no direct effect on the probability 
of any accident. AST methods have been utilized in the analysis of 
the limiting design basis accidents, as follows: loss of coolant 
accident, fuel handling accident, main steam line break, steam 
generator tube rupture, control rod ejection accident, and locked 
rotor accident. The results of the analyses, which include the 
proposed changes to the Technical Specifications, demonstrate that 
the dose consequences of these limiting events are within regulatory 
limits.
    Reactor coolant specific activity is not an initiator for any 
accident previously evaluated. The Completion Time when reactor 
coolant gross activity is not within limit is not an initiator for 
any accident previously evaluated. The current variable limit on 
primary coolant iodine concentration is not an initiator to any 
accident previously evaluated. As a result, the proposed change does 
not significantly increase the probability of an accident. The 
proposed change will limit reactor coolant noble gases to 
concentrations consistent with the accident analyses. The proposed 
change to the Completion Time has no impact on the consequences of 
any design basis accident since the consequences of an accident 
during the extended Completion Time are the same as the consequences 
of an accident during the current Completion Time. As a result, the 
consequences of any accident previously evaluated are not 
significantly increased.
    The Spent Fuel Pool Special Ventilation System is no longer 
credited for filtration or isolation. The Containment Penetrations 
TS is being replaced with a TS on Decay Time, which requires that 
recently irradiated fuel (<50 hours) cannot be handled. The 
Ventilation Filter Testing Program TS is being revised to reflect 
changes to filter testing. As a result of these TS changes, the 
probability or consequences of an accident previously evaluated are 
not significantly increased.
    Based on the above, the proposed changes do 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.
    With this change, PINGP proposes to implement 10 CFR 50.67, 
alternative source term methodologies, implement approved industry 
improved Standard Technical Specification traveler, TSTF-490, and 
revise TS 3.3.7, ``Spent Fuel Pool Special Ventilation System 
Actuation Instrumentation,'' TS 3.7.12, ``Auxiliary Building Special 
Ventilation System,'' TS 3.7.13, ``Spent Fuel Pool Special 
Ventilation System,'' TS 3.9.4, ``Containment Penetrations,'' TS 
5.5.9, ``Ventilation Filter Testing Program,'' TS 5.5.14, 
``Containment Leakage Rate Testing Program,'' and TS 5.5.16, 
``Control Room Habitability Program.''
    The AST methodology is not an accident initiator, as it is a 
method used to estimate resulting accident doses. The proposed 
operation of plant systems affected by this change does not create 
the possibility of a new or different kind of accident previously 
evaluated. Changes that are proposed to plant equipment (ventilation 
systems) pertain to accident mitigation equipment. The operation or 
mis-operation of these ventilation systems do not initiate any 
accidents. The radiological consequence analyses demonstrate that 
the proposed changes are acceptable. The results of the analyses, 
which include the proposed changes to the Technical Specifications, 
demonstrate that the dose consequences of these limiting events are 
within regulatory limits.
    The proposed change in specific activity limits does not alter 
any physical part of the plant nor does it affect any plant 
operating parameter. The change does not create the potential of a 
new or different kind of accident from any accident previously 
evaluated.
    Based on the above, the proposed changes do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    With this change, PINGP proposes to implement 10 CFR 50.67, 
alternative source term methodologies, implement approved industry 
improved Standard Technical Specification traveler, TSTF-490, and 
revise TS 3.3.7, ``Spent Fuel Pool Special Ventilation System 
Actuation Instrumentation,'' TS 3.7.12, ``Auxiliary Building Special 
Ventilation System,'' TS 3.7.13, ``Spent Fuel Pool Special 
Ventilation System,'' TS 3.9.4, ``Containment Penetrations,'' TS 
5.5.9, ``Ventilation Filter Testing Program,'' TS 5.5.14, 
``Containment Leakage Rate Testing Program,'' and TS 5.5.16, 
``Control Room Habitability Program.''
    The proposed implementation of the AST methodology is consistent 
with RG 1.183. The radiological consequences of these accidents are 
within the regulatory acceptance criteria associated with the use of 
the AST methodology. The doses at the exclusion area and low 
population zone boundaries and in the control room are consistent 
with the regulatory limits of 10 CFR 50.67 and the guidance of RG 
1.183. The margin of safety for the radiological consequences of 
these accidents is considered to be that provided by meeting the 
applicable regulatory limits, which are set at or below 10 CFR 50.67 
limits.
    The proposed change to revise the limits on noble gas 
radioactivity in the primary coolant is consistent with the 
assumptions in the safety analyses and will ensure the monitored 
values protect the initial assumptions in the safety analyses.
    Based on the above, 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: Peter M. Glass, Assistant General Counsel, 
Xcel Energy

[[Page 17447]]

Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: Robert J. Pascarelli.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: January 29, 2010.
    Description of amendment request: The amendments would change an 
Emergency Action Level (EAL) scheme based on NUREG-0654, ``Criteria for 
Preparation and Evaluation of Radiological Emergency Response Plan and 
Preparedness in Support of Nuclear Power Plants,'' to one based on NEI 
99-01, ``Methodology for Development of Emergency Action Levels,'' 
Revision 4. This would change the methodology for deriving selected 
Notification of Unusual Event values in Table R-1, Gaseous Effluent 
Monitor Classification Thresholds, and deleting EAL RA2.4 which 
evaluates abnormal radiation readings at infrequently accessed areas 
and revise the radiation level threshold values for Reactor Coolant 
System (RCS) letdown indication.
    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:
    Does the proposed amendment involve a significant increase in 
the probability or Consequences of an accident previously evaluated?
    Response: No.
    These changes affect the North Anna [* * *] Power Station 
Emergency Action Levels, but do not alter any of the requirements of 
the Operating License or the Technical Specifications. The proposed 
changes do not modify any plant equipment and do not impact any 
failure modes that could lead to an accident. Additionally, the 
proposed changes have no effect on the consequences of any analyzed 
accident since the changes do not affect any equipment related to 
accident mitigation. Based on this discussion, the proposed 
amendment does not increase the probability or consequence of an 
accident previously evaluated.
    Criterion 2:
    Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    These changes affect the North Anna [* * *] Power Station 
Emergency Action Levels, but do not alter any of the requirements of 
the Operating License or the Technical Specifications. They do not 
modify any plant equipment and there is no impact on the capability 
of the existing equipment to perform their intended functions. No 
system setpoints are being modified. No new failure modes are 
introduced by the proposed changes. The proposed amendment does not 
introduce accident initiator or malfunctions that would cause a new 
or different kind of accident. Therefore, the proposed amendment 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Criterion 3:
    Does the proposed amendment involve a significant reduction in a 
margin of safety?
    Response: No.
    These changes affect the North Anna [* * *] Power Station 
Emergency Action Levels, but do not alter any of the requirements of 
the Operating License or the Technical Specifications. The proposed 
changes do not affect any of the assumptions used in the accident 
analysis, nor do they affect any operability requirements for 
equipment important to plant safety. Therefore, the proposed changes 
will not result in a significant reduction in the margin of safety 
as defined in the bases for technical specifications covered in this 
license amendment request. [Therefore, this 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 
23219.
    NRC Branch Chief: Gloria Kulesa.

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: January 27, 2010.
    Description of amendment request: The proposed license amendment 
request would increase each unit's rated power (RP) level from 2546 
megawatts thermal (MWt) to 2587 MWt, and make Technical Specifications 
changes as necessary to support operation at the uprated power level. 
The proposed change is an increase in RP of approximately 1.6%.
    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 consequence of an accident previously evaluated?
    Response: No.
    The proposed change will increase the Surry Power Station (SPS) 
Units 1 and 2 rated power (RP) from 2546 megawatts thermal (MWt) to 
2587 MWt. Nuclear steam supply system and balance-of-plant systems, 
components and analyses that could be affected by the proposed 
change to the RP were evaluated using revised design parameters. The 
evaluations determined that these structures, systems and components 
are capable of performing their design function at the proposed 
uprated RP of 2587 MWt. An evaluation of the accident analyses 
demonstrates that the applicable analysis acceptance criteria are 
still met with the proposed changes. Power level is an input 
assumption to equipment design and accident analyses, but it is not 
a transient or accident initiator. Accident initiators are not 
affected by the power uprate, and plant safety barrier challenges 
are not created by the proposed changes.
    The radiological consequences of operation at the uprated power 
conditions have been assessed. The proposed change to RP does not 
affect release paths, frequency of release, or the analyzed reactor 
core fission product inventory for any accidents previously 
evaluated in the SPS Updated Final Safety Analysis Report. There is 
a small increase in the reactor coolant activity concentration. 
Structures, systems and components required to mitigate transients 
are capable of performing their design functions with the proposed 
changes, and are thus acceptable. Analyses performed to assess the 
effects of mass and energy releases remain valid. The assessment of 
radiological consequences for operation at the proposed power level 
confirmed that there is not a significant increase for affected 
events.
    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.
    No new accident scenarios, failure mechanisms, or single 
failures are introduced as a result of any proposed changes. The 
ultrasonic flow meter (UFM) being installed to facilitate the 
Measurement Uncertainty Recapture (MUR) power uprate has been 
analyzed, and system failures will not adversely affect any safety-
related system or any structures, systems or components required for 
transient mitigation. Structures, systems and components previously 
required for transient mitigation are still capable of fulfilling 
their intended design functions. The proposed changes have no 
significant adverse affect on any safety-related structures, systems 
or components and do not significantly change the performance or 
integrity of any safety-related system.
    The proposed changes do not adversely affect any current system 
interfaces or create any new interfaces that could result in an 
accident or malfunction of a different kind than previously 
evaluated. Operating at an RP of 2587 MWt does not create any new 
accident initiators or precursors. Credible

[[Page 17448]]

malfunctions are bounded by the current accident analyses of record 
or recent evaluations demonstrating that applicable criteria are 
still met with the proposed changes.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margins of safety associated with the power uprate are those 
pertaining to core thermal power. These include fuel cladding, 
reactor coolant system pressure boundary, and containment barriers. 
Core analyses demonstrate that power uprate implementation will 
continue to meet the current nuclear design basis. Impacts to 
components associated with the reactor coolant system pressure 
boundary structural integrity, and factors such as pressure-
temperature limits, vessel fluence, and pressurized thermal shock 
were determined to be bounded by the current analyses.
    Systems will continue to operate within their design parameters 
and remain capable of performing their intended safety functions 
following implementation of the proposed change. The current SPS 
safety analyses, and the revised design basis radiological accident 
dose calculations, bound the power uprate without significantly 
impacting margins.
    Therefore, this 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: Lillian M. Cuoco, Esq., Senior Counsel, 
Dominion Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 
23219.
    NRC Branch Chief: Gloria Kulesa.

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: January 29, 2010.
    Description of amendment request: The amendments would change an 
Emergency Action Level (EAL) scheme based on NUREG-0654, ``Criteria for 
Preparation and Evaluation of Radiological Emergency Response Plan and 
Preparedness in Support of Nuclear Power Plants,'' to one based on NEI 
99-01, ``Methodology for Development of Emergency Action Levels,'' 
Revision 4. This would change the methodology for deriving selected 
Notification of Unusual Event values in Table R-1, Gaseous Effluent 
Monitor Classification Thresholds, and deleting EAL RA2.4 which 
evaluates abnormal radiation readings at infrequently accessed areas.
    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:
    Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    These changes affect the [* * *] Surry Power Station Emergency 
Action Levels, but do not alter any of the requirements of the 
Operating License or the Technical Specifications. The proposed 
changes do not modify any plant equipment and do not impact any 
failure modes that could lead to an accident. Additionally, the 
proposed changes have no effect on the consequences of any analyzed 
accident since the changes do not affect any equipment related to 
accident mitigation. Based on this discussion, the proposed 
amendment does not increase the probability or consequence of an 
accident previously evaluated.
    Criterion 2:
    Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    These changes affect the [* * *] Surry Power Station Emergency 
Action Levels, but do not alter any of the requirements of the 
Operating License or the Technical Specifications. They do not 
modify any plant equipment and there is no impact on the capability 
of the existing equipment to perform their intended functions. No 
system setpoints are being modified. No new failure modes are 
introduced by the proposed changes. The proposed amendment does not 
introduce accident initiator or malfunctions that would cause a new 
or different kind of accident. Therefore, the proposed amendment 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Criterion 3:
    Does the proposed amendment involve a significant reduction in a 
margin of safety?
    Response: No.
    These changes affect [* * *] the Surry Power Station Emergency 
Action Levels, but do not alter any of the requirements of the 
Operating License or the Technical Specifications. The proposed 
changes do not affect any of the assumptions used in the accident 
analysis, nor do they affect any operability requirements for 
equipment important to plant safety. Therefore, the proposed changes 
will not result in a significant reduction in the margin of safety 
as defined in the bases for technical specifications covered in this 
license amendment request. [Therefore, this 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 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.

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

    Date of amendment request: December 16, 2009.
    Description of amendment request: The proposed changes would revise 
Technical Specification (TS) 3.8.4, ``DC [Direct Current] Sources--
Operating,'' Surveillance Requirement (SR) 3.8.4.2 and SR 3.8.4.5 to 
revise the battery connection resistance acceptance criteria for inter-
cell connections from <= 150E-6 ohms to <= 33E-6 ohms and would add 
connection resistance acceptance criteria for inter-tier connections 
and inter-bank connection of <= 150E-6 ohms.
    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 changes to revise the SR 3.8.4.2 and SR 3.8.4.5 
acceptance criteria for battery connection resistance will not 
challenge the ability of the safety-related batteries to perform 
their safety function. Appropriate monitoring and maintenance will 
continue to be performed on the safety related batteries. Current TS 
testing and monitoring requirements will not be altered.
    The proposed change does not involve a physical change to the 
batteries, nor does it change the safety function of the batteries. 
The proposed TS revision involves no significant changes to the 
operation of any systems or components in normal and accident 
operating conditions and no changes to existing structures, systems 
or components.
    Therefore, this change will not increase the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any previously evaluated?
    Response: No
    The proposed changes to revise the SR 3.8.4.2 and SR 3.8.4.5 
acceptance criteria for battery connection resistance is an increase 
in conservatism, without a change in system

[[Page 17449]]

testing methods, operation, or control. Safety related batteries 
installed in the plant will be required to meet criteria more 
restrictive and conservative than current acceptance criteria and 
standards. The proposed change does not affect the manner in which 
the batteries are tested and maintained, thus there are no new 
failure mechanisms for the system.
    Therefore, this change will 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 margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The proposed changes will not adversely affect 
operation of plant equipment, as the changes being made are more 
restrictive. These changes will not result in a change to the 
setpoints at which protective actions are initiated. Sufficient DC 
capacity to support operation of mitigation equipment is ensured. 
The changes associated with the new battery maintenance and 
monitoring program will ensure that the station batteries are 
maintained in a highly reliable manner. The equipment fed by the DC 
electrical sources will continue to provide adequate power to safety 
related loads in accordance with analysis assumptions.
    Therefore, this 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: Jay Silberg, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, N.W., Washington, DC 20037.
    NRC Branch Chief: Michael T. Markley.

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.

Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414, 
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of amendment request: October 2, 2008.
    Brief description of amendment request: The proposed amendment 
would revise the Technical Specifications (TS) associated with the 
verification of ice condenser door operability and TS surveillance 
requirements 3.6.13.5 and 3.6.13.6.
    Date of publication of individual notice in Federal Register: March 
8, 2010 (75 FR 10513).
    Expiration date of individual notice: Comments April 7, 2010; 
Hearing May 7, 2010.

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: October 2, 2008.
    Brief description of amendment request: The proposed amendment 
would revise the Technical Specifications (TS) associated with the 
verification of ice condenser door operability and TS surveillance 
requirements 3.6.13.5 and 3.6.13.6.
    Date of publication of individual notice in Federal Register: March 
8, 2010 (75 FR 10508).
    Expiration date of individual notice: Comments April 7, 2010; 
Hearing May 7, 2010.

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, 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 email to [email protected].

Carolina Power and Light Company, Docket No. 50-261, H. B. Robinson 
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of application for amendment: June 19, 2009, as supplemented 
by letter dated October 20, 2009.
    Brief description of amendment: The proposed amendment would revise 
Technical Specification 3.3.1, ``Reactor Protection System 
Instrumentation.'' The proposed change revises the requirements related 
to the reactor protection system interlock for the turbine trip input 
to the reactor protection system.
    Date of issuance: March 17, 2010.
    Effective date: Effective as of the date of issuance and shall be 
implemented by the end of Refueling Outage 26.
    Amendment No.: 222.
    Renewed Facility Operating License No. DPR-23: The amendment 
revises the technical specifications.
    Date of initial notice in Federal Register: January 5, 2010 (75 FR 
460).
    The Commission's related evaluation of the amendment is contained 
in a safety evaluation dated March 17, 2010.

[[Page 17450]]

    Public comments received as to proposed no significant hazards 
consideration (NSHC): No.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1 (ANO-1), Pope County, Arkansas

    Date of amendment request: March 13, 2008, as supplemented by 
letter dated February 28, 2010.
    Brief description of amendment: The amendment replaced the current 
ANO-1 Technical Specification 3.4.12, ``RCS [Reactor Coolant System] 
Specific Activity,'' limit on RCS gross specific activity with a new 
limit on RCS noble gas specific activity. The noble gas specific 
activity limit would be based on a new dose equivalent Xe-133 
definition that would replace the current E Bar average disintegration 
energy definition. In addition, the current dose equivalent I-131 
definition would be revised to allow the use of additional thyroid dose 
conversion factors.
    Date of issuance: March 18, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 243.
    Renewed Facility Operating License No. DPR-51: Amendment revised 
the Technical Specifications/license.
    Date of initial notice in Federal Register: May 6, 2008 (73 FR 
25038). The supplemental letter dated February 28, 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 as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 18, 2010.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 2, 2009, as supplemented 
by letter dated June 24, 2009.
    Brief description of amendment: The amendment modified Technical 
Specification (TS) 3.3.1.1, ``Reactor Protective Instrumentation,'' and 
TS 3.3.2.1, ``Engineered Safety Feature Actuation System 
Instrumentation,'' specifically, Table 3.3-1, Table 4.3-1, and Table 
3.3-3, to adopt a mode of applicability for the Logarithmic Power 
Level--High, Pressurizer Pressure--Low, Steam Generator [SG] Pressure--
Low, and the SG Differential Pressure and Level Low functions. These 
changes are consistent with NUREG-1432, Revision 3.0, ``Standard 
Technical Specifications, Combustion Engineering Plants,'' dated June 
2004.
    Date of issuance: March 11, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 289.
    Renewed Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications/license.
    Date of initial notice in Federal Register: June 2, 2009 (74 FR 
26433). The supplemental letter dated June 24, 2009, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register on June 2, 2009 (74 
FR 26433).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 11, 2010.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 19, 2009.
    Brief description of amendment: The amendment relocated the 
Waterford 3 Steam Generator Level--High trip requirements from 
Technical Specification Sections 2.2 and 3/4.3.1 to the Technical 
Requirements Manual (TRM). This change is consistent with Technical 
Specification Task Force (TSTF) 410-A, ``Relocation of Steam Generator 
Level--High Trip to the TRM,'' and Revision 3 of NUREG-1432, ``Standard 
Technical Specifications, Combustion Engineering Plants.''
    Date of issuance: March 18, 2010.
    Effective date: As of the date of issuance and shall be implemented 
90 days from the date of issuance.
    Amendment No.: 225.
    Facility Operating License No. NPF-38: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: December 1, 2009 (74 FR 
62834).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 18, 2010.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2 (Braidwood), Will County, Illinois

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

    Date of application for amendment: December 4, 2008, as 
supplemented by letters dated February 17, 2009; July 27, 2009; 
December 4, 2009; and January 29, 2010.
    Brief description of amendment: The amendments revise Technical 
Specifications (TSs) 1.1, ``Definitions,'' and 3.4.16, ``RCS [Reactor 
Coolant System] Specific Activity,'' and Surveillance Requirements 
3.4.16.1, 3.4.16.2, and 3.4.16.3. The revisions replace the current TS 
3.4.16 limit on RCS gross specific activity with a new limit on RCS 
noble gas-specific activity. The revisions adopt TS Task Force (TSTF) 
Change Traveler, TSTF-490, ``Deletion of E Bar Definition and Revision 
to RCS Specific Activity Tech Spec [sic],'' Revision 0.
    Date of issuance: March 23, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: Braidwood Unit 1-162; Braidwood Unit 2-162; Byron 
Unit No. 1-167; and Byron Unit No. 2-167.
    Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66: 
The amendments revise the TSs and Licenses.
    Date of initial notice in Federal Register: January 27, 2009 (74 FR 
4771).
    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 March 23, 2010.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: March 26, 2009, as supplemented 
by letter dated October 28, 2009.
    Brief description of amendments: The proposed changes would revise 
Technical Specification 3.5.1, ``Emergency Core Cooling Systems (ECCS) 
Operating,'' to delete the existing allowance with the Automatic 
Depressurization System accumulator backup compressed gas system that

[[Page 17451]]

currently allows a completion time of 72 hours to restore bottle 
pressure to >= 500 psig.
    Date of issuance: March 19, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 196/183.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications and License.
    Date of initial notice in Federal Register: September 8, 2009 (74 
FR 46242). The October 28, 2009 supplement, contained clarifying 
information and did not change the NRC staff's initial proposed finding 
of no significant hazards consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 19, 2010.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

    Date of application for amendment: November 6, 2008, supplemented 
by letters dated December 11, 2008, July 2, 2009, October 2, 2009, and 
November 24, 2009.
    Brief description of amendment: The amendment replaces the current 
TMI-1 technical specification limit on Reactor Coolant System (RCS) 
gross specific activity with a new limit on RCS noble gas specific 
activity. The noble gas specific activity limit is based on a new dose 
equivalent Xenon-133 definition that replaces the previous E-Bar 
average disintegration energy definition. In addition, the dose 
equivalent Iodine-131 definition has been revised.
    Date of issuance: March 11, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 272.
    Renewed Facility Operating License No. DPR-50. Amendment revised 
the license and the technical specifications.
    Date of initial notice in Federal Register: March 10, 2009 (74 FR 
10309). The supplements dated December 11, 2008, July 2, 2009, October 
2, 2009, and November 24, 2009, 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 determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 11, 2010.
    No significant hazards consideration comments received: No.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: March 11, 2009, as supplemented by 
letters dated August 12 and December 21, 2009, and March 5, 2010.
    Brief description of amendment: The amendment revised Surveillance 
Requirements 3.8.4.2 and 3.8.4.5 in Technical Specification Section 
3.8.4, ``DC [Direct Current] Sources--Operating,'' by adding a 
parameter of total battery resistance to the values of battery 
connection resistance.
    Date of issuance: March 18, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days of issuance.
    Amendment No.: 236.
    Facility Operating License No. DPR-46: Amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: May 5, 2009 (74 FR 
20752). The supplemental letters dated August 12 and December 21, 2009, 
and March 5, 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 as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 18, 2010.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 30, 2009, as supplemented 
on November 2, 2009.
    Brief description of amendment: The amendment modifies the NMP2 
Technical Specification (TS) 3.8.1, ``AC Sources--Operating,'' to 
remove operating mode restrictions for the performance of certain 
Surveillance Requirements (SRs) pertaining to the Division 3, High 
Pressure Core Spray (HPCS) Emergency Diesel Generator (DG). The testing 
in Modes 1 or 2 were previously prohibited in SR 3.8.1.7, SR 3.8.1.8, 
and SR 3.8.1.10, and in Modes 1, 2, or 3 in SR 3.8.1.9, SR 3.8.1.11, SR 
3.8.1.14, SR 3.8.1.15, and SR 3.8.1.17. The amendment removes these 
Mode restrictions and allows the above SRs to be performed in any 
operating mode for the Division 3 DG. The Mode restrictions remain 
applicable to the other two safety-related (Division 1 and Division 2) 
DGs.
    Date of issuance: March 18, 2010.
    Effective date: As of the date of issuance to be implemented within 
90 days.
    Amendment No.: 133.
    Renewed Facility Operating License No. NPF-069: The amendment 
revises the License and TSs.
    Date of initial notice in Federal Register: June 16, 2009 (74 FR 
28577).
    The supplemental letter dated November 2, 2009, 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 staff's initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 18, 2010.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, 
Alabama

    Date of application for amendments: July 27, 2009.
    Description of amendment request: The amendments revised the 
Technical Specifications to change Surveillance Requirement 3.6.1.3, 
``Primary Containment Isolation Valves,'' to eliminate unnecessary 
local leak rate tests.
    Date of issuance: March 22, 2010.
    Effective date: Date of issuance, to be implemented within 60 days.
    Amendment Nos.: 277, 304, and 263.
    Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68: 
Amendments revised the Operating License and Technical Specifications.
    Date of initial notice in Federal Register: October 20, 2009 (74 FR 
53781).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 22, 2010.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 20, 2009, as supplemented 
by letters dated December 10, 2009, and January 19, 2010.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 5.5.16, ``Containment Leakage Rate Testing 
Program.'' The revision reflects a one-time extension of

[[Page 17452]]

the current containment Type A leak rate test (integrated leak rate 
test or ILRT) interval requirement of Title 10 of the Code of Federal 
Regulations (10 CFR) Part 50, Appendix J, ``Primary Reactor Containment 
Leakage Testing for Water-Cooled Power Reactors,'' Option B, 
``Performance Based Requirements,'' from 10 years to 15 years. The 
amendment allows the next ILRT to be performed no later than October 
25, 2014.
    Date of issuance: March 17, 2010.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 195.
    Facility Operating License No. NPF-30: The amendment revised the 
Operating License and Technical Specifications.
    Date of initial notice in Federal Register: August 25, 2009 (74 FR 
42931). The supplemental letters dated December 10, 2009, and January 
19, 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 as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 17, 2010.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 25th day of March 2010.

    For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2010-7451 Filed 4-5-10; 8:45 am]
BILLING CODE 7590-01-P