[Federal Register Volume 75, Number 229 (Tuesday, November 30, 2010)]
[Notices]
[Pages 74091-74099]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2010-29941]


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NUCLEAR REGULATORY COMMISSION

[NRC-2010-0367]


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 (NRC or the 
Commission) 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 November 3, 2010, to November 17, 2010. The 
last biweekly notice was published on November 16, 2010 (75 FR 70032).

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), 50.92, this means that operation of the facility 
in accordance with the proposed amendment would not (1) Involve a 
significant increase in the probability or consequences of an accident 
previously evaluated; or (2) create the possibility of a new or 
different kind of accident from any accident previously evaluated; or 
(3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules, 
Announcements and Directives Branch (RADB), TWB-05-B01M, Division of 
Administrative Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be faxed to the RADB at 301-492-3446. 
Documents may be examined, and/or copied for a fee, at the NRC's Public 
Document Room (PDR), located at One White Flint North, Room O1-F21, 
11555 Rockville Pike (first floor), Rockville, Maryland 20852.
    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, Room O1-F21, 
11555 Rockville Pike (first floor), Rockville, Maryland 20854. Publicly 
available records will be accessible from the Agencywide

[[Page 74092]]

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 requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held. If 
the final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards consideration, any hearing held 
would take place before the issuance of any amendment.
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC E-Filing rule (72 FR 49139, 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
ten (10) days prior to the filing deadline, the participant should 
contact the Office of the Secretary by e-mail at 
[email protected], or by telephone at 301-415-1677, to request (1) 
a digital ID certificate, which allows the participant (or its counsel 
or representative) to digitally sign documents and access the E-
Submittal server for any proceeding in which it is participating; and 
(2) advise the Secretary that the participant will be submitting a 
request or petition for hearing (even in instances in which the 
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the 
Secretary will establish an electronic docket for the hearing in this 
proceeding if the Secretary has not already established an electronic 
docket.
    Information about applying for a digital ID certificate is 
available on NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing 
the E-Submittal server are detailed in NRC's ``Guidance for Electronic 
Submission,'' which is available on the agency's public Web site at 
http://www.nrc.gov/site-help/e-submittals.html. Participants may 
attempt to use other software not listed on the Web site, but should 
note that the NRC's E-Filing system does not support unlisted software, 
and the NRC Meta System Help Desk will not be able to offer assistance 
in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with NRC guidance 
available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an e-mail notice confirming receipt of the document. The 
E-Filing system also distributes an e-mail notice that provides access 
to the document to the NRC Office of the General Counsel and any others 
who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must

[[Page 74093]]

apply for and receive a digital ID certificate before a hearing 
request/petition to intervene is filed so that they can obtain access 
to the document via the E-Filing system.
    A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC Web site 
at http://www.nrc.gov/site-help/e-submittals.html, by e-mail at 
[email protected], or by a toll-free call at 1-866-672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 
20852, Attention: Rulemaking and Adjudications Staff. Participants 
filing a document in this manner are responsible for serving the 
document on all other participants. Filing is considered complete by 
first-class mail as of the time of deposit in the mail, or by courier, 
express mail, or expedited delivery service upon depositing the 
document with the provider of the service. A presiding officer, having 
granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant 
to an order of the Commission, or the presiding officer. Participants 
are requested not to include personal privacy information, such as 
social security numbers, home addresses, or home phone numbers in their 
filings, unless an NRC regulation or other law requires submission of 
such information. With respect to copyrighted works, except for limited 
excerpts that serve the purpose of the adjudicatory filings and would 
constitute a Fair Use application, participants are requested not to 
include copyrighted materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Non-timely filings 
will not be entertained absent a determination by the presiding officer 
that the petition or request should be granted or the contentions 
should be admitted, based on a balancing of the factors specified in 10 
CFR 2.309(c)(1)(i)-(viii).
    For further details with respect to this license amendment 
application, see the application for amendment which is available for 
public inspection at the Commission's PDR, located at One White Flint 
North, Public File Area O1F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the ADAMS Public Electronic Reading Room on the Internet at the NRC Web 
site, 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].

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: September 24, 2010.
    Description of amendment request: The proposed amendment would 
revise the Fermi 2 Radiological Emergency Response Preparedness (RERP) 
Plan to increase the staff augmentation times for Technical Support 
Center-related functions from 30 to 60 minutes and for Emergency 
Operations Facility-related functions from 60 to 90 minutes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed extension of staff augmentation times has no effect 
on normal plant operation or on any accident initiator. The change 
affects the response to radiological emergencies under the Fermi 2 
Radiological Emergency Response Preparedness (RERP) Plan. The 
ability of the emergency response organization to respond adequately 
to radiological emergencies has been evaluated. Improvements have 
been made to equipment, procedures, and training since initial 
approval of the Fermi 2 Emergency Plan that have resulted in a 
significant increase in the on-shift capabilities and knowledge such 
that there would be no degradation or loss of Emergency Plan 
function as a result of the proposed change. A functional analysis 
was also performed on the effect of the proposed change on the 
timeliness of performing major tasks for the major functional areas 
of the RERP Plan. The analysis concluded that extension of staff 
augmentation times would not significantly affect the ability to 
perform the required tasks.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change affects the required response times for 
supplementing onsite personnel in response to a Radiological 
emergency. It has been evaluated and determined not to significantly 
affect the ability to perform that function. It has no effect on the 
plant design or on the normal operation of the plant and does not 
affect how the plant is physically operated under emergency 
conditions. The extension of staff augmentation times in the RERP 
Plan does not affect the plant Operating, Abnormal Operating, or 
Emergency Operating procedures which are performed by plant staff 
during all plant conditions.
    Therefore, since the proposed change does not affect the design 
or method of operation of the plant, it does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The proposed change does not affect plant design or method of 
operation. 10 CFR 50.47(b) and 10 CFR Part 50, Appendix E establish 
emergency planning standards that require adequate staffing, 
satisfactory performance of key functional areas and critical tasks; 
and timely augmentation of the response capability. Since the 
initial NRC approval of the Emergency Plan, there have been 
improvements in the technology used to support the RERP functions 
and in the capabilities of onsite personnel. A functional analysis 
was performed on the effect of the proposed change on the timeliness 
of performing major tasks for the functional areas of the RERP Plan. 
The analysis concluded that an increase in staff augmentation times 
would not significantly affect the ability to perform the required 
RERP tasks. Thus, the proposed change has been determined not to 
adversely affect the ability to meet the emergency planning 
standards as described in 10 CFR 50.47(b) and 10 CFR Part 50, 
Appendix E.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this

[[Page 74094]]

review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David G. Pettinari, Attorney--Corporate 
Matters, 688 WCB, Detroit Edison Company, One Energy Plaza, Detroit, 
Michigan 48226-1279.
    NRC Branch Chief: Robert J. Pascarelli.

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

    Date of amendment request: January 6, 2010, as supplemented by 
letters dated August 20, 2010, and October 14, 2010.
    Description of amendment request: The proposed amendment would 
enable PBAPS, Units 2 and 3, to possess byproduct and special nuclear 
material from Limerick Generating Station (LGS), Units 1 and 2. 
Specifically, the revised license paragraph would permit storage of 
low-level radioactive waste (LLRW) from LGS in the PBAPS LLRW Storage 
Facility. The PBAPS LLRW Storage Facility currently provides storage 
for LLRW generated at PBAPS.
    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 (NSHC), 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 is an amendment to the PBAPS FOLs [Facility 
Operating Licenses] that will enable PBAPS to receive and store 
Class B/C LLRW from LGS in the PBAPS LLRWSF [Low Level Radioactive 
Waste Storage Facility]. This proposed change does not impact any 
initiators or precursors of previously analyzed accidents. The 
storage of Class B/C LLRW from LGS does not impact the failure of 
any plant structures, systems, or components. The proposed change 
does not have a detrimental impact on the integrity of any plant 
structure, system, or component that initiates an analyzed event. 
The proposed change does not affect any active or passive failure 
mechanisms that could lead to an accident. The PBAPS LLRWSF is not 
safety related, and is not used for plant shutdown resulting from 
accident or nonstandard operational conditions.
    The proposed change does not significantly increase the 
consequences of postulated design basis events (i.e., seismic, 
flood, tornado, fire, and container drop events), in that the 
postulated impact of these events remains well below regulatory 
requirements (i.e., less than 10 percent of 10 CFR Part 100, 
``Reactor Site Criteria'' acceptance criteria).
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No.
    The proposed change is an amendment to the PBAPS FOLs that will 
enable PBAPS to receive and store Class B/C LLRW from LGS in the 
PBAPS LLRWSF. The proposed amendment does not involve any change to 
the plant equipment or system design functions. EGC has verified 
that the storage of Class B/C LLRW from LGS in the PBAPS LLRWSF does 
not affect the ability of the PBAPS LLRWSF to perform its design 
function, including compliance with NRC regulatory requirements and 
guidance. No new accident initiators are introduced by this 
amendment.
    Therefore, the proposed amendment 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 
the margin of safety?
    Response: No.
    The proposed change is an amendment to the PBAPS FOLs that will 
enable PBAPS to receive and store Class B/C LLRW from LGS in the 
PBAPS LLRWSF. The proposed amendment does not involve any change to 
plant equipment or system design functions. The margin of safety is 
established through the design of the plant structures, systems, and 
components, the parameters within which the plant is operated, and 
the setpoints for the actuation of equipment relied upon to respond 
to an event. The proposed amendment does not affect the PBAPS safety 
limits or setpoints at which protective actions are initiated.
    The proposed amendment does not significantly increase the dose 
rate at the exterior wall of the LLRWSF, the nearest restricted area 
boundary, and the nearest residence when the LLRWSF is filled to 
capacity with Class B/C LLRW. Therefore, these dose rates will 
remain within limits specified in 10 CFR Part 20 and 40 CFR Part 
190.
    Additionally, the potential radiological impact of a postulated 
design basis container drop accident is less than 10 percent of the 
10 CFR Part 100 acceptance criteria.
    Therefore the margin of safety is not reduced by the proposed 
change.

    The 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 NSHC.
    Attorney for licensee: Mr. J. Bradley Fewell, Associate General 
Counsel, Exelon Generation Company LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Harold K. Chernoff.

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

    Date of amendment request: June 25, 2010, as supplemented by letter 
dated August 16, 2010.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Surveillance Requirement (SR) 
3.6.1.3, ``Primary Containment Isolation Valves (PCIVs),'' and SR 
3.6.1.5, ``Reactor Building-to-Suppression Chamber Vacuum Breakers,'' 
to modify the required level for the liquid nitrogen storage tank.
    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 (NSHC), which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed TS changes to increase the level in the liquid 
nitrogen storage tank from >= 16 inches water column to a level of 
>= 22 inches water column, or equivalent volume of >= 124,000 scf 
[standard cubic feet] at 250 psig, is necessary in order to correct 
a non-conservative TS value. Increasing the level is intended to 
ensure continued operability of the PCIVs (SR 3.6.1.3.1) and Reactor 
Building-to-Suppression Chamber Vacuum Breakers (SR 3.6.1.5.1) via 
the SGIG [safety grade instrument gas] system. The non-conservative 
TS condition was identified based on a re-analysis of the liquid 
nitrogen storage tank operation. The leakage allowance that was 
previously assumed was not based on a rigorous empirical value. The 
re-analysis of the leakage allowance assumes more reasonable system 
leakage based on operational data. Exelon determined that the 
current PBAPS, Units 2 and 3, TS SR value for the minimum level in 
the liquid nitrogen storage tank of >= 16 inches water column is 
non-conservative and that the guidance of Nuclear Regulatory 
Commission (NRC) Administrative Letter 98-10, ``Dispositioning of 
Technical Specifications that are Insufficient to Assure Plant 
Safety,'' applies. Exelon has implemented administrative controls to 
maintain the amount of nitrogen in the liquid nitrogen storage tank 
at a level of > 22 inches water column in support of SGIG system 
operation.
    Exelon is submitting this License Amendment Request to address 
this non-conservative condition. The proposed TS

[[Page 74095]]

changes do not introduce new equipment or new equipment operating 
modes, nor do the proposed changes alter existing system 
relationships. The proposed changes do not affect plant operation, 
design function or any analysis that verifies the capability of a 
system, structure or component (SSC) to perform a design function. 
Further, the proposed changes do not increase the likelihood of the 
malfunction of any SSC or impact any analyzed accident. 
Consequently, the probability or consequences of an accident 
previously evaluated are not affected.
    Therefore, the proposed amendments 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 TS change to increase the level in the liquid 
nitrogen storage tank from >= 16 inches water column to a level of 
>= 22 inches water column, or equivalent volume of >= 124,000 scf at 
250 psig, for the PCIVs (SR 3.6.1.3.1) and Reactor Building-to-
Suppression Chamber Vacuum Breakers (SR 3.6.1.5.1) is needed to 
correct a non-conservative value based on a revised analysis. The 
proposed TS changes do not alter the design function or operation of 
any SSC. There is no new system component being installed, no 
construction of a new facility, and no performance of a new test or 
maintenance function. The proposed TS changes do not create the 
possibility of a new credible failure mechanism or malfunction. The 
proposed changes do not modify the design function or operation of 
any SSC. Further, the proposed changes do not introduce new accident 
initiators. Consequently, the proposed changes cannot create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Therefore, the proposed amendments do not create the possibility 
of a new or different kind of accident from any accident previously 
analyzed.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed TS changes to increase the level in the liquid 
nitrogen storage tank from >= 16 inches water column to a level of 
>= 22 inches water column, or equivalent volume of >= 124,000 scf at 
250 psig, for the PCIVs (SR 3.6.1.3.1) and Reactor Building-to-
Suppression Chamber Vacuum Breakers (SR 3.6.1.5.1) are necessary to 
correct an existing non-conservative TS value. The proposed TS 
changes are needed based on a revised analysis that utilizes 
empirical data for nitrogen system uses and losses. The proposed 
changes do not exceed or alter a design basis or a safety limit for 
a parameter established in the PBAPS, Units 2 and 3, Updated Final 
Safety Analysis Report (UFSAR) or the PBAPS, Units 2 and 3, Renewed 
Facility Operating License (FOL). Consequently, the proposed changes 
do not result in a reduction in the margin of safety.
    Therefore, the proposed amendments 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 NSHC.
    Attorney for licensee: Mr. J. Bradley Fewell, Associate General 
Counsel, Exelon Generation Company LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Harold K. Chernoff.

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

    Date of amendment request: September 22, 2010.
    Description of amendment request: The proposed amendment would 
relocate the list of pumps, fans, and valves in Technical Specification 
(TS) 4.5.1.1b, Sequence and Power Transfer Test, to the Three Mile 
Island, Unit 1 (TMI-1) Updated Final Safety Analysis Report. In 
addition, TS 4.5.1.2b, TS 4.5.2.2a, and TS 4.5.2.2b refer to this test 
and are proposed for revision to reflect the proposed change to TS 
4.5.1.1b.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below with an NRC edit in brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of any accident previously 
evaluated?
    Response: No.
    The proposed license amendment does not add, delete or modify 
plant equipment. The proposed changes are administrative in nature. 
The proposed amendment would relocate the list of pumps, fans and 
valves in Technical Specification (TS) 4.5.1.1b, Sequence and Power 
Transfer Test, to the TMI-1 Updated Final Safety Analysis Report 
(UFSAR) Section 8.3, Tests and Inspections.
    The proposed changes relocate surveillance requirement details 
that are not required by 10 CFR 50.36, and are [partially] 
consistent with standard technical specifications, NUREG-1430, 
``Standard Technical Specifications Babcock and Wilcox Plants.'' The 
proposed changes do not change current surveillance requirements. 
The subject list of pumps, fans and valves that will be relocated to 
the UFSAR Section 8.3 will continue to be administratively 
controlled and future changes will be controlled under 10 CFR 50.59.
    The probability of an accident is not increased by these 
proposed changes because the Sequence and Power Transfer Test is not 
an initiator of any design basis event. Additionally, the proposed 
changes do not involve any physical changes to plant structures, 
systems, or components (SSCs), or the manner in which these SSCs are 
operated, maintained, or controlled. The consequences of an accident 
will not be increased because the proposed administrative changes to 
the Sequence and Power Transfer Test and Sequence Test will continue 
to provide a high degree of assurance that the Electric Power System 
will meet its safety related function.
    Therefore, 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 previously evaluated?
    Response: No.
    The proposed changes do not alter the physical design, safety 
limits, safety analyses assumptions, or the manner in which the 
plant is operated or tested. The proposed changes are administrative 
in nature and the surveillance requirements remain the same. 
Accordingly, the proposed changes do not introduce any new accident 
initiators, nor do they reduce or adversely affect the capabilities 
of any plant SSC in the performance of their safety function.
    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 the 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 
changes to SSCs or operating and testing procedures associated with 
the proposed amendment.
    The proposed changes do not impact the assumptions of any design 
basis accident, and do not alter assumptions relative to the 
mitigation of an accident or transient event. The proposed changes 
are administrative in nature and the surveillance requirements 
remain the same.
    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, with the NRC edit noted above incorporated, 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: J. Bradley Fewell, Esquire, Associate 
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Harold K. Chernoff.

[[Page 74096]]

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

    Date of amendment request: September 24, 2010.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.4.1.2.3, to allow up to two Main 
Steam Safety Valves (MSSVs) per steam generator to be inoperable with 
no required reduction in power level. It would also revise the required 
maximum overpower trip setpoints for any additional inoperable MSSVs 
consistent with the plant transient analysis. The proposed change 
requires that with less than four MSSVs associated with either steam 
generator operable, the plant would be required to be brought to the 
hot shutdown condition.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below, with NRC edits in brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of any accident previously 
evaluated?
    Response: No.
    The proposed amendment is not a change to the plant structures, 
systems, or components. There is no increase to the likelihood of 
Main Steam Safety Valve (MSSV) related failures. The MSSVs are 
relied upon to mitigate the effects of Updated Final Safety Analysis 
Report (UFSAR) Chapter 14 design basis events including the loss of 
load (turbine trip), which is the limiting event for secondary 
system overpressure. Analyses, performed in accordance with NRC 
approved methods, have demonstrated that with reduced MSSV 
availability and following the specified power level restrictions, 
the MSSVs will continue to limit the secondary system pressure to 
less than 110 percent of the design pressure of the Once Through 
Steam Generators (OTSGs) and the Main Steam (MS) System as required 
by [the American Society of Mechanical Engineers] ASME code. 
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 previously evaluated?
    Response: No.
    The proposed amendment is not a change to the plant structures, 
systems, or components (SSCs). Furthermore, within the current 
licensing basis, the MSSVs are accident mitigation SSCs. The current 
licensing basis does not [explicitly] include consideration of a 
MSSV failure as an event initiator [and a failed open MSSV has been 
shown to be bounded by the larger maximum break size analysis 
presented in the TMI-1 UFSAR]. The proposed amendment will not 
fundamentally alter or create any new operator actions. 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 the margin of safety?
    Response: No.
    The limiting event for secondary system overpressure is a loss 
of load event (turbine trip). The event has been analyzed for 
varying MSSVs out of service, using NRC approved methods. The 
results of the analysis demonstrate that the existing design 
acceptance criteria (i.e., MS and OTSG pressure remain less than 110 
percent of the design pressure) are met for all combinations of 
inoperable MSSVs and initial power levels described in the proposed 
change. 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, including the edits listed above, 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: J. Bradley Fewell, Esquire, Associate 
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Harold K. Chernoff.

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

    Date of amendment request: January 27, 2010.
    Description of amendment request: The proposed changes would amend 
Renewed Facility Operating Licenses DPR-24 and DPR-27 for the Point 
Beach Nuclear Plant, Units 1 and 2, respectively. The proposed 
amendment consists of changes to Technical Specification 3.8.3, 
``Diesel Fuel Oil and Starting Air.''
    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.
    This proposed change increases the minimum required amount of 
stored diesel fuel in the associated common fuel oil storage tank 
[FOST] for two standby emergency power sources to start, load to 
their respective loading limits and to operate continuously up to a 
maximum of 48 hours. An increase in the minimum required fuel oil 
volume required in the fuel oil storage tanks does not increase the 
probability or consequences of an accident previously evaluated.
    [Limiting Condition for Operation] LCO 3.8.3 Condition A, 
currently requires that one or more standby emergency power sources 
have >= 11,000 gallons of fuel when the associated [emergency diesel 
generator] EDG is declared operable. The proposed change increases 
the amount of stored fuel to >= 24,000 gallons for two standby EDGs. 
It further adds new Required Action A.2 if the FOST stored capacity 
falls below the minimum required values. The proposed change also 
accounts for instrument indicator loop uncertainty values for 
unusable volume.
    New LCO [3.8.3] Condition B, addresses the case of one EDG 
operating in either Train ``A'' or Train ``B.'' The new condition 
specifies that the minimum volume of diesel fuel required to support 
continued operation of a single EDG for 48 hours at rated load is >= 
13,000 gallons. This proposed change also accounts for instrument 
indicator loop uncertainty values for unusable volume.
    [Surveillance Requirement] SR 3.8.3.1 is revised to reflect the 
increased amount of diesel fuel required to be maintained to support 
operation of the EDGs following recalculation of required values.
    Following implementation of this proposed change, there will be 
no change in the ability of the EDGs to supply maximum post-accident 
load demands for 48 hours. The proposed minimum volume of fuel, >= 
24,000 gallons for two EDGs and >= 13,000 gallons for one EDG per 
train, ensures that a 48-hour supply of fuel is available when the 
associated standby emergency power source is required to be 
operable.
    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 EDGs and the associated support systems, such as the fuel 
oil storage and transfer systems, are designed to mitigate accidents 
and are not accident initiators. Following this change, the EDGs 
will continue to supply the required maximum post-accident load 
demand. The current 48-hour fuel supply requirements will be 
maintained following this change. The new required fuel oil volumes 
are within the capacities of the fuel oil storage tanks.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    There are two underground fuel oil storage tanks on site. Each 
tank has a capacity of approximately 35,000 gallons and each common 
fuel tank supports one EDG train.

[[Page 74097]]

Fuel can be manually transferred from one tank to another via a 
cross-connect valve. Sufficient fuel is maintained between the two 
tanks to allow one EDG to operate continuously at the required load 
for seven (7) days. At the proposed minimum required level, which is 
>= 24,000 gallons in the common fuel oil storage tanks for two 
standby emergency power sources, one tank could provide enough fuel 
for two EDGs in either Train A or Train B to continue operation for 
great than 48 hours. At the proposed minimum required level, which 
is >= 13,000 gallons in each fuel oil storage tanks, one tank could 
provide enough fuel for one EDG in Train A and Train B to continue 
operation for greater than 48 hours.
    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: William Blair, Senior Attorney, NextEra 
Energy Point Beach, LLC, P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Robert J. Pascarelli.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama

    Date of amendment request: August 27, 2010.
    Description of amendment request: The proposed amendment would add 
a new Action to Technical Specification (TS) 3.7.3, ``Control Room 
Emergency Ventilation (CREV) System,'' to permit one or more CREV 
subsystems to be inoperable for up to 90 days when the inoperability is 
due to inoperable CREV System High Efficiency Particulate Air (HEPA) 
filter and/or charcoal absorbers. The proposed TS changes also include 
an administrative change to correct errors in Unit 2 TS page header 
information that occurred during issuance of TS pages for a previous 
amendment.
    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.
    UFSAR [Updated Finale Safety Analysis Report] Chapter 14, 
``Plant Safety Analysis,'' evaluates operational transients and 
accidents that result in radiological releases that affect control 
room occupants. UFSAR section 14.6, ``Analysis of Design Basis 
Accidents--Uprated,'' evaluates accidents that release fission 
products to the environment. The CREV System is not an accident 
initiator for any of the accidents described. The CREV System 
processes outside air needed to provide ventilation and 
pressurization for control room habitability to limit the control 
room dose during accidents evaluated in the UFSAR. Without crediting 
the performance of the HEPA filter or charcoal adsorbers, the 
analyses results concludes that the 30[-]day integrated post-
accident doses in the control room are within the limits of 5 rem 
TEDE [total effective dose equivalent], as specified in 10 CFR 50.67 
and GDC [General Design Criterion]-19. The control room dose 
increase is less than 10 percent; leaving more than 60 percent 
remaining margin to the regulatory limit.
    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 CREV System is a ventilation system that filters outside air 
used to pressurize the control rooms to provide a protected 
environment from which operators can control the unit during 
airborne challenges from radioactivity during accident conditions. 
The CREV System does not initiate accidents. The proposed amendment 
allows the CREV HEPA filters and charcoal adsorbers to be repaired 
or replaced without shutting down the operating unit(s). No new 
modes of operation are introduced.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Analyses associated with the prior approval of Alternate Source 
Term methodology for design basis accident dose consequences 
previously did not credit the CREV System charcoal adsorbers. Recent 
analyses have been performed to assess the post-accident 30-day 
control room dose removing credit for the CREV System HEPA filter. 
The results indicate a minimal increase in dose consequences (9.5 
percent increase) due to removing credit for the CREV System HEPA 
filter. Even with no credit for either the CREV System HEPA filter 
or CREV System charcoal filter, the resultant control room dose 
maintains more than 60 percent margin to the regulatory limit of 5 
rem TEDE. As such there is no reduction in a margin of safety for 
any duration of inoperability of the CREV System HEPA filter or 
charcoal adsorbers. While the HEPA filter and charcoal adsorbers are 
not credited for accident mitigation, they remain required by the 
BFN TS for compliance with the LCO 3.7.3, ``Control Room Emergency 
Ventilation (CREV) System,'' further minimizing any potential 
reduction in a margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, 6A West Tower, Knoxville, Tennessee 37902.
    NRC Branch Chief: Douglas A. Broaddus.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: October 21, 2010.
    Description of amendment request: The proposed action involves the 
inclusion of the Westinghouse Best-Estimate (BE) Large Break Loss-of-
Coolant Accident (LBLOCA) analysis methodology using the Automated 
Statistical Treatment of Uncertainty Method (ASTRUM) for the analysis 
of LBLOCA to the list of methodologies approved for reference in the 
Core Operating Limits Report (COLR) in Technical Specification (TS) 
5.6.5.b. This action also removes four obsolete COLR references that 
supported North Anna Improved Fuel (NAIF) product, Westinghouse Vantage 
5, since this product is not planned to be used in future North Anna 
cores.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

[Criterion 1]

    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    No physical plant changes are being made as a result of using 
the Westinghouse Best Estimate Large Break LOCA (BE-LBLOCA) analysis 
methodology. The proposed TS change simply involves updating the 
references in TS 5.6.5.b, Core Operating Limits Report (COLR), to 
reference the Westinghouse BE-LBLOCA analysis methodology, which is 
an NRC approved methodology, and to delete unnecessary references. 
Therefore, the probability of LOCA occurrence is not affected by the 
change. Further, the consequences of a LOCA are not increased, since 
the BE-LBLOCA analysis has demonstrated that the performance of the 
Emergency Core Cooling

[[Page 74098]]

System (ECCS) continues to conform to the criteria contained in 10 
CFR 50.46, ``Acceptance Criteria for Emergency Core Cooling Systems 
for Light-Water Nuclear Power Reactors.'' No other accident 
consequence is potentially affected by this change.
    Systems will continue to be operated in accordance with current 
design requirements under the new analysis, therefore no new 
components or system interactions have been identified that could 
lead to an increase in the probability of any accident previously 
evaluated in the Updated Final Safety Analysis Report (UFSAR). No 
changes were required to the Reactor Protection System (RPS) or 
Engineering Safety Features (ESF) setpoints because of the new 
analysis methodology.
    An analysis of the LBLOCA accident for North Anna Units 1 and 2 
has been performed with the Westinghouse BE-LBLOCA analysis 
methodology using ASTRUM. The analysis was performed in compliance 
with the NRC conditions and limitations as identified in WCAP-1 
6009- P-A. Based on the analysis results, it is concluded that the 
North Anna Units 1 and 2 continue to satisfy the limits prescribed 
by 10 CFR 50.46.
    There are no changes to assumptions of the radiological dose 
calculations. Hence, there is no increase in the predicted 
radiological consequences of accidents postulated in the UFSAR.
    Therefore, neither the probability of occurrence nor the 
consequences of an accident previously evaluated is significantly 
increased.

[Criterion 2]

    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The use of the Westinghouse BE-LBLOCA analysis methodology with 
ASTRUM does not impact any of the applicable design criteria and 
pertinent licensing basis criteria continue to be met. Demonstrated 
adherence to the criteria in 10 CFR 50.46 precludes new challenges 
to components and systems that could introduce a new type of 
accident. Safety analysis evaluations have demonstrated that the use 
of Westinghouse BE-LBLOCA analysis methodology with ASTRUM is 
acceptable. Design and performance criteria continue to be met and 
no new single failure mechanisms have been created. The use of the 
Westinghouse BE-LBLOCA analysis methodology with ASTRUM does not 
involve any alteration to plant equipment or procedures that would 
introduce any new or unique operational modes or accident 
precursors. Furthermore, no changes have been made to any RPS or ESF 
actuation setpoints. Based on this review, it is concluded that no 
new accident scenarios, failure mechanisms, or limiting single 
failures are introduced as a result of the proposed changes.
    Therefore, the possibility for a new or different kind of 
accident from any accident previously evaluated is not created.

[Criterion 3]

    Does this change involve a significant reduction in a margin of 
safety?
    Response: No.
    It has been demonstrated that the analytical technique used in 
the Westinghouse BE-LBLOCA analysis methodology using ASTRUM 
realistically describes the expected behavior of the reactor system 
during a postulated LOCA. Uncertainties have been accounted for as 
required by 10 CFR 50.46. A sufficient number of LOCAs with 
different break sizes, different locations, and other variations in 
properties have been considered to provide assurance that the most 
severe postulated LOCAs have been evaluated. The analysis has 
demonstrated that the acceptance criteria contained in 10 CFR 50.46 
continue to be satisfied.
    Therefore, it is concluded that this change does not involve a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 
23219.
    NRC Branch Chief: Gloria Kulesa.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for a Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) The 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland 20854. Publicly available records will be accessible from the 
Agencywide Documents Access and Management System (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, 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].

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos. 
1, 2, and 3, Maricopa County, Arizona

    Date of application for amendment: April 29, 2010.
    Brief description of amendment: The amendments adopted Nuclear 
Regulatory Commission (NRC)-approved TS Task Force (TSTF) Standard 
Technical Specification change traveler TSTF-491, Revision 2, ``Removal 
of Main Steam and Main Feedwater Valve Isolation Times from Technical 
Specifications.'' The isolation times will be located outside of the 
TSs in a document subject to control by the 10 CFR 50.59 process.
    Date of issuance: November 5, 2010.
    Effective Date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: Unit 1--181; Unit 2--181; Unit 3--181.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendment revised the Operating Licenses and Technical Specifications.
    Date of initial notice in the Federal Register: July 27, 2010 (75 
FR 44024).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 5, 2010.
    No significant hazards consideration comments received: No.

[[Page 74099]]

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

    Date of application for amendment: June 23, 2010.
    Brief description of amendment: Current Technical Specification 
(TS) 6.5.8, ``Inservice Testing Program,'' contains references to the 
American Society of Mechanical Engineers (ASME) Boiler and Pressure 
Vessel Code, Section XI as the source of requirements for the inservice 
testing (IST) of ASME Code Class 1, 2, and 3 pumps and valves. The 
amendment deleted the references to Section XI of the Code and 
incorporated references to the ASME Code for Operation and Maintenance 
of Nuclear Power Plants (ASME OM Code). The amendment also indicates 
that there may be some nonstandard frequencies utilized in the IST 
Program in which the provisions of Surveillance Requirement (SR) 3.0.2 
are applicable. The changes are consistent with Technical Specification 
Task Force (TSTF) Technical Change Travelers TSTF-479-A, ``Changes to 
Reflect Revision of 10 CFR 50.55a,'' and TSTF-497-A, ``Limit Inservice 
Testing Program SR 3.0.2 Application to Frequencies of 2 Years or 
Less.''
    Date of issuance: November 5, 2010.
    Effective Date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 291.
    Renewed Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications/license.
    Date of initial notice in the Federal Register: August 10, 2010 (75 
FR 48375).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 5, 2010.
    No significant hazards consideration comments received: No.

Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, 
Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2, Somervell County, 
Texas

    Date of amendment request: May 27, 2010, as supplemented by letter 
dated August 26, 2010.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 3.8.3, ``Diesel Fuel Oil, Lube Oil, and Starting 
Air,'' by relocating the current stored diesel fuel oil and lube oil 
numerical volume and level requirements from the TSs to the TS Bases so 
that it may be modified under licensee control. The TSs have been 
modified so that the stored diesel fuel oil and lube oil inventory will 
require that a 7-day supply be available for each diesel generator. 
Condition A and Condition B in the Action table and Surveillance 
Requirements (SRs) 3.8.3.1 and 3.8.3.2 are also revised to reflect the 
above change. The changes are consistent with NRC-approved Revision 1 
to Technical Specification Task Force (TSTF) Improved Standard 
Technical Specification Change Traveler TSTF-501, ``Relocate Stored 
Fuel Oil and Lube Oil Volume Values to Licensee Control.'' The 
availability of the TS improvement was announced in the Federal 
Register on May 26, 2010, as part of the consolidated line item 
improvement process.
    Date of issuance: November 4, 2010.
    Effective Date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: Unit 1--153; Unit 2--153.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in the Federal Register: August 10, 2010 (75 
FR 48376). The supplemental letter dated August 26, 2010, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 4, 2010.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of application for amendments: December 17, 2009.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) for Limiting Condition for Operations 
3.1.2 ``Reactivity Anomalies'' changing Surveillance Requirement 
3.1.2.1 methodology.
    Date of issuance: November 4, 2010.
    Effective Date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 263 and 207.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: 
Amendments revised the licenses and the TSs.
    Date of initial notice in Federal Register: February 23, 2010.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 4, 2010.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 19th day of November 2010.

    For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2010-29941 Filed 11-29-10; 8:45 am]
BILLING CODE 7590-01-P