Biweekly Notice: Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 59259-59269 [E9-27406]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 74, Number 220 (Tuesday, November 17, 2009)]
[Notices]
[Pages 59259-59269]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E9-27406]


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NUCLEAR REGULATORY COMMISSION

[NRC-2009-0498]


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 October 22, 2009 to November 4, 2009. The 
last biweekly notice was published on November 3, 2009 (74 FR 56882).

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 the Chief, Rulemaking 
and Directives Branch (RDB), 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 RDB 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

[[Page 59260]]

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 petitioner/requestor 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 
petitioner/requestor 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 petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor 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 petitioner/requestor to relief. A petitioner/
requestor 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, which the NRC 
promulgated in August 28, 2007 (72 FR 49139). 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 petitioner/requestor 
should contact the Office of the Secretary by e-mail at 
hearing.docket@nrc.gov, or by calling (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/or (2) 
creation of an electronic docket for the proceeding (even in instances 
in which the petitioner/requestor (or its counsel or representative) 
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms Viewer\TM\ to 
access the Electronic Information Exchange (EIE), a component of the E-
Filing system. The Workplace Forms Viewer\TM\ is free and is available 
at https://www.nrc.gov/site-help/e-submittals/install-viewer.html. 
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.
    Once a petitioner/requestor has obtained a digital ID certificate, 
had a docket created, and downloaded the EIE viewer, it 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 filer submits its documents through EIE. To be timely, an 
electronic filing must be submitted to the EIE 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 
EIE 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 through the ``Contact Us'' link 
located on the NRC Web site at https://www.nrc.gov/site-help/e-submittals.html or by calling the NRC Meta-System Help Desk, which is 
available between 8 a.m. and 8 p.m., Eastern Time, Monday through 
Friday, excluding government holidays. The Meta-System Help Desk can be 
contacted by telephone at 1-866-672-7640 or by e-mail at 
MSHD.Resource@nrc.gov.
    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.
    Non-timely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission, the presiding 
officer, or the Atomic Safety and Licensing Board that the request and/
or petition should be granted and/or the contentions should be 
admitted, based on a

[[Page 59261]]

balancing of the factors specified in 10 CFR 2.309(c)(1)(i)-(viii).
    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, an Atomic Safety and Licensing Board, or 
a 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 submissions.
    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 Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: September 9, 2009.
    Description of amendment request: The proposed amendment revises 
Technical Specification (TS) 3/4.9.7, ``Crane Travel--Fuel Handling 
Building,'' to permit certain operations needed for dry cask storage of 
spent nuclear fuel. Specifically, the proposed change to this TS (while 
continuing to prohibit travel of a heavy load over irradiated fuel 
assemblies in the spent fuel pool) would permit travel of loads in 
excess of 2,000 pounds (lbs) over a transfer cask containing irradiated 
fuel assemblies, provided a single-failure-proof handling system is 
used.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The FHB [fuel handling building] cask crane will be upgraded to 
meet the applicable single-failure-proof criteria of NUREG 0554 
(Reference 7.10 [NUREG-0554, Single-Failure-Proof Cranes for Nuclear 
Power Plants, U.S. Nuclear Regulatory Commission, May 1979]) and 
NUREG 0612 (Reference 7.13 [NUREG-0612, Control of Heavy Loads at 
Nuclear Power Plants, U.S. Nuclear Regulatory Commission, July 1980 
(ADAMS Accession No. ML070250180)] for the modification of the 
existing non single-failure-proof crane. Due to the reliability of 
this upgraded handling system, a load drop accident will not be 
considered a credible event. While loads in excess of 2000 lbs shall 
continue to be prohibited from travel over irradiated fuel 
assemblies in the spent fuel pool by the WF3 [Waterford 3] Technical 
Specifications, heavy loads will be permitted to travel over 
irradiated fuel assemblies in a transfer cask, using a single-
failure-proof handling system as described in NUREG-0800 Section 
9.1.5 Paragraph III.4.C (Reference 7.9 [NUREG-0800 Section 9.1.5 
Rev. 1, Standard Review Plan for Overhead Heavy Load Handling 
Systems, March 2007 (ADAMS Accession No. ML062260190)]), to enable 
the conduct of dry cask storage loading/unloading operations. 
Specifically, this will enable the MPC [multi-purpose canister] lid 
and its associated lifting apparatus to travel over irradiated fuel 
assemblies in a MPC basket. The probability of dropping a load that 
weighs in excess of 2000 lbs onto an irradiated fuel assembly is not 
increased as a result of the reliability of the single-failure-proof 
handling system.
    The proposed change does not affect the consequences of any 
accidents previously evaluated in the WF3 UFSAR [Updated Final 
Safety Analysis Report] (Reference 7.1 [Waterford Steam Electric 
Station Unit No. 3, Updated Final Safety Analysis Report, Revision 
302, December 2008]). The change involves the travel of heavy loads 
over irradiated fuel assemblies in a transfer cask using a single-
failure-proof handling system. Under these circumstances, no new 
load drop accidents are postulated and no changes to the 
probabilities or consequences of accidents previously evaluated are 
involved.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Section 9.1 of the WF3 UFSAR evaluates fuel storage and handling 
operations. Section 15.7.3.4 of the WF3 UFSAR discusses the analysis 
of design basis fuel handling accidents involving drop of an 
irradiated assembly resulting in multiple fuel rod failures and 
consequent release of radioactivity. The change involves the travel 
of heavy loads over irradiated fuel assemblies in a transfer cask 
using a single-failure-proof handling system. Under these 
circumstances, no new or different load drop accidents are 
postulated to occur and there are no changes in any of the load drop 
accidents previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The revised Technical Specification changes do not involve a 
reduction in any margin of safety. Technical Specification 3/4.9.7 
currently prohibits travel of heavy loads over irradiated fuel 
assemblies in the FHB. Proposed changes to this specification will 
continue to restrict FHB cask crane movements so that travel of 
heavy loads over irradiated fuel assemblies in the FHB are not 
permitted, with the single exception of heavy loads over irradiated 
fuel assemblies in a transfer cask, in order to enable dry cask 
storage operations. This operation is only permitted when the heavy 
load is handled using a single-failure-proof handling system. Due to 
the reliability of this upgraded handling system that complies with 
the guidance of NUREG-0800 Section 9.1.5 Paragraph III.4.C 
(Reference 7.9) for a single-failure-proof handling system, a load 
drop accident is not considered a credible event. Under these 
circumstances, no new load drop accidents are postulated and no 
reductions in margins of safety are involved.

    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: Terence A. Burke, Associate General 
Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: June 30, 2009.
    Description of amendment request: The proposed amendment would 
modify a Surveillance Requirement (SR) regarding the start time tests 
for the Division 3 Emergency Diesel Generator (EDG) to provide 
consistency with existing similar Technical Specification (TS) SRs and 
the time provided in the licensing basis emergency core cooling system 
analyses.
    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?

[[Page 59262]]

    Response: No.
    The proposed amendment corrects and makes consistent the 
acceptance criteria for the [Perry Nuclear Power Plant] PNPP TS SR 
pertaining to the Division 3 EDG. The EDGs mitigate the consequences 
of previously evaluated accidents involving a loss of offsite power. 
The EDGs are used to support mitigation of the consequences of an 
accident, but they are not considered as the initiator of any 
previously analyzed accident.
    The proposed amendment will continue to ensure the EDGs perform 
their function when called upon to mitigate the consequences of 
events. The proposed revision to the TS SRs will continue to 
maintain the capability of the Division 3 [High Pressure Core Spray] 
HPCS system to respond within the times assumed in the Emergency 
Core Cooling System (ECCS) analyses.
    The proposed amendment does not affect the design of the EDGs, 
the interfaces between the EDGs and other plant systems, or the 
function and reliability of the EDGs. Thus, the EDGs will continue 
to be capable of performing their accident mitigation function and 
there is no impact to the radiological consequences determined in 
any accident analysis.
    As such, the proposed amendment continues to provide adequate 
assurance of an operable EDG and does not involve any increase to 
the probability or to the consequences of any accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change is an amendment that introduces no new mode 
of plant operation and it does not involve physical modification to 
the plant. New equipment is not installed with the proposed 
amendment, nor does the proposed amendment cause existing equipment 
to be operated in a new or different manner.
    Since the proposed amendment does not involve a change to the 
plant design or operation, no new system interactions are created by 
this change. The proposed amendment does not produce any parameters 
or conditions that could contribute to the initiation of accidents 
different from those already evaluated in the Updated Safety 
Analysis Report. The change to the affected TS SR does not affect 
the assumed accident performance of the EDG, nor any plant 
structure, system or component 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 change is an amendment that does not impact EDG 
performance as incorporated in the design basis analyses, including 
the capability for the EDG to attain and maintain required voltage 
and frequency for accepting and supporting plant safety loads should 
an EDG start signal be received. The operability of the EDG 
continues to be determined as required to provide emergency power to 
plant equipment that mitigates the consequences of a transient or 
accident, and maintains the HPCS system's capability to respond 
within the time assumed in the accident analyses.
    The proposed amendment does not introduce changes to setpoints 
or limits established in the accident analysis. As a result of the 
above considerations, it is concluded that implementation of 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: David W. Jenkins, Attorney, FirstEnergy 
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Stephen J. Campbell.

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: September 14, 2009.
    Description of amendment request: The proposed amendment would 
correct editorial items in the Technical Specifications (TS) and the 
Facility Operating License (FOL).
    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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to TS and the FOL are administrative in 
nature that correct typographical errors, correct format errors, 
correct inconsistencies between Units, or delete historical 
requirements that have expired. These changes do not affect the 
intent of any TS requirements.
    The proposed change does not have any impact on structures, 
systems and components (SSCs) of the plant, and no effect on plant 
operations. The proposed change does not impact any accident 
initiators or analyzed events or assumed mitigation of accident or 
transient events. They do not involve the addition or removal of any 
equipment, or any design changes to the facility. Therefore, this 
proposed change does not represent a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes to TS and the FOL are administrative in 
nature that correct typographical errors, correct format errors, 
correct inconsistencies between Units, or delete historical 
requirements that have expired. These changes do not affect the 
intent of any TS requirements.
    The proposed change does 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. Additionally, there is no 
change in the types or increases in the amounts of any effluent that 
may be released off-site and there is no increase in individual or 
cumulative occupational exposure. Therefore, the proposed changes do 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes to TS and the FOL are administrative in 
nature that correct typographical errors, correct format errors, 
correct inconsistencies between Units, or delete historical 
requirements that have expired. These changes do not affect the 
intent of any TS requirements.
    The proposed change incorporates corrections to the TS and FOL 
and results in improved accuracy of these licensing documents. There 
is no change to any design basis, licensing basis or safety limit, 
no change to any parameters; consequently no safety margins are 
affected. Therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
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. 50-272, Salem Nuclear Generating Station, 
Unit No. 1, Salem County, New Jersey

    Date of amendment request: September 21, 2009.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 6.8.4.f, ``Primary Containment Leakage 
Rate Testing Program,'' to allow a one-time extension of the 
containment Type A integrated leakage rate test interval from 10 to 15 
years.

[[Page 59263]]

    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 would revise Technical Specification (TS) 
6.8.4.f, ``Primary Containment Leakage Rate Testing Program,'' to 
permit a one-time extension of the containment Type A Integrated 
Leak Rate Test (ILRT) from ten to fifteen years.
    The function of the containment is to isolate and contain 
fission products released from the reactor coolant system following 
a design basis Loss of Coolant Accident (LOCA) and to confine the 
postulated release of radioactive material to within limits. The 
test interval associated with the performance of containment leakage 
testing is not an initiating event for any accident previously 
evaluated. There are no physical changes being made to the 
containment structure and no change made to the containment 
allowable leakage rate specified in Technical Specifications.
    During the extended test interval, containment integrity will 
continue to be assured by programs for local leak rate testing and 
containment inspections are routinely performed as required by [the 
American Society of Mechanical Engineers (ASME) Boiler and Pressure 
Vessel Code (Code)] which demonstrates the structural integrity of 
the primary containment. The proposed changes do not affect 
performance of the containment, reactor operations or accident 
analysis.
    The risk assessment of the proposed change has concluded that 
there is not a significant increase in the consequences of an 
accident as measured by the Large Early Release Frequency, 
Population Dose, and Conditional Containment Failure Frequency. 
These results show that an ILRT test extension will not represent a 
significant increase in the consequences of an accident.
    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 for a one-time, five-year extension of the 
Type A test makes no physical changes to the plant or to plant 
operations. No credible new failure mechanisms, malfunctions or 
accident initiators are being introduced by the proposed change.
    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 integrity of the containment penetrations and isolation 
valves is verified through Type B and Type C local leak rate tests 
(LLRTs) and the overall leak tight integrity of the containment is 
verified by a Type A ILRT, as required by [Title 10 of the Code of 
Federal Regulations (10 CFR), Part 50], Appendix J, ``Primary 
Reactor Containment Leakage Testing for Water-Cooled Power 
Reactors.'' The proposed change does not affect the method or 
acceptance criteria for Type A, B and C testing. During the extended 
test interval, containment inspections performed in accordance with 
the requirements of the [ASME Code], Section XI, ``Inservice 
Inspection,'' and 10 CFR 50.65, ``[Requirements for monitoring the 
effectiveness of maintenance at nuclear power plants],'' provide 
assurance that the containment will not degrade in a manner that is 
only detectable by Type A testing.
    The effect of the proposed change on Large Early Release 
Frequency, person-rem, and Conditional Containment Failure Frequency 
was determined not to be significant.
    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.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego 
County, California

    Date of amendment request: August 10, 2009.
    Description of amendment request: The amendments would revise 
Technical Specification 3.7.5, ``Auxiliary Feedwater (AFW) System,'' to 
allow a 7-day Completion Time for the turbine-driven AFW pump if the 
inoperability occurs in MODE 3, following a refueling outage and if 
MODE 2 had not been entered. This change is based on the U.S. Nuclear 
Regulatory Commission (NRC)-approved Technical Specification Task Force 
(TSTF) traveler, TSTF-340, Revision 3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed amendment to Technical Specification 3.7.5 would 
allow a seven day Completion Time for Condition A for the turbine-
driven Auxiliary Feedwater (AFW) pump if the inoperability occurs in 
MODE 3 following a refueling outage, if MODE 2 had not been entered. 
Extending the Completion Time does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated because: (1) The proposed amendment does not 
represent a change to the system design, (2) the proposed amendment 
does not prevent the safety function of the AFW system from being 
performed, since the other fully redundant essential trains are 
required to be operable, (3) the proposed amendment does not alter, 
degrade, or prevent action described or assumed in any accident 
described in the San Onofre Nuclear Generating Station (SONGS) 
Updated Final Safety Analysis Report (UFSAR) from being performed 
since the other trains of AFW are required to be operable, (4) the 
proposed amendment does not alter any assumptions previously made in 
evaluating radiological consequences, and (5) the proposed amendment 
does not affect the integrity of any fission product barrier. No 
other safety related equipment is affected by the proposed change.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed amendment to Technical Specification 3.7.5 would 
allow a seven day Completion Time for Condition A for the turbine-
driven AFW pump if the inoperability occurs in MODE 3 following a 
refueling outage, if MODE 2 had not been entered. Extending the 
Completion Time does not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because: (1) The proposed amendment does not represent a change to 
the system design, (2) the proposed amendment does not alter how 
equipment is operated or the ability of the system to deliver the 
required AFW flow, and (3) the proposed amendment does not affect 
any other safety related equipment.
    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 change involve a significant reduction in a 
margin of safety?

[[Page 59264]]

    Response: No.
    The proposed changes do not involve a significant reduction in a 
margin of safety.
    The SONGS safety analysis credits AFW pump delivery of 500 
[gallons per minute] gpm at a steam generator pressure of 1097 
[pounds per square inch absolute] psia and 700 gpm at a steam 
generator pressure of 890 psia to meet Accident Analysis flow 
requirements.
    The proposed amendment to Technical Specification 3.7.5 would 
allow a seven day Completion Time for Condition A for the turbine-
driven AFW pump if the inoperability occurs in MODE 3 following a 
refueling outage, if MODE 2 had not been entered. Extending the 
Completion Time does not involve a significant reduction in a margin 
of safety because: (1) During a return to power operations following 
a refueling outage, decay heat is at its lowest levels, (2) the 
other AFW trains are required to be OPERABLE when MODE 3 is entered, 
[and] (3) the motor-driven AFW train can provide sufficient flow to 
remove decay heat and cool the unit to Shutdown Cooling System entry 
conditions from power operations.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Branch Chief: Michael T. Markley.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of amendment request: October 20, 2009.
    Description of amendment request: The proposed amendment would 
delete paragraph d of Technical Specification 5.2.2, ``Unit Staff,'' 
superseded by Title 10 of the Code of Federal Regulations Part 26, 
Subpart I.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change removes Technical Specification (TS) 
restrictions on working hours for personnel who perform safety 
related functions. The TS restrictions are superseded by the worker 
fatigue requirements in 10 CFR Part 26. The proposed change does not 
impact the physical configuration or function of plant structures, 
systems, or components (SSCs) or the manner in which SSCs are 
operated, maintained, modified, tested, or inspected. Worker fatigue 
is not an initiator of any accident previously evaluated. Worker 
fatigue is not an assumption in the consequence mitigation of any 
accident previously evaluated.
    Therefore, it is concluded that this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change removes TS restrictions on working hours for 
personnel who perform safety related functions. The TS restrictions 
are superseded by the worker fatigue requirements in 10 CFR Part 26. 
Working hours will continue to be controlled in accordance with NRC 
requirements. The new rule allows for deviations from controls to 
mitigate or prevent a condition adverse to safety or as necessary to 
maintain the security of the facility. This ensures that the new 
rule will not unnecessarily restrict working hours and thereby 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    The proposed change does not alter the plant configuration, 
require new plant equipment to be installed, alter accident analysis 
assumptions, add any initiators, or effect the function of plant 
systems or the manner in which systems are operated, maintained, 
modified, tested, or inspected.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change removes TS restrictions on working hours for 
personnel who perform safety related functions. The TS restrictions 
are superseded by the worker fatigue requirements in 10 CFR Part 26. 
The proposed change does not involve any physical changes to plant 
or alter the manner in which plant systems are operated, maintained, 
modified, tested, or inspected. The proposed change does not alter 
the manner in which safety limits, limiting safety system settings 
or limiting conditions for operation are determined. The safety 
analysis acceptance criteria are not affected by this change. The 
proposed change will not result in plant operation in a 
configuration outside the design basis. The proposed change does not 
adversely affect systems that respond to safely shutdown the plant 
and to maintain the plant in a safe shutdown condition.
    Removal of plant-specific TS administrative requirements will 
not reduce a margin of safety because the requirements in 10 CFR 
Part 26 are adequate to ensure that worker fatigue is managed.
    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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: L. Raghavan.

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

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile 
Point Nuclear Station, Unit No. 1, Oswego Count, New York

    Date of application for amendment: July 2, 2009, as supplemented 
October 5, 2009.
    Brief description of amendment: The proposed amendment would revise 
the Technical Specifications (TS) by removing position indication for 
the relief valves from TS 3.6.11, ``Accident Monitoring 
Instrumentation.'' The proposed amendment would also correct an 
editorial error in the title of Table 4.6.11.
    Date of publication of individual notice in Federal Register: 
October 14, 2009 (74 FR 52826).
    Expiration date of individual notice: December 14, 2009.

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

    Date of application for amendment: September 18, 2009.
    Brief description of amendment: The proposed amendment would modify 
Technical Specification 3.2.9.1 and

[[Page 59265]]

4.2.7.1, ``Primary Coolant System Pressure Isolation Values,'' to 
incorporate requirements that are consistent with Section 3.4.5 of the 
Improved Standard TSs, NUREG-1433, Revision 3.
    Date of publication of individual notice in Federal Register: 
October 14, 2009 (74 FR 52824).
    Expiration date of individual notice: December 14, 2009.

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.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
Cliffs Independent Spent Fuel Storage Installation, Docket No. 72-8, 
Calvert County, Maryland

    Date of application for amendments: January 22, 2009, as 
supplemented by letters dated February 26, April 8, June 25, July 27, 
October 15, 19, 25 (two letters) 26, and 28, 2009.
    Brief description of amendments: The amendments conform the 
licenses to reflect the direct transfer of Calvert Cliffs Nuclear Power 
Plant, Inc. to Calvert Cliffs Nuclear Power Plant, LLC, as approved by 
Commission Order dated October, 2009. Transfer of the license will also 
authorize Calvert Cliffs Nuclear Power Plant, LLC to store spent fuel 
in the Calvert Cliffs independent spent fuel storage installation.
    Date of issuance: October 30, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 295 and 271.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the License.
    Date of initial notice in Federal Register: May 7, 2009 (74 FR 
21413).
    The letters dated February 26, April 8, June 25, July 27, October 
15, October 19, October 25 (two letters), October 26, and October 28, 
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 amendments is contained 
in a Safety Evaluation dated October 30, 2009.
    No significant hazards consideration comments received: The NRC 
received comments from a member of the public on May 22, 2009. The 
comments did not provide any information additional to that in the 
application, nor did they provide any information contradictory to that 
provided in the application.

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 14, 2008.
    Brief description of amendments: The amendments implemented 
Technical Specification Task Force (TSTF) Changes Travelers TSTF-479, 
Revision 0, ``Changes to Reflect Revision of [Title 10 of the Code of 
Federal Regulations] 10 CFR 50.55a,'' and TSTF-497, Revision 0, ``Limit 
Inservice Testing [IST] Program SR [Surveillance Requirements] 3.0.2 
Application to Frequencies of 2 Years or Less.'' TSTF-479 and TSTF-497 
revised the Technical Specification Administrative Controls section 
pertaining to requirements for the IST Program, consistent with the 
requirements of 10 CFR 50.55a(f)(4) for pumps and valves which are 
classified as American Society of Mechanical Engineers, Boiler and 
Pressure Vessel Code, Class 1, Class 2, and Class 3.
    Date of issuance: October 30, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 252 and 247.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the licenses and the technical specifications. The amendment 
also authorizes revisions to the Updated Facility Safety Analysis 
Report.
    Date of initial notice in Federal Register: April 7, 2009 (74 FR 
15769).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 30, 2009.
    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: October 8, 2008, supplemented 
by letter dated May 5, 2009.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) by removing and updating portions of the 
TSs which are out of date or are obsolete including footnotes and 
references.
    Date of issuance: October 30, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 253 and 248.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the licenses and the TSs.
    Date of initial notice in Federal Register: April 7, 2009 (74 FR 
15769). The supplement dated May 5, 2009 provided additional 
information that clarified the application, did not expand

[[Page 59266]]

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 October 30, 2009.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., (Entergy) Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: March 25, 2009.
    Brief description of amendment: The amendment added two Emergency 
Core Cooling System (ECCS) valves to Technical Specifications (TS) 
Surveillance Requirement (SR) 3.5.2.1 for checking valve position every 
7 days. The TS SR is designed to verify that ECCS valves whose single 
failure could cause loss of the ECCS function are in the required 
position with ac power removed so that misalignment or single failure 
cannot prevent completion of the ECCS function.
    Date of issuance: October 29, 2009.
    Effective date: As of the date of issuance, and shall be 
implemented prior to entering Mode 4 during startup from 2R19.
    Amendment No.: 263.
    Facility Operating License Nos. DPR-26 and DPR-64: The amendment 
revised the License and the Technical Specifications.
    Date of initial notice in Federal Register: May 19, 2009 (74 FR 
23444).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 29, 2009.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 5, 2009.
    Brief description of amendment: The proposed amendment would revise 
the Technical Specification (TS) Section 6.7.C to change requirements 
related to the schedule for performing the 10 CFR Part 50, Appendix J, 
Type A test. Specifically, the proposed change would change the TS from 
requiring the test ``no later than April 2010'' to ``prior to startup 
from the April 2010 refuel outage.''
    Date of Issuance: October 28, 2009.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 240.
    Facility Operating License No. DPR-28: Amendment revised the 
License and Technical Specifications.
    Date of initial notice in Federal Register: June 30, 2009 (74 FR 
31320).
    The Commission's related evaluation of the amendment is contained 
in a safety evaluation dated October 28, 2009.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 22, 2007, as supplemented by 
letters dated January 12 and October 22, 2009.
    Brief description of amendment: The amendment added a new license 
condition 2.c.(10) on the control room envelope (CRE) habitability 
program; revised the Technical Specification (TS) requirements related 
to the CRE habitability in TS 3.7.9, ``Control Room Emergency 
Ventilation System (CREVS)''; and added a new administrative controls 
program, TS 5.5.5, ``Control Room Envelope Habitability Program.'' 
These changes are consistent with the NRC-approved Industry/TS Task 
Force (TSTF) change traveler TSTF-448, Revision 3, ``Control Room 
Envelope Habitability.'' The availability of this TS improvement was 
published in the Federal Register on January 17, 2007 (72 FR 2022), as 
part of the Consolidated Line Item Improvement Process.
    Date of issuance: October 29, 2009.
    Effective date: As of its date of issuance and shall be implemented 
within 30 days from the implementation of the Alternate Source Term 
license Amendment No. 238.
    Amendment No.: 239.
    Renewed Facility Operating License No. DPR-51: Amendment revised 
the Technical Specifications/license.
    Date of initial notice in Federal Register: December 18, 2007 (72 
FR 71708). The supplemental letters dated January 12 and October 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 October 29, 2009.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 22, 2007, as 
supplemented by letter dated January 12, 2009.
    Brief description of amendment: The amendment added a new license 
condition 2.c.(11) on the control room envelope (CRE) habitability 
program; revised Technical Specification (TS) requirements related to 
the CRE habitability in TS 3/4.7.6, ``Control Room Emergency 
Ventilation and Air Conditioning System''; and added a new 
administrative controls program, TS 6.5.12, ``Control Room Envelope 
Habitability Program.'' These changes are consistent with the NRC-
approved Industry/TS Task Force (TSTF) change traveler TSTF-448, 
Revision 3, ``Control Room Envelope Habitability.'' The availability of 
this TS improvement was published in the Federal Register on January 
17, 2007 (72 FR 2022), as part of the Consolidated Line Item 
Improvement Process.
    Date of issuance: October 29, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of the implementation of the Alternate 
Source Term license Amendment No. 238 for Arkansas Nuclear One, Unit 
No. 1.
    Amendment No.: 288.
    Renewed Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications/license.
    Date of initial notice in Federal Register: December 18, 2007 (72 
FR 71710). The supplemental letter dated January 12, 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 October 29, 2009.
    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: June 3, 2009, as supplemented by letters 
dated September 22 and October 6, 2009.
    Brief description of amendment: The amendment modified the 
departure from nucleate boiling ratio (DNBR) safety limit in Technical 
Specification

[[Page 59267]]

(TS) 2.1.1.1, ``DNBR,'' based upon the Combustion Engineering 16x16 
Next Generation Fuel design and the associated departure from nucleate 
boiling correlations.
    Date of issuance: November 3, 2009.
    Effective date: As of the date of issuance and shall be implemented 
after the current cycle (Cycle 16) is completed and prior to the start 
of Cycle 17.
    Amendment No.: 224.
    Facility Operating License No. NPF-38: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: July 14, 2009 (74 FR 
34047). The supplements dated September 22 and October 6, 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 November 3, 2009.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 9, 2008, as supplemented by 
letters dated March 30, 2009 and September 4, 2009.
    Brief description of amendment: The amendment revised Surveillance 
Requirement 4.2.D to decrease the frequency of performing control rod 
drive rod notch testing from weekly to once per 31 days.
    Date of issuance: October 22, 2009.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 275.
    Renewed Facility Operating License No. DPR-16: The amendment 
revised the License and Technical Specifications.
    Date of initial notice in Federal Register: August 12, 2008 (73 FR 
46928). The supplements dated March 30, 2009 and September 4, 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.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 22, 2009.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power 
Station, Unit No. 1, DeWitt County, Illinois

    Date of application for amendment: April 22, 2009.
    Brief description of amendment: The amendment would revise the 
inservice testing (IST) requirements from the American Society of 
Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code, 
Section XI, to the ASME Code for Operation and Maintenance of Nuclear 
Power Plants (OM Code) and applicable addenda. This change would 
eliminate the ASME Code inconsistency between the IST program and the 
TS as required by Title 10 of the Code of Federal Regulations (10 CFR) 
50.55a(f)(5)(ii). Additionally, the amendment would extend the 
applicability of surveillance requirement (SR) 3.0.2 provisions to 
other normal and accelerated frequencies specified as 2 years or less 
in the IST program. Finally, the amendment will remove the phrase 
``including applicable supports'' from TS Section 5.5.6. TS Section 
5.5.6, IST Program, and the associated TS Bases would be revised under 
this TS amendment.
    Date of issuance: October 30, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 189.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: August 11, 2009 (74 FR 
40238).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 30, 2009.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of application for amendments: October 9, 2007,