Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 1644-1653 [2011-218]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 76, Number 7 (Tuesday, January 11, 2011)]
[Notices]
[Pages 1644-1653]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2011-218]


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NUCLEAR REGULATORY COMMISSION

[NRC-2011-0005]


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 December 16 to December 29, 2010. The last 
biweekly notice was published on December 28, 2010 (75 FR 81667).

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) 50.92, this means that operation of the facility 
in accordance with the proposed amendment would not (1) Involve a 
significant increase in the probability or consequences of an accident 
previously evaluated; or (2) create the possibility of a new or 
different kind of accident from any accident previously evaluated; or 
(3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules, 
Announcements and Directives Branch (RADB), TWB-05-B01M, Division of 
Administrative Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be faxed to the RADB at 301-492-3446. 
Documents may be examined, and/or copied for a fee, at the NRC's Public 
Document Room (PDR), located at One White Flint North, Room O1-F21, 
11555 Rockville Pike (first floor), Rockville, Maryland.
    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license. 
Requests for a hearing and a petition for leave to intervene shall be 
filed in accordance with the Commission's ''Rules of Practice for 
Domestic Licensing Proceedings'' in 10 CFR part 2. Interested person(s) 
should consult a current copy of 10 CFR 2.309, which is available at 
the Commission's PDR, located at One White Flint North, Room O1-F21, 
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly 
available records will be accessible from

[[Page 1645]]

the Agencywide Documents Access and Management System's (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, 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, any hearing held 
would take place before the issuance of any amendment.
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC E-Filing rule (72 FR 49139, 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
ten (10) days prior to the filing deadline, the participant should 
contact the Office of the Secretary by e-mail at 
hearing.docket@nrc.gov, or by telephone at (301) 415-1677, to request 
(1) a digital ID certificate, which allows the participant (or its 
counsel or representative) to digitally sign documents and access the 
E-Submittal server for any proceeding in which it is participating; and 
(2) advise the Secretary that the participant will be submitting a 
request or petition for hearing (even in instances in which the 
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the 
Secretary will establish an electronic docket for the hearing in this 
proceeding if the Secretary has not already established an electronic 
docket.
    Information about applying for a digital ID certificate is 
available on NRC's public Web site at 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 EIE, users will be required to install a Web 
browser plug-in from the NRC Web site. Further information on the Web-
based submission form, including the installation of the Web browser 
plug-in, is available on the NRC's public Web site at 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 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

[[Page 1646]]

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://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant 
to an order of the Commission, or the presiding officer. Participants 
are requested not to include personal privacy information, such as 
social security numbers, home addresses, or home phone numbers in their 
filings, unless an NRC regulation or other law requires submission of 
such information. With respect to copyrighted works, except for limited 
excerpts that serve the purpose of the adjudicatory filings and would 
constitute a Fair Use application, participants are requested not to 
include copyrighted materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Non-timely filings 
will not be entertained absent a determination by the presiding officer 
that the petition or request should be granted or the contentions 
should be admitted, based on a balancing of the factors specified in 10 
CFR 2.309(c)(1)(i)-(viii).
    For further details with respect to this license amendment 
application, see the application for amendment which is available for 
public inspection at the Commission's PDR, located at One White Flint 
North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. Publicly available records will be accessible from the ADAMS 
Public Electronic Reading Room on the Internet at the NRC Web site, 
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.

Calvert Cliffs Nuclear Power Plant, LLC, Docket No. 50-318, Calvert 
Cliffs Nuclear Power Plant, Unit 2, Calvert County, Maryland

    Date of amendment request: October 4, 2010.
    Description of amendment request: The proposed amendment revises 
Calvert Cliffs Technical Specification 5.5.16, ``Containment Leakage 
Rate Testing Program'' to allow a one-time extension of the Type A 
Integrated Leakage Rate test interval for no more than 5 years.
    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?
    No.
    This proposed one-time extension of the Type A test interval 
from 10 years to 15 years does not increase the probability of an 
accident since there are no design or operating changes involved and 
the test is not an accident initiator. The proposed extension of the 
test interval does not involve a significant increase in the 
consequences of an accident since research documented in NUREG-1493 
has found that, generically, fewer than 3% of the potential 
containment leak paths are not identified by Types B and C testing. 
Calvert Cliffs, through testing and containment inspections, also 
provides a high degree of assurance that the Containment will not 
degrade in a manner detectable only by a Type A test. Inspections 
required by the American Society of Mechanical Engineers Boiler and 
Pressure Vessel Code are performed to identify containment 
degradation that could affect leak tightness.
    Therefore, this proposed change does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No.
    This proposed one-time extension of the Type A test interval 
from 10 years to 15 years does not involve any design or operational 
changes that could lead to a new or different kind of accident from 
any accident previously evaluated. The test itself is not changing 
and will be performed after a longer interval. 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.
    Therefore, this 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?
    No.
    The proposed one-time extension of the Type A test interval from 
10 years to 15 years does not involve a significant reduction in the 
margin of safety of the containment's ability to maintain its 
integrity during a design basis accident. The generic study of the 
increase in the Type A test interval, NUREG-1493, concluded there is 
an imperceptible increase in the plant risk associated with 
extending the test interval out to 20 years. Further, the extended 
test interval would have a minimal effect on this risk since Types B 
and C testing detect 97% of potential leakage paths. For the 
requested change in the Calvert Cliffs Integrated Leakage Rate Test 
interval, it was determined that the risk contribution of leakage 
will increase 0.07% (based on change in offsite dose). This change 
is considered very small and does not represent a significant 
reduction in the margin of safety.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

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

[[Page 1647]]

Constellation Generation Group, LLC, 750 East Pratt Street, 17th floor, 
Baltimore, MD 21202.
    NRC Branch Chief: Nancy L. Salgado.

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

    Date of amendment request: November 8, 2010.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TS) to eliminate provisions allowing 
the High Pressure Coolant Injection (HPCI) system and the Reactor Core 
Isolation Cooling (RCIC) system to be aligned to the suppression pool 
when required instrument channels are inoperable. In this 
configuration, the HPCI and RICI systems would not be capable of 
mitigating some plant events. Also, an administrative change to the TS 
Table of Contents is proposed.
    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 does not significantly increase the 
probability of an accident since it does not involve a change to any 
plant equipment that initiates a plant accident. The proposed 
amendment is more restrictive than the current TS in that it no 
longer allows the HPCI and RCIC systems to be aligned to the 
suppression pool when required instrument channels are inoperable. 
The change requires HPCI and RCIC to be declared inoperable within 
one hour when the associated trip functions are not operable. The 
change also updates the TS Table of Contents. The HPCI system is 
credited to mitigate small break loss-of-coolant accidents and the 
RCIC System is not credited for accident mitigation. The proposed 
change ensures the systems are aligned consistent with station 
analysis assumptions. 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 change does not involve any physical alteration of 
plant equipment and does not change the method by which any safety-
related system performs its function. The proposed amendment is more 
restrictive than the current technical specifications in that it no 
longer allows the HPCI and RCIC systems to be aligned to the 
suppression pool when required instrument channels are inoperable. 
The change requires HPCI and RCIC to be declared inoperable within 
one hour when the associated trip functions are not operable. The 
change also updates the TS Table of Contents. No new or different 
types of equipment will be installed and the basic operation of 
installed equipment is unchanged. The methods governing plant 
operation and testing remain consistent with current safety analysis 
assumptions. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment is more restrictive than the current 
technical specifications in that it no longer allows the HPCI and 
RCIC systems to be aligned to the suppression pool when required 
instrument channels are inoperable. This ensures that safety margins 
established in station safety analysis are maintained. The proposed 
amendment does not involve a physical modification of the plant and 
does not change the design or function of any component or system. 
The proposed amendment is more restrictive than the current TS in 
that it no longer allows the HPCI and RCIC systems to be aligned to 
the suppression pool when required instrument channels are 
inoperable. The change requires the HPCI and RCIC systems to be 
declared inoperable within one hour when the associated trip 
functions are not operable. The change also updates the TS Table of 
Contents. This ensures analyzed safety margins are maintained. 
Therefore, operation of VY in accordance with the proposed amendment 
will not involve a significant reduction in the margin to safety.

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

Exelon Generation Company, LLC, Docket No. 50-219, Oyster Creek Nuclear 
Generating Station, Ocean County, New Jersey

    Date of amendment request: June 25, 2010.
    Description of amendment request: The amendment would revise the 
Oyster Creek Nuclear Generating Station Technical Specifications (TSs) 
governing actions to be taken if a single emergency diesel generator 
(EDG) is inoperable. Specifically, the proposed amendment would remove 
the requirement to test the other EDG daily. Instead, the licensee 
would be required to either test the other EDG once or determine that 
it is not inoperable due to a common cause failure.
    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 changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.]
    The proposed changes are associated with the testing 
requirements of the two Emergency Diesel Generators (EDGs). The 
changes will eliminate unnecessary EDG testing requirements that 
contribute to potential mechanical degradation of the EDGs. The 
changes are based on the NRC guidance and recommendations provided 
in Generic Letter (GL) 93-05, ``Line-Item Technical Specifications 
Improvement to Reduce Surveillance Requirements for Testing During 
Power Operation,'' and GL 94-01, ``Removal of Accelerated Testing 
and Special Reporting Requirements for Emergency Diesel 
Generators,'' and are consistent with NUREG-1433, ``Standard 
Technical Specifications, General Electric Plants, BWR/4.'' These 
proposed changes implement a recommendation promulgated in NUREG-
1366, ``Improvements To Technical Specifications Surveillance 
Requirements'' to curtail daily testing of remaining operable diesel 
generator[s] when one of the required diesel generators is 
inoperable except for when a valid concern (e.g., potential for 
common cause failure) is posed.
    The probability of an accident is not increased by these changes 
because the EDGs are not initiators of any design basis event. 
Additionally, the proposed changes do not involve any physical 
changes to plant systems, structures, or components (SSC[s]), or the 
manner in which these SSC[s] are maintained [ ]. The surveillance 
testing required for the limiting condition for operation for one 
EDG inoperable will be eliminated for the operable EDG when the 
inoperability is not due to a common cause failure. The EDG 
reliability will thereby be potentially increased by reducing the 
stresses on the EDG caused by unnecessary testing while maintaining 
the requirement to perform a single test if a common cause failure 
potentially exists. The consequences of an accident will not be 
increased because the proposed changes to the EDG surveillance 
requirements will continue to provide a high degree of assurance 
that their operability is maintained.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. [The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.]

[[Page 1648]]

    The proposed changes do not alter the physical design, safety 
limits, or safety analysis assumptions associated with the operation 
of the plant. Accordingly, the proposed changes do not introduce any 
new accident initiators, nor do they reduce or adversely affect the 
capabilities of any plant structure or system in the performance of 
their safety function.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. [The proposed changes do not involve a significant reduction 
in the margin of safety.]
    The proposed changes modify the EDG accelerated testing 
requirements, are consistent with NRC guidance, and [potentially] 
improve EDG reliability. There are no changes being made to the 
current periodic surveillance requirements. The proposed changes do 
not impact the assumptions of any design basis accident, and do not 
alter assumptions relative to the mitigation of an accident or 
transient event.
    Testing the operable EDG every day for the duration of the 
inoperable EDG inspection (i.e., 7 days) may be too excessive and 
may lead to degradation of the EDG and possibly result in [the] 
potential for unnecessary shutdowns. By reducing the possibility of 
degradation from this excessive testing, the margin of safety is 
[not significantly affected.]

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, and with the changes noted above in square 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: Mr. J. Bradley Fewell, Associate General 
Counsel, Exelon Generation Company LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Harold Chernoff.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, 
Beaver Valley Power Station, Unit 2 (BVPS-2), Beaver County, 
Pennsylvania

    Date of amendment request: February 26, 2010.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) by expanding the scope of the 
steam generator (SG) tubesheet inspections using the F* inspection 
methodology to the SG cold-leg tubesheet region for BVPS-2.
    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?
    No. The proposed change modifies the BVPS-2 Technical 
Specifications to expand the scope of steam generator [SG] tubesheet 
inspections using the F* inspection methodology to the SG cold-leg 
tubesheet region based on WCAP-16385-P, Revision 1. Of the various 
accidents previously evaluated in the BVPS-2 Updated Final Safety 
Analysis Report (UFSAR), the proposed change only affects the SG 
tube rupture (SGTR) event evaluation and the postulated steam line 
break (SLB) accident evaluation. Loss-of-coolant accident (LOCA) 
conditions cause a compressive axial load to act on the tube. 
Therefore, since the LOCA tends to force the tube into the tubesheet 
rather than pull it out, it is not a factor in this amendment 
request. Another faulted load consideration is a safe shutdown 
earthquake (SSE); however, the seismic analysis of Model 51M SGs has 
shown that axial loading of the tubes is negligible during an SSE.
    For the SGTR event, the required structural margins of the steam 
generator tubes will be maintained by the presence of the tubesheet. 
Tube rupture is precluded for cracks in the tube expansion region 
due to the constraint provided by the tubesheet. Therefore, 
Regulatory Guide (RG) 1.121, ``Bases for Plugging Degraded PWR 
[pressurized-water reactor] Steam Generator Tubes,'' margins against 
burst are maintained for both normal and postulated accident 
conditions.
    The F* length supplies the necessary resistive force to preclude 
pullout loads under both normal operating and accident conditions. 
The contact pressure results from the tube expansion process used 
during manufacturing and from the differential pressure between the 
primary and secondary side. The proposed changes do 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 an SGTR or SLB accident.
    The consequences of an SGTR event are affected by the primary-
to-secondary leakage flow during the event. Primary-to-secondary 
leakage flow through a postulated broken tube is not affected by the 
proposed change since the tubesheet enhances the tube integrity in 
the region of the expansion by precluding tube deformation beyond 
its initial expanded outside diameter. The resistance to both tube 
rupture and collapse is strengthened by the tubesheet in that 
region. At normal operating pressures, leakage from primary water 
stress corrosion cracking (PWSCC) below the F* distance 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.
    SLB leakage is limited by leakage flow restrictions resulting 
from the crack and tube-to-tubesheet contact pressures that provide 
a restricted leakage path above the indications and also limit the 
degree of crack face opening compared to free span indications. The 
total leakage (i.e., the combined leakage for all such tubes) meets 
the industry performance criterion, plus the combined leakage 
developed by any other alternate repair criteria, and will be 
maintained below the maximum allowable SLB leak rate limit, such 
that off-site doses are maintained less than 10 CFR [Title 10 of the 
Code of Federal Regulation] [Part] 100 guideline values and the 
limits evaluated in the BVPS-2 UFSAR.
    Therefore, based on the above evaluation, 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?
    No. The proposed changes do not introduce any changes or 
mechanisms that create the possibility of a new or different kind of 
accident. Tube bundle integrity will continue to be maintained for 
all plant conditions upon implementation of the F* methodology to 
the cold-leg tubesheet region.
    The proposed changes do not introduce any new equipment or any 
change to existing equipment. No new effects on existing equipment 
are created nor are any new malfunctions introduced.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The proposed changes maintain the required structural 
margins of the SG tubes for both normal and accident conditions. NRC 
Regulatory Guide (RG) 1.121 is used as the basis in the development 
of the F* methodology for determining that SG tube integrity 
considerations are maintained within acceptable limits. Regulatory 
Guide 1.121 describes a method acceptable to the NRC staff for 
meeting General Design Criteria 14, 15, 31, and 32. Regulatory Guide 
1.121 describes the limiting safe conditions of tube wall 
degradation beyond which tubes with unacceptable cracking, as 
established by inservice inspection, should be removed from service 
or repaired. 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 primarily axially oriented cracking located within the 
tubesheet, tube burst is precluded due to the presence of the 
tubesheet. WCAP-16385-P, Revision 1, defines a length, F*, of 
degradation-free expanded tubing that provides the necessary 
resistance to tube pullout due to the pressure-induced forces (with 
applicable safety factors applied). Expansion of the application of 
the F* criteria to the cold-leg tubesheet region 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 F* 
criteria.
    Plugging of the steam generator tubes reduces the reactor 
coolant flow margin for core cooling. Expansion of the F* 
methodology to the cold-leg tubesheet region at BVPS-2 will result 
in maintaining the

[[Page 1649]]

margin of flow that may have otherwise been reduced by tube 
plugging.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The Nuclear Regulatory Commission (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 W. Jenkins, FirstEnergy Nuclear 
Operating Company, FirstEnergy Corporation, 76 South Main Street, 
Akron, OH 44308.
    NRC Branch Chief: Nancy L. Salgado.

FirstEnergy Nuclear Operating Company (FENOC), et al., Docket No. 50-
440, Perry Nuclear Power Plant, Unit No. 1 (PNPP), Lake County, Ohio

    Date of amendment request: October 21, 2010.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) 2.1.1, ``Reactor Core SLs,'' by 
incorporating revised safety limit minimum critical power ratio 
(SLMCPR) values resulting from a plant-specific analysis performed for 
PNPP Cycle 14 core.
    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 SLMCPR value will continue to ensure that during 
normal operation and abnormal operational transients, at 99.9 
percent of all fuel rods in the core do not experience transition 
boiling if the limit is not violated, thereby preserving the fuel 
cladding integrity. The proposed TS changes do not involve any 
modifications or operational changes to system, structures, or 
components (SSC). The proposed TS changes do not affect any 
postulated accident precursors, do not affect any accident 
mitigating systems, and do no introduce any new accident initiation 
mechanisms. Therefore, the proposed TS changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed TS changes do not involve any new modes of 
operation, any changes to setpoints, or any plant modifications. The 
proposed SLMCPR values do not result in the creation of any new 
precursors to an accident. Therefore, the proposed TS changes do not 
create the possibility of an accident of a different kind than 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed SLMCPR value will continue to ensure that during 
normal operation and abnormal operational transients, at 99.9 
percent of all fuel rods in the core do not experience transition 
boiling if the limit is not violated, thereby preserving the fuel 
cladding integrity. The proposed TS changes do involve modifications 
or operational changes that could adversely affect the function or 
performance of a SSC. The proposed TS changes do not affect any 
postulated accident precursors, do not affect any accident 
mitigating systems, and do not introduce any new accident initiation 
mechanisms. Therefore, the proposed TS changes do not involve a 
significant reduction in margin of safety.

    The U.S. Nuclear Regulatory Commission (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 W. Jenkins, Attorney, FirstEnergy 
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Robert D. Carlson.

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

    Date of amendment request: August 5, 2010.
    Description of amendment request: The proposed amendment would 
modify the Callaway Plant, Unit 1, Technical Specifications (TS) by 
relocating specific surveillance frequencies to a licensee-controlled 
program with the guidance of Nuclear Energy Institute (NEI) 04-10, 
``Risk-Informed Technical Specifications Initiative 5b, Risk-Informed 
Method for Control of Surveillance Frequencies.''
    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 any accident previously 
evaluated?
    Response: No.
    The proposed change relocates the specified frequencies for 
periodic surveillance requirements to licensee control under a new 
Surveillance Frequency Control Program [(SFCP)]. Surveillance 
frequencies are not an initiator to any accident previously 
evaluated. As a result, the probability of any accident previously 
evaluated is not significantly increased. The systems and components 
required by the technical specifications for which the surveillance 
frequencies are relocated are still required to be operable, meet 
the acceptance criteria for the surveillance requirements, and be 
capable of performing any mitigation function assumed in the 
accident analysis. As a result, the consequences of any accident 
previously evaluated are not significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No.
    No new or different accidents result from utilizing the proposed 
change. The changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements. The changes do not alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumptions and current plant operating practice.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    The design, operation, testing methods, and acceptance criteria 
for systems, structures, and components (SSCs), specified in 
applicable codes and standards (or alternatives approved for use by 
the NRC) will continue to be met as described in the plant licensing 
basis (including the Final Safety Analysis Report and Bases to TS), 
since these are not affected by changes to the surveillance 
frequencies. Similarly, there is no impact to safety analysis 
acceptance criteria as described in the plant licensing basis. To 
evaluate a change in the relocated surveillance frequency, [the 
licensee] will perform a probabilistic risk evaluation using the 
guidance contained in NRC approved NEI 04-10, Rev. 1 in accordance 
with the TS SFCP. NEI 04-10, Rev. 1, methodology provides reasonable 
acceptance guidelines and methods for evaluating the risk increase 
of proposed changes to surveillance frequencies consistent with 
Regulatory Guide 1.177.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the

[[Page 1650]]

amendment request involves no significant hazards consideration.
    Attorney for licensee: John O'Neill, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: Michael T. Markley.

ZionSolutions LLC, Docket Nos. 50-295 and 50-304, Zion Nuclear Power 
Station (Zion), Units 1 and 2, Lake County, Illinois

    Date of amendment request: November 15, 2010.
    Description of amendment request: The proposed amendments would 
delete license conditions that impose specific requirements for the 
decommissioning trust agreement. In lieu of the license conditions, 
ZionSolutions will directly implement the requirements of 10 CFR 
50.75(h)(1) through (h)(3). ZionSolutions will provide a revised trust 
agreement as required by 10 CFR 50.75(h)(1)(iii) within 60 days of NRC 
approval of this proposal. The licensee has stated that the trust 
agreement will conform with 10 CFR 50.75(h) and ZionSolutions will take 
no action under the existing trust agreement in the interim that would 
be inconsistent with the provisions of the regulation.
    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 change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendments alter the requirements for the 
decommissioning trust fund. These revisions of the financial 
assurance requirements do not involve any changes to any structures, 
systems or components (SSCs) or any method of operation, maintenance 
or testing. The proposed amendments will continue to provide 
assurance that adequate decommissioning funding is maintained. 
Changes to the terms of the trust fund will not alter previously 
evaluated Defueled Safety Analysis Report (DSAR) design basis 
accident assumptions, add any accident initiators, or affect the 
function of the plant SSCs as to how they are operated, maintained, 
modified, tested, or inspected.
    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 evaluated?
    Response: No.
    Implementation of the proposed changes to decommissioning trust 
fund requirements will have no impact upon the design function of 
any SSC. Modifying the precise language of the administrative 
controls on the fund in the trust agreement does not result in the 
need for any new or different DSAR design basis accident analyses. 
It does not introduce new equipment that could create a new or 
different kind of accident, and no new equipment failure modes are 
created. As a result, no new accident scenarios, failure mechanisms, 
or limiting single failures are introduced as a result of the 
proposed amendments.
    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 margin of safety is associated with the confidence in the 
ability of the fission product barriers to limit the level of 
radiation to the public. The proposed amendments would not alter any 
SSC functions and would not alter the way the plant is operated. The 
amendments do not alter the way in which financial assurance for 
decommissioning is achieved. The proposed amendments would not 
introduce any new uncertainties associated with any safety limit. 
The proposed amendments would have no impact upon the structural 
integrity of the fuel cladding or any other barrier to fission 
product release. There would be no reduction in the effectiveness of 
the fission product barriers to limit the level of radiation to the 
public. 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: Russ Workman, Deputy General Counsel, 
EnergySolutions, 423 West 300 South, Suite 200, Salt Lake City, UT 
84101.
    NRC Branch Chief: Bruce Watson.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management System (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, 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.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos. 
1, 2, and 3, Maricopa County, Arizona

    Date of application for amendment: April 8, 2010.
    Brief description of amendment: The amendments deleted redundant 
reporting and operational restriction provisions from Technical 
Specification (TS) Section 2.2, ``Safety Limit Violations,'' consistent 
with Technical Specification Task Force (TSTF) change traveler TSTF-5-
A, Revision 1, ``Delete Safety Limit Violation Notification 
Requirements,'' and replaced plant-specific titles with generic titles 
in TS Section 5.2.1, ``Onsite and Offsite Organizations,'' consistent 
with TSTF-65-A, Revision 1, ``Use of Generic Titles for Utility 
Positions.''
    Date of issuance: December 29, 2010.

[[Page 1651]]

    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: Unit 1--183; Unit 2--183; Unit 3--183.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: July 27, 2010 (75 FR 
44022).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 29, 2010.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 21, 2009, as supplemented 
March 3 and July 28, 2010.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) Section 6.9.1.6 to add NRC approved Topical Report 
(TR) EMF-2310(P)(A), ``SRP Chapter 15 Non-LOCA Methodology for 
Pressurized Water Reactors,'' to the Core Operating Limits Report 
methodologies list. This change will allow the use of thermal-hydraulic 
analysis code S-RELAP5 for Final Safety Analysis Report (FSAR) Chapter 
15 non-loss-of-coolant accident (LOCA) transients in the HNP safety 
analyses. TR EMF-2310(P)(A), Revision 0, was approved by the NRC on May 
11, 2001, for the application of the S-RELAP5 thermal-hydraulic 
analysis computer code to FSAR Chapter 15 non-LOCA transients. EMF-
2310(P)(A), Revision 1, approved by the NRC on May 19, 2004, updated 
Section 5.6 of the TR.
    Date of issuance: December 23, 2010.
    Effective date: Effective as of the date of issuance and shall be 
implemented within 60 days.
    Amendment No.: 135.
    Renewed Facility Operating License No. NPF-63: The amendment 
revises the TSs and facility operating license.
    Date of initial notice in Federal Register: November 10, 2009 (74 
FR 58060). The supplements dated March 3, and July 28, 2010, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a safety evaluation dated December 23, 2010.
    No significant hazards consideration comments received: No.

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: December 14, 2009, as 
supplemented by letters dated September 8, 2010, and October 28, 2010.
    Brief description of amendments: The amendments revised the 
Technical Specifications by revising Surveillance Requirements 3.8.4.3 
and 3.8.4.6. These TS SRs address battery connection resistance values.
    Date of issuance: December 20, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 262, 258.
    Renewed Facility Operating License Nos. NPF-35 and NPF-52: 
Amendments revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: August 10, 2010 (75 FR 
48375). The supplements dated September 8, 2010, and October 28, 2010, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 20, 2010.
    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: December 14, 2009, as 
supplemented by letters dated September 8, 2010, and October 28, 2010.
    Brief description of amendments: The amendments revised the 
Technical Specifications by revising Surveillance Requirements 3.8.4.2 
and 3.8.4.5. These TS SRs address battery connection resistance values.
    Date of issuance: December 20, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 260, 240.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: August 10, 2010 (75 FR 
48375). The supplements dated September 8, 2010, and October 28, 2010, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 20, 2010.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2 (Braidwood), Will County, Illinois 
Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2 
(Byron), Ogle County, Illinois

    Date of application for amendment: December 16, 2009, as 
supplemented by letters dated April 26 and October 25, 2010.
    Brief description of amendment: The amendments revise Technical 
Specifications Section 5.6.5, ``Core Operating Limits Report,'' to 
replace the existing reference for the large break loss-of-coolant 
accident (LOCA) analysis methodology with a reference to WCAP-16009-P-
A, Revision 0, ``Realistic Large Break LOCA Evaluation Methodology 
Using the Automated Statistical Treatment of Uncertainty Method,'' 
January 2005.
    Date of issuance: December 21, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: Braidwood Unit 1--164; Braidwood Unit 2--164; Byron 
Unit No. 1--170; and Byron Unit No. 2--170.
    Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66: 
The amendments revise the TSs and Licenses.
    Date of initial notice in Federal Register: February 23, 2010 (75 
FR 8141). The supplemental letters dated April 26, and October 25, 
2010, contained clarifying information, did not change the initial no 
significant hazards consideration determination, and did not expand the 
scope of the original Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 21, 2010.

[[Page 1652]]

    No significant hazards consideration comments received: No.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit 1 and 2, St.. Lucie County, Florida.

    Date of application for amendments: December 14, 2009, as 
supplemented on July 30, 2010.
    Brief description of amendments: Amendment modifies Technical 
Specification (TS) 3/4 .4.10 ``Structural Integrity,'' in Unit 1 (TS 3/
4.4.11 in Unit 2), TS 3.3.3.8, ``Accident Monitoring Instrumentation,'' 
in Unit 1 (TS 3.3.3.6 in Unit 2), TS 6.4.1, ``Training,'' in Units 1 
and 2, and several administrative changes in the TSs for both units . 
The changes delete the Structural Integrity TS, update Accident 
Monitoring Instrumentation requirements and make various administrative 
TS changes.
    Date of Issuance: December 28, 2010.
    Effective Date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 210, 159.
    Renewed Facility Operating License Nos. DPR-67 and NPF-16: 
Amendments revised the TSs.
    Date of initial notice in Federal Register: April 20, 2010 (75 FR 
20638). The supplement dated July 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 as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 28, 2010.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket Nos. 50-354, 50-272 and 50-311, Hope Creek 
Generating Station and Salem Nuclear Generating Station, Unit 1 and 2, 
Salem County, New Jersey

    Date of application for amendments: March 25, 2010.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TSs) associated with reactor coolant system 
(RCS) structural integrity requirements for Hope Creek Generating 
Station (HCGS) and Salem Nuclear Generating Station (Salem), Unit Nos. 
1 and 2. Specifically, the amendments revise the TSs to: (1) Delete the 
RCS structural integrity requirements contained in HCGS TS 3/4.4.8, 
Salem Unit 1 TS 3/4.4.10, and Salem Unit 2 TS 3/4.4.11; (2) relocate 
the augmented inservice inspection requirements for the reactor coolant 
pump flywheel, currently contained in Salem Unit 1 surveillance 
requirement (SR) 4.4.10.1.1 and Salem Unit 2 SR 4.4.11.1, to a new 
program in TS 6.8.4.k; and (3) delete the augmented inservice 
inspection program requirements for the steam generator channel heads 
currently contained in Salem Unit 1 SR 4.4.10.1.2.
    Date of issuance: December 15, 2010.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment Nos.: 186, 298 and 281.
    Facility Operating License Nos. NPF-57, DPR-70 and DPR-75: The 
amendments revised the TSs and the Licenses.
    Date of initial notice in Federal Register: June 15, 2010 (75 FR 
33843).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 15, 2010.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 26, 2010 (TS 09-05).
    Brief description of amendment: The amendments revised the 
Technical Specification (TS) Table 3.3-1, ``Reactor Trip System 
Instrumentation,'' Functional Unit 5, ``Intermediate Range, Neutron 
Flux,'' to resolve an oversight regarding the operability requirements 
for the intermediate range neutron flux channels. The amendments added 
an action to TS Table 3.3-1 to define that the provisions of 
Specification 3.0.3 are not applicable above 10 percent of thermal 
rated power with the number of operable intermediate range neutron flux 
channels two less than the minimum channels operable requirement.
    Date of issuance: December 21, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 328, 321.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revised the License and Technical Specifications