Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations; Biweekly Notice, 39975-39985 [2010-16879]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 75, Number 133 (Tuesday, July 13, 2010)]
[Notices]
[Pages 39975-39985]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2010-16879]


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NUCLEAR REGULATORY COMMISSION

[NRC-2010-0248]


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

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 June 17, 2010, to June 30, 2010. The last 
biweekly notice was published on June 29, 2010 (75 FR 37471).

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), Sec.  50.92, this means that operation of the 
facility in accordance with the proposed amendment would not: (1) 
Involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules, 
Announcements and Directives Branch (RADB), TWB-05-B01M, Division of 
Administrative Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be faxed to the RADB at 301-492-3446. 
Documents may be examined, and/or copied for a fee, at the NRC's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license. 
Requests for a hearing and a petition for leave to intervene shall be 
filed in accordance with the Commission's ``Rules of Practice for 
Domestic Licensing Proceedings'' in 10 CFR part 2. Interested person(s) 
should consult a current copy of 10 CFR 2.309, which is available at 
the Commission's PDR, located at One White Flint North, Public File 
Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, 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

[[Page 39976]]

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. e.t. 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 (866) 672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., e.t., 
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

[[Page 39977]]

their filings, unless an NRC regulation or other law requires 
submission of such information. With respect to copyrighted works, 
except for limited excerpts that serve the purpose of the adjudicatory 
filings and would constitute a Fair Use application, participants are 
requested not to include copyrighted materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Non-timely filings 
will not be entertained absent a determination by the presiding officer 
that the petition or request should be granted or the contentions 
should be admitted, based on a balancing of the factors specified in 10 
CFR 2.309(c)(1)(i)-(viii).
    For further details with respect to this license amendment 
application, see the application for amendment which is available for 
public inspection at the Commission's PDR, located at One White Flint 
North, Public File Area O1F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the ADAMS Public Electronic Reading Room on the Internet at the NRC Web 
site, 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.

Duke Energy Carolinas, LLC, et al., Docket No. 50-414, Catawba Nuclear 
Station, Unit 2, York County, South Carolina

    Date of amendment request: April 28, 2010.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 5.5.9 to exclude portions of the Steam 
Generator (SG) tube from periodic SG tube inspections and plugging or 
repair. In addition, reporting requirement changes are proposed to TS 
5.6.8. This submittal is requesting a one-cycle approval for the 
Catawba Nuclear Station, Unit 2, End of Cycle 17 Refueling Outage and 
subsequent Cycle 18 operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Criterion 1: Does the proposed amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response: No. The proposed changes to TS 5.5.9, TS 5.6.8, and 
the Facility Operating License have no significant effect upon 
accident probabilities or consequences. Of the various accidents 
previously evaluated, the following are limiting with respect to the 
proposed changes as discussed in this amendment request:
     SG Tube Rupture evaluation
     Steam Line Break/Feed Line Break evaluation
     Locked Rotor evaluation
     Control Rod Ejection evaluation
    Loss of Coolant Accident conditions cause a compressive axial 
load to act on the tube. Therefore, since this accident 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; however, the seismic 
analysis of Model D5 SGs (the SGs at Catawba) has shown that axial 
loading of the tubes is negligible during this event.
    At normal operating pressures, leakage from Primary Water Stress 
Corrosion Cracking (PWSCC) below 16.95 inches from the top of the 
tubesheet 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.
    For the SG Tube Rupture event, tube rupture is precluded for 
cracks in the hydraulic expansion region due to the constraint 
provided by the tubesheet. Therefore, the margin against tube burst/
pullout is maintained during normal and postulated accident 
conditions and the proposed change does not result in a significant 
increase in the probability of a tube rupture. SG Tube Rupture 
consequences are not affected by the primary to secondary leakage 
flow during the event, as primary to secondary leakage flow through 
a postulated tube that has been pulled out of the tubesheet is 
essentially equivalent to that from a severed tube. Therefore, the 
proposed change does not result in a significant increase in the 
consequences of a tube rupture.
    The probability of a Steam Line Break/Feed Line Break, Locked 
Rotor, and Control Rod Ejection are not affected by the potential 
failure of a SG tube, as the failure of a tube is not an initiator 
for any of these events. In WCAP-17072-P, leakage is modeled as flow 
through a porous medium via the use of the Darcy equation. The 
leakage model is used to develop a relationship between operational 
leakage and leakage at accident conditions that is based on 
differential pressure across the tubesheet and the viscosity of the 
fluid. A leak rate ratio was developed to relate the leakage at 
operating conditions to leakage at accident conditions. The fluid 
viscosity is based on fluid temperature and it has been shown that 
for the most limiting accident, the fluid temperature does not 
exceed the normal operating temperature. Therefore, the viscosity 
ratio is assumed to be 1.0 and the leak rate ratio is a function of 
the ratio of the accident differential pressure and the normal 
operating differential pressure.
    The leakage factor of 2.65 for Catawba Unit 2 for a postulated 
Steam Line Break/Feed Line Break has been calculated as shown in 
WCAP-17072-P, as supplemented. The leakage factor has been increased 
to 3.27 per additional Westinghouse analysis specific to Catawba. 
Therefore, Catawba Unit 2 will apply a factor of 3.27 to the normal 
operating leakage associated with the tubesheet expansion region in 
the Condition Monitoring assessment and Operational Assessment. 
Through application of the limited tubesheet inspection scope, the 
proposed operating leakage limit provides assurance that excessive 
leakage (i.e., greater than accident analysis assumptions) will not 
occur. No leakage factor will be applied to the Locked Rotor or 
Control Rod Ejection due to their short duration, since the 
calculated leak rate ratio is less than 1.0. Therefore, the proposed 
change does not result in a significant increase in the consequences 
of these accidents.
    For the Condition Monitoring assessment, the component of 
leakage from the prior cycle from below the H* distance will be 
multiplied by a factor of 3.27 and added to the total leakage from 
any other source and compared to the allowable accident induced 
leakage limit. For the Operational Assessment, the difference in the 
leakage between the allowable leakage and the accident induced 
leakage from sources other than the tubesheet expansion region will 
be divided by 3.27 and compared to the observed operational leakage.
    Based on the above, the performance criteria of NEI 97-06, 
Revision 2 and RG [Regulatory Guide] 1.121 continue to be met and 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.

    Criterion 2: Does the proposed amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response: No. The proposed changes to TS 5.5.9, TS 5.6.8, and 
the Facility Operating License do not introduce any changes or 
mechanisms that create the possibility of a new or different kind of 
accident. Tube bundle integrity is expected to be maintained for all 
plant conditions upon implementation of the one-cycle alternate 
repair criteria. The proposed change does 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, based on the above evaluation, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.

    Criterion 3: Does the proposed amendment involve a significant 
reduction in a margin of safety?
    Response: No. The proposed changes to TS 5.5.9, TS 5.6.8, and 
the Facility Operating License maintain the required structural 
margins of the SG tubes for both normal and accident conditions. NEI 
97-06, Revision 2 and RG 1.121 are used as the basis in the 
development of the limited tubesheet inspection depth methodology 
for determining that SG tube integrity considerations are maintained 
within acceptable limits. RG 1.121 describes a

[[Page 39978]]

method acceptable to the NRC staff for meeting GDC [General Design 
Criteria] 14, 15, 31, and 32 by reducing the probability and 
consequences of a SG Tube Rupture. RG 1.121 concludes that by 
determining the limiting safe conditions for tube wall degradation, 
the probability and consequences of a SG Tube Rupture are reduced. 
This RG uses safety factors on loads for tube burst that are 
consistent with the requirements of Section III of the ASME 
[American Society of Mechanical Engineers] Code [Boiler and Pressure 
Vessel Code].
    For axially oriented cracking located within the tubesheet, tube 
burst is precluded due to the presence of the tubesheet. For 
circumferentially oriented cracking, WCAP-17072-P defines a length 
of degradation-free expanded tubing that provides the necessary 
resistance to tube pullout due to the pressure-induced forces, with 
applicable safety factors applied. Application of the limited hot 
and cold leg tubesheet inspection criteria will preclude 
unacceptable primary to secondary leakage during all plant 
conditions. The methodology for determining leakage as described in 
WCAP-17072-P shows that significant margin exists between an 
acceptable level of leakage during normal operating conditions that 
ensures meeting the accident induced leakage assumption and the TS 
leakage limit.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Kate Nolan, Associate General Counsel 
and Managing Attorney, Duke Energy Carolinas, LLC, 422 South Church 
Street, Mail Code--EC07H, P.O. Box 1244, Charlotte, NC 28201-1244.
    NRC Branch Chief: Gloria Kulesa.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Miami-Dade County, Florida

    Date of amendment request: June 25, 2009, as supplemented May 21, 
2010.
    Description of amendment request: Revise the licensing bases to 
adopt the alternative source term as allowed in Title 10 of the Code of 
Federal Regulations, Sec.  50.67.
    An application that addressed similar issues was previously 
submitted on June 25, 2009, and noticed in the Federal Register (FR) on 
December 29, 2009 (74 FR 68870). Due to certain changes in the 
specifics stated in the May 21, 2010, supplement, from those proposed 
in the June 25, 2009, application, this is a renotice that includes 
those changes. Below is the no significant hazards consideration 
determination (NSHCD) for the changes in the May 21, 2010 supplemental. 
The original NSHCD as published in the FR December 29, 2009, still 
applies to the June 25, 2009 application.
    Basis for proposed NSHCD: 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 amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes to [technical specification] TS 4.4.8 will 
only provide for better assurance of required sampling and analysis 
of the reactor coolant system specific activity during thermal power 
changes and transient conditions (MODES 1, 2, 3, and 4). This will 
ensure potential consequences of a [defined-basis accident] DBA are 
bounded by the approved accident analyses.
    The proposed changes to TS 3/4.7.5 itemize the system 
operability requirements and appropriate actions in the event that 
those requirements are not satisfied. These actions include actions 
to be taken during the allowed outage times (AOTs) specified in the 
actions to bring the system back into compliance with the system 
operability requirements. The actions also provide for restoration 
of the inoperable component or in some cases provide for placing and 
maintaining it in a safe condition until it can be restored. The 
actions may include compensatory measures that require initiation of 
mitigating actions involving operator action to manually align and 
place into service a compensatory filtration unit in the event that 
the normal filtration train is out-of-service. These compensatory 
measures are required to be taken within 24 hours compared to the 
current allowed outage time of 84 hours for system inoperability 
without any compensatory measures specified. Moreover, consistent 
with the current Turkey Point TS and TSTF-448 AOTs, manually 
aligning the compensatory filter within 24 hours to maintain 
[control room emergency ventilation system] CREVS operability is 
acceptable in order to ensure control room operations will be 
protected from analyzed radiological hazards. The other action 
statements for inoperability of a redundant active component provide 
for an AOT of 7 days consistent with the Westinghouse Standard 
Technical Specification. They are based on the low probability of 
occurrence of a DBA challenging the Control Room Habitability during 
this time period and the continued capability of the remaining 
system components to perform the required CREVS safety function.
    The proposed changes have no effect on the probability of an 
accident previously evaluated as they do not affect any accident 
initiators. The proposed changes have no significant effect on the 
consequences of an accident previously evaluated as they either 
provide for better monitoring of plant operating parameters or for 
compensatory actions to be taken for out-of-service equipment not 
previously available. Design changes to enhance the system 
capabilities will be made to the same design and quality standards 
as the existing CREVS. System modifications required to support 
these proposed changes are evaluated under the 10 CFR 50.59 program 
and are enhancements to the mitigation strategies.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes to TS 4.4.8 will only provide for better 
assurance of required sampling and analysis of the reactor coolant 
system specific activity during MODES 1, 2, 3, and 4. The proposed 
modifications to the plant configuration will be fully qualified to 
the appropriate design requirements to assure their required 
function is available for accident mitigation. Additionally, 
functions of other equipment required for accident mitigation are 
also not adversely impacted. Design changes to enhance the system 
capabilities will be made to the same design and quality standards 
as the existing CREVS. The proposed changes to TS 3/4.7.5 will 
provide for better specification of system operability requirements 
and appropriate actions in the event that those requirements are not 
satisfied. The proposed changes have no effect on accident 
precursors or initiators and only enhance mitigation capabilities 
with regard to protecting control room personnel from radiological 
hazards.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The proposed changes to TS 4.4.8 will only provide for better 
assurance of required sampling and analysis of the reactor coolant 
system specific activity during thermal power changes and transient 
conditions (MODES 1, 2, 3, and 4). No plant system or component 
design or operational requirements are affected by these changes.
    The proposed changes to TS 3/4.7.5 will provide for better 
specification of system operability requirements and appropriate 
actions in the event that those requirements are not satisfied. The 
proposed increase in the specified AOT for inoperability of CREVS 
components from 84 hours to 7 days is considered insignificant as it 
is consistent with the Westinghouse Standard Technical Specification 
and based on the low probability of occurrence of a DBA challenging 
the Control Room Habitability during this time period and the 
continued capability of the remaining system components to perform 
the required CREVS safety function. Moreover, consistent with the 
current Turkey Point TS and TSTF-448 AOTs, manually aligning the 
compensatory filter within 24 hours to maintain CREVS operability is 
an acceptable margin of safety to ensure control room operations 
will be protected from analyzed radiological hazards. The proposed 
changes provide for compensatory actions to be taken for out-of-
service equipment that were not previously

[[Page 39979]]

available and thus enhance existing mitigation capabilities with 
regard to protecting control room personnel from radiological 
hazards.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.
    Based on the above discussion, FPL has determined that the 
proposed change does not involve a significant hazards 
consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of Sec.  50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Branch Chief: Douglas A. Broaddus.

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

    Date of amendment request: May 14, 2010.
    Description of amendment request: The proposed changes would revise 
the Seabrook Station Technical Specifications (TSs) governing the 
Containment Enclosure Emergency Air Cleanup System. Specifically, the 
proposed change would insert a requirement that if both trains of the 
system are inoperable, at least one train must be returned to operable 
status within 24 hours or begin a shutdown of the reactor. Currently, 
since there are no limiting conditions for operation proscribed actions 
in the event two trains are inoperable, TS 3.0.3 requires a shutdown 
within 6 hours.
    Basis for proposed NSHC determination: As required by 10 CFR 
50.91(a), the licensee has provided its analysis of the issue of NSHC, 
which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change 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 changes 
neither adversely affect accident initiators or precursors, nor 
alter design assumptions. The proposed changes do not alter or 
prevent the ability of operable [SSCs] to perform their intended 
function to mitigate the consequences of an initiating event within 
the assumed acceptance limits.
    This change is a revision to the technical specifications 
(TS[s]) for the containment enclosure emergency air cleanup system 
(CEEACS), which is a mitigation system designed to prevent 
uncontrolled releases of radioactivity into the environment. The 
change would allow intermittent opening of the containment enclosure 
boundary under administrative controls. These controls would ensure 
that the opening will be quickly sealed to maintain the validity of 
the licensing basis analyses of accident consequences. The proposed 
change adds a new action requirement that would allow 24 hours to 
restore the containment enclosure boundary in the event that both 
trains of the CEEACS are inoperable due to an inoperable containment 
enclosure boundary. The proposed 24 hour completion time is 
reasonable based on the low probability of a design basis accident 
occurring during this time period and the use of preplanned 
compensatory measures. The CEEACS is not an initiator or precursor 
to any accident previously evaluated. Therefore, the probability of 
any accident previously evaluated is not increased.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change will not impact the accident analysis. The 
changes will not alter the requirements of the CEEACS or its 
function during accident conditions, and no new or different 
accidents result from the proposed changes to the TS[s]. 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 significant 
change in the method of plant operation. The changes do not alter 
assumptions made in the safety analysis. Therefore, this request 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel cladding, reactor coolant 
system pressure boundary, and containment structure) to limit the 
level of radiation dose to the public. The proposed changes do 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, will not relax any safety system 
settings, and will not relax the bases for any limiting conditions 
for operation. The safety analysis acceptance criteria are not 
affected by this change. The proposed change will not result in 
plant operation in a configuration outside the design bases. The 
proposed change does not adversely affect systems that respond to 
safely shutdown the plant and to maintain the plant in a safe 
shutdown condition. Therefore, these proposed changes do not involve 
a significant reduction in a margin of safety.

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

Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket No. 50-220, Nine 
Mile Point Nuclear Station Unit No. 1 (NMP1), Oswego County, New York

    Date of amendment request: March 18, 2010.
    Description of amendment request: The proposed amendment would 
revise the NMP1 Technical Specifications (TSs) for inoperable snubbers 
by removing TS 3/4.6.4, ``Shock Suppressors (Snubbers),'' and would 
also add a new Limiting Condition for Operation (LCO) 3.0.8.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to relocate TS 3/4.6.4 to station procedures 
is administrative in nature and does not involve the modification of 
any plant equipment or affect basic plant operation. Snubber 
operability and surveillance requirements will be contained in the 
station procedures to ensure design assumptions for accident 
mitigation are maintained.
    The proposed change to add LCO 3.0.8 allows a delay time for 
entering a supported system TS when the inoperability is due solely 
to an inoperable snubber if risk is assessed and managed. Entrance 
into TS actions or delaying entrance into actions is not an 
initiator of any accident previously evaluated. Consequently, the 
probability of an accident previously evaluated is not significantly 
increased. The consequences of an accident while relying on 
allowance provided by proposed LCO 3.0.8 are no different than the 
consequences of an accident while relying on the current TS required 
actions in effect without the allowance provided by proposed LCO 
3.0.8.
    Revision of TS Table of Contents to reflect deletion of TS 3/
4.6.4 is administrative in nature and therefore does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of

[[Page 39980]]

accident from any accident previously evaluated?
    Response: No.
    The proposed change to relocate TS 3/4.6.4 to station procedures 
is administrative and does not involve any physical alteration of 
plant equipment. The proposed change does not change the method by 
which any safety related system performs its function. As such, 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.
    The proposed change to add LCO 3.0.8 does not involve a physical 
alteration of the plant (no new or different type of equipment will 
be installed). Allowing delay times for entering supported system 
TSs when inoperability is due solely to inoperable snubbers, if risk 
is assessed and managed, will not introduce new failure modes or 
effects.
    Revision of TS Table of Contents to reflect deletion of TS 3/
4.6.4 is administrative in nature and therefore does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change to relocate TS 3/4.6.4 to station procedures 
is administrative in nature, does not negate any existing 
requirement, and does not adversely affect existing plant safety 
margins or the reliability of the equipment assumed to operate in 
the safety analysis. As such, there are no changes being made to 
safety analysis assumptions, safety limits or safety system settings 
that would adversely affect plant safety as a result of the proposed 
change. Margins of safety are unaffected by requirements that are 
retained, but relocated from the TSs to station procedures.
    The proposed change to add LCO 3.0.8 to TSs allows a delay time 
before declaring supported TS systems inoperable when the associated 
snubber(s) cannot perform the required safety function. The proposed 
change retains an allowance in the current NMPI TSs while upgrading 
it to be more conservative for snubbers supporting multiple trains 
or sub-systems of an associated system. The updated TS will continue 
to provide an adequate margin of safety for plant operation upon 
incorporation of LCO 3.0.8. The station design and safety analysis 
assumptions provide margin in the form of redundancy to account for 
periods of time when system capability is reduced.
    Revision of TS Table of Contents to reflect deletion of TS 3/
4.6.4 is administrative in nature and therefore does not involve a 
significant reduction in a margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Carey W. Fleming, Senior Counsel, 
Constellation Energy Nuclear Group, LLC, 100 Constellation Way, Suite 
200C, Baltimore, MD 21202.
    NRC Branch Chief: Nancy L. Salgado.

Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket No. 50-220, Nine 
Mile Point Nuclear Station Unit No. 1 (NMP1), Oswego County, New York

    Date of amendment request: March 22, 2010.
    Description of amendment request: The proposed amendment would 
revise the NMP1 Technical Specifications (TSs) section 4.3.7 
``Containment Spray System,'' by modifying the testing frequency for 
the Surveillance Requirement (SR) 4.3.7.b, ``Nozzles,'' from ``at least 
once per operating cycle * * * '' to ``following maintenance that could 
result in nozzle blockage.'' Additional wording changes would be made 
to the SR to make it more consistent with the corresponding Standard 
TS, SR 3.6.1.7.4.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change modifies the SR to verify that the 
Containment Spray System (CSS) drywell and torus spray nozzles are 
unobstructed after maintenance that could introduce material 
resulting in nozzle blockage. The requirement to test the headers 
will be removed as well as the type of test to be used. Since the 
opening within the pipes is much larger than the nozzles, they are 
not likely to become obstructed unless the nozzles become 
obstructed. The spray nozzles and headers are not assumed to be 
initiators of any previously analyzed accident. Therefore, the 
proposed change does not increase the probability of any accident 
previously evaluated. The spray nozzles are used in the accident 
analyses to mitigate design basis accidents. The revised SR to 
verify system operability following maintenance is considered 
adequate to ensure operability of the CSS. Since the system will 
still be able to perform its accident mitigation function, the 
consequences of accidents previously evaluated are not 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 amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change revises the SR to verify that the CSS 
nozzles are unobstructed after maintenance that could result in 
nozzle blockage. The requirement to test the headers will be removed 
as well as the type of test to be used. The spray nozzles and 
headers are not assumed to be initiators of any previously analyzed 
accident. The change does not introduce a new mode of plant 
operation and does not involve a physical modification to the plant. 
The change will not introduce new accident initiators or impact the 
assumptions made in the safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change revises the frequency for performance of the 
SR to verify that the CSS nozzles are unobstructed. The frequency is 
changed from ``once per operating cycle'' to ``following maintenance 
that could result in nozzle blockage.'' The requirement to test the 
headers will be removed as well as the type of test to be used. The 
revised testing requirement, along with the foreign material 
exclusion program, the normal environmental conditions for the 
system, and the remote physical location of the spray nozzles, 
provide assurance that the spray nozzles and headers will remain 
unobstructed. As the spray nozzles and headers are expected to 
remain unobstructed and able to perform their post-accident 
mitigation function, plant safety is not significantly affected.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Carey W. Fleming, Senior Counsel, 
Constellation Energy Nuclear Group, LLC, 100 Constellation Way, Suite 
200C, Baltimore, MD 21202.
    NRC Branch Chief: Nancy L. Salgado.

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

    Date of amendment request: March 30, 2010, as supplemented on June 
1, 2010.

[[Page 39981]]

    Description of amendment request: The proposed amendment would 
revise the NMP2 Technical Specification (TS) section 3.8.1, ``AC 
Sources--Operating,'' to extend the Completion Time (CT) for an 
inoperable Division 1 or Division 2 diesel generator (DG) from 72 hours 
to 14 days.
    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 TS change to increase the CT for an inoperable 
Division 1 or Division 2 DG from 72 hours to 14 days does not affect 
the design, function, operational characteristics, or reliability of 
the DGs. The DGs are designed to mitigate the consequences of 
previously evaluated accidents and, as such, are not accident 
initiators.
    Extending the CT for an inoperable DG will not significantly 
affect the capability of the DGs to perform their accident 
mitigation safety functions or adversely affect DG or offsite power 
availability. The consequences of previously evaluated accidents 
will not be significantly affected since the remaining DGs 
supporting the redundant Engineered Safety Feature (ESF) systems 
will continue to be available to perform the accident mitigation 
functions as designed.
    Both a deterministic evaluation and a risk impact assessment 
were performed to support the proposed DG CT extension. The 
deterministic evaluation concluded that the defense-in-depth 
philosophy will be maintained with the proposed DG CT extension. The 
current TS and 10 CFR 50.65 (Maintenance Rule) programmatic 
requirements and additional administrative controls provide 
assurance that a loss of offsite power occurring concurrent with an 
inoperable DG will not result in a complete loss of function of 
critical systems. The duration of the proposed DG CT is determined 
considering that there is a minimal possibility that an accident 
will occur while a component is removed from service. A risk impact 
assessment was performed which concluded that the increase in plant 
risk due to the increased DG CT is small and consistent with the 
guidance contained in Regulatory Guide 1.177, ``An Approach for 
Plant-Specific, Risk-Informed Decisionmaking: Technical 
Specifications.''
    Based on the above discussion, it is concluded that the proposed 
amendment does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment does not alter the design, configuration, 
or method of operation of the plant, and does not alter any safety 
analysis inputs or assumptions. The proposed extended DG CT will not 
reduce the number of DGs below the minimum required for safe 
shutdown or accident mitigation. No new component failure modes, 
system interactions, or accident responses will be created that 
could result in a new or different kind of accident from any 
accident previously evaluated.
    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 extension of the DG CT remains consistent with 
codes and standards applicable to the onsite alternating current 
(AC) sources, except that the extension deviates from the 
recommendations of Regulatory Guide 1.93, ``Availability of Electric 
Power Sources.'' The proposed amendment is justified based on the 
results of a deterministic evaluation and a risk impact assessment. 
These demonstrate that the defense-in-depth philosophy will be 
maintained and the increase in plant risk is small and consistent 
with the guidance contained in Regulatory Guide 1.177.
    The DG reliability and availability are monitored and evaluated 
with respect to Maintenance Rule performance criteria to assure DG 
out of service times do not degrade operational safety over time. 
Furthermore, extension of the DG CT does not affect any safety 
analysis inputs or assumptions and will not erode the reduction in 
severe accident risk that was achieved with implementation of the 
Station Blackout (SBO) rule (10 CFR 50.63). The SBO coping analysis 
is unaffected by the CT extension since the DGs are not assumed to 
be available during the coping period. The assumptions used in the 
coping analysis regarding DG reliability are unaffected since 
preventive maintenance and testing will continue to be performed to 
maintain the reliability assumptions.
    Based on the above discussion, it is concluded that the proposed 
amendment does not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Carey W. Fleming, Senior Counsel, 
Constellation Energy Nuclear Group, LLC, 100 Constellation Way, Suite 
200C, Baltimore, MD 21202
    NRC Branch Chief: Nancy L. Salgado.

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.

Carolina Power & 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: May 29, 2009, as supplemented by 
letters dated November 14, 2008, December 11, 2008, August 13, 2009, 
August 28, 2009, October 9, 2009, February 4, 2010, and April 5, 2010.

[[Page 39982]]

    Brief description of amendment: The proposed amendment transitions 
the existing fire protection program to a risk-informed, performance-
based program based on National Fire Protection Association Standard 
805 (NFPA 805), ``Performance-Based Standard for Fire Protection for 
Light Water Reactor Electric Generating Plants,'' 2001 Edition, in 
accordance with Title 10 of the Code of Federal Regulations, Sec.  
50.48(c). NFPA 805 allows the use of performance-based methods, such as 
fire modeling and fire risk evaluations, to demonstrate compliance with 
the nuclear safety performance criteria.
    Date of issuance: June 28, 2010.
    Effective date: Effective as of the date of issuance and shall be 
implemented within 180 days, contingent upon completion of the items 
identified in section 2.9 of the associated NRC Safety Evaluation.
    Amendment No.: 133.
    Renewed Facility Operating License No. NPF-63: The amendment 
revises the Technical Specifications and Facility Operating License.
    Date of initial notice in Federal Register: June 19, 2009 (74 FR 
29241). The supplements dated November 14, 2008, December 11, 2008, 
August 13, 2009, August 28, 2009, October 9, 2009, February 4, 2010, 
and April 5, 2010 provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): No.
    The Commission's related evaluation of the amendment and final NSHC 
determination are contained in a safety evaluation dated June 28, 2010.

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: October 2, 2008, as 
supplemented by letters dated August 25, 2009, and October 23, 2009.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TSs) associated with the verification of ice 
condenser door operability and TS surveillance requirements 3.6.13.5 
and 3.6.13.6.
    Date of issuance: June 28, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 256 and 251.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: March 8, 2010 (75 FR 
10513). The supplements dated August 25, 2009, and October 23, 2009, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 28, 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: September 2, 2008, as 
supplemented by letters dated June 18, 2009, July 8, 2009, August 13, 
2009, September 8, 2009, November 10, 2009 and March 8, 2010.
    Brief description of amendments: The amendments revised the 
technical specifications to allow manual operation of the containment 
spray system and to revise the upper and lower limits of the refueling 
water storage tank.
    Date of issuance: June 28, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 257 and 252.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the licenses and the Technical Specifications.
    Date of initial notice in Federal Register: April 7, 2009 (74 FR 
15767). The supplements dated June 18, 2009, July 8, 2009, August 13, 
2009, September 8, 2009, November 10, 2009, and March 8, 2010, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the NRC staff's original proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 28, 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: October 2, 2008, as 
supplemented by letters dated August 25, 2009, and October 23, 2009.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) associated with the verification of ice 
condenser door operability and TS surveillance requirements 3.6.13.5 
and 3.6.13.6.
    Date of issuance: June 28, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 256 and 236.
    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: March 8, 2010 (75 FR 
10508). The supplements dated August 25, 2009, and October 23, 2009, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 28, 2010.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point 
Nuclear Generating Unit No. 3, Westchester County, New York

    Date of application for amendment: July 23, 2009.
    Brief description of amendment: The amendment removed the local 
refueling water storage tank level indication from Technical 
Specification Surveillance Requirement 3.5.4.5.
    Date of issuance: June 28, 2010.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 241.
    Facility Operating License No. DPR-64: The amendment revised the 
License and the Technical Specifications.
    Date of initial notice in Federal Register: October 6, 2009 (74 FR 
51329).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 28, 2010.
    No significant hazards consideration comments received: No.

[[Page 39983]]

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

    Date of application for amendment: June 30, 2009, as supplemented 
by letter dated May 24, 2010.
    Brief description of amendment: This amendment revises the 
Surveillance Requirement (SR) regarding the start time tests for the 
Division 3 Emergency Diesel Generator to provide consistency with 
existing similar Technical Specification (TS) 3.8.1 ``AC Sources--
Operating'' SRs and the