[Federal Register Volume 74, Number 16 (Tuesday, January 27, 2009)]
[Notices]
[Pages 4764-4781]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E9-1568]


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NUCLEAR REGULATORY COMMISSION

[NRC-2009-0016]


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 staff) 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 December 31, 2008 to January 13, 2009. The 
last biweekly notice was published on January 13, 2009 (74 FR 1712).

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 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.

[[Page 4765]]

    Written comments may be submitted by mail to the Chief, Rulemaking, 
Directives and Editing Branch, 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. Copies of written comments 
received may be examined at the Commission'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 01F21, 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 within 60 days, 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 set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC E-Filing rule, which the NRC 
promulgated in August 28, 2007 (72 FR 49139). The E-Filing process 
requires participants to submit and serve all adjudicatory documents 
over the Internet or in some cases to mail copies on electronic storage 
media. Participants may not submit paper copies of their filings unless 
they seek a waiver in accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
five (5) days prior to the filing deadline, the petitioner/requestor 
must 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 Viewer \TM\ to 
access the Electronic Information Exchange (EIE), a component of the E-
Filing system. The Workplace Forms Viewer\TM\ is free and is available 
at 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

[[Page 4766]]

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 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. The help electronic filing Help Desk can be 
contacted by telephone at 1-866-672-7640 or by e-mail at 
[email protected].
    Participants who believe that they have a good cause for not 
submitting documents electronically must file a motion, 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 petition 
and/or request 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. 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.
    For further details with respect to this amendment action, 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 01F21, 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. 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, Units 1, 2, 
and 3, Maricopa County, Arizona
    Date of amendment request: July 2, 2008.
    Description of amendment request: The amendments would revise 
Technical Specification (TS) 4.2.2, ``Control Element Assemblies,'' to 
support replacement of the full strength control element assemblies 
(CEAs) with a new design beginning with the 14th refueling outage 
(U3R14) for Palo Verde Nuclear Generating Station (PVNGS), Unit 3 in 
the spring of 2009. Additionally, Arizona Public Service Company (APS) 
will be updating the TS by removing the registered trademark 
``Inconel'' while retaining the generic terminology ``Alloy 625'' and 
deleting the references to part-length CEAs in TS 4.2.2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Replacement of full-strength compression sleeve control element 
assemblies with full-strength silver (Ag)-indium (In)-Cadmium (Cd) 
control element assemblies.
    Response: No.
    The proposed change involves a new design for the full-strength 
Control Element Assemblies (CEA) that replaces a portion of B4C 
pellets (including the compression sleeve) in the tips of the CEA 
fingers with hollow silver-indium-cadmium slugs.
    The following events are related to inadvertent movement of the 
CEAs; however, they are not initiated by the CEAs.
     Uncontrolled Control Element Assembly Withdrawal from a 
Subcritical or Low (Hot Zero) Power Condition.
     Uncontrolled Control Element Assembly Withdrawal at 
Power.
     Single Full-Strength Control Element Assembly Drop.
     Control Element Assembly Ejection.
    These previously analyzed accidents are initiated by the failure 
of plant structures, systems, or components (SSC) other than the CEA 
itself. The proposed change to the CEA design does not have a 
detrimental impact on the integrity of any plant SSC that initiates 
an analyzed event. Additionally, the CEAs mitigate other events. In 
these events, the chrome plating on the portion of the clad exterior 
and the added weight has been conservatively accounted for in the 
SCRAM [safety control rod axe man] calculation. The change does not 
adversely affect the protective and mitigative capabilities of the 
plant, nor does the change affect the initiation or probability of 
occurrence of any accident. The SSCs will continue to perform their 
intended safety functions.
    The proposed change in CEA design has resulted in a slight (less 
than 1%) reduction of total reactivity.
    Computer modeling events which exhibit sensitivity to time 
dependent rod worth (sheared shaft/seized rotor, loss of flow from 
SAFDL [specified acceptable fuel design limits] and total loss of 
reactor coolant flow) demonstrate that all acceptance criteria 
continued to be met.
    Therefore this change will not significantly increase the 
probability or consequences of any accident previously evaluated.
    The removal of the registered trademark name ``Inconel''.
    Response: No.
    This change is considered editorial. Inconel is a registered 
trademark of Special Metals Corporation, while Alloy 625 is a 
generic alloy designation from the Unified Numbering System. 
Retaining the already referenced term ``Alloy 625'' does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated, as the material properties and 
application of Alloy 625 have not changed.
    Deletion of the references to part-length control element 
assemblies.
    Response: No.
    This change is considered editorial. The removal of this 
information does not involve a significant increase in the 
probability or consequences of an accident previously evaluated as 
the part-length CEAs were

[[Page 4767]]

replaced in accordance with License Amendment 152, dated March 23, 
2004 (Agency Document Access and Management System (ADAMS) Accession 
No. ML040860573) and the information is no longer applicable.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Replacement of full-strength compression sleeve control element 
assemblies with full-strength silver(Ag)-indium(In)-Cadmium(Cd) 
control element assemblies.
    Response: No.
    There are three differences in the replacement CEAs as compared 
to the current CEAs.
    First, there is a very slight change in the outside diameter of 
a portion of the cladding on the replacement CEAs due to chrome 
plating on the lower portion of cladding. Analysis demonstrates that 
this change will not cause interference between the CEA cladding and 
the guide tube inside diameter in the buffer region. Secondly, there 
is a slight increase in weight with the Ag-In-Cd CEAs. However, this 
difference has been analyzed with respect to the performance 
capability of the CEDMs [Control Element Drive Mechanisms] and found 
to be within design capabilities and design analyses. Finally, the 
upper edges of the spider bosses have been chamfered to prevent 
damage to the self-latching mechanisms that can occur if the CEA 
hangs up when lifting through the upper guide structure cut outs. 
This change is for ease of maintenance and has no impact on 
operation of the CEAs.
    Therefore, the Ag-In-Cd CEAs are identical to the compression 
sleeve CEAs in terms of form, fit and function and the proposed 
change will not introduce any new failure mechanisms, malfunctions, 
or accident initiators not already considered in the design and 
licensing bases. The possibility of a new or different malfunction 
of safety-related equipment is not created. No new accident 
scenarios, transient precursors, or limiting single failures are 
introduced as a result of these changes. There will be no adverse 
effects or challenges imposed on any safety-related system as a 
result of these changes. Therefore, the possibility of a new or 
different accident from any accident previously evaluated is not 
created as a result of any dimensional change.
    The removal of the registered trademark name ``Inconel''.
    Response: No.
    This change is considered editorial. Inconel is a registered 
trademark of Special Metals Corporation, while Alloy 625 is a 
generic alloy designation from the Unified Numbering System. 
Retaining the already referenced term ``Alloy 625'' does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated, as the material properties and 
application of Alloy 625 have not changed.
    Deletion of the references to part-length control element 
assemblies.
    Response: No.
    This change is considered editorial. The removal of this 
information does not create the possibility of a new or different 
kind of accident from any accident previously evaluated as the part-
length CEAs were replaced in accordance with License Amendment 152, 
dated March 23, 2004 (Agency Document Access and Management System 
(ADAMS) Accession No. ML040860573) and the information is no longer 
applicable.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Replacement of full-strength compression sleeve control element 
assemblies with full-strength silver(Ac)-indium(In)-Cadmium(Cd) 
control element assemblies.
    Response: No.
    Reactor core safety limits are established in the PVNGS 
Technical Specifications to prevent overheating of the fuel and 
cladding that would result in the release of fission products to the 
reactor coolant during steady state operation, normal operational 
transients, and anticipated operational occurrences. The margin to 
these safety limits is not affected by the CEA design changes under 
consideration.
    Overheating of the fuel is prevented by maintaining steady 
state, peak linear heat rate (LHR) below the level at which fuel 
centerline melting occurs. If the local LHR is high enough to cause 
the fuel centerline temperature to reach the melting point of the 
fuel, expansion of the pellet caused by centerline melting may cause 
the pellet to stress the cladding to the point of failure, allowing 
an uncontrolled release of activity to the reactor coolant.
    Compliance with the DNBR [departure from nucleate boiling ratio] 
and fuel centerline melt specified acceptable fuel design limits 
(SAFDLs) is assured through the CEA insertion limits and alignment 
technical specifications, and through the power distribution limit 
technical specifications.
    There is no change to the operation of the full-strength CEAs 
due to the change from compression sleeve CEAs to Ag-In-Cd CEAs. 
Since the Ag-In-Cd CEAs may be used to control power distribution 
similar to the compression sleeve CEAs, power distributions will 
still be controlled and maintained within the limits necessary to 
assure SAFDLs are met.
    The proposed change in CEA design has resulted in a slight (less 
than 1%) reduction in total reactivity.
    Computer modeling results of events which exhibit sensitivity to 
time dependent rod worth (sheared shaft/seized rotor, loss of flow 
from SAFDL and total loss of reactor coolant flow) demonstrate that 
all acceptance criteria continued to be met.
    Therefore, since SAFDLs continue to be met, the change from 
compression sleeve CEAs to Ag-In-Cd CEAs does not involve a 
significant reduction in a margin of safety.
    The removal of the registered trademark name ``Inconel''.
    Response: No.
    The removal of the registered trademark name ``Inconel'' [ ] is 
considered editorial. Inconel is a registered trademark of Special 
Metals Corporation, while Alloy 625 is a generic alloy designation 
from the Unified Numbering System. Retaining the already referenced 
term ``Alloy 625'' does not involve a significant reduction in the 
margin of safety as the material properties and application of Alloy 
625 have not changed.
    Deletion of the references to part-length control element 
assemblies.
    Response: No.
    This change is considered editorial. The removal of this 
information does not involve a significant reduction in the margin 
of safety as the part-length CEAs were replaced in accordance with 
Amendment 152, dated March 23, 2004 (Agency Document Access and 
Management System (ADAMS) Accession No. ML040860573) and the 
information is no longer applicable.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Michael G. Green, Senior Regulatory Counsel, 
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695, 
Phoenix, Arizona 85072-2034.
    NRC Branch Chief: Michael T. Markley.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina
    Date of amendments request: October 6, 2008.
    Description of amendments request: The proposed change would remove 
work hour controls and/or references to the NRC Generic Letter 82-12 
from the administrative control sections of the technical 
specifications. On April 17, 2007, the NRC approved a final rule that 
amended 10 CFR Part 26 and, among other changes, established 
requirements for managing worker fatigue at operating nuclear power 
plants. Subpart I, ``Managing Fatigue,'' specifically addresses 
managing worker fatigue by designating individual break requirements, 
work hour limits, and annual reporting requirements. Subpart I was 
published in the Federal Register on March 31, 2008 (73 FR 16966), with 
a required implementation period of 18 months. Compliance is, 
therefore, required by October 1, 2009. In order to support compliance 
with 10 CFR Part 26, Subpart I, the licensee is proposing to remove 
these work hour controls from Technical Specification 5.2.2.e at the 
Brunswick Steam Electric Plant, Units 1 and 2.
    Basis for proposed no significant hazards consideration 
determination Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the

[[Page 4768]]

licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes remove TS [technical specification] 
controls on working hours for personnel who perform safety related 
functions. The TS controls are superseded by the worker fatigue 
requirements in 10 CFR Part 26. Removal of the TS requirements will 
be performed concurrently with the implementation of the 10 CFR Part 
26, Subpart I requirements. The proposed changes do not impact the 
physical configuration or function of plant structures, systems, or 
components (SSCs) or the manner in which SSCs are operated, 
maintained, modified, tested, or inspected. The proposed changes do 
not impact the initiators or assumptions of analyzed events, nor do 
they impact the mitigation of accidents or transient events.
    Therefore, it is concluded that these changes do 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 changes remove TS controls on working hours for 
personnel who perform safety related functions. The TS controls are 
superseded by the worker fatigue requirements in 10 CFR Part 26. 
Work hours will continue to be controlled in accordance with NRC 
requirements. The new rule allows for deviations from controls to 
mitigate or prevent a condition adverse to safety or as necessary to 
maintain the security of the facility. This ensures that the new 
rule will not restrict work hours and thereby create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not alter plant configuration, require 
that new plant equipment be installed, alter assumptions made about 
accidents previously evaluated, add any initiators, or effect the 
function of plant systems or the manner in which systems are 
operated, maintained, modified, tested, or inspected.
    Therefore, it is concluded that this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes remove TS controls on working hours for 
personnel who perform safety related functions. The TS controls are 
superseded by the worker fatigue requirements in 10 CFR Part 26. The 
proposed changes do not involve any physical changes to plant or the 
manner in which plant systems are operated, maintained, modified, 
tested, or inspected. The proposed changes do not alter the manner 
in which safety limits, limiting safety system settings or limiting 
conditions for operation are determined. The safety analysis 
acceptance criteria are not affected by this change. The proposed 
changes will not result in plant operation in a configuration 
outside the design basis. The proposed changes will not adversely 
affect systems that respond to safely shutdown the plant and to 
maintain the plant in a safe shutdown condition.
    Removal of plant-specific TS administrative requirements will 
not reduce a margin of safety because the requirements in 10 CFR 
Part 26 are adequate to ensure that worker fatigue is managed. 
Therefore, it is concluded that these 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: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, NC 27602.
    NRC Branch Chief: Thomas H. Boyce.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina
    Date of amendment request: October 6, 2008.
    Description of amendments request: The proposed change would remove 
work hour controls and/or references to the NRC Generic Letter 82-12 
from the administrative control sections of the technical 
specifications. On April 17, 2007, the NRC approved a final rule that 
amended 10 CFR Part 26 and, among other changes, established 
requirements for managing worker fatigue at operating nuclear power 
plants. Subpart I, ``Managing Fatigue,'' specifically addresses 
managing worker fatigue by designating individual break requirements, 
work hour limits, and annual reporting requirements. Subpart I was 
published in the Federal Register on March 31, 2008 (73 FR 16966), with 
a required implementation period of 18 months. Compliance is, 
therefore, required by October 1, 2009. In order to support compliance 
with 10 CFR Part 26, Subpart I, the licensee is proposing to remove 
these work hour controls from Technical Specification 5.2.2.e at the H. 
B. Robinson Steam Electric Plant, Unit 2.
    Basis for proposed no significant hazards consideration 
determination Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes remove TS [technical specification] 
controls on working hours for personnel who perform safety related 
functions. The TS controls are superseded by the worker fatigue 
requirements in 10 CFR Part 26. Removal of the TS requirements will 
be performed concurrently with the implementation of the 10 CFR Part 
26, Subpart I requirements. The proposed changes do not impact the 
physical configuration or function of plant structures, systems, or 
components (SSCs) or the manner in which SSCs are operated, 
maintained, modified, tested, or inspected. The proposed changes do 
not impact the initiators or assumptions of analyzed events, nor do 
they impact the mitigation of accidents or transient events.
    Therefore, it is concluded that these changes do 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 changes remove TS controls on working hours for 
personnel who perform safety related functions. The TS controls are 
superseded by the worker fatigue requirements in 10 CFR Part 26. 
Work hours will continue to be controlled in accordance with NRC 
requirements. The new rule allows for deviations from controls to 
mitigate or prevent a condition adverse to safety or as necessary to 
maintain the security of the facility. This ensures that the new 
rule will not restrict work hours and thereby create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not alter plant configuration, require 
that new plant equipment be installed, alter assumptions made about 
accidents previously evaluated, add any initiators, or affect the 
function of plant systems or the manner in which systems are 
operated, maintained, modified, tested, or inspected.
    Therefore, it is concluded that this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes remove TS controls on working hours for 
personnel who perform safety related functions. The TS controls are 
superseded by the worker fatigue requirements in 10 CFR Part 26. The 
proposed changes do not involve any physical changes to the plant or 
the manner in which plant systems are operated,

[[Page 4769]]

maintained, modified, tested, or inspected. The proposed changes do 
not alter the manner in which safety limits, limiting safety system 
settings or limiting conditions for operation are determined. The 
safety analysis acceptance criteria are not affected by this change. 
The proposed changes will not result in plant operation in a 
configuration outside the design basis. The proposed changes will 
not adversely affect systems that respond to safely shut down the 
plant and to maintain the plant in a safe shutdown condition.
    Removal of plant-specific TS administrative requirements will 
not reduce a margin of safety because the requirements in 10 CFR 
Part 26 are adequate to ensure that worker fatigue is managed. 
Therefore, it is concluded that these 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: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, NC 27602.
    NRC Branch Chief: Thomas H. Boyce.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina
    Date of amendment request: October 6, 2008.
    Description of amendment request: The proposed change would remove 
work hour controls and/or references to the NRC Generic Letter 82-12 
from the administrative control sections of the technical 
specifications. On April 17, 2007, the NRC approved a final rule that 
amended 10 CFR Part 26 and, among other changes, established 
requirements for managing worker fatigue at operating nuclear power 
plants. Subpart I, ``Managing Fatigue,'' specifically addresses 
managing worker fatigue by designating individual break requirements, 
work hour limits, and annual reporting requirements. Subpart I was 
published in the Federal Register on March 31, 2008 (73 FR 16966), with 
a required implementation period of 18 months. Compliance is, 
therefore, required by October 1, 2009. In order to support compliance 
with 10 CFR Part 26, Subpart I, the licensee is proposing to remove 
these work hour controls from Technical Specification 6.2.2.f at the 
Shearon Harris Nuclear Power Plant, Unit 1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes remove TS [technical specification] 
controls on working hours for personnel who perform safety related 
functions. The TS controls are superseded by the worker fatigue 
requirements in 10 CFR Part 26. Removal of the TS requirements will 
be performed concurrently with the implementation of the 10 CFR Part 
26, Subpart I requirements. The proposed changes do not impact the 
physical configuration or function of plant structures, systems, or 
components (SSCs) or the manner in which SSCs are operated, 
maintained, modified, tested, or inspected. The proposed changes do 
not impact the initiators or assumptions of analyzed events, nor do 
they impact the mitigation of accidents or transient events.
    Therefore, it is concluded that these changes do 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 changes remove TS controls on working hours for 
personnel who perform safety related functions. The TS controls are 
superseded by the worker fatigue requirements in 10 CFR Part 26. 
Work hours will continue to be controlled in accordance with NRC 
requirements. The new rule allows for deviations from controls to 
mitigate or prevent a condition adverse to safety or as necessary to 
maintain the security of the facility. This ensures that the new 
rule will not restrict work hours and thereby create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not alter plant configuration, require 
that new plant equipment be installed, alter assumptions made about 
accidents previously evaluated, add any initiators, or affect the 
function of plant systems or the manner in which systems are 
operated, maintained, modified, tested, or inspected.
    Therefore, it is concluded that this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes remove TS controls on working hours for 
personnel who perform safety related functions. The TS controls are 
superseded by the worker fatigue requirements in 10 CFR Part 26. The 
proposed changes do not involve any physical changes to the plant or 
the manner in which plant systems are operated, maintained, 
modified, tested, or inspected. The proposed changes do not alter 
the manner in which safety limits, limiting safety system settings 
or limiting conditions for operation are determined. The safety 
analysis acceptance criteria are not affected by this change. The 
proposed changes will not result in plant operation in a 
configuration outside the design basis. The proposed changes will 
not adversely affect systems that respond to safely shut down the 
plant and to maintain the plant in a safe shutdown condition.
    Removal of plant-specific TS administrative requirements will 
not reduce a margin of safety because the requirements in 10 CFR 
Part 26 are adequate to ensure that worker fatigue is managed. 
Therefore, it is concluded that these 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: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, NC 27602.
    NRC Branch Chief: Thomas H. Boyce.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas
    Date of amendment request: November 13, 2008.
    Description of amendment request: The proposed change will modify 
Technical Specification (TS) 3.3.1.1, ``Reactor Protective 
Instrumentation.'' Specifically, Table 4.3-1 and the associated Notes 7 
and 8 will be revised to clarify and streamline the reactor coolant 
system (RCS) flow verification requirements associated with the 
departure from nucleate boiling ratio (DNBR) reactor trip signal.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The CPC [Core Protection Calculator] reactor protective function 
is not considered an accident initiator. The primary function is to 
initiate an automatic reactor trip signal when specific plant 
conditions are reached, thereby limiting the consequences of an 
accident. The proposed change acts to eliminate unnecessary 
conservatisms and accordingly increase operational margin by 
eliminating the requirement to use

[[Page 4770]]

calorimetric flow measurement in the CPC flow verification. This 
method of verification will normally only be used in the future 
during periods when the COLSS [Core Operating Limits Supervisory 
System] RCP [Reactor Coolant Pump] [Delta] p flow measurement is 
unavailable. Regardless of the method of verification used, the CPC 
will continue to be verified to have an indicated RCS flow equal to 
or conservative relative to the measured RCS flow on a once per 12-
hour basis. In so doing, the CPC will continue to act to generate a 
reactor trip on low DNBR as originally designed in order to ensure 
the DNBR reactor core Safety Limit is not exceeded.
    The relocation of measurement uncertainty references to the TS 
Bases does not reduce the requirements to account for uncertainties 
in any Limiting Safety System Setting (LSSS) designed to protect 
reactor core Safety Limits. The necessary uncertainties will 
continue to be applied as required and will be controlled in 
accordance with TS 6.5.14, Technical Specification Bases Control 
Program, and station procedures.
    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 does not result in any physical plant 
modifications or changes in the way the plant is operated. In 
addition, the CPCs are unrelated to any type of accident initiator 
previously evaluated.
    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 increases operating margin when the COLSS 
RCP [Delta]p flow measurement is available for use while unaffecting 
the CPC ability to initiate an automatic reactor trip on low DNBR 
prior to the DNBR reactor core safety limit being exceeded. 
Relocating the references to measurement uncertainties to the TS 
Bases likewise has no impact on the CPC design function and the 
uncertainties will continue to be applied as required and controlled 
in accordance with TS 6.5.14, Technical Specification Bases Control 
Program, and station procedures.
    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.
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: December 8, 2008
    Description of amendment request: The proposed amendment adds a 
license condition to allow a one-time extension of surveillance 
requirements involving the 18-month channel calibration and logic 
system functional tests for one channel of the reactor water level 
instrumentation system. The extension is to account for the effects of 
rescheduling the next refueling outage from early to late 2009.
    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 requested action is a one-time extension to the performance 
interval of certain TS [Technical Specification] surveillance 
requirements. The performance of the surveillances, or the failure 
to perform the surveillances, is not a precursor to an accident. 
Performing the surveillances or failing to perform the surveillances 
does not affect the probability of an accident. Therefore, the 
proposed delay in performance of the surveillance requirements in 
this amendment request does not increase the probability of an 
accident previously evaluated.
    A delay in performing the surveillances does not result in a 
system being unable to perform its required function. Additionally, 
the defense in depth of the system design provides additional 
confidence that the safety function is maintained. In the case of 
this one-time extension request, the relatively short period of 
additional time that the systems and components will be in service 
before the next performance of the surveillance will not affect the 
ability of those systems to operate as designed. Therefore, the 
systems required to mitigate accidents will remain capable of 
performing their required function. No new failure modes have been 
introduced because of this action and the consequences remain 
consistent with previously evaluated accidents. Therefore, the 
proposed delay in performance of the surveillance requirement in 
this amendment request does not involve a significant increase in 
the consequences of an accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment does not involve a physical alteration of 
any system, structure, or component (SSC), or a change in the way 
any SSC is operated. The surveillance intervals of the level 
instrumentation are currently evaluated for 30 months, which bounds 
the requested interval extension. The proposed amendment does not 
involve operation of any SSCs in a manner or configuration different 
from those previously recognized or evaluated. No new failure 
mechanisms will be introduced by the one-time surveillance extension 
being requested.
    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 amendment is a one-time extension of the 
performance-interval of certain TS surveillance requirements. 
Extending the surveillance requirements does not involve a 
modification of any TS Limiting Conditions for Operation. Extending 
the surveillance frequency does not involve a change to any limit on 
accident consequences specified in the license or regulations. 
Extending the surveillance frequency does not involve a change to 
how accidents are mitigated or a significant increase in the 
consequences of an accident. Extending the surveillance frequency 
does not involve a change in a methodology used to evaluate 
consequences of an accident. Extending the surveillance frequency 
does not involve a change in any operating procedure or process. The 
surveillance intervals of the level instrumentation are currently 
evaluated for 30 months which bounds the requested interval 
extension. The components involved in this request have exhibited 
reliable operation based on the results of the most recent 
performances of their 18-month surveillance requirements and the 
associated functional surveillances.
    Based on the limited additional period of time that the systems 
and components will be in service before the surveillance is next 
performed, as well as the operating experience that these 
surveillances are typically successful when performed, it is 
reasonable to conclude that the margin of safety associated with the 
surveillance requirement will not be affected by the requested 
extension.
    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.

[[Page 4771]]

    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.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois
    Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 
and 2, Ogle County, Illinois.
    Date of amendment request: December 4, 2008.
    Description of amendment request: The proposed amendments would 
revise Technical Specifications (TSs) 1.1, ``Definitions,'' and 3.4.16, 
``RCS Specific Activity,'' and Surveillance Requirements 3.4.16.1 and 
3.4.16.3. The proposed changes would replace the current TS 3.4.16 
limit on reactor coolant system (RCS) gross specific activity with a 
new limit on RCS noble gas specific activity. The noble gas specific 
activity limit would be based on a new dose equivalent Xe-133 
definition that would replace the current E Bar average disintegration 
energy definition. In addition, the current dose equivalent I-131 
definition would be reformatted. The availability of this TS revision 
was announced in the Federal Register on March 15, 2007 (72 FR 12217) 
as part of the consolidated line item improvement process. The licensee 
affirmed the applicability of the model no significant hazards 
consideration determination in its application.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration 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.
    Reactor coolant specific activity is not an initiator for any 
accident previously evaluated. The Completion Time when primary coolant 
gross activity is not within limit is not an initiator for any accident 
previously evaluated. The current variable limit on primary coolant 
iodine concentration is not an initiator to any accident previously 
evaluated. As a result, the proposed change does not significantly 
increase the probability of an accident. The proposed change will limit 
primary coolant noble gases to concentrations consistent with the 
accident analyses. The proposed change to the Completion Time has no 
impact on the consequences of any design basis accident since the 
consequences of an accident during the extended Completion Time are the 
same as the consequences of an accident during the Completion Time. As 
a result, the consequences of any accident previously evaluated are not 
significantly increased.
    Criterion 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 in specific activity limits does not alter any 
physical part of the plant nor does it affect any plant operating 
parameter. The change does not create the potential for a new or 
different kind of accident from any previously calculated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.
    The proposed change revises the limits on noble gas radioactivity 
in the primary coolant. The proposed change is consistent with the 
assumptions in the safety analyses and will ensure the monitored values 
protect the initial assumptions in the safety analyses.
    The Nuclear Regulatory Commission (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 amendments involve no 
significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Russell Gibbs.
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power 
Station, Unit No. 1, DeWitt County, Illinois
    Date of amendment request: September 2, 2008.
    Description of amendment request: The proposed amendments would 
relocate Surveillance Requirements (SR) 3.8.3.6 from the technical 
specifications (TSs) to a licensee-controlled document. SR 3.8.3.6 
requires Emergency Diesel Generator fuel oil storage tanks to be 
drained, sediment removed, and cleaned on a 10-year interval. The 
change is consistent with the current revision (i.e., Rev. 3) of the 
Improved Standard Technical Specifications (ISTS), NUREG 1434, 
``Standard Technical Specifications General Electric Plants, BWR/6.'' 
The SR was removed from the ISTS under Technical Specification Task 
Force (TSTF) Traveler No. 2, ``Relocate the 10-Year Sediment Cleaning 
of the Fuel Oil Storage Tank to Licensee Control,'' approved by the 
Nuclear Regulatory Commission on July 16, 1998.
    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 FOSTs [fuel oil storage tanks] provide the storage for the 
DG [diesel generator] fuel oil, assuring an adequate volume is 
available for each DG to operate for seven days in the event of a 
loss of offsite power concurrent with a loss of coolant accident. 
The relocation of the SR to drain and clean the FOSTs to a licensee-
controlled document will not impact any of the previously analyzed 
accidents. Sediment in the tank, or failure to perform this SR, does 
not necessarily result in an inoperable storage tank. Fuel oil 
quantity and quality are assured by other TS SRs that remain 
unchanged. These SRs help ensure tank sediment is minimized and 
ensure that any degradation of the tank wall surface that results in 
fuel oil volume reduction is detected and corrected in a timely 
manner. Future changes to the licensee-controlled document will be 
evaluated pursuant to the requirements of 10 CFR 50.59, ``Changes, 
tests, and experiments,'' to ensure that such changes do not result 
in more than a minimal increase in the probability or consequences 
of an accident previously evaluated.
    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, and configuration or the manner in which the plant is 
operated and maintained. The proposed change does not adversely 
affect the ability of structures, systems or components (SSCs) to 
perform their intended safety function to mitigate the consequences 
of an initiating event within the assumed acceptance limits.
    The proposed change does not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of any accident previously evaluated. 
Further, the proposed change does not increase the types and amounts 
of radiological effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures.
    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 TS change does not involve the addition or 
modification of any plant equipment. Also, the proposed change will

[[Page 4772]]

not alter the design configuration, or method of operation of plant 
equipment beyond its normal functional capabilities. The 
requirements retained in the TS continue to require testing of the 
diesel fuel oil to ensure the proper functioning of the DGs. The 
proposed TS change does not create any new credible failure 
mechanisms, malfunctions or accident initiators.
    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 proposed change does not alter or exceed a design basis or 
safety limit. The requirements retained in the TS continue to 
require testing of the diesel fuel oil to ensure the DGs are able to 
perform their intended function.
    Therefore, the proposed changes does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: Russell Gibbs.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2), 
Beaver County, Pennsylvania
    Date of amendment request: September 24, 2008.
    Description of amendment request: The proposed amendment would 
modify Technical Specifications (TSs) to allow the BVPS-2 containment 
spray additive sodium hydroxide (NaOH) to be replaced by sodium 
tetraborate (NaTB).
    Basis for proposed no significant hazards consideration 
determination: As required by10 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.
    Use of NaTB in lieu of NaOH would not involve a significant 
increase in probability of a previously evaluated accident because 
the containment spray additive is not an initiator of any analyzed 
accident. The NaTB would be stored and delivered by a passive method 
that does not have potential to affect plant operations. Any 
existing NaOH delivery system equipment which remains in place but 
is removed from service would meet existing seismic, electrical and 
containment isolation requirements. Therefore the change in 
additive, including removal of NaOH equipment from service, would 
not result in any failure modes that could initiate an accident.
    The spray additive is used to mitigate the consequences of a 
LOCA [loss-of-coolant accident]. Use of NaTB as an additive in lieu 
of NaOH would not involve a significant increase in the consequences 
of a previously evaluated accident because the amount of NaTB 
specified in the proposed TS would achieve a pH of 7 or greater, 
consistent with the current licensing basis. This pH is sufficient 
to achieve long-term retention of iodine by the containment sump 
fluid for the purpose of reducing accident related radiation dose 
following a LOCA.
    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.
    Regarding the proposed use of NaTB in lieu of NaOH, the NaTB 
would be stored and delivered by a passive method that does not have 
potential to affect plant operations. Any existing NaOH delivery 
system equipment remaining in place but which is removed from 
service would meet existing seismic, electrical and containment 
isolation requirements. Hydrogen generation would not be 
significantly impacted by the change. Therefore, no new failure 
mechanisms, malfunctions, or accident initiators would be introduced 
by the proposed change and it would 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.
    Since the quantity of NaTB specified in the amended TS would 
reduce the potential for undesirable chemical effects while 
achieving radiation dose reductions, corrosion control and hydrogen 
generation effects that are comparable to NaOH, the proposed change 
does not involve a significant reduction in a margin of safety. The 
primary function of an additive is to reduce loss of coolant 
accident consequences by controlling the amount of iodine fission 
products released to containment atmosphere from reactor coolant 
accumulating in the sump during a LOCA. Because the amended 
technical specifications would achieve a pH of 7 or greater using 
NaTB, dose related safety margins would not be significantly 
reduced. Use of NaTB reduces the potential for undesirable chemical 
effects that could interfere with recirculation flow through the 
sump strainers. Any existing NaOH delivery system equipment which 
remains in place but is removed from service would meet existing 
seismic, electrical and containment isolation requirements and would 
not interfere with operation of the existing containment or 
containment spray system.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, FirstEnergy Nuclear 
Operating Company, FirstEnergy Corporation, 76 South Main Street, 
Akron, OH 44308.
    NRC Branch Chief: Mark G. Kowal.
FirstEnergy Nuclear Operating Company (FENOC), et al., Docket No. 50-
440, Perry Nuclear Power Plant, Unit No. 1 (PNPP), Lake County, Ohio
    Date of amendment request: November 18, 2008
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) 5.5.6 to incorporate Technical 
Specification Task Force (TSTF) Travelers TSTF-479, ``Changes to 
Reflect Revision of 10 CFR 50.55a,'' and TSTF 497, ``Limit Inservice 
Testing Program SR [Surveillance Requirement] 3.0.2 Application to 
Frequencies of 2 Years or Less.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment revises TS 5.5.6, ``Inservice Testing 
Program,'' for consistency with 10 CFR 50.55a(f)(4) requirements 
regarding inservice testing of pumps and valves. The proposed 
amendment incorporates revisions to the ASME Code that result in a 
net improvement in the measures for testing pumps and valves. The 
proposed changes do 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 changes do 
not represent a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not involve a modification to the 
physical configuration of the plant. There is no new equipment to be 
installed or a change in the methods governing normal plant 
operation. The proposed change will not impose any new or different 
requirements or introduce a new

[[Page 4773]]

accident initiator, accident precursor, or malfunction mechanism. 
Additionally, there is no change in the types or increases in the 
amounts of any effluent that may be released off-site and there is 
no increase in individual cumulative occupational exposure. 
Therefore, the proposed change does not create the possibility of an 
accident of a different kind than previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment revises TS 5.5.6, ``Inservice Testing 
Program,'' for consistency with the requirements of 10 CFR 
50.55a(f)(4) regarding the inservice testing of pumps and valves. 
The proposed amendment incorporates revisions to the ASME Code that 
result in a net improvement in the measures for testing pumps and 
valves. The safety function of the affected pumps and valves will be 
maintained. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy 
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Russell Gibbs.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit 3 Nuclear Generating Plant, Citrus County, Florida
    Date of amendment request: October 6, 2008.
    Description of amendment request: The proposed change would remove 
work hour controls and/or references to the NRC Generic Letter 82-12 
from the administrative control sections of the technical 
specifications. On April 17, 2007, the NRC approved a final rule that 
amended 10 CFR Part 26 and, among other changes, established 
requirements for managing worker fatigue at operating nuclear power 
plants. Subpart I, ``Managing Fatigue,'' specifically addresses 
managing worker fatigue by designating individual break requirements, 
work hour limits, and annual reporting requirements. Subpart I was 
published in the Federal Register on March 31, 2008 (73 FR 16966), with 
a required implementation period of 18 months. Compliance is, 
therefore, required by October 1, 2009. In order to support compliance 
with 10 CFR Part 26, Subpart I, the licensee is proposing to remove 
these work hour controls from Technical Specification 5.2.2.e at the 
Crystal River Unit 3 Nuclear Generating Plant.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes remove TS [technical specification] 
controls on working hours for personnel who perform safety related 
functions. The TS controls are superseded by the worker fatigue 
requirements in 10 CFR Part 26. Removal of the TS requirements will 
be performed concurrently with the implementation of the 10 CFR Part 
26, Subpart I requirements. The proposed changes do not impact the 
physical configuration or function of plant structures, systems, or 
components (SSCs) or the manner in which SSCs are operated, 
maintained, modified, tested, or inspected. The proposed changes do 
not impact the initiators or assumptions of analyzed events, nor do 
they impact the mitigation of accidents or transient events.
    Therefore, it is concluded that these changes do 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 changes remove TS controls on working hours for 
personnel who perform safety related functions. The TS controls are 
superseded by the worker fatigue requirements in 10 CFR Part 26. 
Work hours will continue to be controlled in accordance with NRC 
requirements. The new rule allows for deviations from controls to 
mitigate or prevent a condition adverse to safety or as necessary to 
maintain the security of the facility. This ensures that the new 
rule will not restrict work hours and thereby create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not alter plant configuration, require 
that new plant equipment be installed, alter assumptions made about 
accidents previously evaluated, add any initiators, or effect the 
function of plant systems or the manner in which systems are 
operated, maintained, modified, tested, or inspected.
    Therefore, it is concluded that this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes remove TS controls on working hours for 
personnel who perform safety related functions. The TS controls are 
superseded by the worker fatigue requirements in 10 CFR Part 26. The 
proposed changes do not involve any physical changes to plant or the 
manner in which plant systems are operated, maintained, modified, 
tested, or inspected. The proposed changes do not alter the manner 
in which safety limits, limiting safety system settings or limiting 
conditions for operation are determined. The safety analysis 
acceptance criteria are not affected by this change. The proposed 
changes will not result in plant operation in a configuration 
outside the design basis. The proposed changes will not adversely 
affect systems that respond to safely shut down the plant and to 
maintain the plant in a safe shutdown condition.
    Removal of plant-specific TS administrative requirements will 
not reduce a margin of safety because the requirements in 10 CFR 
Part 26 are adequate to ensure that worker fatigue is managed. 
Therefore, it is concluded that these 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: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, NC 27602.
    NRC Branch Chief: Thomas H. Boyce.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit 3 Nuclear Generating Plant, Citrus County, Florida
    Date of amendment request: December 17, 2008.
    Description of amendments request: The proposed change would revise 
the Crystal River Unit 3 Improved Technical Specifications 
Administrative Controls, Section 5.6, to revise the Inservice Testing 
Program to incorporate the Technical Specification Task Force (TSTF) 
Standard TS Change Traveler, TSTF-479, Revision 0, ``Changes to Reflect 
Revision of 10 CFR 50.55a,'' and TSTF-497, Revision 0, ``Limit 
Inservice Testing Program SR 3.0.2 Application to Frequencies of 2 
Years or Less.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    4. Does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change revises the CR-3 [Crystal River Unit 3] ITS 
[Improved Technical Specifications], Section 5.6.2.9, ``Inservice 
Testing Program,'' for consistency with the requirements of 10 CFR 
50.55a(f)(4) regarding the inservice testing of pumps and

[[Page 4774]]

valves which are classified as ASME [American Society of Mechanical 
Engineers] 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 change does not impact any accident initiators or 
analyzed events or assumed mitigation of accident or transient 
events. The proposed change does not involve the addition or removal 
of any equipment, or any design changes to the facility. Therefore, 
this proposed change does not involve an increase in probability or 
consequences of an accident previously evaluated.
    5. Does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed change revises the CR-3 ITS, Section 5.6.2.9, 
``Inservice Testing Program,'' for consistency with the requirements 
of 10 CFR 50.55a(f)(4) regarding the inservice testing of 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 change does not involve a modification to the 
physical configuration of the plant (i.e., no new equipment will be 
installed) or involve a change in the methods governing normal plant 
operation. The proposed change will not introduce a new accident 
initiator, accident precursor, or malfunction mechanism. 
Additionally, there is no change in types or increases in the 
amounts of any effluents that may be released offsite and there is 
no increase in individual or cumulative occupational exposure. 
Therefore, the proposed change does not create the possibility of an 
accident of a different kind than previously evaluated.
    6. Does not involve a significant reduction in a margin of 
safety[.]
    The proposed change revises the CR-3 ITS, Section 5.6.2.9, 
``Inservice Testing Program,'' for consistency with the requirements 
of 10 CFR 50.55a(f)(4) regarding the inservice testing of pumps and 
valves which are classified as ASME Code Class 1, Class 2, and Class 
3. The proposed change does not involve a modification to the 
physical configuration of the plant (i.e., no new equipment will be 
installed) or change the methods governing normal plant operation. 
The proposed change incorporates revisions to the ASME Code that 
result in a net improvement in the measures for testing pumps and 
valves. The safety function of the affected pumps and valves will be 
maintained. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, NC 27602.
    NRC Branch Chief: Thomas H. Boyce.
Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue 
County, Minnesota
    Date of amendment request: November 4, 2008.
    Description of amendment request: The proposed amendments would 
make changes to the Technical Specifications to increase the 24 month 
test load for the Unit 1 Emergency Diesel Generators (EDGs), D1 and D2, 
reduce the monthly test load for the Unit 2 EDGs, D5 and D6, and reduce 
the 24 month test loads for the Unit 2 EDGs.
    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 license amendment request proposes to increase a portion of 
the Prairie lsland Nuclear Generating Plant Unit 1 emergency diesel 
generator's 24-month test loading, reduce the Unit 2 emergency 
diesel generators' monthly test loading which demonstrates Technical 
Specification operability and revise the 24-month test to require 
the Unit 2 emergency diesel generators to operate for at least 2 
hours at 100-110% of the continuous rated loading and the remainder 
of the 24-hour test at or above 4000 kW. The proposed test loads 
will continue to assure that the emergency diesel generators have 
the necessary reliability and availability for the design basis 
accidents and station blackout events.
    The emergency diesel generators are required to be operable in 
the event of a design basis accident coincident with a loss of 
offsite power to mitigate the consequences of the accident. They are 
also the alternate AC source for a station blackout on the other 
Prairie lsland Nuclear Generating Plant unit. The emergency diesel 
generators are not accident initiators and therefore these changes 
do not involve a significant increase in the probability of an 
accident previously evaluated.
    The accident analyses assume that at least one safeguards bus is 
provided with power either from the offsite sources or the emergency 
diesel generators. The Technical Specification changes proposed in 
this license amendment request will continue to assure that the 
emergency diesel generators have the capacity and capability to 
assume their maximum auto-connected loads. Thus, the changes 
proposed in this license amendment request do not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    The changes proposed in this license amendment 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.
    This license amendment request proposes to increase a portion of 
the Prairie Island Nuclear Generating Plant Unit 1 emergency diesel 
generator's 24-month test loading, reduce the Unit 2 emergency 
diesel generators' monthly test loading which demonstrates Technical 
Specification operability and revise the 24-month test to require 
the Unit 2 emergency diesel generators to operate for at least 2 
hours at 100-110% of the continuous rated loading and the remainder 
of the 24-hour test at or above 4000 kW. The proposed test loads 
will continue to assure that the emergency diesel generators have 
the necessary reliability and availability for the design basis 
accidents and station blackout events.
    The proposed Technical Specification changes do not involve a 
change in the plant design, system operation, or the use of the 
emergency diesel generators. The proposed changes require the Unit 1 
emergency diesel generators to be tested at increased loads and 
allow the Unit 2 emergency diesel generator to be tested at reduced 
loads which envelope the required safety function loads. These 
revised loads continue to demonstrate the capability and capacity of 
the emergency diesel generators to perform their required functions. 
There are no new failure modes or mechanisms created due to testing 
the emergency diesel generators at the proposed test loading. 
Testing of the emergency diesel generators at the proposed test 
loadings does not involve any modification in the operational limits 
or physical design of plant systems. There are no new accident 
precursors generated due to the proposed test loadings.
    The Technical Specification changes proposed in this license 
amendment do not create the possibility of a new or different kind 
of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    This license amendment request proposes to increase a portion of 
the Prairie Island Nuclear Generating Plant Unit 1 emergency diesel 
generator's 24-month test loading, reduce the Unit 2 emergency 
diesel generators' monthly test loading which demonstrates Technical 
Specification operability and revise the 24-month test to require 
the Unit 2 emergency diesel generators to operate for at least 2 
hours at 100-110% of the continuous rated loading and the remainder 
of the 24-hour test at or above 4000 kW. The proposed test loads 
will continue to assure that the emergency diesel generators have 
the necessary reliability and availability for the design basis 
accidents and station blackout events.

[[Page 4775]]

    The proposed Technical Specification changes will continue to 
demonstrate that the emergency diesel generators meet the Technical 
Specification definition of operability, that is, the proposed tests 
will demonstrate that the emergency diesel generators will perform 
their safety function and the necessary emergency diesel generator 
attendant instrumentation, controls, cooling, lubrication and other 
auxiliary equipment required for the emergency diesel generators to 
perform their safety function loads are also tested at these 
proposed loadings. The proposed testing will also continue to 
demonstrate the capability and capacity of the emergency diesel 
generators to supply their required loss of offsite power loads 
coincident with station blackout loads from the opposite unit. Since 
the proposed surveillance testing will continue to demonstrate 
operability, and the capability and capacity to supply their 
required loss of offsite power coincident with opposite unit station 
blackout loads, the proposed Technical Specification changes do not 
involve a significant reduction in a margin of safety.
    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 requests involve no significant hazards consideration.
    Attorney for licensee: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: Lois M. James.
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: March 27, 2008, as supplemented by a 
letter December 19, 2008.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) requirements related to 
control building envelope habitability in TS Section 3.7.3 Control Room 
Emergency Ventilation (CREV) System, and add TS Section 5.5.13, Control 
Building Envelope Habitability Program, to the Administrative Section 
of the TSs. The licensee has included conforming technical changes to 
the TS Bases. The proposed revision to the Bases also includes 
editorial and administrative changes to reflect applicable changes to 
the corresponding TS Bases, which were made to improve clarity, conform 
to the latest information and references, correct factual errors, and 
achieve more consistency with the standard TS NUREGs. The proposed 
revision to the TS and associated Bases is similar to the TSTF-448, 
Revision 3. The supplement contains additional information related to 
smoke and chemical effects and addresses the associated proposed 
revision to TS Section 3.7.3, TS Section 5.5.13 and TS Bases 3.7.3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed Technical Specification change involve a 
significant increase in the probability or consequences of an 
accident previously evaluated?
    No. The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change revises the TS for the CRE emergency 
ventilation system, which is a mitigation system designed to 
minimize unfiltered air leakage into the CRE and to filter the CRE 
atmosphere to protect the CRE occupants in the event of accidents 
previously analyzed. An important part of the CRE emergency 
ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated. Therefore, the probability of any accident 
previously evaluated is not increased. Performing tests to verify 
the operability of the CRE boundary and implementing a program to 
assess and maintain CRE habitability ensure that the CRE emergency 
ventilation system is capable of adequately mitigating radiological 
consequences to CRE occupants during accident conditions, and that 
the CRE emergency ventilation system will perform as assumed in the 
consequence analyses of design basis accidents. Thus, the 
consequences of any accident previously evaluated are not increased. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed Technical Specification change create the 
possibility of a new or different kind of accident from any accident 
previously evaluated?
    No. The proposed change does not impact the accident analysis. 
The proposed change does not alter the required mitigation 
capability of the CRE emergency ventilation system, or its 
functioning during accident conditions as assumed in the licensing 
basis analyses of design basis accident radiological consequences to 
CRE occupants. No new or different accidents result from performing 
the new surveillance or following the new program. The proposed 
change does not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed) or a 
significant change in the methods governing normal plant operation. 
The proposed change does not alter any safety analysis assumptions 
and is consistent with current plant operating practice. Therefore, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does the proposed Technical Specification change involve a 
significant reduction in a margin of safety?
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition. 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: Thomas H. Boyce.
Tennessee Valley Authority, Docket No. 50-260, Browns Ferry Nuclear 
Plant, Unit 2, Limestone County, Alabama
    Date of amendment request: December 22, 2008 (TS-463-T).
    Description of amendment request: The proposed amendment would, on 
a one-time basis, extend several Technical Specification (TS) 
surveillance frequencies approximately 45 days.
    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 requested action is a one-time extension to the performance 
interval of a limited number of TS surveillance requirements. The 
performance of these surveillances, or the failure to perform these 
surveillances, is not a precursor to an accident. Performing these 
surveillances or

[[Page 4776]]

failing to perform these surveillances does not affect the 
probability of an accident. Therefore, the proposed delay in 
performance of the surveillance requirements in this amendment 
request does not increase the probability of an accident previously 
evaluated.
    A delay in performing these surveillances does not result in a 
system being unable to perform its required function. In the case of 
this one-time extension request, the relatively short period of 
additional time that the systems and components will be in service 
before the next performance of the surveillance will not affect the 
ability of those systems to operate as designed. Therefore, the 
systems required to mitigate accidents will remain capable of 
performing their required function. No new failure modes have been 
introduced because of this action and the consequences remain 
consistent with previously evaluated accidents. Therefore, the 
proposed delay in performance of the surveillance requirements in 
this amendment request does not involve a significant increase in 
the consequences of an accident.
    Therefore, operation of the facility in accordance with the 
proposed license amendment would 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 does not involve a physical alteration of 
any system, structure, or component (SSC) or a change in the way any 
SSC is operated. The proposed amendment does not involve operation 
of any SSCs in a manner or configuration different from those 
previously recognized or evaluated. No new failure mechanisms will 
be introduced by the one-time surveillance requirement extensions 
being requested.
    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.
    The proposed amendment is a one-time extension of the 
performance interval of a limited number of TS surveillance 
requirements. Extending these surveillance requirements does not 
involve a modification of any TS Limiting Conditions for Operation. 
Extending these surveillance requirements does not involve a change 
to any limit on accident consequences specified in the license or 
regulations. Extending these surveillance requirements does not 
involve a change to how accidents are mitigated or a significant 
increase in the consequences of an accident. Extending these 
surveillance requirements does not involve a change in a methodology 
used to evaluate consequences of an accident. Extending these 
surveillance requirements does not involve a change in any operating 
procedure or process.
    The instrumentation and components involved in this request have 
exhibited reliable operation based on the results of the most recent 
performance of their 24-month surveillance requirements.
    Based on the limited additional period of time that the systems 
and components will be in service before the surveillances are next 
performed, as well as the operating experience that these 
surveillances are typically successful when performed, it is 
reasonable to conclude that the margins of safety associated with 
these surveillance requirements will not be affected by the 
requested extension.

    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: Thomas Boyce.
Tennessee Valley Authority, Docket No. 50 390, Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee
    Date of amendment request: August 1, 2008, as supplemented November 
25 and December 31, 2008 (2 letters).
    Description of amendment request: The proposed amendment would 
revise the following: (1) Technical Specification (TS) 4.2.1, ``Fuel 
Assemblies,'' and TS Surveillance Requirements 3.5.1.4, 
``Accumulators,'' and 3.5.4.3, ``RWST [Refueling Water Storage Tank],'' 
to increase the maximum number of Tritium Producing Burnable Absorber 
Rods (TPBARs) that can be irradiated per cycle from 400 to 704.
    An application that addressed similar issues was previously 
submitted on August 1, 2008, and notice of that application was 
provided in the Federal Register on November 12, 2008 (73 FR 66946). 
Due to certain changes in the specifics of the December 31, 2008, 
revision from those proposed in the August 1, 2008, application, as 
supplemented on November 25 and December 31, 2008, the application is 
being renoticed in its entirety. This notice supersedes the notice 
published in the Federal Register on November 12, 2008.
    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 modifies the maximum number of TPBARs in the 
core. The required boron concentration for the cold leg accumulators 
(CLAs) and RWST remains unchanged. The current boron concentration 
has been demonstrated to maintain the required accident mitigation 
safety function for the CLAs and RWST with the higher number of 
TPBARs and this will be verified for each core that contains TPBARs 
as part of the normal reload analysis. The CLAs and RWST safety 
function is to mitigate accidents that require the injection of 
borated water to cool the core and to control reactivity. These 
functions are not potential sources for accident generation and the 
modification of the number of TPBARs will not increase the potential 
for an accident. Therefore, the possibility of an accident is not 
increased by the proposed changes. The current boron concentration 
levels are supported by the proposed number of TPBARs in the core. 
Since the current boron concentration levels will continue to 
maintain the safety function of the CLAs and RWST in the same manner 
as currently approved, the consequences of an accident are not 
increased by the proposed changes.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change only modifies the maximum number of TPBARs 
in the core. The boron concentrations for accident mitigation 
functions of the CLAs and RWST remain unchanged. These functions do 
not have a potential to generate accidents as they only serve to 
perform mitigation functions associated with an accident. The 
proposed modification will maintain the mitigation function in an 
identical manner as currently approved. There are no plant equipment 
or operational changes associated with the proposed revision. 
Therefore, since the CLA and RWST functions are not altered and the 
plant will continue to operate without change, the possibility of a 
new or different kind of an accident is not created.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    This change proposes a change to the maximum number of TPBARs in 
the core. The boron concentration requirements that support the 
accident mitigation functions of the CLAs and RWST remain unchanged. 
The proposed change does not alter any plant equipment or components 
and does not alter any setpoints utilized for the actuation of 
accident mitigation system or control functions. The proposed number 
of TPBARs, in conjunction with the current boron concentration 
values, has been demonstrated to provide an adequate level of 
reactivity control for accident mitigation and this will be verified 
for each core that contains TPBARs as part of the normal reload 
analysis. 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 review, it appears that the three

[[Page 4777]]

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 Acting Branch Chief: P. Milano.

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.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Plant, 
Van Buren County, Michigan
    Date of amendment request: November 25, 2008.
    Brief description of amendment request: The proposed amendment 
would revise Appendix A, Technical Specifications (TS), as they apply 
to the spent fuel pool (SFP) storage requirements in TS section 3.7.16 
and the criticality requirements for the Region I SFP and north tilt 
pit fuel storage racks, in TS section 4.3.1.1.
    Date of publication of individual notice in Federal Register: 
January 2, 2009 (74 FR 123).
    Expiration date of individual notice: February 3, 2009.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri
    Date of amendment request: December 1, 2008.
    Description of amendment request: By letter dated October 31, 2008, 
the Nuclear Regulatory Commission issued Amendment No. 186, to Callaway 
Plant, Unit 1, Facility Operating License No. NPF-30. The amendment 
allowed a one-time extension of the allowed outage time (completion 
time) for each of the two essential service water (ESW) trains (ESW 
Train A and Train B) from 72 hours to 14 days. The extended completion 
time was requested to support planned replacement of the underground 
carbon steel piping with new high density polyethylene (HDPE) piping 
for ESW Train A and ESW Train B during plant operation. The amendment 
was issued with a requirement to complete the replacement of carbon 
steel piping with HDPE piping for both ESW trains by December 31, 2008. 
By its application dated December 1, 2008, the licensee informed NRC 
that it had experienced significant delays in completing the 
replacement of underground piping/conduit due, in part, to underground 
obstructions during excavation, a longer refueling outage (Refuel 16) 
than anticipated, a forced outage at the beginning of Cycle 17, 
switchyard maintenance, and other equipment and personnel issues. 
However, the replacement of ESW Train A carbon steel piping was 
completed by the required date of December 31, 2008, but the 
replacement of ESW Train B carbon steel piping was deferred. 
Consequently, the licensee proposed to extend the implementation date 
for completion of replacement of carbon steel piping for ESW Train B 
from December 31, 2008, to April 30, 2009.
    Date of publication of individual notice in Federal Register: 
December 23, 2008 (73 FR 78858).
    Expiration date of individual notice comment period: January 22, 
2009.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management 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].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit No. 1, DeWitt County, Illinois
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
    Date of application for amendments: June 20, 2008.
    Brief description of amendments: The amendments conform the 
licenses to reflect the direct transfer of AmerGen Energy Company, 
LLC's ownership and operating authority for Clinton Power Station, Unit 
No. 1, Oyster Creek Nuclear Generating Station (Oyster Creek), and 
Three Mile Island Nuclear Station, Unit 1, to Exelon Generation 
Company, LLC, (ECG) as approved by Commission Order dated December 23, 
2008. Transfer of the license for Oyster Creek will also authorize EGC 
to store spent fuel in the Oyster Creek independent spent fuel storage 
installation.
    Date of issuance: January 8, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.

[[Page 4778]]

    Amendment Nos.: CPS-183, Oyster Creek-271, and TMI-1-267.
    Facility Operating License Nos. NPF-62, DPR-16, and DPR-50: The 
amendments revised the Technical Specifications and Licenses.
    Date of initial notice in Federal Register: August 26, 2008 (73 FR 
50368). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 23, 2008.
    No significant hazards consideration comments received: The NRC 
received three comments on August 27, 2008, one for each plant's 
initial notice. The comments did not provide any information additional 
to that in the application, nor did they provide any information 
contradictory to that provided in the application.
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power 
Station, Kewaunee County, Wisconsin
    Date of application for amendment: April 4, 2008.
    Brief description of amendment: The amendment revised the Technical 
Specifications by removing the operability and surveillance 
requirements for the shield building ventilation (SBV) and auxiliary 
building special ventilation filter train heaters, and reducing the 
operating time required to verify the SBV system operability from 10 
hours to 15 minutes.
    Date of issuance: December 30, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 201.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 3, 2008 (73 FR 
31720) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 30, 2008.
    No significant hazards consideration comments received: No.
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power 
Station, Kewaunee County, Wisconsin
    Date of application for amendment: April 14, 2008, as supplemented 
by letter dated October 17, 2008.
    Brief description of amendment: The amendment adds a new footnote 
to Kewaunee Technical Specifications Table 3.5-4, ``Instrument 
Operating Conditions for Isolation Functions.'' The new footnote allows 
the main steam line isolation circuitry to be inoperable when both main 
steam isolation valves are closed and deactivated.
    Date of issuance: January 12, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 202.
    Facility Operating License No. DPR-43: Amendment revised the 
operating license and Technical Specifications.
    Date of initial notice in Federal Register: June 17, 2008 (73 FR 
34340) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 12, 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: December 11, 2007, as 
supplemented December 18, 2008.
    Brief description of amendments: The amendments revised the 
Technical Specifications sections to allow the bypass test times and 
Completion Times (CTs) for Limiting Condition for Operation (LCOs) 
3.3.1, ``Reactor Trip System (RTS) Instrumentation;'' 3.3.2, 
``Engineered Safety Feature Actuation System (ESFAS) Instrumentation;'' 
3.3.6, ``Containment Air Release and Addition Isolation 
Instrumentation,'' and 3.3.9, ``Boron Dilution Mitigation System 
(BDMS).''
    The proposed license amendment request (LAR) adopts changes as 
described in Westinghouse Commercial Atomic Power (WCAP) topical report 
WCAP-14333-P-A, Revision 1, ``Probabilistic Risk Analysis of the 
Reactor Protection System and Engineered Safety Features Actuation 
System Test Times and Completion Times,'' issued October 1998 and 
approved by U.S. Nuclear Regulatory Commission (NRC) letter dated July 
15, 1998. Implementation of the proposed changes is consistent with 
Technical Specification Task Force (TSTF) Traveler TSTF-418, Revision 
2, ``RPS [Reactor Protection System] and ESFAS Test Times and 
Completion Times (WCAP-14333).'' The NRC approved TSTF-418, Revision 2, 
by letter dated April 2, 2003.
    In addition, the proposed LAR adopts changes as described in WCAP-
15376-P-A, Revision 1, ``Risk-Informed Assessment of the RTS and ESFAS 
Surveillance Test Intervals and Reactor Trip Breaker Test and 
Completion Times,'' issued March 2003, as approved by NRC letter dated 
December 20, 2002. Implementation of the proposed changes is consistent 
with TSTF Traveler  TSTF-411, Revision 1, ``Surveillance Test 
Interval Extension for Components of the Reactor Protection System 
(WCAP-15376).'' The NRC approved TSTF-411, Revision 1, by letter dated 
August 30, 2002. The licensee also requested additional changes not 
specifically included in the above topical reports. These changes will 
be evaluated in a future amendment.
    Date of issuance: December 22, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 247 and 240.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: March 25, 2008 (73 FR 
15783). The supplement dated December 18, 2008, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 22, 2008.
    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: December 11, 2007, as 
supplemented by letter dated December 18, 2008.
    Brief description of amendments: The amendments revised the 
Technical Specification sections to allow the bypass test times and 
Completion Times for Limiting Condition for Operation 3.3.1, ``Reactor 
Trip System (RTS) Instrumentation'' and 3.3.2, ``Engineered Safety 
Feature Actuation System (ESFAS) Instrumentation.''
    By letter dated December 30, 2008 (Agencywide Documents Access and 
Management System Accession No. ML0083460216), the NRC issued Amendment 
No. 247 and Amendment No. 240 for Catawba Units 1 and 2, respectively, 
for all the proposed changes approved by the NRC in TSTFs 411 and 418. 
The December 30, 2008, amendment stated that the following changes 
would be evaluated in a future amendment:
    Surveillance requirement (SR) 3.3.1.5, Safety injection input from 
ESFAS, Condition J, Feedwater isolation with low average core 
temperature coincident with reactor trip P-4, SR 3.3.2.2, turbine

[[Page 4779]]

trip and feedwater isolation for steam generator water level high high.
    (P-14), SR 3.3.2.4 turbine trip and feedwater isolation for steam 
generator water level high high (P-14), and SR 3.3.2.5 turbine trip and 
feedwater isolation for low average core temperature trip coincident 
with reactor trip P-4.
    This amendment approves the above changes.
    Date of issuance: January 9, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 248 and 241.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: March 25, 2008 (73 FR 
15783). The supplement dated December 18, 2008, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the NRC 
staff's original proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 9, 2009.
    No significant hazards consideration comments received: No.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont
    Date of application for amendment: February 6, 2008, as 
supplemented by letter dated July 29, 2008.
    Brief description of amendment: The amendment revised the 
Surveillance Requirements (SRs) for control rod exercising from weekly 
to monthly in Technical Specification (TS) 4.3.A.2, revise verification 
of control rod coupling integrity as described in TS 4.3.B.1, revise 
the scram insertion time Limiting Conditions for Operation (LCOs) and 
SRs as described in TS 3.3.C and 4.3.C, and enhance TS 3.3.D and 4.3.D, 
the LCO and SR for Control Rod Accumulators.
    Date of issuance: January 7, 2009.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 233.
    Facility Operating License No. DPR-28: Amendment revised the 
License and Technical Specifications.
    Date of initial notice in Federal Register: March 11, 2008 (73 FR 
13024). The supplemental letter dated July 29, 2008, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination. The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated January 7, 2009.
    No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas
    Date of amendment request: July 30, 2008, as supplemented by letter 
dated October 2, 2008.
    Brief description of amendment: The amendment revises the current 
TS 3.6.6.3 surveillance requirements for sodium hydroxide (NaOH) 
concentration. Specifically, the amendment changes the surveillance 
requirements of the NaOH tank solution concentration from between 5.0 
weight (wt.) percent and 16.5 wt. percent to between 6.0 wt. percent 
and 8.5 wt. percent.
    Date of issuance: January 13, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: Unit 1--234.
    Renewed Facility Operating License No. DPR-51: Amendment revised 
the License and Technical Specifications.
    Date of initial notice in Federal Register: November 4, 2008, (73 
FR 65694). The supplement dated October 2, 2008, 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 January 13, 2009.
    No significant hazards consideration comments received: No.
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, Claiborne 
County, Mississippi
    Date of application for amendment: June 30, 2008.
    Brief description of amendment: The amendment (1) deleted Technical 
Specification (TS) surveillance requirement (SR) 3.1.3.2 and revised SR 
3.1.3.3; (2) removed the reference to SR 3.1.3.2 from Required Action 
A.2 of TS 3.1.3, ``Control Rod OPERABILITY''; (3) clarified the 
requirement to fully insert all insertable rods for the limiting 
condition for operation in TS 3.3.1.2 Required Action E.2, ``Source 
Range Monitoring Instrumentation''; and (4) revised Example 1.4-3 in 
Section 1.4, ``Frequency,'' to clarify the applicability of the 1.25 
surveillance test interval extension.
    Date of issuance: December 31, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No: 180.
    Facility Operating License No. NPF-29: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: August 26, 2008 (73 FR 
50359).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 31, 2008.
    No significant hazards consideration comments received: No.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3, Nuclear Generating Plant, Citrus County, Florida
    Date of application for amendment: January 17, 2008.
    Brief description of amendment: The amendment revises the Crystal 
River, Unit 3 Improved Technical Specification Surveillance Requirement 
3.7.5.2, ``Emergency Feedwater System,'' to align the text for the 
emergency feedwater system surveillance frequency with the text in the 
Technical Specifications Task Force Standard Technical Specification 
Change Traveler-101, Revision 0 and the NRC technical report, NUREG-
1430, Volume 1, Revision 3, ``Standard Technical Specifications Babcock 
and Wilcox Plants--Specification.''
    Date of issuance: January 9, 2009.
    Effective date: Date of issuance, to be implemented within 60 days.
    Amendment No.: 231.
    Facility Operating License No. DPR-72: Amendment revises the 
technical specifications.
    Date of initial notice in Federal Register: May 20, 2008 (73 FR 
29163).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 9, 2009.
    No significant hazards consideration comments received: No.

[[Page 4780]]

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 application for amendments: December 17, 2007, as 
supplemented by letters dated October 2, and November 18, 2008.
    Brief description of amendments: The amendments increase the 
completion times (CTs) for required actions related to Technical 
Specifications (TS) 3.5.2, regarding the Emergency Core Cooling System, 
and 3.6.6, regarding the Containment Spray and Cooling Systems from 72 
hours to 14 days. In addition, invalid notes were deleted from TSs 
3.5.2 and 3.6.6 and new notes were added to specify the limitations on 
the use of the 14-day extended CT.
    Date of issuance: December 31, 2008.
    Effective date: As of its date of issuance and shall be implemented 
within 180 days from the date of issuance.
    Amendment Nos.: Unit 1--202; Unit 2--203.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: January 29, 2008 (73 FR 
5227). The supplement(s) dated October 2 and November 18, 2008, 
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 December 31, 2008.
    No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
    Date of application for amendments: July 7, 2008.
    Brief description of amendments: The amendments revised the 
Technical Specification (TS) testing frequency for the Surveillance 
Requirement (SR) in TS 3.1.4, ``Control Rod Scram Times.'' The change 
revised the frequency of SR 3.1.4.2, control rod scram time testing, 
from ``120 days cumulative operation in Mode 1'' to ``200 days 
cumulative operation in Mode 1.'' These changes are based on TS Task 
Force (TSTF) change traveler TSTF-460 (Revision 0) that has been 
approved generically for the Boiling-Water Reactor (BWR) Standard TS, 
NUREG-1433 (BWR/4) and NUREG-1434 (BWR/6) by revising the frequency of 
SR 3.1.4.2, control rod scram time testing, from ``120 days cumulative 
operation in MODE 1'' to ``200 days cumulative operation in MODE 1.''
    Date of issuance: January 2, 2009.
    Effective date: January 2, 2009.
    Amendment Nos.: 249 for Unit 1 and 228 for Unit 2
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the License and Technical Specifications.
    Date of initial notice in Federal Register: October 7, 2008 (73 FR 
58675).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 2, 2009.
    No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
    Date of application for amendments: July 7, 2008.
    Brief description of amendments: The amendment adopted the Nuclear 
Regulatory Commission (NRC) approved Technical Specification Task Force 
(TSTF) change traveler TSTF-475, (Revision 1), ``Control Rod Notch 
Testing Frequency and SRM [Source Range Monitor] Insert Control Rod 
Action,'' to change the Standard Technical Specifications (STS) for 
General Electric (GE) Plants (NUREG-1433, BWR/4 to the plant-specific 
TS, that allows: (1) Revising the frequency of Surveillance Requirement 
(SR) 3.1.3.2, notch testing of fully withdrawn control rod, from ``7 
days after the control rod is withdrawn and THERMAL POWER is greater 
than the LPSP of RWM'' to ``31 days after the control rod is withdrawn 
and THERMAL POWER is greater than the LPSP [Low Power Set Point] of the 
RWM [Rod With Minimizer]'', and (2) revising Example 1.4-3 in Section 
1.4 ``Frequency'' to clarify that the 1.25 surveillance test interval 
extension in SR 3.0.2 is applicable to time periods discussed in NOTES 
in the ``SURVEILLANCE'' column in addition to the time periods in the 
``FREQUENCY'' column.
    Date of issuance: January 2, 2009.
    Effective date: January 2, 2009.
    Amendment Nos.: 250 for Unit 1 and 229 for Unit 2.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the License and Technical Specifications.
    Date of initial notice in Federal Register: October 7, 2008 (73 FR 
58675).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 2, 2009.
    No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50 390, Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee
    Date of application for amendment: September 19, 2008.
    Brief description of amendment: The amendment modifies the Final 
Safety Analysis Report by requiring an inspection of the ice condenser 
within 24 hours of experiencing a seismic event greater than or equal 
to an operating basis earthquake within the 5-week period after ice 
basket replenishment has been completed to confirm that adverse ice 
fallout has not occurred that could impede the ability of the ice 
condenser lower inlet doors to open.
    Date of issuance: January 6, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days of issuance.
    Amendment No.: 73.
    Facility Operating License No. NPF-90: Amendment authorizes 
revision to the FSAR.
    Date of initial notice in Federal Register: November 4, 2008 (73 FR 
65698).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 6, 2009.
    No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50 390, Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee
    Date of application for amendment: March 27, 2008, as supplemented 
September 26, 2008.
    Brief description of amendment: The amendment revises the allowable 
value listed for Function 3, ``Containment Purge Exhaust Radiation 
Monitors,'' in Table 3.3.6-1, ``Containment Vent Isolation 
Instrumentation,'' of the limited condition for operation 3.3.6.
    Date of issuance: January 8, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 74.
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications and License.
    Date of initial notice in Federal Register: May 6, 2008 (73 FR 
25047). The supplement dated September 26,

[[Page 4781]]

2008, 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 January 8, 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 amendment: December 17, 2007, as 
supplemented on July 22, 2008, September 26, 2008, and November 25, 
2008.
    Brief description of amendment: These amendments revised Technical 
Specification (TS) 3.8.3 to allow a one-time extended 14-day completion 
time (CT) for each of the two underground diesel fuel oil storage tanks 
(FOST) to permit removal of the current coating and to recoat the tanks 
in preparation for use of ultra-low sulfur diesel fuel oil. The change 
revised the TS to extend the CT associated with an inoperable emergency 
diesel generator FOST from 7 days to 14 days, applicable once for each 
of the two tanks.
    Date of issuance: December 31, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 254 and 235.
    Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments 
change the licenses and the technical specifications.
    Date of initial notice in Federal Register: January 15, 2008 (73 FR 
2552). The supplements dated July 22, 2008, September 26, 2008, and 
November 25, 2008, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated December 31, 2008.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 15th day of January 2009.

    For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. E9-1568 Filed 1-26-09; 8:45 am]
BILLING CODE 7590-01-P