Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 55125-55136 [2011-22541]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 76, Number 172 (Tuesday, September 6, 2011)]
[Notices]
[Pages 55125-55136]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2011-22541]


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NUCLEAR REGULATORY COMMISSION

[NRC-2011-0205]


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

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 August 11, 2011 to August 24, 2011. The last 
biweekly notice was published on August 23, 2011 (76 FR 52699).

ADDRESSES: Please include Docket ID NRC-2011-0205 in the subject line 
of your comments. Comments submitted in writing or in electronic form 
will be posted on the NRC Web site and on the Federal rulemaking Web 
site https://www.regulations.gov. Because your comments will not be 
edited to remove any identifying or contact information, the NRC 
cautions you against including any information in your submission that 
you do not want to be publicly disclosed.
    The NRC requests that any party soliciting or aggregating comments 
received from other persons for submission to the NRC inform those 
persons that the NRC will not edit their comments to remove any 
identifying or contact information, and therefore, they should not 
include any information in their comments that they do not want 
publicly disclosed.
    You may submit comments by any one of the following methods:
     Federal Rulemaking Web Site: Go to https://www.regulations.gov and search for documents filed under Docket ID NRC-
2011-0205. Address questions about NRC dockets to Carol Gallagher 301-
492-3668; e-mail Carol.Gallagher@nrc.gov.
     Mail comments to: Chief, Rules, Announcements, and 
Directives Branch (RADB), Office of Administration, Mail Stop: TWB-05-
B01M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.
     Fax comments to: RADB at 301-492-3446.
    You can access publicly available documents related to this notice 
using the following methods:
     NRC's Public Document Room (PDR): The public may examine 
and have copied, for a fee, publicly available documents at the NRC's 
PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, 
Rockville, Maryland 20852.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): Publicly available documents created or received at the NRC 
are accessible electronically through ADAMS in the NRC Library at 
https://www.nrc.gov/reading-rm/adams.html. From this page, the public 
can gain entry into ADAMS, which provides text and image files of the 
NRC's public documents. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC's PDR reference staff at 1-800-397-4209, 301-415-4737, or by e-mail 
to pdr.resource@nrc.gov.
     Federal Rulemaking Web Site: Public comments and 
supporting materials related to this notice can be found at https://www.regulations.gov by searching on Docket ID: NRC-2011-0205.

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

[[Page 55126]]

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.
    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 NRC's PDR, located at One White Flint North, Room O1-F21, 11555 
Rockville Pike (first floor), Rockville, Maryland 20852. NRC 
regulations are accessible electronically from the NRC Library on the 
NRC Web site at https://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 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by e-mail at hearing.docket@nrc.gov, 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 https://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 
https://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 https://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

[[Page 55127]]

https://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 https://www.nrc.gov/site-help/e-submittals.html, by e-mail at 
MSHD.Resource@nrc.gov, 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 
https://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 NRC's PDR, located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. Publicly available documents created or received at the NRC are 
accessible electronically through ADAMS in the NRC Library at https://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 pdr.resource@nrc.gov.

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

    Date of amendment request: July 12, 2011.
    Description of amendment request: The proposed license amendments 
would revise Technical Specification (TS) 3.4.5, ``Reactor Coolant 
System (RCS) Leakage Detection Instrumentation,'' to define a new time 
limit for restoring inoperable RCS leakage detection instrumentation to 
operable status and establish alternate methods of monitoring RCS 
leakage when one or more required monitors are inoperable. These 
proposed changes would be consistent with Standard Technical 
Specifications Change Traveler (TSTF)-514, ``Revise BWR Operability 
Requirements and Actions for RCS Leakage Instrumentation.'' The 
availability of TSTF-514 was announced in the Federal Register on 
December 17, 2010 (75 FR 79048), as part of the consolidated line item 
improvement process.
    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 primary 
containment atmosphere gaseous radioactivity 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.
    Therefore, it is concluded that this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No
    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 primary 
containment atmosphere gaseous radioactivity monitor. The proposed 
change does not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed) or a change in 
the methods governing normal plant operation.
    Therefore, it is concluded that the proposed change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No

[[Page 55128]]

    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 primary 
containment atmosphere gaseous radioactivity monitor. Reducing the 
amount of time the plant is allowed to operate with only the primary 
containment atmosphere gaseous radioactivity 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.
    Therefore, it is concluded that the proposed change does not 
involve a significant reduction in a margin of safety.

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

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

    Date of amendment request: April 29, 2011.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.4.15, ``RCS [Reactor Coolant 
System] Leakage Detection Instrumentation,'' to define a new time limit 
for restoring inoperable RCS leakage detection instrumentation to 
operable status; establish alternate methods of monitoring RCS leakage 
when one or more required monitors are inoperable; and make TS Bases 
changes which reflect the proposed changes and more accurately reflect 
the contents of the facility design basis related to operability of the 
RCS leakage detection instrumentation. New Condition C is applicable 
when the reactor building atmosphere gaseous radioactivity monitor is 
the only operable TS-required monitor. New Condition C Required Actions 
require analyzing grab samples of the reactor building atmosphere every 
12 hours and restoring another monitor within 7 days. These changes are 
consistent with NRC-approved Revision 3 to Technical Specification Task 
Force (TSTF) Standard Technical Specification (STS) Change Traveler 
TSTF-513, ``Revise PWR [Pressurized-Water Reactor] Operability 
Requirements and Actions for RCS Leakage Instrumentation.'' The 
availability of this TS improvement was announced in the Federal 
Register on January 3, 2011 (76 FRN 189), as part of the consolidated 
line item improvement process (CLIIP).
    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 reactor 
building 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.
    Therefore, it is concluded that 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 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 reactor 
building 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.
    Therefore, it is concluded that the proposed change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Does the Proposed Change Involve a Significant Reduction in a 
Margin of Safety?
    Response: No.
    The proposed change 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 reactor 
building atmosphere gaseous radiation monitor. Reducing the amount 
of time the plant is allowed to operate with only the reactor 
building 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.
    Therefore, it is concluded that the proposed change does not 
involve a significant reduction in a margin of safety.

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

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

    Date of amendment request: April 6, 2011.
    Description of amendment request: The proposed amendment would 
modify the actions to be taken when the atmospheric gaseous 
radioactivity monitor is the only operable reactor coolant leakage 
detection instrument. The modified actions require additional, more 
frequent monitoring of other indications of Reactor Coolant System 
(RCS) leakage and provide appropriate time to restore another leakage 
detection instrument to operable status. This change is consistent with 
the U.S. Nuclear Regulatory Commission (NRC) approved safety evaluation 
on Technical Specification Task Force (TSTF) Traveler, TSTF-514-A, 
Revision 3, ``Revised BWR [boiling-water reactor] Operability 
Requirements and Actions for RCS Leakage Instrumentation'' dated 
November 24, 2010.
    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, with NRC edits in brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No
    The proposed changes [ ] modify the time allowed for the plant 
to operate when the only Operable RCS leakage detection

[[Page 55129]]

instrumentation monitor is the atmospheric 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.
    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 changes [ ] modify the time allowed for the plant 
to operate when the only Operable RCS leakage detection monitor is 
the atmospheric gaseous radiation monitor. The proposed changes do 
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.
    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 changes [ ] increase the time allowed for the plant 
to operate when the only Operable RCS leakage detection 
instrumentation monitor is the atmospheric gaseous radiation monitor 
from 24 hours to 7 days. Increasing the amount of time the plant is 
allowed to operate with only the atmospheric gaseous radiation 
monitor Operable does not significantly decrease the margin of 
safety due to the addition of compensatory Required Actions to 
analyze grab samples of the primary containment atmosphere once per 
12 hours and monitor Reactor Coolant System leakage by 
administrative means once per 12 hours. The overall likelihood that 
an increase in RCS leakage will be detected before it potentially 
results in gross failure is maintained with the addition of the 
Required Actions.
    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, including the edits in brackets above, 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: J. Bradley Fewell, Esquire, Associate 
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Richard B. Ennis, Acting.

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

    Date of amendment request: June 2, 2011.
    Description of amendment request: The proposed amendment would 
modify Technical Specification Limiting Condition for Operation 3.1.2, 
``Reactor Anomalies,'' to allow performance of the surveillance on a 
comparison of predicted to actual (or monitored) effective core 
reactivity (Keff). The reactivity anomaly verification is 
currently determined by a comparison of predicted vs. actual control 
rod density.
    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 with changes by the NRC staff 
noted in brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed Technical Specifications changes do not 
[substantively] affect any plant systems, structures, or components 
designed for the prevention or mitigation of previously evaluated 
accidents. The amendment would only change how the reactivity 
anomaly surveillance is performed. Verifying that the core 
reactivity is consistent with predicted values ensures that accident 
and transient safety analyses remain valid. This amendment changes 
the Technical Specification requirements such that, rather than 
performing the surveillance by comparing predicted to actual control 
rod density, the surveillance is performed by a direct comparison of 
keff. Present day on-line core monitoring systems, such 
as the one in use at Peach Bottom Atomic Power Station (PBAPS), 
Units 2 and 3 are capable of performing the direct measurement of 
reactivity.
    Therefore, since the reactivity anomaly surveillance will 
continue to be performed by a viable method, the proposed amendment 
does not involve a significant increase in the probability or 
consequence of a previously evaluated accident.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    This Technical Specifications amendment request does not 
[substantively change] the operation, testing, or maintenance of any 
safety-related, or otherwise important to safety systems. All 
systems important to safety will continue to be operated and 
maintained within their design bases. The proposed changes to the 
reactivity anomaly Technical Specifications will only provide a new, 
more efficient method of detecting an unexpected change in core 
reactivity.
    Since all systems continue to be operated within their design 
bases, no new failure modes are introduced and the possibility of a 
new or different kind of accident is not created.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    This proposed Technical Specifications amendment proposes to 
change the method for performing the reactivity anomaly surveillance 
from a comparison of predicted to actual control rod density to a 
comparison of predicted to actual keff. The direct 
comparison of keff provides a technically superior method 
of calculating any differences in the expected core reactivity. The 
reactivity anomaly surveillance will continue to be performed at the 
same frequency as is currently required by the Technical 
Specifications, only the method of performing the surveillance will 
be changed. Consequently, core reactivity assumptions made in safety 
analyses will continue to be adequately verified.
    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, including the changes made by the NRC staff as noted in 
brackets, 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: J. Bradley Fewell, Esquire, Associate 
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: G. Edward Miller, Acting.

Florida Power Corporation, et al. (FPC), Docket No. 50-302, Crystal 
River Unit 3 Nuclear Generating Plant (CR-3), Citrus County, Florida

    Date of amendment request: December 20, 2010, as supplemented by 
the July 20, 2011 letter.
    Description of amendments request: FPC will be constructing and 
operating an on-site independent spent fuel storage installation at CR-
3, as a general licensee under the provisions of 10 CFR part 72, 
Subpart K to maintain full-core offload capacity in the spent fuel 
pools. The spent fuel pools are located in the CR-3 Auxiliary Building 
(AB). In support of future dry shielded canister/transfer cask loading 
operations, FPC is replacing the existing AB overhead crane with a new 
single failure proof crane designed in accordance with American Society 
of Mechanical Engineers (ASME) NOG-1-2004, ``Rules

[[Page 55130]]

for Construction of Overhead and Gantry Cranes (Top Running Bridge, 
Multiple Girder).'' The licensee requested NRC approval of the 
following:
    1. An exception to ASME NOG-1-2004 pertaining to the application of 
tornado wind and tornado generated missile loading to auxiliary 
building overhead crane (FHCR-5) and its support structure.
    2. Revisions to the CR-3 Final Safety Analysis Report (FSAR) 
Sections 5.1.1.1.h and 9.6.1.5.a.5 to specifically identify the design 
parameters for FHCR-5 and its support structure.
    3. Deletion of a commitment in FSAR Section 9.6.3.1, ``Spent Fuel 
Assembly Removal,'' due to the expansion of spent fuel storage over 
that originally credited in the CR-3 Safety Evaluation Report dated 
July 5, 1974.
    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.
    The proposed LAR [license amendment request] does not involve 
plant equipment used to operate or shut down the reactor or in the 
mitigation of accidents described in Chapter 14 of the FSAR. FHCR-5 
will be restricted from movement over fuel stored in either of the 
spent fuel pools by administrative controls and designated safe load 
paths when moving spent fuel casks, and it will be single failure 
proof so a cask load drop accident affecting stored spent fuel is 
prevented. The change provides justification for an exception to a 
Code requirement pertaining to the design and qualification of the 
new single failure proof crane in the AB. The new crane will meet 
the design specifications in ASME NOG-1-2004, with the exception of 
Section 4134(c). The change also includes a commitment not to 
operate the crane if an Approaching or Potential Tropical Storm, an 
Approaching or Potential Hurricane, or a Tornado Watch or Warning 
has been declared for the site. The revised FSAR description of the 
crane will meet the intent of the original description and will 
ensure the crane will exceed the design requirements of the original 
design. With the replacement of the crane, the occurrence of a cask 
load drop accident is not considered credible. As a result, the 
proposed change does not increase the probability or consequences of 
a load drop accident previously evaluated that could impact stored 
fuel and/or pool structural integrity.
    Therefore, the proposed change does not involve significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The power generation portion of the plant is unaffected by the 
proposed change, which is limited to the design and analysis of a 
new overhead crane in the AB. The location and design functions of 
the AB overhead crane remain as they are currently described in the 
CR-3 FSAR. Overall, the design of the crane is being enhanced to 
single failure proof in order to reduce the likelihood of an 
uncontrolled lowering of the load due to an unforeseen malfunction 
or subcomponent failure. Portions of the design and analysis of the 
crane require NRC approval because they deviate from the NRC-
endorsed design code for single failure proof cranes and the CR-3 
licensing basis. The new single failure proof crane will be used to 
move a loaded or unloaded transfer cask between the cask loading 
pit, the decontamination pit, and the transfer trailer in the truck 
bay. Any credible event involving the fuel handling evolutions are 
bounded by existing FSAR analyses.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does not involve a significant reduction in a margin of 
safety.
    This proposed LAR involves the replacement of the existing non-
single failure proof AB overhead crane with a new single failure 
proof crane. The new crane will meet the design specifications found 
in ASME NOG-1-2004, with the exception of Section 4134(c). ASME NOG-
1-2004 has been endorsed by the NRC in Regulatory Issue Summary 
(RIS) 2005-25, Supplement 1, ``Clarification of NRC Guidelines for 
Control of Heavy Loads,'' as an acceptable means of meeting the 
criteria in NUREG-0554, ``Single Failure Proof Cranes for Nuclear 
Power Plants.'' The ASME NOG-1-2004 design code has been found by 
the NRC to provide adequate protection and safety margin against the 
uncontrolled lowering of the lifted load. The occurrence of a cask 
load drop accident is considered not credible when the load is 
lifted with a single failure proof lifting system meeting the 
guidance in NUREG-0612, ``Control of Heavy Loads at Nuclear Power 
Plants,'' Section 5.1.6, ``Single Failure Proof Handling Systems.'' 
As a result, the proposed change has no adverse impact on new fuel, 
stored spent fuel, cooling capacity of the pool, or structural 
integrity of the pool. Similarly, the margin of safety for the 
operation and safe shutdown of the plant will not be affected by the 
proposed change.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

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

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

    Date of amendment request: July 14, 2011.
    Description of amendment request: The proposed change would replace 
the Technical Specification (TS) required 10-year surveillance 
frequency for testing the containment spray nozzles in accordance with 
TS surveillance 4.6.2.1.d with an event-based frequency. Specifically, 
verification that the spray nozzle is unobstructed would only be 
required following activities that could result in nozzle blockage.
    Basis for proposed no significant hazards consideration (NSHC) 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of NSHC, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The spray nozzles and the associated containment spray system 
(CBS) are designed to perform accident mitigation functions. The 
proposed change to reduce the frequency and remove specific details 
of surveillance testing that verifies the spray nozzles are 
unobstructed does not impact the physical function of plant 
structures, systems, or components (SSCs) or the manner in which 
SSCs perform their design function. The proposed change neither 
adversely affects accident initiators or precursors, nor alters 
design assumptions. The proposed change does not alter or prevent 
the ability of operable SSCs to perform their intended function to 
mitigate the consequences of an initiating event within assumed 
acceptance limits. The capability of the CBS system to perform its 
accident mitigation functions is not adversely affected by the 
proposed change.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed change will not impact the accident analysis. The 
change does not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed), a significant 
change in the method of plant operation, or new operator actions. 
The change does not make any physical modifications to the CBS 
system, changes to setpoints, or changes to the method of delivering 
borated water to the CBS spray nozzles. The proposed change will not 
introduce failure modes that could result in a new accident, and the 
change does not alter assumptions made in the safety analysis.

[[Page 55131]]

    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel cladding, reactor coolant 
system pressure boundary, and containment structure) to limit the 
level of radiation dose to the public. The proposed change does not 
involve a significant change in the method of plant operation, and 
no accident analyses will be affected by the proposed changes. 
Additionally, the proposed changes will not relax any criteria used 
to establish safety limits and will not relax any safety system 
settings. The safety analysis acceptance criteria are not affected 
by this change. The proposed change will not result in plant 
operation in a configuration outside the design basis. The proposed 
change does not adversely affect systems that respond to safely shut 
down the plant and to maintain the plant in a safe shutdown 
condition.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves NSHC. Attorney for licensee: M.S. Ross, 
Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-
0420.
    NRC Branch Chief: Harold K. Chernoff.

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

    Date of amendment request: July 26, 2011.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TSs). Specifically, the proposed 
change would revise the minimum indicated nitrogen cover pressure 
specified for the accumulators in TS Surveillance Requirement (SR) 
3.5.1.3 from 617 psig (pounds per square inch, gauge) to 626 psig. The 
proposed change is necessary to account for the uncertainty associated 
with the accumulator pressure indication instrumentation. Currently, in 
accordance with NRC Administrative Letter 98-10, ``Dispositioning of 
Technical Specifications that Are Insufficient to Assure Plant 
Safety,'' VEGP is administratively controlling the minimum indicated 
accumulator pressure to greater than or equal to 626 psig. In addition, 
an editorial error in the text of TS SR 3.6.2.1 would also be 
corrected.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No
    The proposed amendment revises the minimum indicated nitrogen 
cover pressure specified for the SI [safety injection] accumulators 
in SR 3.5.1.3 from 617 psig to 626 psig. In addition, the proposed 
change includes an administrative change to correct an editorial 
error in the text of TS SR 3.6.2.1.
    The SI accumulators are not a precursor to any accident 
previously evaluated. The SI accumulators are used to mitigate the 
consequences of accidents previously evaluated. The proposed change 
to the indicated minimum SI accumulator nitrogen cover pressure 
provides assurance that the requirements of the TS continue to bound 
the acceptance limits of the SI accumulators with respect to the 
assumptions in the LOCA [loss-of-coolant accident] analyses.
    Thus, the proposed change does not affect the probability or the 
consequences of any accident previously evaluated. The proposed 
change to correct an editorial error in the text of SR 3.6.2.1 has 
no impact on the probability or consequences of any accident 
previously evaluated.
    Therefore, it is concluded that 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 revises the minimum indicated nitrogen cover 
pressure specified for the SI accumulators in SR 3.5.1.3 from 617 
psig to 626 psig. In addition, the proposed change includes an 
administrative change to correct an editorial error in the text of 
TS SR 3.6.2.1.
    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 to the requirements of the TS assure that the 
acceptance limits of the SI accumulators with respect to the 
assumptions in the LOCA analyses continue to be met, and correct an 
editorial error in the text of an SR. Thus, the proposed change does 
not adversely affect the design function or operation of any 
structures, systems, and components important to safety.
    Therefore, it is concluded that the proposed change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No
    The proposed change revises the minimum indicated nitrogen cover 
pressure specified for the SI accumulators in SR 3.5.1.3 from 617 
psig to 626 psig. In addition, the proposed change includes an 
administrative change to correct an editorial error in the text of 
TS SR 3.6.2.1.
    The proposed change to the indicated SI accumulator nitrogen 
cover pressure provides assurance that the requirements of the TS 
continue to bound the acceptance limits of the SI accumulators with 
respect to the assumptions in the LOCA analyses. Thus the proposed 
change to the SI accumulator minimum nitrogen cover pressure assures 
the existing margin of safety is maintained. The proposed change to 
correct an editorial error in the text of SR 3.6.2.1 has no impact 
on the margin of safety.
    Therefore, it is concluded that the proposed change does not 
involve a significant reduction in a margin of safety.

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

Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant, 
Unit 2, Hamilton County, Tennessee

    Date of amendment request: July 15, 2011 (TS-SQN-2011-01).
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TSs) requirements for steam 
generator (SG) tube inspections to reflect the replacement steam 
generators (RSGs) to be installed during Sequoyah Nuclear Plant (SQN), 
Unit 2, refueling outage 18 presently scheduled for the fall of 2012. 
Previous changes to the SQN, Unit 2, TSs to reflect the Technical 
Specification Task Force (TSTF) Standard Technical Specification 
Traveler, TSTF-449, ``Steam Generator Tube Integrity,'' Revision 4, 
were approved by Nuclear Regulatory Commission (NRC) on May 22, 2007. 
The changes proposed in this amendment reflect the inspection 
requirements of TSTF-449, Revision 4. The RSG tubes will be made of 
Alloy 690 thermally treated (TT) material, and the existing SGs have 
Alloy 600 tubes. The revisions to TSs are required because the 
inspection frequency for Alloy 690 TT tube material, as defined

[[Page 55132]]

in TSTF-449, differs from the inspection frequency for Alloy 600, and 
the tube repair processes and products in the existing TSs are not 
applicable to the RSGs.
    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 for RSGs continues to implement the current 
SG Program that includes performance criteria which provide 
reasonable assurance that the RSG tubing will retain integrity over 
the full range of operating conditions (including startup, operation 
in the power range, hot standby, cooldown, and all anticipated 
transients included in the design specifications). This change 
removes repair criteria from the SG Program that were approved by 
previous License Amendments for the existing SGs which are not 
applicable to the RSGs. It removes references to use of repairs and 
reporting of repair results in other TS sections. This change 
removes inspection requirements that are designated for specific 
damage conditions in the existing SGs. The change also revises the 
inspection interval for 100 percent inspections of SG tubes and the 
maximum interval for inspection of a single SG consistent with 
Technical Specification Task Force (TSTF) Standard Technical 
Specification Traveler, TSTF-449, ``Steam Generator Tube 
Integrity,'' Revision 4 for the Alloy 690 tube material in the RSGs. 
The revised inspection requirements are based on properties and 
experience with the improved Alloy 690 tube material. The revised 
inspection requirements will result in the same outcome that SG tube 
integrity will continue to be maintained.
    This change continues to implement SG performance criteria for 
tube structural integrity, accident induced leakage, and operational 
leakage for the RSGs. Meeting the performance criteria provides 
reasonable assurance that the RSG tubing will remain capable of 
fulfilling its specific safety function of maintaining reactor 
coolant pressure boundary integrity throughout each operating cycle 
and in the unlikely event of a design basis accident (DBA). The 
performance criteria are only a part of the SG Program required by 
the existing TS. The program, defined by NEI [Nuclear Energy 
Institute] 97-06, ``Steam Generator Program Guidelines,'' includes a 
framework that incorporates a balance of prevention, inspection, 
evaluation, repair, and leakage monitoring. These features will 
continue to be implemented as they are currently approved. The 
proposed changes do not, therefore, significantly increase the 
probability of an accident previously evaluated.
    The consequences of DBAs are, in part, functions of the Dose 
Equivalent 1-131 in the primary coolant and the primary to secondary 
leakage rates resulting from an accident. Therefore, limits are 
included in the TS for Operational Leakage and for Dose Equivalent 
1-131 in the primary coolant to ensure the plant is operated within 
its analyzed condition. The analysis of the limiting DBA assumes 
that the primary to secondary leak rate, after the accident, is 1 
gallon per minute with no more than 150 gallons per day in any one 
SG, and that the reactor coolant activity levels of Dose Equivalent 
1-131 are at the TS values before the accident. The proposed change 
to the SG inspection program does not affect the design of the SGs, 
their method of operation, operational leakage limits, or primary 
coolant chemistry controls. The proposed change does not adversely 
impact any other previously evaluated DBA. In addition, the proposed 
changes do not affect the consequences of a main steam line break, 
rod ejection, a reactor coolant pump locked rotor event, or other 
previously evaluated accident.
    Therefore, the proposed change does not affect the consequences 
of a[n] SG tube rupture accident and the probability of such an 
accident is unchanged.
    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 license amendment does not affect the method of 
operation of the SGs, or the primary or secondary coolant chemistry 
controls. In addition, the proposed amendment does not impact any 
other plant system or component. The change modifies existing SG 
inspection requirements based on the RSG design and the properties 
and experience associated with their improved materials. The revised 
inspection requirements will result in the same outcome that SG tube 
integrity will continue to be maintained.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The SG tubes in pressurized water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
isolate the radioactive fission products in the primary coolant from 
the secondary system. In summary, the safety function of a[n] SG is 
maintained by ensuring the integrity of its tubes. SG tube integrity 
is a function of the design, environment, and the physical condition 
of the tube. The proposed change to the SG inspection program does 
not affect tube design or operating environment. The existing SG 
Program is maintained in this change. The repair criteria that are 
being removed are specific to the existing SGs and are not 
applicable to the RSGs. If tube defects are detected that exceed 
limits in the RSGs, then the tube will be removed from service. The 
effective tube plugging percentage will continue to be tracked for 
all plugging in each SG in accordance with TS Section 6.9.1.16.1 to 
ensure the heat transfer function of the SGs is not adversely 
affected. For the above reasons, the margin of safety is not changed 
and overall plant safety will be enhanced by the proposed change to 
the TS.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, 6A West Tower, Knoxville, Tennessee 37902.
    NRC Branch Chief: Douglas A. Broaddus.

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.

Exelon Generation Company, LLC, and PSEG Nuclear, LLC, Docket No. 50-
278, Peach Bottom Atomic Power Station (PBAPS), Unit 3, York and 
Lancaster Counties, Pennsylvania

    Date of application for amendments: June 28, 2011.
    Brief description of amendment request: The proposed amendment 
would modify the PBAPS, Unit 3, Technical Specification Section 2.1.1 
to revise Safety Limit Minimum Critical Power Ratio values.
    Date of publication of individual notice in Federal Register: 
August 22, 2011 (76 FR 52357).
    Expiration date of individual notice: September 21, 2011 (public 
comments) and October 21, 2011 (hearing requests).

[[Page 55133]]

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 NRC's Public Document Room (PDR), located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. Publicly available documents created or received at the NRC are 
accessible electronically through the Agencywide Documents Access and 
Management System (ADAMS) in the NRC Library at https://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 
pdr.resource@nrc.gov.

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 application