Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 33839-33851 [2010-14063]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 75, Number 114 (Tuesday, June 15, 2010)]
[Notices]
[Pages 33839-33851]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2010-14063]


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NUCLEAR REGULATORY COMMISSION

[NRC-2010-0195]


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 May 20, 2010 to June 2, 2010. The last 
biweekly notice was published on June 1, 2010 (75 FR 30440).

Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in Title 10 of the Code of Federal 
Regulations (10 CFR), Section 50.92, this means that operation of the 
facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to Cindy Bladey, Chief, 
Rules, Announcements, and Directives Branch

[[Page 33840]]

(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 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

[[Page 33841]]

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 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., 
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, 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.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: April 21, 2010.
    Description of amendment request: The proposed amendment would 
revise James A. FitzPatrick Technical Specification (TS) 2.0, ``Safety 
Limits (SLs).'' Specifically, TS 2.1.1.2 would replace the listed 
safety limit minimum critical power ratio values of 1.07 for two 
recirculation loop operation and 1.09 for single recirculation loop 
operation with new values of 1.08 and 1.11, respectively.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The operation of JAF [James A. FitzPatrick Nuclear Power 
Plant] in accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The basis of the Safety Limit Minimum Critical Power Ratio 
(SLMCPR) is to ensure no mechanistic fuel damage is calculated to 
occur if the limit is not violated. The new SLMCPR values preserve 
the existing margin to transition boiling and probability of fuel 
damage is not increased. The derivation of the revised SLMCPR for 
JAF, for incorporation into the Technical Specifications and its use 
to determine plant and cycle-specific thermal limits, has been 
performed using NRC approved methods. These plant-specific 
calculations are performed each operating cycle and, if necessary, 
will require future changes to these values based upon revised core 
designs. The revised SLMCPR values do not change the method of 
operating the plant and have no effect on the probability of an 
accident initiating event or transient.
    Based on the above, JAF has concluded that the proposed change 
will not result in a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The operation of JAF in accordance with the proposed 
amendment will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The proposed changes result only from a specific analysis for 
the JAF core reload design. These changes do not involve any new or 
different methods for operating the facility. No new initiating 
events or transients will result from these changes.
    Based on the above, JAF has concluded that the proposed change 
will not create the possibility of a new or different kind of 
accident from those previously evaluated.
    3. The operation of JAF in accordance with the proposed 
amendment will not involve a significant reduction in a margin of 
safety.

[[Page 33842]]

    The new SLMCPR is calculated using NRC approved methods with 
plant and cycle specific parameters for the current core design. The 
SLMCPR value remains conservative enough to ensure that greater than 
99.9% of all fuel rods in the core will avoid transition boiling if 
the limit is not violated, thereby preserving the fuel cladding 
integrity. The operating MCPR limit is set appropriately above the 
safety limit value to ensure adequate margin when the cycle specific 
transients are evaluated. Accordingly, the margin of safety is 
maintained with the revised values.
    As a result, JAF has determined that the proposed change will 
not result in a significant reduction in a margin of safety.
    On the basis of the above, JAF has determined that operation of 
the facility in accordance with the proposed change does not involve 
a significant hazards consideration as defined in 10 CFR 50.92(c), 
in that it: (1) Does not involve a significant increase in the 
probability or consequences of an accident previously evaluated; (2) 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated; and (3) does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Nancy L. Salgado.

Pacific Gas and Electric Co., Docket No. 50-133, Humboldt Bay Power 
Plant (HBPP), Unit 3 Humboldt County, California

    Date of amendment request: April 9, 2010.
    Description of amendment request: The licensee has proposed 
amending the technical specifications (TS) to delete TS 3.1.3, ``Fuel 
Storage Pool Liner Water Level.'' TS 3.1.3 places restrictions on the 
maximum operating water level in the gap between the stainless steel 
liner covering the inside surface of the fuel storage pool and the pool 
walls. This TS restriction was originally intended to preclude leakage 
from the fuel storage pool to surrounding groundwater, and to preclude 
groundwater seepage into the pool. Additional conforming and editorial 
changes are also 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 change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes eliminate the TS requirement to control the 
operating water level in the gap between the fuel storage pool liner 
and the pool concrete structure and delete the TS sections that are 
applicable to Limiting Conditions for Operation (LCO).
    Elimination of the TS requirements to control liner gap water 
level will not increase the probability or consequences of the 
previously analyzed (and recently updated) fuel storage pool rupture 
accident. The basis of TS 3.1.3 is to preclude both pool leakage to 
the surrounding groundwater and groundwater leakage into the pool. 
The radiological consequences of pool leakage are conservatively 
bounded by the fuel storage pool rupture analysis, an analysis which 
demonstrates that the consequences of a breech of the fuel storage 
pool are insignificant.
    Additionally, the proposed changes will not result in the 
modification of any systems, structures or components and will not 
affect any parameters or conditions that could contribute to the 
initiation of an accident.
    Therefore, the proposed changes will 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.
    The proposed changes will not result in the modification of any 
systems, structures or components and will not affect any parameters 
or conditions that could contribute to the initiation of an 
accident. Therefore, the proposed changes will not create the 
possibility of a new or different kind of accident from those 
previously evaluated
    (3) Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The design basis and accident assumptions within the Humboldt 
Bay Power Plant (HBPP) Unit 3 Defueled Safety Analysis Report and 
the TS relating to spent fuel are no longer applicable. The proposed 
changes do not affect remaining plant operations, nor structures, 
systems, or components supporting decommissioning activities. In 
addition, the proposed changes do not result in a change in initial 
conditions, system response time, or in any other parameter 
affecting the course of a decommissioning activity accident 
analysis. Therefore, the proposed changes will 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: Ms. Jennifer K. Post, Pacific Gas and 
Electric Company, 77 Beale Street, B30A, San Francisco, CA.
    NRC Branch Chief: Bruce Watson (Acting).

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

    Date of amendment request: March 19, 2010.
    Description of amendment request: The proposed amendment would 
modify the Hope Creek Generating Station (HCGS) Technical 
Specifications (TSs) by relocating specific surveillance frequencies to 
a licensee-controlled program, the Surveillance Frequency Control 
Program (SFCP). The proposed changes are based on Nuclear Regulatory 
Commission (NRC)-approved Technical Specification Task Force (TSTF) 
change TSTF-425, Revision 3, ``Relocate Surveillance Frequencies to 
Licensee Control--RITSTF [Risk-Informed TSTF] Initiative 5b'' 
(Agencywide Documents Access and Management System (ADAMS) Package 
Accession No. ML090850642). Plant-specific deviations from TSTF-425 are 
proposed to accommodate differences between the HCGS TSs and the model 
TSs originally used to develop TSTF-425.
    The NRC staff issued a Notice of Availability for TSTF-425 in the 
Federal Register on July 6, 2009 (74 FR 31996). The notice included a 
model safety evaluation and a model no significant hazards 
consideration (NSHC) determination. In its application dated March 19, 
2010, PSEG Nuclear LLC (PSEG, the licensee) provided its analysis of 
the issue of NSHC based on the model NSHC determination for TSTF-425.
    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. 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

[[Page 33843]]

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 change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The 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, PSEG will 
perform a probabilistic risk evaluation using the guidance contained 
in NRC approved NEI 04-10, Rev. 1 in accordance with the TS SFCP. 
[The] 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 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: Vincent Zabielski, PSEG Nuclear LLC-N21, 
P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Harold K. Chernoff.

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

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

    Date of amendment request: March 25, 2010.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TSs) associated with reactor 
coolant system (RCS) structural integrity requirements for Hope Creek 
Generating Station (Hope Creek) and Salem Nuclear Generating Station 
(Salem), Unit Nos. 1 and 2. Specifically, the amendments would: (1) 
Delete the structural integrity requirements contained in TS \3/4\.4.8 
(Hope Creek), TS \3/4\.4.10 (Salem Unit 1), and TS \3/4\.4.11 (Salem 
Unit 2); (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 Section 6.8.4; 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 and Salem Unit 2 SR 4.4.11.2.
    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, with Nuclear Regulatory 
Commission (NRC) staff edits in square brackets:

    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 remove the RCS Structural Integrity TS 
does not impact any mitigation equipment or the ability of the RCS 
pressure boundary to fulfill any required safety function. Since no 
accident mitigation or initiators are impacted by this change, no 
design basis accidents are affected. The removal of the RCS 
Structural Integrity TS eliminates from the TS the redundancy of 
requirements that are already covered by the inspections necessary 
to maintain structural integrity under 10 CFR 50.55a [Section 50.55a 
of Title 10 of the Code of Federal Regulations (10 CFR)].
    Therefore, the proposed changes do not represent a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not involve a modification to the 
physical configuration of the plant (i.e., no new equipment will be 
installed) or change in the methods governing normal plant 
operation. The proposed change will not impose any new or different 
requirements or introduce a new accident initiator, accident 
precursor, or malfunction mechanism. Therefore, this proposed change 
does not create the possibility of an accident of a [new or] 
different kind than previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Removal of the RCS Structural Integrity TS does not reduce the 
controls that are required to maintain the RCS pressure boundary for 
ASME Code [American Society of Mechanical Engineers Boiler and 
Pressure Vessel Code] Class 1, 2, or 3 components. The removal of 
the RCS Structural Integrity TS eliminates from the TS the 
redundancy of requirements that are already covered by the 
inspections necessary to maintain structural integrity under 10 CFR 
50.55a. No equipment or RCS safety margins are impacted due to the 
proposed change[.]
    Therefore, this 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, 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: Vincent Zabielski, PSEG Nuclear LLC-N21, 
P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Harold K. Chernoff.

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

    Date of amendment request: March 23, 2010.
    Description of amendment request: The proposed amendment would 
modify the Salem Nuclear Generating Station, Unit Nos. 1 and 2 (Salem), 
Technical Specifications (TSs) by relocating specific surveillance 
frequencies to a licensee-controlled program, the Surveillance 
Frequency Control Program (SFCP). The proposed changes are based on 
Nuclear Regulatory Commission (NRC)-approved Technical Specification 
Task Force (TSTF) change TSTF-425, Revision 3, ``Relocate Surveillance 
Frequencies to Licensee Control--RITSTF [Risk-Informed TSTF] Initiative 
5b'' (Agencywide Documents Access and Management System (ADAMS) Package 
Accession No. ML090850642). Plant-

[[Page 33844]]

specific deviations from TSTF-425 are proposed to accommodate 
differences between the Salem TSs and the model TSs originally used to 
develop TSTF-425.
    The NRC staff issued a Notice of Availability for TSTF-425 in the 
Federal Register on July 6, 2009 (74 FR 31996). The notice included a 
model safety evaluation and a model no significant hazards 
consideration (NSHC) determination. In its application dated March 23, 
2010, PSEG Nuclear LLC (PSEG, the licensee) provided its analysis of 
the issue of NSHC based on the model NSHC determination for TSTF-425.
    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. 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 change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in 
[a] margin of safety?
    Response: No.
    The 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, PSEG will 
perform a probabilistic risk evaluation using the guidance contained 
in NRC approved NEI [Nuclear Energy Institute] 04-10, Rev. 1 in 
accordance with the TS SFCP. [The] 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 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: Vincent Zabielski, PSEG Nuclear LLC-N21, 
P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Harold K. Chernoff.

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

    Date of amendment request: April 13, 2010.
    Description of amendment request: The proposed amendment would add 
a footnote to Function 8.a in Technical Specification (TS) Table 3.3.2-
1, ``Engineered Safety Feature Actuation System Instrumentation,'' 
concerning the reactor trip P-4 engineered safety feature actuation 
system (ESFAS) interlock. This would specify which functions of the 
interlock are necessary in each mode in order to meet the limiting 
condition for operation. Specifically, the functions of tripping the 
main turbine and isolating main feedwater with a coincident low average 
temperature would no longer be applicable in Mode 3, which is hot 
standby. The amendment would also identify that the function of the P-4 
interlock that allows arming of the steam dump valves and transfers the 
steam dump load rejection (Tavg) controller to the plant 
trip controller is not required in any mode.
    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.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses. Defeating the 
feedwater isolation low Tavg coincident with P-4 function 
will not impact any accidents previously evaluated in the Updated 
Safety Analysis Report (USAR) since feedwater isolation on low 
Tavg coincident with P-4 is not credited. Bypassing the 
turbine trip on reactor trip function will not impact any accidents 
previously evaluated in the USAR since the turbine trip on reactor 
trip function of P-4 is not credited.
    The instrumentation utilized to initiate transfer to the plant 
trip steam dump controller does not serve a primary protective 
function so as to warrant inclusion in the TS. The instrumentation 
does not serve to ensure that the plant is operated within the 
bounds of initial conditions assumed in design basis accident and 
transient analyses. Likewise, the transfer to the plant trip steam 
dump controller instrumentation does not serve as part of the 
primary success path of a safety sequence analysis used to 
demonstrate that the consequence of these events are within the 
appropriate acceptance criteria.
    The ESFAS will continue to function in a manner consistent with 
the accident analysis assumptions and the plant design basis. As 
such, there will be no degradation in the performance of, nor an 
increase in, the number of challenges to equipment assumed to 
function during an accident situation. The proposed changes to the 
TSs do not affect the probability of any event initiators. There 
will be no change to normal plant operating parameters or accident 
mitigation capabilities.
    Therefore, this change will not increase the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any previously evaluated?
    Response: No.
    There are no changes in the method by which any safety related 
plant system performs its safety function and the normal manner of 
plant operation is unaffected, other than the proposed allowance to 
defeat feedwater isolation on low Tavg coincident with P-
4 and the proposed allowance to defeat the turbine trip on reactor 
trip function of P-4.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of this change. There will be no adverse effect or challenges 
imposed on any safety related system as a result of this change. 
Therefore,

[[Page 33845]]

the possibility of a new or different type of accident is not 
created. The proposed change does not affect the steam generator 
high-high level trip ESFAS function which initiates feedwater 
isolation and trips the turbine and main feedwater pumps. Therefore, 
this change will not create the possibility of a new or different 
kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    There will be no effect on the manner in which safety limits or 
limiting safety system settings are determined nor will there be any 
effect on those plant systems necessary to assure the accomplishment 
of protection functions. There will be no impact on departure from 
nucleate boiling ratio (DNBR) limits, heat flux hot channel factor 
(FQ(Z)) limits, nuclear enthalpy rise hot channel factor 
(FN[Delta]H) limits, peak centerline 
temperature (PCT) limits, peak local power density or any other 
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: Jay Silberg, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: Michael T. Markley.

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.

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: May 28, 2009, as supplemented 
by letter dated April 5, 2010.
    Brief description of amendments: The amendments revised the 
Technical Specification (TS) 3.8.1, ``AC Sources--Operating,'' to 
restrict voltage limits for the applicable TS 3.8.1 surveillances 
governing the Emergency Diesel Generators.
    Date of issuance: May 27, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 255, 250.
    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 9, 2010 (75 FR 
10825). The supplement dated 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.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 27, 2010.
    No significant hazards consideration comments received: No.

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

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

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of application for amendments: May 18, 2009.
    Brief description of amendments: The proposed changes would revise 
the Technical Specifications (TSs) to adopt Technical Specification 
Task Force (TSTF)-248, ``Revise Shutdown Margin Definition For Stuck 
Rod Exception.'' The TSTF revises the definition of shutdown margin 
(SDM) in the TSs with all control rods verified fully inserted by two 
independent means. It is not necessary to account for a stuck control 
rod in the SDM calculation.
    Date of issuance: May 28, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 254, 249, 255, 235, 367, 369, and 368.
    Renewed Facility Operating License Nos. NPF-35, NPF-52, NPF-9, NPF-
17, DPR-38, DPR-47, and DPR-55: Amendments revised the Operating 
Licenses and the Technical Specifications.
    Date of initial notice in Federal Register: March 9, 2010 (75 FR 
10827).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 28, 2010.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 17, 2009, as supplemented 
by letter dated January 21, 2010.
    Brief description of amendment: The amendment modified (1) 
Technical Specification (TS) 3.8.3, ``Diesel Fuel Oil, Lube Oil, and 
Starting Air,'' to relocate specific numerical values for fuel oil and 
lube oil storage volumes from the TS to the TS Bases, (2) TS 3.8.1, 
``AC [Alternating Current] Sources--Operating,'' to relocate

[[Page 33846]]

specific values for the day tank fuel oil volumes from the TS to the TS 
Bases, and (3) TS 5.5.9, ``Diesel Fuel Oil Testing Program,'' to 
relocate the specific standard for particulate concentration testing of 
fuel oil from the TS to the TS Bases.
    Date of issuance:
    Effective Date: As of its date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 215.
    Facility Operating License No. NPF-21: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: November 3, 2009 (74 FR 
56884). The supplemental letter dated January 21, 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 May 27, 2010.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear 
Plant, Van Buren County, Michigan

    Date of application for amendment: March 31, 2010, supplemented by 
letter dated May 13, 2010.
    Brief description of amendment: The amendment adds a new license 
condition 2.C(4) to Palisades Nuclear Plant, renewed facility license 
No. DPR-20. This license condition would state that performance of 
Technical Specification (TS) surveillance requirement (SR) 3.1.4.3 is 
not required for control rod drive 22 through cycle 21 or until the 
next entry into Mode 3. The amendment consists of changes to TS by 
addition of a note in SR 3.1.4.3, stating: ``Not required to be 
performed or met for control rod 22 during cycle 21 provided control 
rod 22 is administratively declared immovable, but trippable and 
Condition D is entered for control rod 22.''
    Date of issuance: June 2, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 15 days.
    Amendment No.: 239.
    Facility Operating License No. DPR-20: Amendment revised the 
Technical Specifications and license.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): The notice provided an opportunity to submit 
comments on the Commission's proposed NSHC determination. No comments 
have been received. The notice also provided an opportunity to request 
a hearing by June 13, 2010, which is within 60 days of the individual 
notice published on April 14; but indicated that if the Commission 
makes a final NSHC determination, any such hearing would take place 
after issuance of the amendment.
    Date of initial Individual notice in Federal Register: April 14, 
2010 (75 FR 19428), followed by the repeat biweekly notice in the 
Federal Register on May 4, 2010 (75 FR 23818).
    The Commission's related evaluation of the amendment, state 
consultation, and final NSHC determination are contained in a Safety 
Evaluation dated June 2, 2010.
    Attorney for licensee: Mr. William Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White 
Plains, NY 10601.
    NRC Branch Chief: Robert J. Pascarelli.

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

    Date of application for amendment: May 15, 2009, as supplemented by 
letters dated November 10, 2009, and May 17, 2010.
    Brief description of amendment: The amendment modified Technical 
Specification (TS) 6.6.5, ``Core Operating Limits Report (COLR),'' to 
minimize the number of U.S. Nuclear Regulatory Commission (NRC)-
approved references consistent with the guidance provided in NRC 
Generic Letter 88-16, ``Removal of Cycle-Specific Parameter Limits from 
Technical Specifications,'' dated October 3, 1988. This also fulfills 
the commitment made in the licensee's letter to the NRC dated March 11, 
2008, ``Response to Request for Additional Information License 
Amendment Request to Revise Technical Specification 6.6.5, Core 
Operating Limits Report.''
    Date of issuance: May 25, 2010.
    Effective Date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 290.
    Renewed Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications/license.
    Date of initial notice in Federal Register: June 30, 2009 (74 FR 
31321). The supplemental letters dated November 10, 2009, and May 17, 
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 May 25, 2010.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 22, 2009, as supplemented by letter 
dated May 17, 2010.
    Brief description of amendment: The amendment modified Technical 
Specification 6.9.1.11 to minimize the number of references that 
reflect U.S. Nuclear Regulatory Commission (NRC)-approved methods used 
in establishing the Core Operating Limits Report (COLR) parameter 
limits, consistent with the guidance provided in NRC Generic Letter 88-
16, ``Removal of Cycle-Specific Parameter Limits from Technical 
Specifications,'' dated October 3, 1988.
    Date of issuance: May 25, 2010.
    Effective Date: As of the date of issuance and shall be implemented 
90 days from the date of issuance.
    Amendment No.: 226.
    Facility Operating License No. NPF-38: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: June 30, 2009 (74 FR 
31322). The supplemental letter dated May 17, 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 May 25, 2010.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 29, 2008, supplemented 
by letters dated May 6, 2009, June 23, 2009, August 21, 2009, September 
17, 2009, October 15, 2009, and November 11, 2009.
    Brief description of amendment: The changes revise the TMI-1 
technical

[[Page 33847]]

specifications (TSs) to reflect design changes resulting from the 
planned Control Rod Drive Control System digital upgrade project. In 
addition, the proposed amendment revises the TS to remove all 
references to the Axial Power Shaping Rods to reflect changes resulting 
from their elimination from the TMI-1 reactor.
    Date of issuance: May 27, 2010.
    Effective Date: Immediately, and shall be implemented prior to 
exiting cold shutdown from the fall 2011 (T1R19) refueling outage.
    Amendment No.: 273.
    Renewed Facility Operating License No. DPR-50. Amendment revised 
the license and the technical specifications. The supplements dated May 
6, 2009, June 23, 2009, August 21, 2009, September 17, 2009, October 
15, 2009, and November 11, 2009, provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the NRC staff's original 
proposed no significant hazards determination.
    Date of initial notice in Federal Register: March 10, 2009 (74 FR 
10308).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 27, 2010.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412 Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2), 
Beaver County, Pennsylvania

    Date of application for amendment: June 11, 2009.
    Brief description of amendment: The amendments will (1) modify 
Technical Specifications (TSs) to eliminate Surveillance Requirement 
(SR) 3.3.2.9, which verifies that the Engineered Safety Feature 
Actuation System Response Times are within the limits for the 
recirculation spray pumps, (2) revise Section 1.4 of the TSs to add 
clarification to Notes associated with SRs in accordance with Technical 
Specification Task Force Traveler, TSTF 475-A, Revision 1, ``Control 
Rod Notch Testing Frequency and SRM [Source Range Monitor] Insert 
Control Rod Action,'' (3) revise the BVPS-1 operating license to remove 
a License Condition for recommended inspections of steam generator 
repairs, and (4) make some editorial changes to the operating license 
pages.
    Date of issuance: May 20, 2010.
    Effective Date: As of the date of issuance, and shall be 
implemented within 30 days from the date of issuance.
    Amendment Nos.: 285 and 171.
    Facility Operating License Nos. DPR-66 and NPF-73: The amendments 
revised the License, TSs, Appendix B, and Appendix D.
     Date of initial notice in Federal Register: December 15, 2009 (74 
FR 66385).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 20, 2010.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie 
Plant, Unit No. 2, St. Lucie County, Florida

    Date of application for amendment: May 22, 2009, as supplemented 
June 22, 2009.
    Brief description of amendment: Revises Technical Specification 
3.1.3.4, related to requirements for Control Element Assembly drop 
time.
    Date of Issuance: May 31, 2010.
    Effective Date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 158.
    Renewed Facility Operating License No. NPF-16: Amendment revised 
the license and the Technical Specifications.
    Date of initial notice in Federal Register: June 1, 2009 (74 FR 
26261). The supplement dated June 22, 2009, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 31, 2010.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: April 17, 2009, as supplemented 
by letter dated January 19, 2010.
    Brief description of amendments: Amend Renewed Operating Licenses 
DPR-24 and DPR-27 for Point Beach Nuclear Plant Units 1 and 2, 
respectively, to reflect a change in the legal name of the licensee 
from ``FPL Energy Point Beach, LLC'' to ``NextEra Energy Point Beach, 
LLC'' and correct a typographical error in Appendix C from ``FPLE Group 
Capital'' to ``FPL Group Capital.''
    Date of issuance: May 13, 2010.
    Effective Date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 237, 241.
    Renewed Facility Operating License Nos. DPR-24 and DPR-27: 
Amendments revised the Technical Specifications/License.
    Date of initial notice in Federal Register: July 14, 2009 (74 FR 
34048) as supplemented by March 3, 2010 (75 FR 9619).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 13, 2010.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 24, 2009, as supplemented 
by letter dated December 21, 2009.
    Brief description of amendments: The amendments modify the 
technical specification requirements to control room habitability in 
accordance with Technical Specification Task Force (TSTF)-448, Revision 
3, ``Control Room Habitability.''
    Date of issuance: May 20, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days.
    Amendment Nos.: 195, 184.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Facility Operating Licenses and the Technical 
Specifications.
    Date of initial notice in Federal Register: September 8, 2009 (74 
FR 46243). The supplemental letter contained clarifying information and 
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 May 20, 2010.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company et al., Docket No. 52-011, Vogtle 
Electric Generating Plant ESP Site, Burke County, Georgia

    Date of amendment request: April 20, 2010, as supplemented April 23 
and 28, May 5, 10, 13, and 20, 2010.
    Description of amendment request: The amendment revised the Vogtle 
Electric Plant (VEGP) ESP Site Safety Analysis Report (SSAR) to allow 
the use of Category 1 and 2 backfill material

[[Page 33848]]

from additional onsite areas that were not specifically id