[Federal Register Volume 76, Number 75 (Tuesday, April 19, 2011)]
[Notices]
[Pages 21917-21928]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2011-9177]


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NUCLEAR REGULATORY COMMISSION

[NRC-2011-0082]


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 24, 2011, to April 6, 2011. The last 
biweekly notice was published on April 5, 2011 (76 FR 18801).

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.

[[Page 21918]]

Under the Commission's regulations in Title 10 of the Code of Federal 
Regulations (10 CFR), Sec.  50.92, this means that operation of the 
facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules, 
Announcements and Directives Branch (RADB), TWB-05-B01M, Division of 
Administrative Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be faxed to the RADB at 301-492-3446. 
Documents may be examined, and/or copied for a fee, at the NRC's Public 
Document Room (PDR), located at One White Flint North, Room O1-F21, 
11555 Rockville Pike (first floor), Rockville, Maryland 20852.
    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license. 
Requests for a hearing and a petition for leave to intervene shall be 
filed in accordance with the Commission's ``Rules of Practice for 
Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested person(s) 
should consult a current copy of 10 CFR 2.309, which is available at 
the Commission's PDR, located at One White Flint North, Room O1-F21, 
11555 Rockville Pike (first floor), Rockville, Maryland 20852. Publicly 
available records will be accessible from the Agencywide Documents 
Access and Management System's (ADAMS) Public Electronic Reading Room 
on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to 
intervene is filed by the above date, the Commission or a presiding 
officer designated by the Commission or by the Chief Administrative 
Judge of the Atomic Safety and Licensing Board Panel, will rule on the 
request and/or petition; and the Secretary or the Chief Administrative 
Judge of the Atomic Safety and Licensing Board will issue a notice of a 
hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held. If 
the final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards consideration, then any hearing 
held would take place before the issuance of any amendment.
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC E-Filing rule (72 FR 49139, 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
ten

[[Page 21919]]

(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 identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a request or 
petition for hearing (even in instances in which the participant, or 
its counsel or representative, already holds an NRC-issued digital ID 
certificate). Based upon this information, the Secretary will establish 
an electronic docket for the hearing in this proceeding if the 
Secretary has not already established an electronic docket.
    Information about applying for a digital ID certificate is 
available on 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 the Electronic Information Exchange System, 
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 1-866-672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, 
Maryland, 20852, Attention: Rulemaking and Adjudications Staff. 
Participants filing a document in this manner are responsible for 
serving the document on all other participants. Filing is considered 
complete by first-class mail as of the time of deposit in the mail, or 
by courier, express mail, or expedited delivery service upon depositing 
the document with the provider of the service. A presiding officer, 
having granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
http://ehd1.nrc.gov/EHD/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. With 
respect to copyrighted works, except for limited excerpts that serve 
the purpose of the adjudicatory filings and would constitute a Fair Use 
application, participants are requested not to include copyrighted 
materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Non-timely filings 
will not be entertained absent a determination by the presiding officer 
that the petition or request should be granted or the contentions 
should be admitted, based on a balancing of the factors specified in 10 
CFR 2.309(c)(1)(i)-(viii).
    For further details with respect to this license amendment 
application, see the application for amendment which is available for 
public inspection at the Commission's PDR, located at One White Flint 
North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland 20852. Publicly available records will be accessible from the 
ADAMS Public Electronic Reading Room on the Internet at the NRC Web 
site, http://www.nrc.gov/reading-rm/adams.html. Persons who do not have 
access to ADAMS or who encounter problems in accessing the documents 
located in ADAMS, should contact the NRC PDR Reference staff at 1-800-
397-4209, 301-415-4737, or by e-mail to [email protected].

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

    Date of amendments request: October 25, 2010.
    Description of amendments request: The amendment would revise 
Technical Specification Limiting Condition for Operation (LCO) 3.0.5 to 
provide clarification as to when the LCO can be invoked in order to 
perform required testing to demonstrate OPERABILITY of equipment.

[[Page 21920]]

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to LCO 3.0.5 more clearly specifies the 
situations when LCO 3.0.5 can be applied. In some Technical 
Specifications, the steps taken to comply with ACTIONS involve the 
placement of redundant or alternate equipment or trains into 
service, or the repositioning (e.g., opening or closing) or 
components. The proposed change would allow the use of LCO 3.0.5 in 
situations such as these. This proposed change does not, however, 
change the intent of LCO 3.0.5. The purpose of LCO 3.0.5 remains to 
provide an exception to LCO 3.0.2, to not comply with the applicable 
Required Action(s) while performing required testing to demonstrate 
the OPERABILITY of either equipment being returned to service or the 
OPERABILITY of other equipment.
    The proposed change does not affect any analyzed accident 
initiators, nor does it change the units' ability to successfully 
respond to any previously evaluated accident. As a result, there is 
also no change to existing radiological assumptions used in the 
accident evaluations. In addition this proposed change does not 
change the operation or maintenance performed on operating 
equipment.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change to LCO 3.0.5 more clearly specifies the 
situations when LCO 3.0.5 can be applied. In some Technical 
Specifications, the steps taken to comply with ACTIONS involve the 
placement of redundant or alternate equipment or trains into 
service, or the repositioning (e.g., opening or closing) or 
components. The proposed change would allow the use of LCO 3.0.5 in 
situations such as these. This proposed change does not, however, 
change the intent of LCO 3.0.5. The purpose of LCO 3.0.5 remains to 
provide an exception to LCO 3.0.2, to not comply with the applicable 
Required Action(s) while performing required testing to demonstrate 
the OPERABILITY of either equipment being returned to service or the 
OPERABILITY of other equipment.
    The proposed change does not involve a modification to the 
physical configuration of the units nor does it involve any change 
in the methods governing normal plant operation. The proposed change 
does not impose any new or different requirements or introduce a new 
accident initiator, accident precursor, or malfunction mechanism. 
Additionally there is no change in the types or increase in the 
amounts of any effluent that may be released offsite and there is no 
increase in individual or cumulative occupational exposure.
    Therefore the proposed change does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change to LCO 3.0.5 more clearly specifies the 
situations when LCO 3.0.5 can be applied. In some Technical 
Specifications, the steps taken to comply with ACTIONS involve the 
placement of redundant or alternate equipment or trains into 
service, or the repositioning (e.g., opening or closing) or 
components. The proposed change would allow the use of LCO 3.0.5 in 
situations such as these. This proposed change does not, however, 
change the intent of LCO 3.0.5. The purpose of LCO 3.0.5 remains to 
provide an exception to LCO 3.0.2, to not comply with the applicable 
Required Action(s) while performing required testing to demonstrate 
the OPERABILITY of either equipment being returned to service or the 
OPERABILITY of other equipment.
    The proposed change does not involve any modification to the 
physical configuration of the operating units and does not alter 
equipment operation. As such, the safety functions of plant 
equipment and their response to any analyzed accident scenario are 
unaffected by this proposed change and thus there is no reduction in 
the margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety for the operation of each unit.

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

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

    Date of amendments request: March 22, 2011.
    Description of amendments request: The proposed amendment would 
revise the Technical Specifications (TSs) to define a new time limit 
for restoring inoperable reactor coolant system (RCS) leakage detection 
instrumentation to operable status. The proposed TS changes are 
consistent with TS Task Force (TSTF)-513, ``Revise PWR [pressurized-
water reactor] Operability Requirements and Actions for RCS Leakage 
Instrumentation.''
    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 clarifies the operability requirements for 
the RCS leakage detection instrumentation and reduces the time 
allowed for the plant to operate when the only TS required operable 
RCS leakage detection instrumentation monitor is the containment 
atmosphere gaseous radiation monitor. The monitoring of RCS leakage 
is not a precursor to any accident previously evaluated. The 
monitoring of RCS leakage is not used to mitigate the consequences 
of any accident previously evaluated. The plant specific variation 
to this license amendment request, to insert the Note ``Not required 
until 12 hours after establishment of steady state operation'' into 
applicable portions of the Technical Specification is administrative 
in nature. As a result, its inclusion does not impact any plant 
equipment's ability to perform its required functions. Therefore, it 
is concluded that the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change clarifies the operability requirements for 
the RCS leakage detection instrumentation and reduces the time 
allowed for the plant to operate when the only TS required operable 
RCS leakage detection instrumentation monitor is the containment 
atmosphere gaseous radiation monitor. 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 
normal plant operation. The proposed change maintains sufficient 
continuity and diversity of leak detection capability that the 
probability of piping evaluated and approved for leak-before-break 
progressing to pipe rupture remains extremely low. The plant 
specific variation to this license amendment request, to insert the 
Note ``Not required until 12 hours after establishment of steady 
state operation'' into applicable portions of the Technical 
Specification also does not involve a physical alteration of the 
plant or change in how plant equipment is operated. Therefore, it is 
concluded that the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.

[[Page 21921]]

    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The proposed change clarifies the operability requirements for 
the RCS leakage detection instrumentation and reduces the time 
allowed for the plant to operate when the only TS required operable 
RCS leakage detection instrumentation monitor is the containment 
atmosphere gaseous radiation monitor. Reducing the amount of time 
the plant is allowed to operate with only the containment atmosphere 
gaseous radiation monitor operable increases the margin of safety by 
increasing the likelihood that an increase in RCS leakage will be 
detected before it potentially results in gross failure. The plant 
specific variation to this license amendment request, to insert the 
Note ``Not required until 12 hours after establishment of steady 
state operation'' into applicable portions of the Technical 
Specification provides clarification as it reflects the time 
necessary for plant conditions to stabilize in order to ensure an 
accurate water inventory can be obtained.
    Therefore, it is concluded that the proposed changes do not 
involve a significant reduction in a margin of safety. The NRC staff 
has reviewed the licensee's analysis and, based on this review, it 
appears that the three standards of 10 CFR 50.92(c) are satisfied. 
Therefore, the NRC staff proposes to determine that the amendments 
request involves no significant hazards consideration.

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

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

    Date of amendment request: October 20, 2010.
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TS) description of fuel assemblies 
specified in TS 4.2.1. Additionally, changes are requested to the 
analytical methods referenced in TS 5.6.5.b. The changes to TS 5.6.5.b 
includes the addition of AREVA topical reports, BAW-10240(P)(A), 
``Incorporation of M5\TM\ Properties in Framatome ANP Approved 
Methods,'' and EMF-2328(P)(A), ``PWR Small Break LOCA Evaluation Model 
S-RELAP5 Based,'' and the deletion of nine analytical methods that were 
previously approved but are no longer planned to be used, and therefore 
have not been analyzed for acceptability for M5\TM\ (M5) alloy fuel.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The Proposed Change Does Not Involve a Significant Increase 
in the Probability or Consequences of an Accident Previously 
Evaluated.
    The proposed license amendment adds a Nuclear Regulatory 
Commission approved analytical method, BAW-10240(P)(A), 
``Incorporation of M5\TM\ Properties in Framatome ANP Approved 
Methods,'' used to determine the core operating limits, to Technical 
Specification (TS) 5.6.5.b and changes the description of fuel 
assemblies specified in TS 4.2.1 to allow use of the M5 alloy. The 
proposed amendment does not affect the acceptance criteria for any 
Final Safety Analysis Report (FSAR) safety analysis analyzed 
accidents or anticipated operational occurrences. The proposed 
amendment does not involve operation of the required structures, 
systems or components (SSCs) in a manner different from those 
previously recognized or evaluated. As such, the proposed amendment 
does not increase the probability or consequences of an accident.
    In addition, the proposed license amendment adds NRC approved 
methodology EMF-2328(P)(A), ``PWR Small Break LOCA Evaluation Model, 
S-RELAP5 Based.'' This change, by itself, does not impact the 
current design bases. The proposed change enables the use of new 
methodologies to re-analyze small break loss-of-coolant accidents. 
Revised analyses may either result in continued conformance within 
design bases, or may change the design bases. If design bases 
changes result from a revised analysis, then the specific design 
changes will be evaluated in accordance with HBRSEP, Unit 2, design 
change procedures and 10 CFR 50.59. Further, this part of the change 
does not involve physical changes to any plant structure, system, or 
component.
    In addition, the proposed license amendment deletes nine 
analytical methods that were previously approved and listed in 
Section 5.6.5.b, but are no longer planned to be used. This change 
is administrative in nature as it removes methodologies that have 
become obsolete and hence have not been analyzed for acceptability 
with M5 fuel.
    Therefore, operation of the facility in accordance with the 
proposed amendment would 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 Previously Evaluated.
    Use of M5 fuel will not result in changes in the operation or 
configuration of the facility. Topical reports BAW-10227(P)(A) and 
BAW-10240(P)(A) evaluate the material properties of the M5 alloy and 
conclude that they are similar or better than those of zircaloy-4. 
Therefore, M5 fuel rod cladding will perform similarly to those 
fabricated from zircaloy-4, thus precluding the possibility of the 
fuel becoming an accident initiator and causing a new or different 
type of accident. No new failure mechanisms will be introduced by 
the changes being requested.
    The proposed addition of EMF-2328(P)(A) does not involve any 
physical alteration of plant systems, structures, or components, 
other than allowing for fuel design in accordance with NRC-approved 
methodologies. No new or different equipment is being installed. No 
installed equipment is being operated in a different manner. There 
is no change to the parameters within which the plant is normally 
operated or in the setpoints that initiate protective or mitigative 
actions. As a result, no new failure modes are being introduced by 
introduction of this methodology.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The Proposed Change Does Not Involve a Significant Reduction 
in the Margin of Safety.
    The proposed change will not involve a significant reduction in 
the margin of safety because it has been demonstrated that the 
material properties of the M5 alloy are not significantly different 
from those of zircaloy-4. M5 alloy is expected to perform similarly 
or better than zircaloy-4 for all normal operating and accident 
scenarios, including both loss-of-coolant accident (LOCA) and non-
LOCA scenarios. The proposed changes do not affect the acceptance 
criteria for any FSAR safety analysis analyzed accidents or 
anticipated operational occurrences. All required safety limits 
would continue to be analyzed using methodologies approved by the 
Nuclear Regulatory Commission.
    There is no impact on any margin of safety resulting from the 
incorporation of these new topical reports into the Technical 
Specifications. If design basis changes result from a revised 
analysis that uses these new methodologies, the specific design 
changes will be evaluated in accordance with HBRSEP, Unit 2, design 
change procedures and 10 CFR 50.59. Any potential reduction in the 
margin of safety would be evaluated for that specific design change.
    Therefore, the proposed amendment would not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Branch Chief: Douglas A. Broaddus.

[[Page 21922]]

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

    Date of amendment request: January 13, 2011.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) to change the description of fuel 
assemblies specified in TS 5.3.1 and add the AREVA NP Inc., topical 
report, BAW-10240(P)(A), ``Incorporation of M5\TM\ Properties in 
Framatome ANP Approved Methods,'' to the referenced analytical methods 
in administrative TS 6.9.1.6.2 to allow the use of M5\TM\ alloy for 
fuel rod cladding.
    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 license amendment adds a NRC approved analytical 
method, BAW-10240(P)(A), ``Incorporation of M5\TM\ Properties in 
Framatome ANP Approved Methods,'' used to determine the core 
operating limits, to TS 6.9.1.6.2 and changes the description of 
fuel assemblies specified in TS 5.3.1 to allow use of the M5\TM\ 
alloy. The proposed amendment does not affect the acceptance 
criteria for any Final Safety Analysis Report (FSAR) safety analysis 
analyzed accidents and anticipated operational occurrences. As such, 
the proposed amendment does not increase the probability or 
consequences of an accident. The proposed amendment does not involve 
operation of the required structures, systems or components (SSCs) 
in a manner or configuration different from those previously 
recognized or evaluated. Therefore, operation of the facility in 
accordance with the proposed amendment would 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.
    Use of M5\TM\ clad fuel will not result in changes in the 
operation or configuration of the facility. Topical Report BAW-10240 
describes, by reference, that the material properties of the M5\TM\ 
alloy are similar to or better than those of Zircaloy-4. Therefore, 
since M5\TM\ fuel rod cladding will perform similarly to those 
fabricated from Zircaloy-4, the possibility of the fuel becoming an 
accident initiator and causing a new or different type of accident 
is precluded. Since the material properties of M5\TM\ alloy are 
similar to or better than those of Zircaloy-4, there will be no 
significant changes in the types of any effluents that may be 
released off-site. There will not be a significant increase in 
occupational or public radiation exposure. The proposed amendment 
does not involve operation of any required SSCs in a manner or 
configuration different from those previously recognized or 
evaluated. No new failure mechanisms will be introduced by the 
changes being requested. 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 change will not involve a significant reduction in 
the margin of safety because it has been demonstrated that the 
material properties of the M5\TM\ alloy are not significantly 
different from those of Zircaloy-4. M5\TM\ alloy is expected to 
perform similarly to or better than Zircaloy-4 for all normal 
operating and accident scenarios, including both loss-of-coolant 
accident (LOCA) and non-LOCA scenarios. The proposed changes do not 
affect the acceptance criteria for any FSAR safety analysis analyzed 
accidents or anticipated operational occurrences. All required 
safety limits will continue to be analyzed using methodologies 
approved by the NRC. Therefore, the proposed amendment would not 
involve a significant reduction in a margin of safety. margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Branch Chief: Douglas A. Broaddus.

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

    Date of amendment request: January 20, 2011.
    Description of amendment request: The proposed change would revise 
H. B. Robinson Steam Electric Plant Technical Specifications (TS) 
Section 3.8.3, ``Diesel Fuel Oil and Starting Air,'' and Section 3.8.5, 
``DC Sources--Shutdown.'' The proposed change to TS 3.8.3 revises a 
nonconservative air receiver tank pressure to a value consistent with 
vendor recommendations. The proposed change to TS 3.8.5 corrects an 
editorial error related to TS formatting.
    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.
    The proposed change to TS 3.8.3 revises a non-conservative value 
in the current TS for EDG air start pressure. The proposed value is 
consistent with vendor recommendations and will ensure that the 
intent of the TS requirement is met. Therefore, the proposed change 
will provide improved assurance that the EDGs will be able to meet 
their safety function.
    The proposed change to TS 3.8.5 is an editorial correction and 
there will be no actual changes to plant design or operation.
    Therefore, operation of the facility in accordance with the 
proposed amendment would 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 Previously Evaluated.
    As described above, the proposed change to TS 3.8.3 provides 
improved assurance that the EDGs will be able to meet their safety 
function. No new failure modes are introduced. Therefore, no new 
accident initiators or precursors are introduced by the proposed 
change.
    The proposed change to TS 3.8.5 is an editorial correction and 
there will be no actual changes to plant design or operation.
    Therefore, operation of the facility in accordance with the 
proposed amendment would 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.
    As described above, the proposed change to TS 3.8.3 provides 
improved assurance that the EDGs will be able to meet their safety 
function of mitigating events that involve a loss of offsite power. 
Therefore, the proposed change will preserve any margin of safety.
    The proposed change to TS 3.8.5 is an editorial correction and 
there will be no actual changes to plant design or operation.

    Therefore, operation of the facility in accordance with the 
proposed amendment would 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: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy

[[Page 21923]]

Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 
27602.
    NRC Branch Chief: Douglas A. Broaddus.

Northern States Power Company--Minnesota, Docket No. 50-263, Monticello 
Nuclear Generating Plant (MNGP), Wright County, Minnesota

    Date of amendment request: February 7, 2011.
    Description of amendment request: The licensee proposed to amend 
the MNGP Technical Specifications (TS), revising Surveillance 
Requirement 3.5.1.7 regarding the Emergency Core Cooling System (ECCS) 
core spray flow from a minimum of 2800 gpm to a minimum of 2835 gpm. 
The licensee considers the current minimum flow rate requirement as 
non-conservative.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration (NSHC) analysis. The NRC staff reviewed the licensee's 
NSHC analysis and has prepared its own as follows:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The purpose of the minimum core spray flow rate requirement is 
to ensure that the ECCS will perform as designed. Of the postulated 
accidents and transients previously analyzed in the MNGP Updated 
Safety Analysis Report, none of them were postulated to be initiated 
by the ECCS performing as designed.
    Furthermore, the consequences of the previously analyzed 
accidents were not postulated to be exacerbated by the ECCS 
performing as designed. Accordingly, the probability of occurrence 
and the consequences of the previously analyzed accidents would not 
be affected in any way by the proposed amendment to the TS.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment does not involve any physical alteration 
of the plant (no new or different type of equipment will be 
installed) nor does it change methods and procedures governing plant 
operation. The proposed amendment will not impose any new or 
eliminate any old requirements. Therefore, the proposed amendment 
does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment will not have any effect on previously 
used safety analysis methods, scenarios, acceptance criteria, or 
assumptions. 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 
its own analysis, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the proposed amendment involves no significant hazards 
consideration.
    Attorney for the licensee: Peter M. Glass, Assistant General 
Counsel, Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 
55401.
    NRC Branch Chief: Robert J. Pascarelli.

Virginia Electric and Power Company, Docket No. 50-281, Surry Power 
Station, Unit 2, Surry County, Virginia

    Date of amendment request: December 16, 2010.
    Description of amendment request: This amendment request proposes 
to revise Technical Specification (TS) 6.4.Q, ``Steam Generator (SG) 
Program,'' to exclude portions of the SG tube below the top of the SG 
tubesheet from periodic tube inspections for Unit 2 during Refueling 
Outage 23 and the subsequent operating cycle. This amendment request 
also proposes to revise TS 6.6.A.3, ``Steam Generator Tube Inspection 
Report,'' to provide reporting requirements specific to Unit 2 for the 
temporary alternate repair criteria.
    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 previously analyzed accidents are initiated by the failure 
of plant structures, systems, or components. The proposed change 
that alters the steam generator inspection/repair criteria and the 
steam generator inspection reporting criteria does not have a 
detrimental impact on the integrity of any plant structure, system, 
or component that initiates an analyzed event. The proposed change 
will not alter the operation of, or otherwise increase the failure 
probability of any plant equipment that initiates an analyzed 
accident.
    Of the applicable accidents previously evaluated, the limiting 
transients with consideration to the proposed change to the steam 
generator tube inspection and repair criteria are the steam 
generator tube rupture (SGTR) event and the steam line break (SLB) 
postulated accidents.
    During the SGTR event, the required structural integrity margins 
of the steam generator tubes and the tube-to-tubesheet joint over 
the H* distance will be maintained. Tube rupture in tubes with 
cracks within the tubesheet is precluded by the constraint provided 
by the tube-to-tubesheet joint. This constraint results from the 
hydraulic expansion process, thermal expansion mismatch between the 
tube and tubesheet, and from the differential pressure between the 
primary and secondary side. Based on this design, the structural 
margins against burst, as discussed in Regulatory Guide (RG) 1.121, 
``Bases for Plugging Degraded PWR [Pressurized-Water Reactor] Steam 
Generator Tubes,'' are maintained for both normal and postulated 
accident conditions.
    The proposed change has no impact on the structural or leakage 
integrity of the portion of the tube outside of the tubesheet. The 
proposed change maintains structural integrity of the steam 
generator tubes and does not affect other systems, structures, 
components, or operational features. Therefore, the proposed change 
results in no significant increase in the probability of the 
occurrence of a SGTR accident.
    At normal operating pressures, leakage from primary water stress 
corrosion cracking below the proposed limited inspection depth is 
limited by both the tube-to-tubesheet crevice and the limited crack 
opening permitted by the tubesheet constraint. Consequently, 
negligible normal operating leakage is expected from cracks within 
the tubesheet region. The consequences of an SGTR event are affected 
by the primary to secondary leakage flow during the event. However, 
primary to secondary leakage flow through a postulated broken tube 
is not affected by the proposed changes since the tubesheet enhances 
the tube integrity in the region of the hydraulic expansion by 
precluding tube deformation beyond its initial hydraulically 
expanded outside diameter. Therefore, the proposed changes do not 
result in a significant increase in the consequences of a SGTR.
    The consequences of a steam line break (SLB) are also not 
significantly affected by the proposed changes. During a SLB 
accident, the reduction in pressure above the tubesheet on the shell 
side of the steam generator creates an axially uniformly distributed 
load on the tubesheet due to the reactor coolant system pressure on 
the underside of the tubesheet. The resulting bending action 
constrains the tubes in the tubesheet thereby restricting primary to 
secondary leakage below the midplane.
    Primary to secondary leakage from tube degradation in the 
tubesheet area during the limiting accident (i.e., a SLB) is limited 
by flow restrictions. These restrictions result from the crack and 
tube-to-tubesheet contact pressures that provide a restricted 
leakage path above the indications and also limit the degree of 
potential crack face opening as compared to free span indications.
    The probability of a SLB is unaffected by the potential failure 
of a steam generator tube as the failure of the tube is not an 
initiator for a SLB event.
    The leakage factor of 2.03 is a bounding value for all SGs, both 
hot and cold legs, in

[[Page 21924]]

Table 9-7 of WCAP-17092-P. Also as shown in Table 9-7 of WCAP-17092-
P, for Surry for a postulated SLB, a leakage factor of 1.80 has been 
calculated. However, for Surry, a more conservative leakage factor 
of 2.03 will be applied to the normal operating leakage associated 
with the tubesheet expansion region in the condition monitoring (CM) 
assessment and the operational assessment (OA). Specifically, for 
the CM assessment, the component of leakage from the prior cycle 
from below the H* distance will be multiplied by a factor of 2.03 
and added to the total leakage from any other source and compared to 
the allowable accident induced leakage limit. For the OA, the 
difference in the leakage between the allowable leakage and the 
accident induced leakage from sources other than the tubesheet 
expansion region will be divided by 2.03 and compared to the 
observed operational leakage.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change that alters the steam generator inspection/
repair criteria and the steam generator inspection reporting 
criteria does not introduce any new equipment, create new failure 
modes for existing equipment, or create any new limiting single 
failures. Plant operation will not be altered, and all safety 
functions will continue to perform as previously assumed in accident 
analyses.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The proposed change that alters the steam generator inspection/
repair criteria and the steam generator inspection reporting 
criteria maintains the required structural margins of the steam 
generator tubes for both normal and accident conditions. NEI 
[Nuclear Energy Institute] 97-06, Revision 2, ``Steam Generator 
Program Guidelines,'' and RG 1.121 are used as the bases in the 
development of the limited tubesheet inspection depth methodology 
for determining that steam generator tube integrity considerations 
are maintained within acceptable limits. RG 1.121 describes a method 
acceptable to the NRC for meeting GDC [General Design Criteria] 14, 
``Reactor Coolant Pressure Boundary,'' GDC 15, ``Reactor Coolant 
System Design,'' GDC 31, ``Fracture Prevention of Reactor Coolant 
Pressure Boundary,'' and GDC 32, ``Inspection of Reactor Coolant 
Pressure Boundary,'' by reducing the probability and consequences of 
a SGTR. RG 1.121 concludes that by determining the limiting safe 
conditions for tube wall degradation the probability and 
consequences of a SGTR are reduced. This RG uses safety factors on 
loads for tube burst that are consistent with the requirements of 
Section III of the American Society of Mechanical Engineers (ASME) 
Code.
    For axially oriented cracking located within the tubesheet, tube 
burst is precluded due to the presence of the tubesheet. For 
circumferentially oriented cracking, the H* analysis, documented in 
Section 4 of the license amendment request, defines a length of 
degradation free expanded tubing that provides the necessary 
resistance to tube pullout due to the pressure induced forces, with 
applicable safety factors applied. Application of the limited hot 
and cold leg tubesheet inspection criteria will preclude 
unacceptable primary to secondary leakage during all plant 
conditions. The methodology for determining leakage provides for 
large margins between calculated and actual leakage values in the 
proposed limited tubesheet inspection depth criteria.
    Therefore, the proposed change does not involve a significant 
reduction in any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219.
    NRC Branch Chief: Gloria Kulesa.

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.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-449, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: May 18, 2010, as supplemented by letter 
dated March 1, 2011.
    Brief description of amendment request: The proposed amendment 
would revise Technical Specification (TS) 6.8.3.I, ``Containment Post-
Tensioning System Surveillance Program.'' TS 6.8.3.I states that the 
containment post-tensioning system surveillance program shall be in 
accordance with American Society of Mechanical Engineers (ASME) Code, 
Section XI, Subsection IWL, 1992 Edition with 1992 Addenda, as 
supplemented by 10 CFR 50.55a(b)(2)(viii).
    The proposed amendment removes the specific year of the applicable 
Code edition consistent with Revision 3.1 of NUREG-1431, ``Standard 
Technical Specifications, Westinghouse Plants'' and will allow for 
future updates to the surveillance program when the applicable code 
edition changes without requiring additional TS changes.
    Date of publication of individual notice in the Federal Register: 
March 22, 2011 (76 FR 16012).
    Expiration date of individual notice: April 21, 2011 (public 
comments); May 23, 2011 (hearing requests).

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant 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

[[Page 21925]]

Assessment as indicated. All of these items are available for public 
inspection at the Commission's Public Document Room (PDR), located at 
One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland 20852. Publicly available records will be 
accessible from the Agencywide Documents Access and Management System 
(ADAMS) Public Electronic Reading Room on the internet at the NRC Web 
site, http://www.nrc.gov/reading-rm/adams.html. If you do not have 
access to ADAMS or if there are problems in accessing the documents 
located in ADAMS, contact the PDR Reference staff at 1-800-397-4209, 
301-415-4737 or by e-mail to [email protected].

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 application for amendments: March 31, 2010, as supplemented 
by letter dated November 30, 2010.
    Brief description of amendments: The amendments revised the 
Technical Specifications to relocate specific surveillance frequencies 
to a licensee-controlled program using a risk-informed justification.
    Date of issuance: March 29, 2011.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 263, 259.
    Renewed Facility Operating License Nos. NPF-35 and NPF-52: 
Amendments revised the licenses and the technical specifications.
    Date of initial notice in the Federal Register: November 16, 2010 
(75 FR 70034). The supplement dated November 30, 2010, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 29, 2011.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: March 24, 2010, as supplemented 
by letters dated November 18, 2010, and March 2, 2011.
    Brief description of amendments: The amendments revised the 
Technical Specifications to relocate specific surveillance frequencies 
to a licensee-controlled program using a risk-informed justification.
    Date of issuance: March 29, 2011.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 261, 241.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the licenses and the technical specifications.
    Date of initial notice in the Federal Register: November 16, 2010 
(75 FR 70035).
    The supplements dated November 18, 2010, and March 2, 2011, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 29, 2011.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 15, 2010, as supplemented 
by letters dated August 30, 2010, September 21, 2010, January 31, 2011, 
and February 18, 2011.
    Brief description of amendment: This amendment request would modify 
the Technical Specifications to revise the setpoint and setpoint 
tolerances for safety relief valves (SRVs) and spring safety valves 
(SSVs) and support the plant modifications associated with the 
replacement of (1) four Target Rock two-stage SRVs with three-stage 
SRVs, and (2) two existing Dresser 3.749 inch throat diameter SSVs with 
Dresser 4.956 inch throat diameter SSVs.
    Date of issuance: March 28, 2011.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 235.
    Facility Operating License No. DPR-35: The amendment revised the 
License and Technical Specifications.
    Date of initial notice in the Federal Register: May 4, 2010 (75 FR 
23812).
    The supplemental letters dated August 30, 2010, September 21, 2010, 
January 31, 2011, and February 18, 2011, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated March 28, 2011.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendment: August 31, 2010.
    Brief description of amendments: The amendments involve 
administrative changes to the Technical Specifications (TSs). The 
changes involve: (1) Making an editorial change to Limerick Generating 
Station (LGS) Unit 1 TS Limiting Condition for Operation (LCO) 3.3.1, 
Action b; (2) making an editorial change to LGS Units 1 and 2 TS Table 
3.3.1-1, Actions 2 and 9; (3) making the layout and format of LGS Unit 
1 TS LCO 3.6.5.3 Action requirements consistent with the LGS Unit 2 LCO 
Action requirements for the same TS; and (4) adding a reference to the 
minimum required number of operable main turbine bypass valves and the 
turbine bypass system response time to the core operating limits 
documented in the Core Operating Limits Report as specified in LGS, 
Units 1 and 2, TS 6.9.1.9.
    Date of issuance: March 31, 2011.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment Nos.: 200 and 161.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the license and the technical specifications.
    Date of initial notice in the Federal Register: November 2, 2010 
(75 FR 67402).
    The Commission's related evaluation of the amendment is contained 
in Safety Evaluation dated March 31, 2011.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-353, Limerick Generating 
Station, Unit 2, Montgomery County, Pennsylvania

    Date of application for amendment: December 15, 2010, as 
supplemented on February 17, 2011, and March 17, 2011.
    Brief description of amendment: The changes revise the Technical 
Specification (TS) relating to the Safety Limit Minimum Critical Power 
Ratios (SLMCPRs). The changes result from a cycle specific analysis 
performed to support the operation of Limerick

[[Page 21926]]

Generating Station, Unit 2, in the upcoming Cycle 12. Specifically, the 
TS changes will revise the SLMCPRs contained in TS 2.1 for two 
recirculation loop operation and single recirculation loop operation to 
reflect the changes in the cycle specific analysis. The new SLMCPRs are 
calculated using Nuclear Regulatory Commission-approved methodology 
described in NEDE 24011-P-A, General Electric Standard Application for 
Reactor Fuel, Revision 17.
    Date of issuance: April 5, 2011.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 162.
    Facility Operating License No. NPF-85. The amendment revised the 
license and the technical specifications.
    Date of initial notice in the Federal Register: February 1, 2011 
(76 FR 5620). The supplements dated February 17, 2011, and March 17, 
2011, clarified the application, did not expand the scope of the 
application as originally noticed, and did not change the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 5, 2011.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company (IandM), Docket No. 50-316, Donald C. 
Cook Nuclear Plant, Unit 2, Berrien County, Michigan

    Date of application for amendment: March 19, 2009, as supplemented 
on November 20, 2009, February 24, March 11, and March 25, 2011.
    Brief description of amendment: The amendment adopts a new analysis 
of a large-break loss-of-coolant accident, and revises the Technical 
Specifications to reflect this new analysis, which was performed using 
a plant-specific adaptation of the NRC-approved methodology set forth 
in Westinghouse Topical Report WCAP-16009-P-A, ``Realistic Large-Break 
LOCA Evaluation Methodology Using the Automated Statistical Treatment 
of Uncertainty Method (ASTRUM).''
    Date of issuance: March 31, 2011.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 297.
    Facility Operating License No. DPR-74: Amendment revised the 
Renewed Operating License and Technical Specifications.
    Date of initial notice in the Federal Register: August 11, 2009 (74 
FR 40238).
    The supplemental information contained clarifying information, did 
not change the scope of the license amendment request, did not change 
the NRC staff's initial proposed finding of no significant hazards 
consideration determination, and did not expand the scope of the 
original Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 31, 2011.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 1, 2010.
    Brief description of amendments: The amendments revised the 
inspection scope and repair requirements in Technical Specification 
(TS) 5.5.9, ``Unit 1 Model D76 and Unit 2 Model D5 Steam Generator (SG) 
Program,'' to exclude portions of the Comanche Peak Nuclear Power Plant 
(CPNPP), Unit 2, Model D5 SG tubes below the top of the SG tubesheet 
from periodic SG tube inspections. In addition, the amendments revised 
TS 5.6.9, ``Unit 1 Model D76 and Unit 2 Model D5 Steam Generator Tube 
Inspection Reports,'' to provide reporting requirements specific to 
CPNPP, Unit 2, for the temporary alternate repair criteria. The changes 
are applicable only to CPNPP, Unit 2, during Refueling Outage 12 and 
the subsequent operating cycle.
    Date of issuance: April 6, 2011.
    Effective date: As of the date of issuance and shall be implemented 
prior to Mode 4 entry during startup from Unit 2 Refueling Outage 12.
    Amendment Nos.: Unit 1--154; Unit 2--154.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in the Federal Register: February 1, 2011 
(76 FR 5622).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 6, 2011. No significant hazards 
consideration comments received: No.

NextEra Energy, Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of application for amendments: April 7, 2009, as supplemented 
by letters dated June 17 (two letters), September 11, September 25, 
October 9, November 20 (two letters), November 21 (two letters), 
November 30, December 8, and December 16 of 2009; and January 7, 
January 8, January 22, February 11, February 25, March 3, April 15, 
April 22, April 28, July 8, July 28, August 2, August 9, August 24, 
October 15, November 1, November 12 (two letters), November 30, and 
December 21 of 2010. The proposed changes were originally included as 
part of the April 7, 2009, extended power uprate (EPU) license 
amendment request, but subsequently divided into a separate licensing 
action for independent technical review.
    Brief description of amendments: The amendment changes the AFW 
system design and Technical Specifications (TS) 3.7.5, ``Auxiliary 
Feedwater (AFW),'' and TS 3.7.6, ``Condensate Storage Tank (CST),'' 
resulting from (1) modifications to the AFW system to support 
requirements for transients and other accidents at EPU conditions; (2) 
installation of main feedwater isolation valves to support accident 
mitigation by ensuring that containment pressure does not exceed safety 
analysis limits; (3) automatic AFW switchover from a CST suction source 
to a safety-related Service Water source; and (4) instrumentation 
setpoint changes supporting the aforementioned physical modifications. 
The upgrades and modifications to the AFW system are being installed to 
provide additional capacity and reliability for the system. Although 
the proposed changes are also designed to support the requirements for 
transients and other accidents at EPU conditions, the changes for this 
amendment have been evaluated using the current licensing basis.
    Date of issuance: March 25, 2011.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days.
    Amendment Nos.: 238, 242.
    Renewed Facility Operating License Nos. DPR-24 and DPR-27: 
Amendments revise the License, Appendix C, and the Technical 
Specifications.
    Date of initial notice in the Federal Register: September 21, 2010 
(75 FR 57525).
    The supplemental letters contained clarifying information and did 
not change the staff's initial proposed finding of no significant 
hazards consideration.

[[Page 21927]]

    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 25, 2011. No significant hazards 
consideration comments received: No.

NextEra Energy, Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of application for amendments: April 7, 2009, as supplemented 
by letters dated June 17, September 11, September 25, November 20, 
November 30, and December 8 of 2009; February 11, February 25, April 
22, April 30, July 21, July 28, August 2, and September 28 of 2010.
    Brief description of amendments: The amendment changed the 
Technical Specifications to support (1) modifications to the AFW 
system; (2) an EPU to increase plant core thermal power from 1,540 
megawatts thermal (MWt) to 1,800 MWt; and (3) update non-conservative 
RPS and ESFAS setpoints not associated with the EPU. The amendment also 
modified the RPS instrumentation setpoints of TS Table 3.3.1-1 and the 
ESFAS instrumentation setpoints of TS Table 3.3.2-1. The changes 
include both EPU and non-EPU related setpoints. The revised TS 
allowable values have been calculated to account for new analytical 
limits, instrument uncertainties, and instrument drift. The changes 
also include the addition of a new column entitled Nominal Trip 
Setpoint that was added to provide consistency with the TS Table format 
in NUREG 1431, ``Standard Technical Specifications--Westinghouse 
Plants,'' and Technical Specification Task Force (TSTF)-493, Revision 
4, ``Clarify the Application of Setpoint Methodology for Limiting 
Safety System Setting (LSSS) Functions.'' The RPS and ESFAS 
instrumentation uncertainty/setpoint calculations have also been 
revised to eliminate the use of a single-sided reduction factor in the 
total loop error determination for LSSS setpoints.
    Date of issuance: March 25, 2011.
    Effective date: As of the date of issuance, and shall be 
implemented prior to Unit 1 startup from the Fall 2011 refueling outage 
(Unit 1) and within 180 days (Unit 2).
    Amendment Nos.: 239 and 243.
    Renewed Facility Operating License Nos. DPR-24 and DPR-27: 
Amendments revised the Technical Specifications/License.
    Date of initial notice in the Federal Register: September 21, 2010 
(75 FR 57524).
    The supplemental letters contained clarifying information and did 
not change the staff's initial proposed finding of no significant 
hazards consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 25, 2011. No significant hazards 
consideration comments received: No.

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

    Date of application for amendment: March 29, 2010, as supplemented 
by letters dated May 28, and September 30, 2010, and two letters dated 
February 14, 2011.
    Brief description of amendment: The amendment modifies the 
Technical Specifications (TSs) to extend the allowed outage time for 
the A and B emergency diesel generators from 72 hours to 14 days.
    Date of issuance: March 25, 2011.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 188.
    Facility Operating License No. NPF-57: The amendment revised the 
TSs and the License.
    Date of initial notice in the Federal Register: June 29, 2010 (75 
FR 37476). The letters dated May, 28, and September 30, 2010, and 
February 14, 2011 (two letters), provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination or expand the application beyond the scope 
of the original Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 25, 2011. No significant hazards 
consideration comments received: No.

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

    Date of amendment request: April 13, 2010, as supplemented by 
letters dated October 13 and December 21, 2010, and January 18, 2011.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) Table 3.3.2-1, ``Engineered Safety Feature Actuation 
System Instrumentation,'' by adding a footnote to Function 8.a 
concerning the reactor trip P-4 engineered safety feature actuation 
system interlock. The footnote specifies which functions of the 
interlock are necessary in each mode in order to meet the limiting 
condition for operation. Specifically, the functions of tripping the 
main turbine and isolating main feedwater with a coincident low average 
temperature would no longer be applicable in MODE 3, which is hot 
standby. Revised TS Table 3.3.2-1 also identifies that the function of 
the P-4 interlock that allows arming of the steam dump valves and 
transfers the steam dump load rejection (Tavg) controller to 
the plant trip controller is not required in any mode.
    Date of issuance: March 30, 2011.
    Effective date: The amendment will be effective upon issuance and 
will be implemented within 90 days from the date of issuance.
    Amendment No.: 194.
    Renewed Facility Operating License No. NPF-42. The amendment 
revised the Operating License and Technical Specifications.
    Date of initial notice in the Federal Register: June 15, 2010 (75 
FR 33844). The supplemental letters dated October 13 and December 21, 
2010, and January 18, 2011, provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the staff's original proposed 
no significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation March 30, 2011. No significant hazards 
consideration comments received: No.

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

    Date of amendment request: November 30, 2010.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 5.5.9, ``Steam Generator (SG) Program,'' to exclude 
portions of the tube below the top of the steam generator tubesheet 
from periodic steam generator tube inspections during Refueling Outage 
18 and the subsequent operating cycle. In addition, TS 5.6.10, ``Steam 
Generator Tube Inspection Report'' will be revised to remove a 
reference to the previous interim alternate repair criteria and to 
provide reporting requirements specific to the temporary alternate 
repair criteria.
    Date of issuance: April 6, 2011.
    Effective date: The amendment is effective upon issuance and will 
be implemented prior to MODE 4 entry during startup from Refueling 
Outage 18.
    Amendment No.: 195.
    Renewed Facility Operating License No. NPF-42. The amendment 
revised

[[Page 21928]]

the Operating License and Technical Specifications.
    Date of initial notice in the Federal Register: February 1, 2011 
(76 FR 5623).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 6, 2011. No significant hazards 
consideration comments received: No.

    Dated at Rockville, Maryland, this 8th day of April 2011.

    For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2011-9177 Filed 4-18-11; 8:45 am]
BILLING CODE 7590-01-P