[Federal Register Volume 74, Number 163 (Tuesday, August 25, 2009)]
[Notices]
[Pages 42926-42940]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E9-20403]


=======================================================================
-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION

[NRC-2009-0363]


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 July 30, 2009 to August 12, 2009. The last 
biweekly notice was published on August 11, 2009 (74 FR 40233).

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, http://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

[[Page 42927]]

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 2007 (72 FR 49139, August 28, 2007). The E-Filing 
process requires participants to submit and serve all adjudicatory 
documents over the Internet, or in some cases to mail copies on 
electronic storage media. Participants may not submit paper copies of 
their filings unless they seek an exemption in accordance with the 
procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
ten (10) days prior to the filing deadline, the petitioner/requestor 
should contact the Office of the Secretary by e-mail at 
[email protected], 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 http://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 http://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 http://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 http://www.nrc.gov/site-help/e-submittals.html or by calling the NRC electronic filing Help Desk, 
which is available between 8 a.m. and 8 p.m., Eastern Time, Monday 
through Friday, excluding government holidays. The toll-free help line 
number is 1-866-672-7640. A person filing electronically may also seek 
assistance by sending an e-mail to the NRC electronic filing Help Desk 
at [email protected].
    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 
http://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

[[Page 42928]]

available records will be accessible from the ADAMS Public Electronic 
Reading Room on the Internet at the NRC Web site, http://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 [email protected].

Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc., 
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, 
Louisiana

    Date of amendment request: June 29, 2009.
    Description of amendment request: The proposed amendment would 
revise the requirements in Technical Specification (TS) 5.5.6, 
``Inservice Testing Program.'' TS 5.5.6 currently contains references 
to the American Society of Mechanical Engineers Boiler and Pressure 
Vessel Code (ASME 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 changes would delete the references to Section Xl 
of the ASME Code and incorporate references to the ASME Code for 
Operation and Maintenance of Nuclear Power Plants (ASME OM Code). In 
addition, the proposed amendment would address the applicability of 
Surveillance Requirement 3.0.2 to other normal and accelerated 
frequencies as 2 years or less in the IST program. These changes are 
consistent with changes identified in the Improved Standard Technical 
Specifications (ISTS) by Technical Specification Task Force Traveler 
(TSTF) Nos. 479 and 497.
    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 the Technical Specification 
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 that result in a net improvement in the measures for 
testing pumps and valves.
    The proposed changes revise TS 5.5.6 for RBS to conform to the 
requirements of 10 CFR 50.55a(f) regarding the IST of pumps and 
valves for the third 10-Year Interval. The current TS reference the 
ASME Boiler and Pressure Vessel Code, Section XI, requirements for 
the IST of ASME Code Class 1, 2, and 3 pumps and valves. The 
proposed changes would reference the ASME OM Code instead. This is 
consistent with 10 CFR 50.55a(f). The proposed changes are 
administrative in nature.
    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, 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?
    Response: No.
    The proposed change revises the Technical Specification 
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 that result in a net 
improvement in the measures for testing pumps and valves.
    The proposed TS changes do not involve physical changes to the 
facility. In addition, the proposed changes have no affect on plant 
configuration, or method of operation of plant structures, systems, 
or components.
    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?
    Response: No.
    The proposed change revises the Technical Specification 
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 that result in a net 
improvement in the measures for testing pumps and valves.
    The change does not involve a physical change to the plant or a 
change in the manner in which the plant is operated or controlled. 
The IST of the Class 1, 2, and 3 pumps and valves continue to meet 
the appropriate requirements.
    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: Terence A. Burke, Associate General 
Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Michael T. Markley.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California

    Date of amendment request: July 3, 2009.
    Description of amendment request: The proposed amendments would 
revise the operability requirements and actions in Technical 
Specification (TS) 3.4.15, ``RCS [Reactor Coolant System] Leakage 
Detection Instrumentation,'' and the associated Bases Section to 
reflect the revised TSs.
    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 reduces the time allowed for the plant to 
operate when the only Technical Specification (TS) 3.4.15 operable 
Reactor Coolant System (RCS) leakage instrumentation monitor is the 
containment atmosphere gaseous radioactivity monitor, and revises 
the basis for operability for the containment sump monitors, 
containment atmosphere particulate radioactivity monitor, 
containment atmosphere gaseous radioactivity monitor, and the 
containment fan cooler unit condensate collection monitor. The 
proposed change increases the allowed operating time when all RCS 
leakage detection system instrumentation is inoperable. The proposed 
change also removes the word ``required'' from TS 3.4.15 Condition 
A, Required Action A.2, Condition B, and Required Action B.2, 
revises TS 3.4.15 Condition A to apply to any containment sump 
monitor, and revises the name of the containment fan cooler unit 
(CFCU) condensate collection monitor in the TS 3.4.15 Actions. The 
monitoring of RCS leakage is not a precursor to any accident 
previously evaluated. The monitoring of RCS leakage is not used to 
mitigate the consequences of any accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant or the addition of new or different type of

[[Page 42929]]

equipment. The change does not involve a change in how the plant is 
operated.
    Therefore, the proposed change does not create the possibility 
of a new or different accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The change that reduces the allowed time of operation with only 
the least accurate containment atmosphere gaseous radiation monitor 
operable increases the margin of safety by increasing the likelihood 
that an increase in RCS leakage will be detected before it 
potentially results in gross failure. For the change that allows a 
limited period of time to restore at least one RCS leakage detection 
monitor to operable status when all leakage detection monitors are 
inoperable, two sources of diverse leakage detection capability are 
required to be provided during the limited period. Allowing a 
limited period of time to restore at least one RCS leakage detection 
instrument to operable status before requiring a plant shutdown 
avoids the situation of putting the plant through a thermal 
transient without RCS leakage monitoring. The change to TS 3.4.15 
Condition A, Required Action A.2, Condition B, Required Action B.2, 
Condition C, and Required Action C.2.2 is consistent with TS 
[Limiting Condition for Operation] 3.4.15 and does not impact the 
RCS leakage instrumentation. The revision to the TS bases for 
operability of the RCS leakage instrumentation monitors does not 
involve a change in the leakage instrumentation and is consistent 
with the original design of the leakage instrumentation.
    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 requests involve no significant hazards consideration.
    Attorney for licensee: Jennifer Post, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Branch Chief: Michael T. Markley.

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: April 9, 2009.
    Description of amendment request: The proposed amendment would 
relocate Technical Specification (TS) requirements pertaining to 
communications during refueling operations (TS 3/4.9.5), manipulator 
crane operability (TS 3/4.9.6), and crane travel (TS 3/4.9.7) to the 
Technical Requirements Manual.
    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 NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The 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 proposed amendment would relocate TS requirements to the 
Technical Requirements Manual (TRM) which is a licensee-controlled 
document. The TS requirements to be relocated relate to control room 
communications during refueling, operability of the manipulator 
crane and auxiliary hoist for movement of control rods or fuel 
assemblies within the reactor pressure vessel, and control of heavy 
loads over fuel assemblies in the fuel storage pool. Once relocated, 
any future changes would be controlled by 10 CFR 50.59. The proposed 
amendment is administrative in nature from the standpoint that the 
current TS requirements would be relocated verbatim to the TRM. 
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 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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Harold K. Chernoff.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: February 3, 2009.
    Description of amendment request: The proposed amendment would 
revise the Operating Licenses to deviate from certain South Texas 
Project Fire Protection Program requirements. The amendment will allow 
the performance of operator manual actions to achieve and maintain safe 
shutdown in the event of a fire in lieu of meeting circuit separation 
protection requirements of Title 10 of the Code of Federal Regulations 
(10 CFR), Part 50, Appendix R, Section III.G.2 for Fire Area 31.
    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 design functions of structures, systems and component[s] are 
not impacted by the proposed change. The proposed change involves 
operator manual actions in response to a fire and will not initiate 
an event. The proposed actions do not increase the probability of 
occurrence of a fire or any other accident previously evaluated.
    The proposed actions are feasible and reliable and demonstrate 
that the unit can be safely shutdown in the event of a fire. No 
significant consequences result from the performance of the proposed 
actions.
    Therefore, the proposed change does not involve a significant 
increase in the

[[Page 42930]]

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 design functions of structures, systems and component[s] are 
not impacted by the proposed amendment. The proposed change involves 
operator manual actions in response to a fire. They do not involve 
new failure mechanisms or malfunctions that can initiate a new 
accident.
    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.
    Adequate time is available to perform the proposed operator 
manual actions to account for uncertainties in estimates of the time 
available and in estimates of how long it takes to diagnose and 
execute the actions. The actions are straightforward and do not 
create any significant concerns. The actions have been verified that 
they can be performed through demonstration and they are 
proceduralized. The proposed actions are feasible and reliable and 
demonstrate that the unit can be safely shutdown in the event of a 
fire.
    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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendment involves no significant hazards consideration.
    Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Branch Chief: Michael T. Markley.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: March 3, 2009.
    Description of amendment request: The proposed change would revise 
the Operating Licenses to deviate from certain South Texas Project Fire 
Protection Program requirements. The amendment will allow the 
performance of operator manual actions to achieve and maintain safe 
shutdown in the event of a fire in lieu of meeting circuit separation 
protection requirements of Title 10 of the Code of Federal Regulations 
(10 CFR), Part 50, Appendix R, Section III.G.2 for Fire Area 27.
    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 design functions of structures, systems and components are 
not impacted by the proposed change. The proposed change involves 
operator manual actions in response to a fire, and will not initiate 
an event. The proposed actions do not increase the probability of 
occurrence of a fire or any other accident previously evaluated.
    The proposed actions are feasible and reliable and demonstrate 
that the unit can be safely shutdown in the event of a fire. No 
significant consequences result from the performance of the proposed 
actions.
    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 design functions of structures, systems and components are 
not impacted by the proposed amendment. The proposed change involves 
operator manual actions in response to a fire. They do not involve 
new failure mechanisms or malfunctions that can initiate a new 
accident.
    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 rendition 
in a margin of safety?
    Response: No.
    Adequate time is available to perform the proposed operator 
manual actions to account for uncertainties in estimates of the time 
available and in estimates of how long it takes to diagnose and 
execute the actions. The actions are straightforward and do not 
create any significant concerns. The actions have been verified that 
they can be performed through demonstration and they are 
proceduralized. The proposed actions are feasible and reliable and 
demonstrate that the unit can be safely shutdown in the event of a 
fare.
    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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendment involves no significant hazards consideration.
    Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    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: July 9, 2009.
    Description of amendment request: The proposed amendment would 
allow use of a dedicated on-line core power distribution monitoring 
system (PDMS) to enhance surveillance of core thermal limits and would 
revise Technical Specification (TS) TS 1.1, ``Definitions,'' TS 3.1.8, 
``Rod Position Indication,'' TS 3.2.1, ``Heat Flux Hot Channel 
Factor,'' TS 3.2.4, ``Quadrant Power Tilt Ratio (QPTR)'', and TS 3.3.1, 
``Reactor Trip System (RTS) Instrumentation.''
    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 Power Distribution Monitoring System (PDMS) performs 
essentially continuous core power distribution monitoring with data 
input from existing plant instrumentation. This system utilizes an 
NRC-approved Westinghouse proprietary computer code, i.e. Best 
Estimate Analyzer for Core Operations--Nuclear (BEACON), to provide 
data reduction for incore flux maps, core parameter analysis, load 
follow, operation simulation, and core prediction. The PDMS does not 
provide any protection or control system function. Fission product 
barriers are not impacted by these proposed changes. The proposed 
changes occurring with PDMS will not result in any additional 
challenges to plant equipment that could increase the probability of 
any previously evaluated accident. The changes associated with the 
PDMS do not affect plant systems such that their function in the 
control of radiological consequences is adversely affected. These 
proposed changes will, therefore, not affect the mitigation of the 
radiological consequences of any accident described in the Updated 
Final Safety Analysis Report (UFSAR).
    Use of the PDMS supports maintaining the core power distribution 
within required limits. Further, continuous on-line monitoring 
through the use of PDMS provides significantly more information 
about the power distributions present in the core than is currently 
available. This result in more time (i.e. earlier determination of 
an adverse condition developing) for operator action prior to having 
an adverse condition develop that could lead to an accident 
condition or to unfavorable initial conditions for an accident.
    Therefore, the proposed change does not involve a significant 
increase in the

[[Page 42931]]

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.
    Other than use of the PDMS to monitor core power distribution, 
implementation of the PDMS and associated Technical Specification 
changes has no impact on plant operations or safety, nor does it 
contribute in any way to the probability or consequences of an 
accident. No safety related equipment, safety function, or plant 
operation will be altered as a result of this proposed change. The 
possibility for a new or different type of accident from any 
accident previously evaluated is not created since the changes 
associated with implementation of the PDMS do not result in a change 
to the design basis of any plant component or system. The evaluation 
of the effects of using the PDMS to monitor core power distribution 
parameters shows that all design standards and applicable safety 
criteria limits are met.
    The proposed changes do not result in any event previously 
deemed incredible being made credible. Implementation of the PDMS 
will not result in any additional adverse condition and will not 
result in any increase in the challenges to safety systems. The 
cycle specific variables required by the PDMS are calculated using 
NRC approved methods. The Technical Specifications will continue to 
require operation within the required core operating limits, and 
appropriate actions will continue to be taken when or if limits are 
exceeded.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of an accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    No margin of safety is adversely affected by the implementation 
of the PDMS. The margins of safety provided by current Technical 
Specification requirements and limits remain unchanged, as the 
Technical Specifications will continue to require operation within 
the core limits that are based on NRC approved reload design 
methodologies. Appropriate measures exist to control the values of 
these cycle specific limits, and appropriate actions will continue 
to be specified and taken for when limits are violated. Such actions 
remain unchanged.
    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.

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

    Date of amendment request: March 20, 2009.
    Description of amendment request: The proposed amendment would 
revise the Operating License No. NPF-30 for Callaway Plant, Unit 1, in 
order to incorporate a change to Technical Specification (TS) 5.5.16, 
``Containment Leakage Rate Testing Program,'' which establishes the 
program for leakage rate testing of the containment, as required by 
Title 10 of Code of Federal Regulations (10 CFR) Section 50.54, 
``Conditions of licenses,'' Subsection (o) and 10 CFR 50, Appendix J, 
``Primary Reactor Containment Leakage Testing for Water-Cooled Power 
Reactors,'' Option B, ``Performance Based Requirements,'' as modified 
by approved exemptions. Specifically, the TS 5.5.16 would be revised to 
reflect a one-time 5-year deferral of the containment Type A integrated 
leak rate test (ILRT) from once in 10 years to once in 15 years.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No
    The proposed change will revise Callaway Plant TS 5.5.16, 
``Containment Leakage Rate Testing Program,'' to reflect a one-time, 
five-year extension for the containment Type A test date to enable 
the implementation of a 15-year test interval. While the containment 
is designed to contain radioactive material that may be released 
from the reactor core following a design basis Loss-of-Coolant 
Accident (LOCA), the test interval associated with Type A testing is 
part of ensuring the plant's ability to mitigate the consequences of 
accidents described in the FSAR [Final Safety Analysis Report] and 
does not involve a precursor or initiator of any accident previously 
evaluated. Thus, the proposed change to the Type A test interval 
cannot increase the probability of an accident previously evaluated 
in the FSAR.
    Type A testing does provide assurance that the containment will 
not exceed allowable leakage rate criteria specified in the TS and 
will continue to perform its design function following an accident. 
However, per NUREG-1493, ``Performance-Based Containment Leak-Test 
Program,'' Type A tests identify only a few potential leakage paths 
that cannot be identified by Type B and C testing. The current Type 
B and C penetration test frequencies for Callaway are established 
based on performance, using the requirements of 10 CFR 50, Appendix 
J, Option B, and the Type B and C testing requirements will not be 
changed as a result of the proposed license amendment. As a result, 
with respect to the consequences of an accident, a risk assessment 
of the proposed change has concluded that there is an insignificant 
increase in total population dose rate and an insignificant increase 
in the conditional containment failure probability.
    Based on the above, the proposed change does not involve 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 change is for a one-time, five-year extension of 
the Type A test for Callaway Plant and will not affect the control 
parameters governing unit operation or the response of plant 
equipment to transient or accident conditions. The proposed change 
does not introduce new equipment, modes of system operation, or 
failure mechanisms.
    Therefore, based on the above, the proposed change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No
    The Callaway Plant containment consists of the concrete 
containment building, its steel liner, and the penetrations through 
this structure. The structure is designed to contain radioactive 
material that may be released from the reactor core following a 
design basis LOCA. Additionally, this structure provides shielding 
from the fission products that may be present in the containment 
atmosphere following accident conditions.
    The containment is a prestressed, reinforced concrete, 
cylindrical structure with a hemispherical dome and a reinforced 
concrete base slab. The inside structure is lined with a carbon 
steel liner to ensure a high degree of leak tightness during 
operating and accident conditions. A post-tensioning system is used 
to prestress the cylindrical shell and dome.
    The concrete containment building is required for structural 
integrity of the containment under Design Basis Accident (DBA) 
conditions. The steel liner and its penetrations establish the 
leakage-limiting boundary of the containment. Maintaining 
operability of the containment will limit leakage of fission product 
radioactivity released from the containment to the environment.
    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 an ILRT, as required by 10 CFR 50, Appendix J, ``Primary 
Reactor Containment Leakage Testing for Water-Cooled Power 
Reactors.''
    The existing 10-year interval at Callaway Plant is based on past 
performance. Previous Type A tests conducted at Callaway Plant

[[Page 42932]]

indicate that leakage from containment has been less than all 10 CFR 
50 Appendix J, Option B, leakage limits.
    The proposed change for a one-time extension of the Type A test 
does not affect the method for Type A, B, or C testing or the test 
acceptance criteria. Type B and C testing will continue to be 
performed at the frequency required by Callaway Plant Technical 
Specifications. The containment inspections that are performed in 
accordance with the requirements of the American Society of 
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section 
XI, ``Inservice Inspection,'' and 10 CFR 50.65, ``Requirements for 
Monitoring the Effectiveness of Maintenance at Nuclear Power 
Plants,'' provide a high degree of a assurance that the containment 
will not degrade in a manner that is only detectable by Type A 
testing.
    In NUREG-1493, ``Performance-Based Containment Leak-Test 
Program,'' the NRC indicated that a 20-year extension for Type A 
testing resulted in an imperceptible increase in risk to the public. 
The NUREG-1493 study also concluded that, generically, the design 
containment leak rate contributes a very small amount to the 
individual risk and that the decrease in Type A testing frequency 
would have a minimal affect on this risk. AmerenUE has conducted 
risk assessments to determine the impact of a one-time change to the 
Callaway Plant Type A test schedule from a baseline value of once in 
10 years to once in 15 years for the risk measures of Large Early 
Release Frequency (LERF), Total Population Dose, and Conditional 
Containment Failure Probability (CCFP). The results of the risk 
assessments indicate that the proposed change to the Callaway Plant 
Type A test schedule has a minimal impact on public risk.
    Based on the above and on previous Type A test results for the 
Callaway Plant containment, the current containment surveillance 
program, and the results of the AmerenUE risk assessment, there is 
no reduction in the effectiveness of the Callaway Plant containment 
as a barrier to the release of the post-accident containment 
atmosphere to the public or to personnel in the Control Room. Thus, 
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: John O'Neill, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: May 4, 2009.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.7.3, ``Main Feedwater Isolation 
Valves (MFIVs) and Main Feedwater Regulating Valves (MFRVs), and Main 
Feedwater Regulating Valve Bypass Valves (MFRVBVs),'' so that the 
limiting condition for operation (LCO) and Applicability more 
accurately reflect the conditions for when the LCO should be applicable 
and more effectively provide appropriate exceptions to the 
Applicability for certain valve configurations. The amendment would 
incorporate other minor changes; the title to TS 3.7.3 and the header 
for each TS page would be revised, and the exception footnotes in TS 
Table 3.3.2-1 of TS 3.3.2, ``ESFAS [Engineered Safety Features 
Actuation System] Instrumentation,'' would be revised to improve the 
application of existing notes and/or incorporate more appropriate notes 
as applicable.
    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 alter any design or operating 
limits, nor do they physically alter safety-related systems, nor do 
they affect the way in which safety-related systems perform their 
functions. The proposed changes do not change accident initiators or 
precursors assumed or postulated in the FSAR [Final Safety Analysis 
Report]-described accident analyses, nor do they alter the design 
assumptions, conditions, and configuration of the facility or the 
manner in which the plant is normally operated and maintained. The 
proposed changes do not alter or prevent the ability of structures, 
systems, and components (SSCs) from performing their intended 
functions to mitigate the consequences of an initiating event within 
the assumed acceptance limits. With specific regard to the proposed 
TS changes, although the changes involve the exceptions contained in 
the Applicability of TS 3.7.3 as well as the notes attached to TS 
Table 3.3.2-1 (which are themselves exceptions), the provisions of 
the exceptions and notes would continue to be based on the premise 
that adequate isolation or isolation capability exists for the main 
feedwater lines, i.e., that the required safety function is 
performed or capable of being performed as required or assumed for 
mitigation of the applicable postulated accidents.
    All accident analysis acceptance criteria will therefore 
continue to be met with the proposed changes. The proposed changes 
will not affect the source term, containment isolation, or 
radiological release assumptions used in evaluating the radiological 
consequences of an accident previously evaluated. The proposed 
changes will not alter any assumptions or change any mitigation 
actions in the radiological consequence evaluations in the FSAR. The 
applicable radiological dose acceptance criteria will continue to be 
met. Overall protection system performance will remain within the 
bounds of the previously performed accident analyses.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No
    There are no proposed design changes, nor are there any changes 
in the method by which any safety-related plant structure, system, 
or component (SSC) performs its specified safety function. The 
proposed changes will not affect the normal method of plant 
operation or change any operating parameters. No equipment 
performance requirements will be affected. The proposed changes will 
not alter any assumptions made in the safety analyses. No new 
accident scenarios, transient precursors, failure mechanisms, or 
limiting single failures will be introduced as a result of this 
amendment. There will be no adverse effect or challenges imposed on 
any safety-related system as a result of this amendment. The 
proposed amendment will not alter the design or performance of the 
7300 Process Protection System, Nuclear Instrumentation System, or 
Solid State Protection System used in the plant protection systems.
    Therefore, the proposed changes do not create the possibility of 
a new or different accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No
    There will be no effect on those plant systems necessary to 
assure the accomplishment of protection functions. There will be no 
impact on the overpower limit, departure from nucleate boiling ratio 
(DNBR) limits, heat flux hot channel factor (FQ), nuclear 
enthalpy rise hot channel factor (F[Delta]H), loss of coolant 
accident peak cladding temperature (LOCA PCT), peak local power 
density, or any other margin of safety. The applicable radiological 
dose consequence acceptance criteria for design-basis transients and 
accidents will continue to be met. The proposed changes do not 
eliminate any surveillances or alter the frequency of surveillances 
required by the Technical Specifications. None of the acceptance 
criteria for any accident analysis will be changed.

[[Page 42933]]

    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: John O'Neill, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: May 4, 2009.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.7.2, ``Main Steam Isolation 
Valves (MSIVs),'' to add the main steam isolation valve bypass valves 
(MSIVBVs) and main steam low point drain isolation valves (MSLPDIVs) to 
the scope of the TS. In addition, the proposed amendment would make 
editorial changes to the title and header on each page of TS 3.7.2, and 
would incorporate other minor changes to revise exception footnote (i) 
in TS Table 3.3.2-1 of TS 3.3.2, ``ESFAS [Engineered Safety Features 
Actuation System] Instrumentation,'' to remove the MSIVs from the 
footnote such that the footnote only addresses the MSIVBVs and 
MSLPDIVs. The MSIVs would be addressed in new exception footnote (k) 
added to TS Table 3.3.2-1.
    The proposed amendment would add new TS 3.7.19, ``Secondary System 
Isolation Valves (SSIVs),'' which would provide limiting conditions for 
operation (LCOs) and surveillance requirements for the SSIVs, steam 
generator chemical injection isolation valves (SGCIIVs), steam 
generator blowdown isolation valves (SGBSIVs), and steam generator 
sample line isolation valves (SGBSSIVs). New Function 10, ``Steam 
Generator Blowdown System and Sample Line Isolation Valve Actuation,'' 
would be added to TS Table 3.3.2-1. The SGBSIVs and SGBSSIVs would be 
addressed in new exception footnote (t) added to Table 3.3.2-1 for 
Function 10.
    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 adds requirements to the TS to ensure that 
systems and components are maintained consistent with the safety 
analysis and licensing basis.
    Requirements are incorporated into the TS for secondary system 
isolation valves. These changes do not involve any design or 
physical changes to the facility, including the SSIVs themselves. 
The design and functional performance requirements, operational 
characteristics, and reliability of the SSIVs are unchanged. There 
is no impact on the design safety function of MSIVs, MSIVBVs, 
MSLPDIVs, MFIVs [main feedwater isolation valves], MFRVs [main 
feedwater regulating valves] or MFRVBVs [MFRV bypass valves] to 
close (either as an accident mitigator or as a potential transient 
initiator). Since no failure mode or initiating condition that could 
cause an accident (including any plant transient) evaluated per the 
FSAR [Final Safety Analysis Report]-described safety analyses is 
created or affected, the change cannot involve a significant 
increase in the probability of an accident previously evaluated.
    With regard to the consequences of an accident and the equipment 
required for mitigation of the accident, the proposed changes 
involve no design or physical changes to components in the main 
steam supply system or feedwater system. There is no impact on the 
design safety function of MSIVs, MSIVBVs, MSLPDIVs, MFIVs, MFRVs, or 
MFRVBVs or any other equipment required for accident mitigation. 
Adequate equipment availability would continue to be required by the 
TS. The consequences of applicable, analyzed accidents (such as a 
main steam line break [or] feedline break) are not impacted by the 
proposed changes.
    The changes to TS 3.3.2, TS Table 3.3.2-1, and exception 
footnotes associated with Table Function 4 and New Function 10 
maintain consistency with the Applicability of revised TS 3.7.2 and 
new TS 3.7.19. Maintaining TS 3.3.2 and TS Table 3.3.2-1 consistent 
with the Applicability of TS 3.7.2 and TS 3.7.19 is consistent with 
the Westinghouse Standard Technical Specifications.
    These changes involve no physical changes to the facility and do 
not adversely affect the availability of the safety functions 
assumed for the MSIVs, MSIVBVs, MSLPDIVs, and SSIVs. Therefore, they 
do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Based on the above considerations, the proposed changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes add requirements to the TS that support or 
ensure the availability of the safety functions assumed or required 
for the MSIVs, MSIVBVs, MSLPDIVs, and SSIVs. The changes do not 
involve a physical alteration of the plant (no new or different type 
of equipment will be installed) or changes in controlling 
parameters. Additional requirements are being imposed, but they are 
consistent with the assumptions made in the safety analysis and 
licensing basis. The addition of Conditions, Required Actions and 
Completion Times to TS for the MSIVBVs, MSLPDIVs, and SSIVs does not 
involve a change in the design, configuration, or operational 
characteristics of the plant. Further, the proposed changes do not 
involve any changes in plant procedures for ensuring that the plant 
is operated within analyzed limits. As such, no new failure modes or 
mechanisms that could cause a new or different kind of accident from 
any previously evaluated are introduced.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed addition of Conditions, Required Actions and 
Completion Times for SSIVs, MSIVBVs, and MSLPDIVs, as well as the 
proposed change to the LCO and Applicability for TS 3.7.2 and the 
proposed new TS 3.7.19 (and the corresponding changes to TS 3.3.2, 
``ESFAS Instrumentation'') does not alter the manner in which safety 
limits or limiting safety system settings are determined. No changes 
to instrument/system actuation setpoints are involved. The safety 
analysis acceptance criteria are not impacted and the proposed 
change will not permit plant operation in a configuration outside 
the design basis. The changes are consistent with the safety 
analysis and licensing basis for the facility.

    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: John O'Neill, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: June 1, 2009.
    Description of amendment request: The proposed amendment would 
revise the Limiting Condition for Operation (LCO) Applicability Note 
for Technical Specification (TS) 3.3.9, ``Boron Dilution Mitigation 
System (BDMS).''

[[Page 42934]]

The LCO Applicability Note would be revised to more explicitly define 
what the term ``during reactor startup'' means in MODES 2 and 3. This 
revision to the Applicability Note is proposed to clarify the 
situations during which the BDMS signal may be blocked.
    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.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since there are 
no design changes. All design, material, and construction standards 
that were applicable prior to this amendment request will be 
maintained. There will be no changes to any design or operating 
limits.
    The proposed change will not adversely affect accident 
initiators or precursors [or] adversely alter the design 
assumptions, conditions, and configuration of the facility or the 
manner in which the plant is operated and maintained. There are no 
design or operating changes to the reactor makeup water system 
(RMWS), the reactor makeup control system (RMCS), or the chemical 
and volume control system (CVCS). There will be no decrease in the 
boron concentration of the boric acid tanks. There will be no 
changes to the BDMS setpoint or the operation of the BDMS, other 
than the limited durations during which flux multiplication signal 
blocking would be allowed. Therefore, there will be no changes that 
would serve to increase the likelihood of occurrence of an 
inadvertent boron dilution event.
    The proposed change will not alter or prevent the ability of 
structures, systems, and components (SSCs) from performing their 
intended functions to mitigate the consequences of an initiating 
event within the applicable acceptance limits. Exceptions to 
Technical Specification requirements are allowed and, in fact, 
rather commonplace when plant operation would otherwise be 
restricted in a manner that is not commensurate with the desired 
safety objective, especially when those exceptions are of short 
duration and are accompanied by compensatory measures.
    The proposed change does not physically alter safety-related 
systems [or] affect the way in which safety-related systems perform 
their functions.
    The inadvertent boron dilution analysis acceptance criteria will 
continue to be met with the proposed change, with consideration 
given to the fact that the current licensing basis analyses do not 
assume concurrent rod withdrawal in the MODES 2 and 3 boron dilution 
analyses. The licensing basis analyses assume that positive 
reactivity insertion is being added by a single method, i.e., boron 
dilution. The MODE 2 licensing basis analysis of an inadvertent 
boron dilution event in FSAR [Final Safety Analysis Report] Section 
15.4.6 assumes that the shutdown banks are fully withdrawn and that 
the control banks are withdrawn to the 0% power rod insertion limits 
depicted in the COLR [Core Operating Limits Report]. The MODE 2 
analysis credits operator action to swap the charging suction source 
after an automatic reactor trip, and corresponding rod insertion, on 
high source range neutron flux. The MODE 3 licensing basis analysis 
credits automatic mitigation by the BDMS with steady state initial 
conditions and static initial rod positions (all shutdown and 
control banks are fully inserted other than the single most reactive 
rod which is assumed to be fully withdrawn) at bounding RCS [reactor 
coolant system] T-avg values at either end of MODE 3. Neither the 
analysis nor the BDMS design basis assumes that the system protects 
against a rod withdrawal event.
    The proposed change will not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of an accident previously evaluated. 
The applicable radiological dose criteria will continue to be met.
    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?
    Response: No.
    There are [neither] proposed design changes nor are there any 
changes in the method by which any safety-related plant structure, 
system, or component (SSC) performs its specified safety function. 
The proposed change will not affect the normal method of plant 
operation or change any operating parameters. Equipment performance 
necessary to fulfill safety analysis missions will be unaffected. 
The proposed change will not alter any assumptions required to meet 
the safety analysis acceptance criteria.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures will be introduced as a 
result of this amendment. There will be no adverse effect or 
challenges imposed on any safety-related system as a result of this 
amendment.
    The proposed amendment will not alter the design or performance 
of the 7300 Process Protection System, Nuclear Instrumentation 
System, or Solid State Protection System used in the plant 
protection systems.
    The proposed change does not, therefore, create the possibility 
of a new or different accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    There will be no effect on those plant systems necessary to 
assure the accomplishment of protection functions. There will be no 
impact on the overpower limit, departure from nucleate boiling ratio 
(DNBR) limits, heat flux hot channel factor (FQ), nuclear 
enthalpy rise hot channel factor (F[Delta]H), loss of coolant 
accident peak cladding temperature (LOCA PCT), peak local power 
density, or any other margin of safety. Mode-specific required 
shutdown margins in the COLR will not be changed. The applicable 
radiological dose consequence acceptance criteria will continue to 
be met.
    The proposed change does not eliminate any surveillances or 
alter the frequency of surveillances required by the Technical 
Specifications. None of the acceptance criteria for any accident 
analysis will be changed.
    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: John O'Neill, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: July 10, 2009.
    Description of amendment request: The proposed amendment would 
delete the Technical Specification (TS) requirements for the 
containment hydrogen recombiners and hydrogen monitors. The proposed TS 
changes support implementation of the revision to Title 10 of the Code 
of Federal Regulations (10 CFR), Section 50.44, ``Standards for 
Combustible Gas Control System in Light-Water-Cooled Power Reactors,'' 
that became effective on October 16, 2003. The proposed changes are 
consistent with Revision 1 of the NRC-approved Industry/Technical 
Specification Task Force (TSTF) Standard Technical Specification Change 
Traveler, TSTF-447, ``Elimination of Hydrogen Recombiners and Change to 
Hydrogen and Oxygen Monitors.''
    The NRC staff issued a notice of opportunity for public comments on 
TSTF-447, Revision 1, published in the Federal Register on August 2, 
2002 (67 FR 50374), soliciting comments on a model safety evaluation 
(SE) and a model no significant hazards consideration (NSHC) 
determination for the elimination of requirements for hydrogen 
recombiners, and hydrogen and oxygen monitors from TS. Based on its 
evaluation of the public comments

[[Page 42935]]

received, the NRC staff made appropriate changes to the models and 
included final versions in a notice of availability published in the 
Federal Register on September 25, 2003 (68 FR 55416), regarding the 
adoption of TSTF-447, Revision 1, as part of the NRC's consolidated 
line item improvement process (CLIIP).
    In addition to the changes related to requirements for the hydrogen 
recombiners and monitors, this amendment application includes four 
unrelated, minor changes to correct typographical errors identified in 
Callaway's TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC adopted by the licensee 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 revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen monitors are no longer required to mitigate design-basis 
accidents and, therefore, the hydrogen monitors do not meet the 
definition of a safety-related component as defined in 10 CFR 50.2. 
RG [Regulatory Guide] 1.97 Category 1 is intended for key variables 
that most directly indicate the accomplishment of a safety function 
for design-basis accident events. The hydrogen monitors no longer 
meet the definition of Category 1 in RG 1.97. As part of the 
rulemaking to revise 10 CFR 50.44 the Commission found that Category 
3, as defined in RG 1.97, is an appropriate categorization for the 
hydrogen monitors because the monitors are required to diagnose the 
course of beyond design-basis accidents.
    The regulatory requirements for the hydrogen monitors can be 
relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3 and removal of 
the hydrogen monitors from TS will not prevent an accident 
management strategy through the use of the SAMGs [severe accident 
management guidelines], the emergency plan (EP), the emergency 
operating procedures (EOP), and site survey monitoring that support 
modification of emergency plan protective action recommendations 
(PARs).
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen monitor requirements, 
including removal of these requirements from TS, does not involve a 
significant increase in the probability or the consequences of any 
accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of 
a New or Different Kind of Accident From Any Previously Evaluated

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, will not result in any failure mode 
not previously analyzed. The hydrogen recombiner and hydrogen 
monitor equipment was intended to mitigate a design-basis hydrogen 
release. The hydrogen recombiner and hydrogen monitor equipment are 
not considered accident precursors, nor does their 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.
    Therefore, this 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 the Margin of Safety

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, in light of existing plant equipment, 
instrumentation, procedures, and programs that provide effective 
mitigation of and recovery from reactor accidents, results in a 
neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the [Three Mile Island], 
Unit 2 accident, can be adequately met without reliance on safety-
related hydrogen monitors.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. Removal of hydrogen monitoring from TS will 
not result in a significant reduction in their functionality, 
reliability, and availability.

    The NRC staff has reviewed the analysis adopted by the licensee 
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: John O'Neill, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: Michael T. Markley.

Virginia Electric and Power Company, Docket No. 50-338 North Anna Power 
Station, Unit No. 1, Louisa County, Virginia

    Date of amendment request: July 23, 2009
    Description of amendment request: The proposed change, a one-time 
extension to the Completion Time (CT) of Technical Specification 3.8.9 
Condition A, will provide an opportunity to fully investigate the 
extent of the damaged breaker and its condition to ensure continued bus 
reliability for the remainder of the operating cycle.
    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 license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    The proposed change does not alter any plant equipment or 
operating practices in such a manner that the probability of an 
accident is significantly increased. The proposed change will not 
alter assumptions relative to the mitigation of an accident or 
transient event. Manual operator actions in the event of an SGTR 
have been identified during the one-time extended CT for the 1J1 
[Motor Control Center] MCC outage. A risk-informed evaluation of 
these operator actions has been performed and the increase in annual 
Core Damage and Large Early Release Frequencies associated with the 
proposed change in the Technical Specification CT are characterized 
as ``small changes'' by Regulatory Guide (RG) 1.174. The Incremental 
Conditional Core Damage and Large Early Release Probabilities [ICCDP 
and ICLERP] associated with the proposed

[[Page 42936]]

Technical Specification CT meet the acceptance criteria in 
Regulatory Guide 1.177.
    The ICCDP and ICLERP are 1.01 E-7 per year and 9.86E-9 per year, 
respectively. These results are below the RG 1.177 limits of 5E-7 
for ICCDP and 5E-8 for ICLERP.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    The systems' design and operation are not affected by the 
proposed change. The safety analysis acceptance criteria stated in 
the Updated Final Safety Analysis Report is not impacted by the 
change. Redundancy and diversity of the electrical distribution 
system will be maintained with the exception of the MCCs 1J 1-2N and 
2S. The proposed change will not allow plant operation in a 
configuration outside the design basis.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The Nuclear Regulatory Commission (NRC) staff has reviewed the 
licensee's analysis and, based on this review, it appears that the 
three standards of 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Counsel, 
Dominion Resources Services, Inc., Millstone Power Station, Building 
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385
    NRC Branch Chief: Undine Shoop.

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.

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

    Date of amendment request: March 20, 2008, as supplemented by 
letters dated May 28, 2008, October 6, 2008, December 17, 2008, and 
February 12, 2009.
    Brief description of amendment request: The proposed amendments 
would revise the McGuire licensing basis by adopting the Alternative 
Source Term (AST) radiological analysis methodology as allowed by 10 
CFR 50.67, Accident Source Term, for the Loss of Coolant Accident. This 
amendment request represents full scope implementation of the AST as 
described in Nuclear Regulatory Commission (NRC) Regulatory Guide 
1.183, ``Alternative Radiological Source Terms for Evaluating Design 
Basis Accidents at Nuclear Power Reactors, Revision 0.''
    Date of publication of individual notice in Federal Register: 
February 27, 2009 (74 FR 9009).
    Expiration date of individual notice: April 28, 2009.

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 amendment request: June 23, 2008.
    Brief description of amendment request: The amendments revise the 
Technical Specifications (TSs) for Catawba Nuclear Station, Units 1 and 
2. This request modifies the subject TS and Bases by changing the logic 
configuration of TS Table 3.3.2-1, ``Engineered Safety Feature 
Actuation System Instrumentation'', Function 5.b. (5), ``Turbine Trip 
and Feedwater Isolation, Feedwater Isolation, Doghouse Water Level--
High High.'' The existing one-out-of-one (\1/1\) logic per train per 
doghouse is being modified to a two-out-of-three (\2/3\) logic per 
train per doghouse. The proposed change will improve the overall 
reliability of this function and will reduce the potential for spurious 
actuations.
    Date of publication of individual notice in Federal Register: 
February 24, 2009 (74 FR 8276).
    Expiration date of individual notice: April 27, 2009.

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

    Date of amendment request: November 13, 2008.
    Brief description of amendment request: The amendment proposes a 
one-cycle revision to the Technical Specifications to incorporate an 
interim alternate repair criterion for steam generator tube repair 
criteria during the End of Cycle 16 refueling outage and subsequent 
cycle 17 operation.
    Date of publication of individual notice in Federal Register: 
February 24, 2009 (74 FR 8278).
    Expiration date of individual notice: April 27, 2009.

Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, 
Comanche Peak Steam Electric Station, Units 1 and 2, Somervell County, 
Texas

    Date of amendment request: June 8, 2009.
    Brief description of amendment request: The proposed amendment 
would revise Technical Specification (TS) 5.5.9.2, ``Unit 1 Model D76 
and Unit 2 Model D5 Steam Generator (SG) Program,'' to exclude portions 
of the CPSES, Unit 2 Model D5 SG below the top of the SG tubesheet from 
periodic SG tube inspections. In addition, the proposed amendment would 
revise TS 5.6.9, ``Unit 1 Model D76 and Unit 2 Model D5 Steam Generator 
Tube Inspection Report,'' to include reporting requirements specific to 
the permanent alternate repair criteria for CPSES, Unit 2. The 
amendment request is supported by Westinghouse WCAP-17072-P, ``H*: 
Alternate Repair Criteria for the Tube Sheet Expansion Region in Steam 
Generators with Hydraulically Expanded Tubes (Model D5),'' May 2009.
    Date of publication of individual notice in Federal Register: July 
23, 2009 (74 FR 36533).
    Expiration date of individual notice: September 21, 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

[[Page 42937]]

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 Systems (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, http://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 [email protected].

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: April 23, 2009.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TSs) by removing working hour restrictions 
from TS 5.2.2 to support compliance with recent revisions to Title 10 
of the Code of Federal Regulations, Part 26, Subpart I. The amendments 
are consistent with the guidance contained in Nuclear Regulatory 
Commission (NRC) approved Technical Specifications Task Force Traveler 
511 (TSTF-511). This TS improvement was made available by the NRC on 
December 30, 2008 (73 FR 79923) as part of the consolidated line item 
improvement process.
    Date of issuance: August 6, 2009.
    Effective date: As of the date of issuance to be implemented with 
the implementation of the new 10 CFR Part 26, Subpart I requirements.
    Amendment Nos.: 292 and 268.
    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 2, 2009 (74 FR 
26430).
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated August 6, 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: July 14, 2008.
    Brief description of amendments: The changes revised Technical 
Specifications (TSs) Section 3.7.10, ``Control Room Area Ventilation,'' 
its associated Bases, and TS Section 5.5 ``Programs and Manuals.'' This 
LAR institutes the Control Room Habitability Program.
    The changes are consistent with NRC-approved Industry Technical 
Specification Task Force (TSTF) Standard Technical Specification Change 
Traveler, TSTF-448, Revision 3, ``Control Room Habitability Program.'' 
The availability of this TS improvement was announced in the Federal 
Register on January 17, 2007, as part of the Consolidated Line-Item 
Improvement Process (CLIIP). The amendments also authorized a change to 
the Catawba Updated Final Safety Analysis Report (UFSAR).
    Date of issuance: July 30, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 250 and 245.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the licenses and the technical specifications.XXX
    Date of initial notice in Federal Register: June 2, 2009 (74 FR 
26431).
    The Commission's related evaluation, State consultation, and final 
no significant hazards consideration determination of the amendments is 
contained in a Safety Evaluation dated July 30, 2009.
    No significant hazards consideration comments received: No.

Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc., 
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, 
Louisiana

    Date of amendment request: January 21, 2009, as supplemented by 
letters dated January 23 and June 22, 2009.
    Brief description of amendment: The amendment modified the 
Technical Specifications (TSs) to adopt U.S. Nuclear Regulatory 
Commission (NRC)-approved TS Task Force (TSTF) change travelers TSTF-
163, TSTF-222, TSTF-230, and TSTF-306, and made two minor 
administrative corrections.
    Date of issuance: August 11, 2009.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 165.
    Facility Operating License No. NPF-47: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: March 24, 2009 (74 FR 
12392). The supplemental letters dated January 23 and June 22, 2009, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 11, 2009.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1 (ANO1), Pope County, Arkansas

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2 (ANO2), Pope County, Arkansas

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

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1 (GGNS), Claiborne 
County, Mississippi

Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian 
Point Nuclear Generating Unit Nos. 2 and 3 (IP2 and IP3), Westchester 
County, New York
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Plant 
(PAL), Van Buren County, Michigan


[[Page 42938]]


Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station (PIL), Plymouth County, Massachusetts

Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc., 
Docket No. 50-458, River Bend Station, Unit 1 (RBS), West Feliciana 
Parish, Louisiana

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3 (W3), St. Charles Parish, Louisiana

    Date of application for amendment: April 27, 2009, as supplemented 
July 10, 2009.
    Brief description of amendment: The amendments deleted those 
portions of the Technical Specifications (TSs) superseded by Title 10 
of the Code of Federal Regulations (10 CFR) Part 26, Subpart I, 
consistent with U.S. Nuclear Regulatory Commission (NRC)-approved TS 
Task Force (TSTF) change traveler TSTF-511, Revision 0, ``Eliminate 
Working Hour Restrictions from TS 5.2.2 to Support Compliance with 10 
CFR Part 26.''
    Date of issuance: August 4, 2009.
    Effective date: As of the date of issuance and shall be implemented 
by October 1, 2009.
    Amendment Nos.: ANO1--237; ANO2--285; JAF--295; GGNS--183; IP2--
261; IP3--240; PAL--238; PIL--233; RBS--164; and W3--221.
    Facility Operating License Nos. DPR-51 (ANO1), NPF-6 (ANO2), DPR-59 
(JAF), NPF-29 (GGNS), DPR-26 (IP2), DPR-64 (IP3), DPR-20 (PAL), DPR-35 
(PIL), NPF-47 (RBS), and NPF-38 (W3): The amendments revised the 
Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: June 2, 2009 (74 FR 
26432). The supplement dated July 10, 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 August 4, 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: September 17, 2008, as supplemented by 
letters dated January 8, March 18, and June 30, 2009.
    Brief description of amendment: The amendment revised the Operating 
License and modified Technical Specification (TS) \3/4\.3.1 and Note 2 
of TS Table 4.3-1. The changes result in the addition of conservatism 
to Core Protection Calculator power indications when calibrations are 
required in certain conditions.
    Date of issuance: August 10, 2009.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 222.
    Facility Operating License No. NPF-38: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: November 4, 2008 (73 FR 
65695). The supplemental letters dated January 8, March 18, and June 
30, 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 August 10, 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

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

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

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: June 9, 2008, as supplemented 
by letter dated March 30, 2009.
    Brief description of amendments: The amendments revise the 
Technical Specification (TS) surveillance requirement (SR) frequency in 
TS 3.1.3, ``Control Rod OPERABILITY.'' The amendments also clarify the 
requirement to fully insert all insertable control rods for the 
limiting condition for operation in TS 3.3.1.2, Required Action E.2, 
``Source Range Monitoring Instrumentation'' (Clinton Power Station 
only). Finally, the amendments revise Example 1.4-3 in Section 1.4, 
``Frequency,'' to clarify the applicability of the 1.25 surveillance 
test interval extension.
    Date of issuance: August 11, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 188, 232/225, 193/180, 272/276, 244/239.
    Facility Operating License Nos. NPF-62, DPR-19, DPR-25, NPF-11, 
NPF-18, DPR-44, DPR-56, DPR-29, DPR-30: The amendments revised the 
Technical Specifications/Licenses.
    Date of initial notice in Federal Register: August 12, 2009 (73 FR 
46928) The March 30, 2009, supplement contained clarifying information 
and did not change the NRC staff's initial proposed finding of no 
significant hazards consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 11, 2009.
    No significant hazards consideration comments received: No.

Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, 
Comanche Peak Steam Electric Station, Unit Nos. 1 and 2 (CPSES), 
Somervell County, Texas

    Date of amendment request: April 1, 2009, as supplemented by letter 
dated July 9, 2009.
    Brief description of amendments: The amendments deleted Technical 
Specification (TS) 5.2.2.d, in TS 5.2.2, ``Unit Staff,'' regarding the 
requirement to develop and implement administrative procedures to limit 
the working hours of personnel who perform safety-related functions. In 
addition, paragraphs e and f of TS 5.2.2 were renumbered to d and e and 
in TS 5.2.2.b the reference to 5.2.2.f was revised to 5.2.2.e to 
reflect the removal of paragraph d of TS 5.2.2. The change is 
consistent with U.S. Nuclear Regulatory Commission (NRC)-approved 
Revision 0 to TS Task Force (TSTF) Improved Technical Specification 
change traveler, TSTF-511, ``Eliminate Working Hour Restrictions from 
TS 5.2.2 to Support Compliance with 10 CFR Part 26.'' The availability 
of this TS improvement was announced in the Federal Register on 
December 30, 2008 (73 FR 79923), as part of the consolidated line item 
improvement process.
    Date of issuance: August 7, 2009.

[[Page 42939]]

    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: Unit 1-148; Unit 2-148.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: May 19, 2009 (74 FR 
23445). The supplemental letter dated July 9, 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 August 7, 2009.
    No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket Nos. 50-220 and 50-410, 
Nine Mile Point Nuclear Station, Unit Nos. 1 and 2 (NMP 1 and 2), 
Oswego County, New York

    Date of application for amendment: February 11, 2009.
    Brief description of amendments: The amendments delete those 
portions of the Technical Specifications (TSs) superseded by Title 10 
of the Code of Federal Regulations (10 CFR), Part 26, Subpart I. This 
change is consistent with Nuclear Regulatory Commission (NRC) approved 
Technical Specification Task Force (TSTF) Improved Standard Technical 
Specifications Change Traveler TSTF-511, Revision 0, ``Eliminate 
Working Hour Restrictions from TS 5.2.2 to Support Compliance with 10 
CFR Part 26.'' These changes were described in a Notice of Availability 
for Consolidated Line Item Improvement Process TSTF-511 published in 
the Federal Register on December 30, 2008 (73 FR 79923).
    Date of issuance: July 27, 2009.
    Effective date: As of the date of issuance to be implemented by 
October 1, 2009.
    Amendment Nos.: 203 and 131.
    Renewed Facility Operating License Nos. DPR-063 and NPF-069: The 
amendments revise the License and TSs.
    Date of initial notice in Federal Register: April 21, 2009 (73 FR 
18255).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 27, 2009.
    No significant hazards consideration comments received: No.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: September 2, 2008.
    Brief description of amendments: The amendment approved the 
licensee's request to incorporate a revision in the Updated Final 
Safety Analysis Report (UFSAR) Section 13.7.2.3, ``PRA Risk 
Categorization,'' to add a separate set of criteria for assessing the 
risk significance of the risk achievement worth values of common cause 
failures as part of the probabilistic risk assessment analysis of the 
risk importance of components.
    Date of issuance: August 12, 2009
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: Unit 1-191; Unit 2-179.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Facility Operating Licenses, and Updated Final Safety 
Analysis Report.
    Date of initial notice in Federal Register: December 2, 2008 (73 FR 
73354).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 12, 2009.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units 1 and 2, Louisa County, Virginia

    Date of application for amendments: February 6, 2009.
    Brief description of amendments: The proposed amendments deleted 
applicable portions of the Technical Specifications (TSs) superseded by 
Part 26, Subpart I of Title 10 of the Code of Federal Regulations (10 
CFR). This change is 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: July 29, 2009.
    Effective date: As of the date of issuance and shall be implemented 
by October 1, 2009.
    Amendment Nos.: 256 and 237.
    Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments 
change the license and the technical specifications.
    Date of initial notice in Federal Register: March 24, 2009 (74 FR 
12396).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 29, 2009.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 14, 2008.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.3.2, ``Engineered Safety Feature Actuation System 
(ESFAS) Instrumentation,'' to extend the Surveillance Frequency on 
selected ESFAS slave relays from 92 days to 18 months.
    Date of issuance: July 30, 2009.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days of the date of issuance.
    Amendment No.: 183.
    Renewed Facility Operating License No. NPF-42. The amendment 
revised the Operating License and Technical Specifications.
    Date of initial notice in Federal Register: October 7, 2008 (73 FR 
58379).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 30, 2009.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 14, 2008, as supplemented by 
letter dated April 10, 2009.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.3.2, ``Engineered Safety Feature Actuation System 
(ESFAS) Instrumentation,'' TS 3.7.2, ``Main Steam Isolation Valves 
(MSIVs),'' and added new TS 3.7.19, ``Secondary System Isolation Valves 
(SSIVs).'' TS 3.7.2 has been revised to add MSIV bypass valves to the 
scope of TS 3.7.2. TS Table 3.3.2-1 has been revised to reflect the 
addition of the MSIV bypass valves to TS 3.7.2 and the associated 
applicability to be consistent with Westinghouse Standard Technical 
Specifications (NUREG-1431, Revision 3.0). TS 3.7.19 has been added to 
include a limiting condition for operation, conditions/required 
actions, and surveillance requirements for the steam generator blowdown 
isolation valves and steam generator blowdown sample isolation valves.

[[Page 42940]]

    Date of issuance: July 31, 2009.
    Effective date: As of the date of issuance and shall be implemented 
prior to startup from Refueling Outage 17.
    Amendment No.: 184.
    Renewed Facility Operating License No.: NPF-42. The amendment 
revised the Operating License and Technical Specifications.
    Date of initial notice in Federal Register: October 7, 2008 (73 FR 
58679). The supplemental letter dated April 10, 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 July 31, 2009.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: March 6, 2009, as supplemented by letter 
dated July 14, 2009.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 5.2.2, ``Unit Staff,'' to eliminate working hour 
restrictions (TS 5.2.2.d) to support compliance with Title 10 of the 
Code of Federal Regulations (10 CFR) Part 26. In addition, paragraphs e 
and f of TS 5.2.2 were renumbered to d and e to reflect the removal of 
paragraph d of TS 5.2.2, and a reference in 5.2.2b was updated to 
reflect the renumbering of 5.2.2f. to 5.2.2e. The request is consistent 
with the guidance contained in U.S. Nuclear Regulatory Commission 
(NRC)-approved TS Task Force (TSTF) change traveler TSTF-511, Revision 
0, ``Eliminate Working Hour Restrictions from TS 5.2.2 to Support 
Compliance with 10 CFR Part 26.''
    Date of issuance: August 7, 2009.
    Effective date: As of its date of issuance and shall be implemented 
by October 1, 2009.
    Amendment No.: 185.
    Renewed Facility Operating License No.: NPF-42. The amendment 
revised the Operating License and Technical Specifications.
    Date of initial notice in Federal Register: April 21, 2009 (74 FR 
18258). The supplemental letter dated July 14, 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 August 7, 2009.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 14th day of August 2009.

    For the Nuclear Regulatory Commission.
Allen G. Howe,
Acting Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. E9-20403 Filed 8-24-09; 8:45 am]
BILLING CODE 7590-01-P