Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 51326-51339 [E9-23780]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 74, Number 192 (Tuesday, October 6, 2009)]
[Notices]
[Pages 51326-51339]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E9-23780]


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NUCLEAR REGULATORY COMMISSION

[NRC-2009-0433]


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

[[Page 51327]]

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 September 10, 2009, to September 23, 2009. 
The last biweekly notice was published on September 22, 2009 (74 FR 
48316).

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 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

[[Page 51328]]

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 ViewerTM 
to access the Electronic Information Exchange (EIE), a component of the 
E-Filing system. The Workplace Forms ViewerTM 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 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 email to pdr.resource@nrc.gov.
Dairyland Power Cooperative, Docket No. 50-409, La Crosse Boiling Water 
Reactor, Genoa, Wisconsin (TAC J00359)
    Date of amendment request: July 28, 2009.
    Description of amendment request: The amendment application 
proposes changes to Technical Specifications, in support of the dry 
cask storage project at La Crosse Boiling Water Reactor. The 
application specifically proposes lower Fuel Element Storage Well water 
level limits and proposes changes to the definition of ``fuel 
handling.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated? 
No.
    The proposed change to the definition of FUEL HANDLING is an 
administrative

[[Page 51329]]

clarification and does not affect the operation of the plant or the 
postulated accidents in any way. The proposed changes to allow lower 
Fuel Element Storage Well (FESW) water level limits do not alter the 
manner in which individual fuel assemblies are moved or alter the 
design function of the FESW or any other structures, systems, and 
components used to ensure safe fuel storage. The total number of 
fuel assembly moves to the Dry Cask Storage System is exactly the 
same as that contemplated during original plant design when fuel was 
assumed to be transported from the plant directly to a disposal 
site. All of the accidents previously evaluated in the La Crosse 
Boiling Water Reactor (LACBWR) Decommissioning Plan have been 
reviewed for impact as a result of the proposed water level changes. 
The proposed changes do not affect the plant in such a manner that 
the likelihood or consequences of any previously evaluated accident 
is increased.
    Therefore, the 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? 
No.
    The proposed change to the definition of FUEL HANDLING is an 
administrative clarification and does not affect the operation of 
the plant in any way. The proposed changes to allow lower FESW water 
level limits do not alter the manner in which individual fuel 
assemblies are moved; or alter the design function of the FESW or 
any other structures, systems, and components used to ensure safe 
fuel storage. All of the accidents previously evaluated in the 
LACBWR Decommissioning Plan have been reviewed for impact as a 
result of the proposed water level changes. The existing accidents 
remain applicable and bounding for the LACBWR facility with the 
proposed changes in place and do not affect the plant in such a 
manner that a new accident has been created.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety? No.
    The proposed change to the definition of FUEL HANDLING is an 
administrative clarification and does not affect plant operation or 
safety margins in any way. The proposed changes to allow lower FESW 
water level limits do not alter the manner in which individual fuel 
assemblies are moved; or alter the design function of the FESW or 
any other structures, systems, and components used to ensure safe 
fuel storage. All of the accidents previously evaluated in the 
LACBWR Decommissioning Plan have been reviewed for impact as a 
result of the proposed water level changes. The likelihood and 
consequences of previously evaluated accidents remain applicable and 
bounding with the proposed changes in place; thus, safety margins 
remain the same.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the 
licensee's analysis and, based on this review, it appears that the 
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC 
staff proposes to determine that the amendment request involves no 
significant hazards consideration.
    NRC Branch Chief: Andrew Persinko.

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

    Date of amendment request: July 23, 2009.
    Description of amendment request: The proposed amendment would 
remove the level indicating instrument from the Technical Specification 
Surveillance Requirement (SR) for the refueling water storage tank, but 
leave the low level alarm function in the SR.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed change revises the existing Indian Point 3 
Refueling Water Storage Tank (RWST) Technical Specification (TS) 
Surveillance Requirement (SR) to remove the level indication 
function for the L-921 instrument loop. Removal of a TS SR for the 
level indication does not increase the probability of an accident 
occurring since it is not an accident initiator and does not 
increase the consequences of an accident since it is not performing 
any mitigating function and is not a post accident instrument. The 
proposed revision will not affect RWST lo-lo level alarm function 
used for operator guidance to begin sequencing to Recirculation Mode 
of Safety Injection during a postulated loss of coolant accident 
(LOCA). There will be no change in equipment qualification 
requirements or changes to the surveillance requirement for the lo-
lo level alarm. 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 accident previously evaluated?
    No. The proposed change removes the RWST level indication 
function from the RWST lo-lo level alarm surveillance requirement 
for the L-921 instrument loop. The proposed change does not involve 
installation of new equipment or modification of existing equipment, 
so that no new equipment failure modes are introduced. Also, the 
proposed change does not result in a change to the way that the 
equipment or facility is operated so that no new accident initiators 
are created. 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?
    No. The proposed change removes the RWST level indication 
function from the RWST io-lo level alarm surveillance requirement 
for the L-921 instrument loop. There is no change to the design 
requirements or the surveillance interval. The proposed change does 
not add the level indicating function elsewhere in the TS because it 
is a local level indication that is only used during normal 
operation and was never a post accident monitoring instrument. 
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: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Nancy L. Salgado.

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

    Date of amendment request: July 31, 2009.
    Description of amendment request: The proposed change would revise 
the JAFNPP Technical Specifications (TSs) Surveillance Requirements 
(SR) for testing of the Residual Heat Removal (RHR) System Shutdown 
Cooling (SDC) mode Containment Isolation, Reactor Pressure--High 
Function by replacing the current requirement to perform TS SR 
3.3.6.1.3, Perform Channel Calibration, with TS SR 3.3.6.1.1 Perform 
Channel Check, SR 3.3.6.1.2, Perform Channel Functional Test, SR 
3.3.6.1.4, Calibrate the Trip Units, and SR 3.3.6.1.5, Perform Channel 
Calibration. These changes are to support a proposed plant modification 
to increase the reliability of SDC isolation logic by changing the 
source of the reactor high pressure input signal.
    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:


[[Page 51330]]


    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The proposed change modifies the SRs that demonstrate the 
operability of the SDC Isolation, Reactor Pressure--High function. 
The current surveillance requirements include a 92-day calibration 
and a 24-month logic system functional test. These surveillance 
requirements are typical for pressure switches installed on 
dedicated process measurement lines. The proposed change in 
surveillance requirements is consistent with the use of ATTS [Analog 
Transmitter Trip System] transmitters installed on shared process 
measurement lines. The proposed surveillance requirements include 
the standard requirements applied to all ATTS equipment and thus 
will result in acceptable demonstration of the operability of the 
SDC Isolation Reactor Pressure--High function.
    The ATTS equipment that will be used for the SDC Isolation, 
Reactor Pressure--High function is classified as safety related and 
is environmentally qualified. The logic input configuration of the 
ATTS equipment will be the same as the configuration of the pressure 
switches. This will assure the same functionality currently 
performed by the pressure switches currently used for the SDC 
Isolation Reactor Pressure--High function. The reliability of the 
ATTS has been proven in other RPS [Reactor Protection System], PCIS 
[Primary Containment Isolation System], and ECCS [Emergency Core 
Cooling System] functions and is comparable to the reliability of 
the pressure switches that currently perform the SDC Isolation, 
Reactor Pressure--High function. Therefore, the consequences of any 
accident mitigated by the SDC Isolation, Reactor Pressure--High 
function will not increase.
    Based on these considerations, the proposed surveillance 
requirement changes do not involve a significant increase in the 
probability or consequences of an accident 'previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: No.
    The proposed change aligns the TS surveillance requirements with 
the type of equipment that will be used to supply the reactor 
pressure input to the SDC Isolation Reactor Pressure--High logic. 
Since the transmitters that will be used to supply the reactor 
pressure input are currently installed equipment there are no new 
accidents introduced by the equipment. The proposed change in SRs 
aligns the requirements with the--requirements currently imposed on 
the equipment in other JAF TS applications. The performance of the 
SDC Isolation, Reactor Pressure--High function, is not altered by 
changing the input source for reactor pressure parameter. Redundant 
power sources within the ATTS assure the functionality of the system 
during all plant operating modes that require the SDC Isolation, 
Reactor Pressure--High function. The proposed change will not 
introduce any new failure modes and, therefore, does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    The TS surveillance requirements that will be imposed on the SDC 
Isolation, Reactor Pressure--High function reflect the equipment 
that will perform that function. The proposed change in surveillance 
requirements will appropriately demonstrate the operability of the 
SDC Isolation, Reactor Pressure--High function.
    Since the proposed changes to the SRs are consistent with the 
SRs for ATTS transmitters in other RPS, PCIS, and ECCS applications 
the proposed requirements have been demonstrated to provide an 
adequate margin of safety. Therefore, the proposed change does not 
involve a significant reduction in any 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.

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

    Date of amendment request: August 5, 2009.
    Description of amendment request: Current Technical Specification 
(TS) 5.5.8, ``Inservice Testing Program,'' contains references to the 
American Society of Mechanical Engineers (ASME) Boiler and Pressure 
Vessel Code, Section XI as the source of requirements for the inservice 
testing (IST) of ASME Code Class 1, 2, and 3 pumps and valves. The 
proposed amendment would delete the references to Section XI of the 
Code and incorporate references to the ASME Code for Operation and 
Maintenance of Nuclear Power Plants (ASME OM Code). The proposed 
amendment would also indicate that there may be some nonstandard 
frequencies utilized in the IST Program in which the provisions of 
Surveillance Requirement 3.0.2 are applicable. The proposed changes are 
consistent with Technical Specification Task Force (TSTF) Technical 
Change Travelers 479-A, ``Changes to Reflect Revision to 10 CFR 
50.55a,'' and 497-A, ``Limit Inservice Testing Program SR 3.0.2 
Application to Frequencies of 2 Years or Less.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises TS 5.5.8, Inservice Testing Program, 
for consistency with the requirements of 10 CFR 50.55a(f)(4) for 
pumps and valves which are classified as American Society of 
Mechanical Engineers (ASME) Code Class 1, Class 2 and Class 3. The 
proposed change incorporates revisions to the ASME Code which is 
consistent with the expectations of 10 CFR 50.55(a).
    The proposed change does not impact any accident initiators or 
analyzed events or assumed mitigation of accident or transient 
events. The proposed change does 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. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change 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 does not introduce a new accident 
initiator, accident precursor, or malfunction mechanism. Therefore, 
this proposed change does not create the possibility of an accident 
or a different kind than previously evaluated.
    3. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises TS 5.5.8, Inservice Testing Program, 
for consistency with the requirements of 10 CFR 50.55a(f)(4) for 
pumps and valves which are classified as ASME Code Class 1, Class 2 
and Class 3. The proposed change incorporates revisions to the ASME 
Code, which is consistent with the expectations of 10 CFR 50.55a. 
The safety function of the affected pumps and valves are maintained. 
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, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the

[[Page 51331]]

amendment request involves no significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: July 30, 2009.
    Description of amendment request: The proposed amendment would 
delete Technical Specification (TS) Section 3.6.3.1, ``Containment 
Atmosphere Dilution (CAD) System,'' to modify containment combustible 
gas control requirements as permitted by Title 10 of the Code of 
Federal Regulations, Part 50 Section 50.44 (10 CFR 50.44). 10 CFR 50.44 
was revised on September 16, 2003, as noticed in the Federal Register 
(68 FR 54123).
    The Nuclear Regulatory Commission (NRC) staff issued a ``Notice Of 
Opportunity To Comment On Model Safety Evaluation, Model No Significant 
Hazards Determination, And Model Application For Licensees that Wish To 
Adopt TSTF-478, Revision 2, `BWR [Boiling-Water Reactor] Technical 
Specification Changes that Implement the Revised Rule for Combustible 
Gas Control,'' in the Federal Register on October 11, 2007 (72 FR 
57970). The notice included a model safety evaluation (SE) and a model 
no significant hazards consideration (NSHC) determination. On November 
21, 2007, the NRC staff issued a notice in the Federal Register (72 FR 
65610) announcing that the model SE and model NSHC determination may be 
referenced in plant-specific applications to adopt the changes. In its 
application dated July 30, 2009, the licensee affirmed the 
applicability of the model NSHC determination which is presented below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC is presented below:

Criterion 1: The proposed change does not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated

    The Containment Atmosphere Dilution (CAD) system is not an 
initiator to any accident previously evaluated. The TS Required 
Actions taken when a drywell cooling system fan is inoperable are 
not initiators to any accident previously evaluated. As a result, 
the probability of any accident previously evaluated is not 
significantly increased.
    The revised 10 CFR 50.44 no longer defines a design-basis 
accident (DBA) hydrogen release and the Commission has subsequently 
found that the DBA loss-of-coolant accident (LOCA) hydrogen release 
is not risk significant. In addition, CAD has been determined to be 
ineffective at mitigating hydrogen releases from the more risk 
significant beyond DBAs that could threaten containment integrity. 
Therefore, elimination of the CAD system will not significantly 
increase the consequences of any accident previously evaluated. The 
consequences of an accident while relying on the revised TS Required 
Actions for drywell cooling system fans are no different than the 
consequences of the same accidents under the current Required 
Actions. 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.

Criterion 2: The proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated

    No new or different accidents result from utilizing the proposed 
change. The proposed change permits physical alteration of the plant 
involving removal of the CAD system. The CAD system is not an 
accident precursor, nor does its existence or elimination have any 
adverse impact on the pre-accident state of the reactor core or post 
accident confinement of radionuclides within the containment 
building from any design basis event. The changes to the TS do not 
alter assumptions made in the safety analysis, but reflect changes 
to the design requirements allowed under the revised 10 CFR 50.44. 
The proposed change is consistent with the revised safety analysis 
assumptions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.

Criterion 3: The proposed change does not involve a significant 
reduction in a margin of safety

    The Commission has determined that the DBA LOCA hydrogen release 
is not risk significant, therefore is not required to be analyzed in 
a facility accident analysis. The proposed change reflects this new 
position and, due to remaining plant equipment, instrumentation, 
procedures, and programs that provide effective mitigation of and 
recovery from reactor accidents, including postulated beyond design 
basis events, does not result in a significant reduction in a margin 
of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    Based on the above, the NRC concludes that the proposed change 
presents no significant hazards consideration under the standards set 
forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no 
significant hazards consideration'' is justified.
    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. J. Bradley Fewell, Associate General 
Counsel, Exelon Generation Company LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Harold K. Chernoff.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station 
(FCS), Unit No. 1, Washington County, Nebraska

    Date of amendment request: May 29, 2009.
    Description of amendment request: The proposed amendment would: (1) 
Revise the definition for Operable-Operability in the FCS Technical 
Specifications (TS); (2) modify the provisions under which equipment 
may be considered operable when either its normal or emergency power 
source is inoperable; (3) delete TS limiting condition for operation 
(LCO) 2.0.1(2); (4) delete diesel generator surveillance requirement 
(SR) 3.7(1)e; and (5) relocate the guidance for inoperable power 
supplies and verifying operability of redundant components into the LCO 
for electrical equipment 2.7, Electrical Systems.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to revise the definition of operable-
operability, modify the provisions under which equipment may be 
considered operable when either its normal or emergency power source 
is inoperable, delete Technical Specification (TS) limiting 
conditions for operation (LCO) 2.0.1(2), and relocate the guidance 
for inoperable power supplies and verifying operability of redundant 
components into the LCO for electrical equipment is more aligned 
with NUREG-1432, Standard Technical Specifications [STS] for 
Combustion Engineering Plants, and does not adversely impact the 
probability of an accident

[[Page 51332]]

previously evaluated. The proposed changes are being made to address 
inconsistencies in guidance provided in TS 2.0.1(2) and TS 2.7(2). 
The proposed change does not affect the operability requirements for 
the emergency diesel generators (EDGs) or the house service 
transformers, and therefore does not impact the consequences of an 
analyzed accident.
    The new requirement added to TS 2.7 provides assurance that a 
loss of offsite power during the period that an EDG (or house 
service transformer) is inoperable, or loss of an EDG during the 
period that a house service transformer is inoperable, or loss of a 
house service transformer during the period that an EDG is 
inoperable, does not result in a complete loss of safety function of 
critical systems; thereby such a loss does not significantly 
increase the probability of an accident.
    Consistent with NUREG 1432, the 4-hour allowed time added to TS 
2.7(2)j for the EDGs, takes into account the capacity and capability 
of the remaining alternating current (AC) sources, a reasonable time 
for repairs, and the low probability of a design basis accident 
(DBA) occurring during this period. On a component basis, single 
failure protection for the required feature's function may have been 
lost; however, function has not been lost.
    Additionally, consistent with NUREG-1432, the 24-hour allowed 
time added to TS 2.7(2)b for the house service transformers takes 
into account the capacity and capability of the remaining AC 
sources, a reasonable time for repairs, and the low probability of a 
DBA occurring during this period.
    The proposed change removes the surveillance requirement (SR) to 
perform an inspection of the EDG on a refueling inspection frequency 
in accordance with the manufacturer's recommendations. This 
inspection is considered a maintenance activity, not an SR, and has 
no impact on the probability of an accident since EDGs are not 
initiators for any analyzed event. Deletion of TS SR 3.7(1)e from 
the TS does not impact the capability of the EDGs to perform their 
accident mitigation functions. The required EDG maintenance 
inspections will continue to be performed in accordance with the 
licensee-controlled EDG maintenance process. The consequences of an 
accident are not impacted because EDG operability is controlled by 
other portions of TS 3.7, which ensures that required surveillances 
are performed. The appropriate LCOs are entered in the event that 
EDG surveillance criteria are not met.
    As a result of redefining ``OPERABLE'' and adding the provision 
to TS 2.7(2)j, the statements ``provided there are no inoperable 
required engineered safeguards components which are redundant'' 
related to the electrical distribution components are being deleted 
from the other 2.7(2) TS for the buses, transformer, and motor 
control center (MCC) for clarification and consistency because these 
statements restrict only to engineered safeguards components. In 
addition, the administrative changes to renumber the existing TS 
sections ``TS 2.0.1(3) to 2.0.1(2)'' and TS 3.7(1)f to TS 3.7(1)e. 
are being made as a result of deletions to previous TS paragraphs 
and are being made for consistency and clarification. Rearranging 
the listing order of the MCCs in TS 2.7(1)f and TS 2.7(2)g in bus 
order clarifies the TS. As such, these editorial changes are not 
initiators of any accidents previously evaluated. As a result, the 
probability of an accident previously evaluated is not affected.
    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 changes do not involve a physical alteration to the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in methods governing normal plant operation. 
The proposed changes to TS 2.0.1(2) and TS 2.7 do not create the 
possibility of a new or different kind of accident since the design 
function of the affected equipment is not changed. No new 
interactions between systems or components are created. No new 
failure mechanisms of associated systems will exist.
    By deleting TS LCO 2.0.1(2) and including the guidance in TS 
2.7, inconsistencies in the existing TS will be eliminated. The new 
requirements added to TS 2.7 will include guidance to declare 
required systems or components without a normal or emergency power 
source available inoperable, when a redundant system or component is 
also inoperable. This provides assurance that a loss of offsite 
power, during the period that an EDG (or house service transformer) 
is inoperable, or loss of an EDG during the period that a house 
service transformer is inoperable (or vice versa), does not result 
in a complete loss of safety function of critical systems.
    No new failure mechanisms would be created. The proposed changes 
do not alter any assumptions made in the safety analyses. For the 
most part, the proposed changes are more aligned with the STS.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes to delete TS 2.0.1(2) and relocate the 
guidance for inoperable power supplies and verifying operability of 
redundant components to TS LCO 2.7(2)j, to delete the statement that 
MCC-3C1 may be inoperable in excess of 8 hours if battery chargers 
No. 1 and No. 2 are operable, and to delete the SR for inspecting 
the DG on a refueling frequency in accordance with the 
manufacturer's recommendations do not alter the manner in which 
safety limits or limiting safety system settings are determined. The 
safety analysis acceptance criteria are not affected by these 
proposed changes. The sources of power credited for design basis 
events are not affected by the proposed changes.
    The proposed changes to modify the provisions under which 
equipment may be considered operable when either its normal or 
emergency power source is inoperable, delete TS LCO 2.0.1(2), and 
relocate the guidance for inoperable power supplies and verifying 
operability of redundant components into the LCO for electrical 
equipment is more aligned with the STS. These changes are being made 
to address inconsistencies in guidance provided in TS 2.0.1(2) and 
TS 2.7(2). The proposed change does not reduce the operability 
requirements for the transformers, buses, MCCs, or EDGs and 
therefore will not result in plant operation in a configuration 
outside of the design basis.
    Further, the proposed change does not change the design function 
of any equipment assumed to operate in the event of an accident.
    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: David A. Repka, Esq., Winston & Strawn, 1700 
K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Michael T. Markley.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2 (SSES Units 1 and 2), Luzerne County, 
Pennsylvania

    Date of amendment request: March 24, 2009, as supplemented by 
letters dated April 24, and September 11, 2009.
    Description of amendment request: The proposed change revises the 
allowable value in the Technical Specification (TS) Table 3.3.5.1-1 
(Function 3.d) for the high-pressure coolant injection (HPCI) automatic 
pump suction transfer from the condensate storage tank (CST) to the 
suppression pool (SP). The present allowable value for this transfer is 
greater than or equal to 36 inches above the CST bottom. The proposed 
change is to increase the allowable value for this transfer to occur at 
greater than or equal to 40.5 inches above the CST bottom.
    Additionally, the proposed amendment also includes an editorial/
administrative change which corrects a typographical error in the SSES 
Units 1 and 2 TS Section 3.10.8.f.
    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:


[[Page 51333]]


    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed change to TS Table 3.3.5.1-1 increases the 
Technical Specification allowable value for the HPCI suction low 
level automatic transfer function from >= 3 6 inches to >= 40.5 
inches above the CST bottom. There are no process setpoint changes 
associated with this TS allowable value change. This TS change does 
not introduce the possibility of an increase in the probability or 
consequences of an accident because the HPCI automatic transfer 
function is not an initiator of any new accidents nor does it 
introduce any new failure modes. The CST is not safety related and 
therefore not credited in any design basis accident analyses. 
However, the CST reserve volume is credited in anticipated 
transients without scram (ATWS), Appendix R and station blackout 
(SBO) evaluations. The reserve volume available in the CST at the 
proposed allowable value of 40.5 inches above the CST bottom remains 
adequate to fully support these HPCI system support functions and 
the change fully supports HPCI system operation. The reserve volume 
is not reduced as a result of the proposed change in the TS 
allowable value since the transfer will still occur at the CST low 
level instrument setpoint of 43.5 inches above tank bottom, which 
remains unchanged.
    The HPCI system automatic transfer function occurs at the point 
in a design basis accident (DBA) when the CST level reaches the low 
level transfer setpoint. This proposed change will require the HPCI 
pump suction to be transferred from the CST to the SP at 40.5 inches 
versus 36 inches above the CST bottom. Currently, the TS allow this 
transfer to occur at 36 inches. This proposed change is conservative 
because it assures the suction transfer will occur while there is 
more water in the tank, thus eliminating the possibility of vortex 
formation and air intrusion to the HPCI pump suction. Since this 
proposed change ensures the HPCI system automatic suction transfer 
function occurs without adversely impacting HPCI system operation, 
it does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed editorial/administrative change is necessary to 
correct a typographical error in the SSES Units 1 and 2 TS Section 
3.10.8.f. This editorial change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. As discussed above, the proposed change to TS Table 3.3.5.1-
1 involves increasing the TS allowable value for the HPCI low level 
automatic transfer function from the CST to the SP at >= 36 inches 
to >= 40.5 inches above the CST tank bottom. This change ensures the 
HPCI automatic transfer function occurs without introducing the 
possibility of vortex formation or air intrusion in the HPCI pump 
suction path. All HPCI system support functions remain unaffected by 
this change. This TS change does not introduce the possibility of a 
new accident because the HPCI automatic transfer function is not an 
initiator of any accident and no new failure modes are introduced. 
There are no new types of failures or new or different kinds of 
accidents or transients that could be created by these changes. 
Therefore, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed editorial/administrative change only corrects a 
typographical error in the SSES Units 1 and 2 TS Section 3.10.8.f. 
This editorial 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?
    No. The margin of safety is established through equipment 
design, operating parameters, and the setpoints at which automatic 
actions are initiated. The proposed change to TS Table 3.3.5.1-1 
involves increasing the allowable level at which the HPCI automatic 
suction transfer from the CST to the SP must occur to avoid the 
possibility of vortex formation or air intrusion into the HPCI pump. 
This change does not result in a change to the level switch 
setpoint, which initiates the HPCI suction transfer from the CST to 
the SP. Although the allowable value for the transfer is now closer 
to the process setpoint for activation of the level switch, this 
reduction in operating margin was reviewed and determined to be 
acceptable. The level switch setpoint tolerances were established 
based on historical instrument data and instrument characteristics. 
These tolerances provide adequate margin to the proposed TS 
allowable value of 40.5 inches above the CST bottom. The tolerances 
further ensure the transfer will occur prior to level reaching the 
technical specification allowable value. Therefore, the proposed 
change does not result in a significant reduction in a margin of 
safety.
    The proposed editorial/administrative change only corrects a 
typographical error in the SSES Units 1 and 2 TS Section 3.10.8.f. 
This editorial change does not result in 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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Branch Chief: Nancy L. Salgado.

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

    Date of amendment request: July 30, 2009.
    Description of amendment request: The proposed amendment would 
relocate Technical Specification (TS) surveillance requirements (SRs) 
for the reactor recirculation system motor-generator (MG) set scoop 
tube stop settings to the Technical Requirements Manual (TRM). 
Specifically, the proposed amendment would relocate TS SR 4.4.1.1.3 to 
the TRM which is a licensee-controlled document. SR 4.4.1.1.3 requires 
that each MG set scoop tube mechanical and electrical stop be 
demonstrated operable with overspeed setpoints less than or equal to 
109% and 107%, respectively, of rated core flow, at least once per 18 
months.
    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. The Nuclear Regulatory Commission (NRC) staff's review 
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 major components in the MG set consist of a motor, fluid 
coupler and a generator. The motor drives the generator through the 
fluid coupler. The speed and output of the generator rise and fall 
as the volume of fluid in the coupler is varied by changing the 
position of the scoop tube. As the generator's output increases or 
decreases, the speed of the recirculation pump follows suit. The 
scoop tube mechanism has both mechanical and electrical overspeed 
stops that limit recirculation flow by limiting the MG set speed. 
The electrical stop actuates first. The mechanical stop is designed 
to prevent the scoop tube motion if the electrical stop fails or to 
mitigate overshoot of the electrical stop. The electrical stops are 
not credited in any of the accident or transient analyses. The 
mechanical stop settings are an input used in the determination of 
the flow dependent minimum critical power ratio (MCPR) and the 
linear heat generation rate (LHGR) or average planar linear heat 
generation rate (APLHGR) operating limits. These operating limits 
are established and documented on a cycle-specific basis in the core 
operating limits report (COLR) in accordance with TS 6.9.1.9. 
Operation within the MCPR, LGHR and APLHGR operating limits is 
required in accordance with TSs 3.2.3, 3.2.4, and 3.2.1, 
respectively.
    Once relocated, any future changes to the surveillance 
requirements for the MG set scoop tube mechanical and electrical 
stop settings would be controlled by 10 CFR 50.59.
    There are no physical plant modifications associated with this 
change. The proposed amendment would not alter the way any 
structure, system, or component (SSC) functions and would not alter 
the way the plant is operated. As such, the proposed

[[Page 51334]]

amendment would have no impact on the ability of the affected SSCs 
to either preclude or mitigate 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 amendment would not change the design function or 
operation of the SSCs involved and would not impact the way the 
plant is operated. As such, the proposed change would not introduce 
any new failure mechanisms, malfunctions, or accident initiators not 
already considered in the design and licensing bases. 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 margin of safety is associated with the confidence in the 
ability of the fission product barriers (i.e., fuel cladding, 
reactor coolant pressure boundary, and containment structure) to 
limit the level of radiation to the public. There are no physical 
plant modifications associated with the proposed amendment. The 
proposed amendment would not alter the way any SSC functions and 
would not alter the way the plant is operated. The proposed 
amendment would not introduce any new uncertainties or change any 
existing uncertainties associated with any safety limit. The 
proposed amendment would have no impact on the structural integrity 
of the fuel cladding, reactor coolant pressure boundary, or 
containment structure. Based on the above considerations, the NRC 
staff concludes that the proposed amendment would not degrade the 
confidence in the ability of the fission product barriers to limit 
the level of radiation to the public. Therefore, the proposed change 
does not involve a significant reduction in a margin of safety.

    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.

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 
County, Maryland
    Date of application for amendments: May 21, 2009.
    Brief description of amendments: The amendments removed the Table 
of Contents from the Technical Specifications and place them under 
licensee control.
    Date of issuance: September 21, 2009.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment Nos.: 293 and 269.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments 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 these amendments is 
contained in a Safety Evaluation dated September 21, 2009.
    No significant hazards consideration comments received: No.
Duke Power Company LLC, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina.
    Duke Power Company LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina.
    Duke Power Company LLC, 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: February 27, 2009.
    Brief description of amendments: The amendments deleted those 
portions of the Technical Specifications (TSs) superseded by the Code 
of Federal Regulations, Part 26, Subpart I. The changes are consistent 
with Nuclear Regulatory Commission (NRC)-approved Revision 0 to 
Technical Specification Task Force (TSTF) Improved Standard Technical 
Specification Change Traveler, TSTF-511, ``Eliminate Working Hour 
Restrictions from TS 5.2.2 to Support Compliance with 10 CFR Part 26.''
    Date of issuance: September 21, 2009.
    Effective date: As of the