Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 31318-31331 [E9-15117]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 74, Number 124 (Tuesday, June 30, 2009)]
[Notices]
[Pages 31318-31331]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E9-15117]


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NUCLEAR REGULATORY COMMISSION

[NRC-2009-0261]


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 June 4, 2009 to June 17, 2009. The last 
biweekly notice was published on June 16, 2009 (74 FR 28575).

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.
    Written comments may be submitted by mail to the Chief, Rulemaking 
and Directives 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, https://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

[[Page 31319]]

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 
hearing.docket@nrc.gov, or by calling (301) 415-1677, to request (1) a 
digital ID certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and/or (2) 
creation of an electronic docket for the proceeding (even in instances 
in which the petitioner/requestor (or its counsel or representative) 
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms ViewerTM 
to access the Electronic Information Exchange (EIE), a component of the 
E-Filing system. The Workplace Forms ViewerTM is free and is 
available at https://www.nrc.gov/site-help/e-submittals/install-viewer.html. Information about applying for a digital ID certificate is 
available on NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/apply-certificates.html.
    Once a petitioner/requestor has obtained a digital ID certificate, 
had a docket created, and downloaded the EIE viewer, it can then submit 
a request for hearing or petition for leave to intervene. Submissions 
should be in Portable Document Format (PDF) in accordance with NRC 
guidance available on the NRC public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the 
time the filer submits its documents through EIE. To be timely, an 
electronic filing must be submitted to the EIE system no later than 
11:59 p.m. Eastern Time on the due date. Upon receipt of a 
transmission, the E-Filing system time-stamps the document and sends 
the submitter an e-mail notice confirming receipt of the document. The 
EIE system also distributes an e-mail notice that provides access to 
the document to the NRC Office of the General Counsel and any others 
who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically may seek assistance through the 
``Contact Us'' link located on the NRC Web site at https://www.nrc.gov/site-help/e-submittals.html or by calling the NRC electronic filing 
Help Desk, which is available between 8 a.m. and 8 p.m., Eastern Time, 
Monday through Friday, excluding government holidays. The electronic 
filing Help Desk can be contacted by telephone at 1-866-672-7640 or by 
e-mail at MSHD.Resource@nrc.gov.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file 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

[[Page 31320]]

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 
https://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant 
to an order of the Commission, an Atomic Safety and Licensing Board, or 
a Presiding Officer. Participants are requested not to include personal 
privacy information, such as social security numbers, home addresses, 
or home phone numbers in their filings, unless an NRC regulation or 
other law requires submission of such information. With respect to 
copyrighted works, except for limited excerpts that serve the purpose 
of the adjudicatory filings and would constitute a Fair Use 
application, participants are requested not to include copyrighted 
materials in their 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, https://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr.resource@nrc.gov.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of amendments request: May 21, 2009.
    Description of amendments request: The amendments would remove the 
Table of Contents (TOC) from the Technical Specifications (TSs) and 
place them under licensee control.
    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?
    No.
    The proposed change is administrative and affects control of a 
document, the TOC, listing the specifications in the plant TSs. 
Transferring control from the Nuclear Regulatory Commission (NRC) to 
CCNPP [Calvert Cliffs Nuclear Power Plant] (the licensee) does not 
affect the operation, physical configuration, or function of plant 
equipment or systems. It does not impact the initiators or 
assumptions of analyzed events; nor does it impact the mitigation of 
accidents or transient events. The change has no impact on, and 
hence cannot increase, 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?
    No.
    The proposed change is administrative and does not alter the 
plant configuration, require installation or new equipment, alter 
assumptions about previously analyzed accidents, or impact the 
operation or function of plant equipment or systems. 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 amendment involve a significant reduction 
in a margin of safety?
    No.
    The proposed change is administrative. The TOC is not required 
by regulation to be in the TS. Removal does not impact any safety 
assumptions or have the potential to reduce a margin of safety as 
described in the TS Bases. The change involves a transfer of control 
of the TOC from the NRC to CCNPP. No change in the technical content 
of the TS specifications is involved. Consequently, transfer from 
the NRC to CCNPP has no impact on the margin of safety, and hence 
cannot involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear 
Generation, Constellation Generation Group, LLC, 750 East Pratt Street, 
17th floor, Baltimore, MD 21202.
    NRC Acting Branch Chief: John Boska.

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

    Date of amendment request: May 5, 2009.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) Section 6.7.C to change 
requirements related to the schedule for performing the 10 CFR Part 50, 
Appendix J, Type A test. Specifically, the proposed change would change 
the TS from requiring the test ``no later than April 2010'' to ``prior 
to startup from the April 2010 refuel outage.''
    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:

    1.0 Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No. The change does not impact the function of any 
structure, system or component that affects the probability of an 
accident or that supports mitigation or consequences of an accident 
previously evaluated. The proposed change involves testing of 
Primary Containment but does not impact containment design or 
performance requirements. The proposed change ensures that the Type 
A test is performed prior to establishing Primary Containment 
following the April 2010 Refuel[ing] Outage. The proposed change 
does not affect reactor operations or accident analysis and there is 
no change to the radiological consequences of a previously analyzed 
accident. The operability requirements for accident mitigation 
systems remain consistent with the licensing and design basis. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2.0 Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No. The proposed change does not involve any physical 
alteration of plant equipment and does not change the method by 
which any safety-related system performs its function. The proposed 
change involves the scheduling of the Type A test and does not alter 
the way the test is performed. Type A tests have been previously 
performed and are well within the design capability of station 
structures, systems or components. No new or different types of 
equipment will be permanently installed or operated. Operation of 
existing installed equipment is unchanged. The methods governing 
plant operation and testing remain consistent with current safety 
analysis assumptions. Therefore, the proposed change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3.0 Does the proposed change involve a significant reduction in 
a margin of safety?
    Response: No. These changes do not change any existing design or 
operational requirements and do not adversely affect existing plant 
safety margins or the reliability of the equipment assumed to 
operate in the safety analysis. The proposed change affects the 
schedule for performing the Type A test and does not affect the way 
the test is

[[Page 31321]]

performed or margins for the existing Primary Containment. As such, 
there are no changes being made to safety analysis assumptions, 
safety limits or safety system settings that would adversely affect 
plant safety as a result of the proposed change. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Acting Branch Chief: John Boska.

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

    Date of amendment request: May 13, 2009.
    Description of amendment request: The proposed change will modify 
the Technical Specification (TS) 2.1.1.1, ``DNBR,'' to revise the 
Departure from Nucleate Boiling Ratio (DNBR) safety limit based upon 
the Combustion Engineering (CE) 16 x 16 Next Generation Fuel (NGF) 
design and the associated Departure from Nucleate Boiling (DNB) 
correlations.
    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.
    No changes to plant equipment or operating procedures are 
required due to the change in the safety limit for DNBR. This change 
does not impact any of the accident initiators. The analyses of the 
reload are performed using NRC [U.S. Nuclear Regulatory Commission] 
approved methodologies to ensure the Specified Acceptable Fuel 
Design Limits (SAFDLs), of which DNBR is one, are not violated. The 
current DNBR setpoint continues to ensure automatic protective 
action is initiated to prevent exceeding the proposed DNBR safety 
limit.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed 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 plant modifications 
or change in the way the plant is designed to function. The proposed 
change is not associated with any accident precursor or initiator. 
The proposed change supports the loading and use of Next Generation 
Fuel (NGF) at ANO-2 [Arkansas Nuclear One, Unit 2] as previously 
approved by the NRC.
    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 use the NRC-approved NGF WSSV-T correlation with the ABB-NV 
correlation to establish a new bounding DNBR safety limit of 1.23, 
preserves the DNBR margin of safety at a 95/95 level. The Core 
Protection Calculator (CPC) DNBR power adjustment addressable 
constant BERR1 is calculated based on the WSSV-T and ABB-NV CHF 
[critical heat flux] correlations such that a CPC trip at a DNBR of 
1.25 using the CE-1 CHF correlation assures that the bounding DNBR 
safety limit of 1.23 for the WSSV-T and ABB-NV CHF correlations will 
not be violated during normal operation and AOOs [anticipated 
operational occurrences] to at least a 95/95 probability/confidence 
level.
    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 
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: May 15, 2009.
    Description of amendment request: The amendment would modify 
Technical Specification (TS) 6.6.5, ``Core Operating Limits Report 
(COLR),'' to minimize the number of U.S. Nuclear Regulatory Commission 
(NRC)-approved references consistent with the guidance provided in NRC 
Generic Letter 88-16, ``Removal of Cycle-Specific Parameter Limits from 
Technical Specifications,'' dated October 3, 1988. This request also 
fulfills the commitment made in the licensee's letter to the NRC dated 
March 11, 2008, ``Response to Request for Additional Information 
License Amendment Request to Revise Technical Specification 6.6.5, Core 
Operating Limits Report.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to the list of NRC-approved methodologies 
listed in TS 6.6.5 are administrative in nature and have no impact 
on any plant configuration or system performance relied upon to 
mitigate the consequences of an accident. Changes to the calculated 
core operating limits may only be made using NRC-approved 
methodologies, must be consistent with all applicable safety 
analysis limits, and are controlled by the 10 CFR 50.59 [Title 10 of 
the Code of Federal Regulations Section 50.59] process.
    The proposed change will minimize and clarify the listing of the 
NRC-approved methodologies that are currently being used in the ANO-
2 [Arkansas Nuclear One, Unit 2] core designs and the determination 
of the operating limits for those cores. Assumptions used for 
accident initiators and/or safety analysis acceptance criteria are 
not altered by the proposed changes.
    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 to the list of topical reports used to 
determine the operating limits has no impact on any plant 
configurations or on system performance that is relied upon to 
mitigate the consequences of an accident. These changes are 
administrative in nature and do not result in a change to the 
physical plant or to the modes of operation defined in the facility 
license.
    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 does not amend the cycle specific parameter 
limits located in the COLR from the values presently required by the 
TS. The individual specifications continue to require operation of 
the plant within the bounds of the limits specified in COLR. The 
proposed change to the list of analytical methods referenced in the 
COLR is administrative in nature.

[[Page 31322]]

    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 
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: May 22, 2009.
    Description of amendment request: The proposed amendment will 
modify the Waterford Steam Electric Station, Unit 3 (Waterford 3), 
Technical Specification (TS) 6.9.1.11 to minimize the number of 
references that reflect U.S. Nuclear Regulatory Commission (NRC)-
approved methods used in establishing the Core Operating Limits Report 
(COLR) parameter limits, consistent with the guidance provided in NRC 
Generic Letter 88-16, ``Removal of Cycle-Specific Parameter Limits from 
Technical Specifications,'' dated October 3, 1988.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to the list of NRC-approved methodologies 
listed in TS 6.9.1.11.1 are administrative in nature and have no 
impact on any plant configuration or system performance relied upon 
to mitigate the consequences of an accident. Changes to the 
calculated core operating limits may only be made using NRC approved 
methodologies, must be consistent with all applicable safety 
analysis limits, and are controlled by the 10 CFR 50.59 [Title 10 of 
the Code of Federal Regulations Section 50.59] process.
    The proposed changes will minimize and clarify the listing of 
the NRC-approved methodologies that are currently being used in the 
Waterford 3 core designs and the determination of the operating 
limits for those cores.
    Assumptions used for accident initiators and/or safety analysis 
acceptance criteria are not altered by the proposed changes.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes to the list of topical reports used to 
determine the operating limits has no impact on any plant 
configurations or on system performance that is relied upon to 
mitigate the consequences of an accident. These changes are 
administrative in nature and do not result in a change to the 
physical plant or to the modes of operation defined in the facility 
license.
    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 changes do not amend the cycle specific parameter 
limits located in the COLR from the values presently required by the 
TS. The individual specifications continue to require operation of 
the plant within the bounds of the limits specified in COLR.
    The proposed changes to the list of analytical methods 
referenced in the COLR are administrative in nature.
    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.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station (DNPS), Units 2 and 3, Grundy County, Illinois; 
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station 
(QCPS), Units 1 and 2, Rock Island County, Illinois

    Date of application for amendment request: April 7, 2009.
    Description of amendment request: The proposed amendment deletes a 
no longer applicable footnote from the DNPS Technical Specifications 
(TS), corrects administrative errors in the titles of analytical 
methods, and deletes historical analytical methods no longer applicable 
in DNPS and QCPS TS. The proposed amendment also deletes a license 
condition from the DNPS and QCPS Renewed Facility Operating License 
(FOL).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    DNPS TS 3.4.5, ``RCS Leakage Detection Instrumentation,'' 
establishes the applicability and requirements for equipment used to 
quantify unidentified reactor coolant system operational leakage 
(i.e., the drywell floor drain sump monitoring system). The proposed 
change deletes a footnote that established a limited duration 
alternative to these requirements for DNPS Unit 3.
    The deletion of the footnote restores DNPS TS 3.4.5 requirements 
to the requirements prior to NRC approval of an emergency license 
amendment, which provided an alternative means to demonstrate TS 
compliance. In that the condition necessitating the footnote (i.e., 
a failed component) has been resolved (i.e., repair of the failed 
component), the footnote is no longer applicable. The proposed 
change will have no effect on any accident initiator or precursor 
previously evaluated and will not change the manner in which the 
plant is operated. Thus, the proposed change does not have any 
effect on the probability of an accident previously evaluated.
    DNPS and QCNPS TS 5.6.5 ``Core Operating Limits Report (COLR),'' 
lists the NRC-approved analytical methods that are used at DNPS and 
QCNPS to determine core operating limits. The proposed changes will 
correct administrative errors in the titles of several analytical 
methods in DNPS and QCNPS TS 5.6.5.b. The proposed changes will also 
delete historical analytical methods from DNPS and QCNPS TS 5.6.5.b 
that are no longer applicable, as well as renumber the remaining 
analytical methods.
    The correction of administrative errors in the titles of 
analytical methods does not change the content or application of the 
methods. Similarly, the deletion of non-applicable analytical 
methods does not affect the ability to accurately model core 
behavior, including the determination of core operating limits, for 
the fuel that is currently loaded in the DNPS and QCNPS reactors. 
Therefore, the proposed changes will have no effect on any accident 
initiator or precursor previously evaluated and will not change the 
manner in which the core is operated. Thus, the proposed changes do 
not have any effect on the probability of an accident previously 
evaluated.
    Finally, the proposed changes will delete a license condition in 
the DNPS Units 2 and 3 and QCNPS Units 1 and 2 Facility

[[Page 31323]]

Operating Licenses (FOLs) that limits the maximum average fuel rod 
burnup to 60 gigawattdays per metric ton of uranium (GWD/MTU) until 
a generic environmental assessment that supports an extended limit 
is approved.
    The proposed deletion of the license condition is justified by 
completion of generic environmental assessments for DNPS and QCNPS 
(i.e., as required by the license condition). As such, these license 
conditions are no longer required or applicable. Therefore, the 
proposed change will have no effect on any accident initiator or 
precursor previously evaluated and will not change the manner in 
which the core is operated. Thus, the proposed changes do not have 
any effect on the probability of an accident previously evaluated.
    The proposed changes to the DNPS TS 3.4.5, DNPS and QCNPS TS 
5.6.5.b, and the deletion of the Renewed FOL license conditions do 
not affect the ability to successfully respond to previously 
evaluated accidents and does not affect the radiological assumptions 
used in the evaluations for both DNPS and QCNPS.
    Thus, the proposed changes will have no effect on the type or 
amount of radiation released, and will have no effect on predicted 
offsite doses in the event of an accident.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes to DNPS TS Section 3.4.5, DNPS and QCNPS TS 
Section 5.6.5, and the proposed deletion of Renewed FOL license 
conditions do not affect the performance of any structure, system, 
or component credited with mitigating any accident previously 
evaluated.
    The deletion of the footnote from DNPS TS 3.4.5 restores TS 
requirements to the requirements prior to NRC approval of an August 
2008 emergency license amendment. The proposed deletion of the 
footnote does not affect the control parameters governing unit 
operation or the response of plant equipment to transient 
conditions. The proposed changes do not introduce any new modes of 
system operation or failure mechanisms.
    The NRC-approved analytical methodologies in TS 5.6.5.b are used 
to accurately model core behavior, including the determination of 
core operating limits, for the fuel that is currently loaded in the 
DNPS and QCNPS reactors. These methodologies do not affect the 
control parameters governing unit operation or the response of plant 
equipment to transient conditions. The proposed changes do not 
introduce any new modes of system operation or failure mechanisms.
    The existing Renewed FOL license condition limits fuel burnup 
until completion of a generic environmental assessment. In June 
2004, the NRC issued NUREG-1437, ``Generic Environmental Impact 
Statement for License Renewal of Nuclear Plants,'' Supplement 16, 
``Quad Cities Nuclear Power Station, Units 1 and 2,'' and Supplement 
17, ``Dresden Nuclear Power Station, Units 2 and 3.'' Based on the 
completion and conclusions of these generic environmental 
assessments for DNPS and QCNPS, the license condition limiting fuel 
burnup for each unit has been satisfied. As such, these license 
conditions are no longer required or applicable.
    The proposed deletion of the license condition does not affect 
the control parameters governing unit operation or the response of 
plant equipment to transient conditions. The proposed changes do not 
introduce any new modes of system operation or failure mechanisms.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The proposed changes to DNPS TS 3.4.5, DNPS and QCNPS TS 
5.6.5.b, and the DNPS and QCNPS Renewed FOLs (i.e., deletion of the 
fuel burnup license condition) will not affect the ability to 
quantify unidentified RCS leakage, accurately model core behavior 
for the currently loaded fuel, and ensure compliance with NRC-
approved LTRs.
    As such, the proposed changes do not modify the safety limits or 
setpoints at which protective actions are initiated and do not 
change the requirements governing operation or availability of 
safety equipment assumed to operate to preserve the margin of 
safety. Therefore, the proposed changes provide an equivalent level 
of protection as that currently provided.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: Russell A. Gibbs.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Miami-Dade County, Florida

    Date of amendment request: April 13, 2009.
    Description of amendment request: The amendment would delete those 
portions of Technical Specifications superseded by 10 CFR Part 26, 
Subpart I. This change is consistent with NRC approved Revision 0 to 
Technical Specification Task Force (TSTF) ``Improved Standard Technical 
Specification Change Traveler, TSTF-511, Eliminate Working Hour 
Restrictions from TS 5.2.2 to support Compliance with 10 CFR Part 26.''
    The NRC staff issued a notice of availability of the model safety 
evaluation and model no significant hazards consideration (NSHC), using 
the consolidated line-item improvement process for referencing in 
license amendment applications in the Federal Register on December 30, 
2008 (73 FR 79923). The licensee affirmed the applicability of the 
following NSHC determination in its application dated April 13, 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:

Criterion 1: The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change removes Technical Specification restrictions 
on working hours for personnel who perform safety related functions. 
The Technical Specification restrictions are superseded by the 
worker fatigue requirements in 10 CFR Part 26.
    Removal of the Technical Specification requirements will be 
performed concurrently with the implementation of the 10 CFR Part 
26, Subpart I, requirements. The proposed change does not impact the 
physical configuration or function of plant structures, systems, or 
components (SSCs) or the manner in which SSCs are operated, 
maintained, modified, tested, or inspected. Worker fatigue is not an 
initiator of any accident previously evaluated. Worker fatigue is 
not an assumption in the consequence mitigation of any accident 
previously evaluated.
    Therefore, it is concluded that this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.

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 removes Technical Specification restrictions 
on working hours for personnel who perform safety related functions. 
The Technical Specification restrictions are superseded by the 
worker fatigue requirements in 10 CFR Part 26. Working hours will 
continue to be controlled in accordance with NRC requirements. The 
new rule allows for deviations from controls to mitigate or prevent 
a condition adverse to safety or as necessary to maintain the 
security of the facility. This ensures that the new rule will not 
unnecessarily restrict working hours and thereby create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed change does not alter the plant configuration, 
require new plant equipment to be installed, alter accident analysis 
assumptions, add any initiators, or

[[Page 31324]]

effect the function of plant systems or the manner in which systems 
are operated, maintained, modified, tested, or inspected. Therefore, 
the proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.

Criterion 3: The Proposed Change Does Not Involve a Significant 
Reduction in a Margin of Safety

    The proposed change removes Technical Specification restrictions 
on working hours for personnel who perform safety related functions. 
The Technical Specification restrictions are superseded by the 
worker fatigue requirements in 10 CFR Part 26. The proposed change 
does not involve any physical changes to the plant or alter the 
manner in which plant systems are operated, maintained, modified, 
tested, or inspected. The proposed change does not alter the manner 
in which safety limits, limiting safety system settings or limiting 
conditions for operation are determined. The safety analysis 
acceptance criteria are not affected by this change. The proposed 
change will not result in plant operation in a configuration outside 
the design basis. The proposed change does not adversely affect 
systems that respond to safely shut down the plant and to maintain 
the plant in a safe shutdown condition. Removal of plant-specific 
Technical Specification administrative requirements will not reduce 
a margin of safety because the requirements in 10 CFR Part 26 are 
adequate to ensure that worker fatigue is managed.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Branch Chief: Thomas H. Boyce.

FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy 
Center (DAEC), Linn County, Iowa

    Date of amendment requests: March 4, 2009.
    Description of amendment requests: The proposed amendments would 
change the Technical Specification (TS) Section 5.5.12 (Primary 
Containment Leakage Rate Testing Program) and change TS Section 3.6.1.3 
(Primary Containment Isolation Valves) to remove the repair criterion 
for Main Steamline Isolation Valves (MSIVs) that fail their as-found 
leakage rate acceptance criterion found in current Surveillance 
Requirement 3.6.1.3.9.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This proposed change to TS 5.5.12 does not modify existing 
structures, systems or components (SSCs) of the plant, and it does 
not introduce new SSCs. It does not change assumptions, methodology, 
likelihood, or results of previously evaluated accidents in the 
Updated Final Safety Analysis Report [UFSAR]. It does not change 
operating procedures or administrative controls that affect the 
functions of SSCs. By excluding Main Steam pathway leakage from Type 
A, and Type B and C test results, this change will make the Primary 
Containment Leakage Rate Testing Program more closely aligned with 
the assumptions used in associated accident dose consequence 
analyses.
    The proposed change [to TS 3.6.1.3] to eliminate the repair 
criterion (i.e., as-left leakage limit) for MSIVs that fail their 
as-found leak test, does not change how the MSIVs function in 
response to any event, nor the likelihood of occurrence of any 
accident previously identified in the UFSAR. Repairing the MSIVs to 
an as-left leakage value, which can be higher than the currently 
specified value in TS that reliably assures the next as-found 
leakage test will be within limits is sufficient to ensure that the 
calculated dose consequences of any event involving MSIV leakage as 
an effluent pathway remain within analyzed limits.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    No new or different accidents result from utilizing the proposed 
changes. The changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. The changes do not alter assumptions made in the safety 
analysis for MSIV performance.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Since Main Steam pathway leakage bypasses the containment and 
its filtration system (Standby Gas Treatment System) during a Loss-
of-Coolant Accident (LOCA), the effect on release to the environment 
is analyzed and specifically accounted for in the DAEC dose analysis 
methodology approved by Amendments 237 and 241. This proposed change 
to exclude Main Steam pathway leakage from Type A, and Type B and C 
test results does not change dose analysis values, and thus does not 
affect actual margin in the dose analysis.
    Similarly, removing the as-left repair criterion for MSIVs from 
the TS has no impact on the assumptions for MSIV leakage used in the 
accident analysis, which are based upon the as-found MSIV leakage 
limit, not the as-left leakage. As long as the as-left leakage value 
gives high confidence that the as-found leakage will remain within 
limits over the next operating cycle until the next as-found leak 
test is conducted, the assumptions of the dose consequence analyses 
are not adversely impacted and the previously calculated results 
remain bounding.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. R. E. Helfrich, Florida Power & Light 
Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Lois M. James.

FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of amendment request: April 17, 2009.
    Description of amendment request: The proposed amendment would 
revise Operating License No. DPR-49 by changing ``FPL Energy Duane 
Arnold, LLC'' to ``NextEra Energy Duane Arnold, LLC,'' where 
appropriate, to reflect the renaming of FPL Energy Duane Arnold, LLC to 
NextEra Energy Duane Arnold, LLC.
    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 request is for administrative changes only. No actual 
facility equipment or accident analyses will be affected by the 
proposed changes. Therefore, this request will have no impact on 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.

[[Page 31325]]

    This request is for administrative changes only. No actual 
facility equipment or accident analyses will be affected by the 
proposed changes and no failure modes not bounded by previously 
evaluated accidents will be created. Therefore, this request will 
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.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel cladding, Reactor Coolant 
System pressure boundary, and containment structure) to limit the 
level of radiation dose to the public. This request is for 
administrative changes only. No actual plant equipment or accident 
analyses will be affected by the proposed changes. Additionally, the 
proposed changes will not relax any criteria used to establish 
safety limits, will not relax any safety system settings, and will 
not relax the bases for any limiting conditions of operation. 
Therefore, these proposed changes 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 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. R. E. Helfrich, Florida Power & Light 
Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Lois M. James.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: June 2, 2009.
    Description of amendment request: The proposed amendment would (1) 
delete Technical Specification (TS) surveillance requirement (SR) 
3.1.3.2 and revise SR 3.1.3.3, (2) remove reference to SR 3.1.3.2 from 
Required Action A.3 of TS 3.1.3, ``Control Rod OPERABILITY,'' and (3) 
revise Example 1.4-3 in TS Section 1.4, ``Frequency,'' to clarify the 
applicability of the 1.25 surveillance test interval extension. The 
changes are in accordance with U.S. Nuclear Regulatory Commission 
(NRC)-approved TS Task Force (TSTF) traveler TSTF-475, Revision 1, 
``Control Rod Notch Testing Frequency and SRM [Source Range Monitor] 
Insert Control Rod Action.''
    The NRC issued a ``Notice of Availability of Model Application 
Concerning Technical Specification Improvement To Revise Control Rod 
Notch Surveillance Frequency, Clarify SRM Insert Control Rod Action, 
and Clarify Frequency Example'' in the Federal Register on November 13, 
2007 (72 FR 63935). In its application dated June 2, 2009, the licensee 
affirmed the applicability of the model no significant hazards 
consideration (NSHC).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC adopted by the licensee is presented below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change generically implements TSTF-475, Revision 1, 
``Control Rod Notch Testing Frequency and SRM Insert Control Rod 
Action.'' TSTF-475, Revision 1 modifies NUREG-1433 (BWR/4) and 
NUREG-1434 (BWR/6) STS. The changes: (1) Revise TS testing frequency 
for surveillance requirement (SR) 3.1.3.2 in TS 3.1.3, ``Control Rod 
OPERABILITY'', (2) clarify the requirement to fully insert all 
insertable control rods for the limiting condition for operation 
(LCO) in TS 3.3.1.2, Required Action E.2, ``Source Range Monitoring 
Instrumentation'' (NUREG-1434 only), and (3) revise Example 1.4-3 in 
Section 1.4 ``Frequency'' to clarify the applicability of the 1.25 
surveillance test interval extension. The consequences of an 
accident after adopting TSTF-475, Revision 1 are no different than 
the consequences of an accident prior to adoption. Therefore, this 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.

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

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. The 
proposed change will not introduce new failure modes or effects and 
will not, in the absence of other unrelated failures, lead to an 
accident whose consequences exceed the consequences of accidents 
previously analyzed. Thus, this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

    TSTF-475, Revision 1 will: (1) [Revise the TS SR 3.1.3.2 
frequency in TS 3.1.3, ``Control Rod OPERABILITY'', (2) clarify the 
requirement to fully insert all insertable control rods for the 
limiting condition for operation (LCO) in TS 3.3.1.2, ``Source Range 
Monitoring Instrumentation,'' and (3)] revise Example 1.4-3 in 
Section 1.4 ``Frequency'' to clarify the applicability of the 1.25 
surveillance test interval extension. [The GE Nuclear Energy Report, 
``CRD Notching Surveillance Testing for Limerick Generating 
Station,'' dated November 2006, concludes that extending the control 
rod notch test interval from weekly to monthly is not expected to 
impact the reliability of the scram system and that the analysis 
supports the decision to change the surveillance frequency.] 
Therefore, the proposed changes in TSTF-475, Revision 1 are 
acceptable and do not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the analysis adopted by the licensee 
and, based upon this review, it appears that the standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the request for amendment involves NSHC.
    Attorney for licensee: Mr. John C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Branch Chief: Michael T. Markley.

Southern Nuclear Operating Company, Inc. (SNC), Docket Nos. 50-321, 50-
366, 50-348, 50-364, 50-424, 50-425, Joseph M. Farley Nuclear Plant, 
Unit Nos. 1 and 2 (FNP), Houston County, Alabama, Edwin I. Hatch 
Nuclear Plant, Unit Nos. 1 and 2 (HNP), Appling County, Georgia, Vogtle 
Electric Generating Plant, Units Nos. 1 and 2 (VEGP), Burke County, 
Georgia

    Date of amendment request: May 19, 2009.
    Description of amendment request: The proposed amendment would 
delete those portions of technical specifications (TS) superseded by 
Title 10 of the Code of Federal Regulations (10 CFR) Part 26, Subpart 
I. This change is consistent with the Nuclear Regulatory Commission 
(NRC)-approved Revision 0 to Technical Specification Task Force (TSTF) 
Traveler, TSTF-511, ``Eliminate Working Hour Restrictions from TS 5.2.2 
to Support Compliance with 10 CFR Part 26.'' The availability of this 
TS improvement was announced in the Federal Register on December 30, 
2008, (73 FR 79923) as part of the consolidated line item improvement 
process.
    Basis for proposed no significant hazards consideration 
determination: SNC has reviewed the no significant hazards 
determination published on December 30, 2008 (73 FR 79925), as part of 
the CLIIP Notice of Availability. SNC has concluded that the 
determination presented in the notice is applicable to FNP, HNP, and 
VEGP. SNC has evaluated the proposed changes to the TS using the 
criteria in 10 CFR 50.92 and has determined that the proposed changes 
do not involve a significant hazards consideration. An analysis of the 
issue of no significant hazards consideration is presented below:

[[Page 31326]]

Criterion 1: The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change removes Technical Specification restrictions 
on working hours for personnel who perform safety related functions. 
The Technical Specification restrictions are superseded by the 
worker fatigue requirements in 10 CFR Part 26. Removal of the 
Technical Specification requirements will be performed concurrently 
with the implementation of the 10 CFR Part 26, Subpart I, 
requirements. The proposed change does not impact the physical 
configuration or function of plant structures, systems, or 
components (SSCs) or the manner in which SSCs are operated, 
maintained, modified, tested, or inspected. Worker fatigue is not an 
initiator of any accident previously evaluated. Worker fatigue is 
not an assumption in the consequence mitigation of any accident 
previously evaluated.
    Therefore, it is concluded that this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.

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 removes Technical Specification restrictions 
on working hours for personnel who perform safety related functions. 
The Technical Specification restrictions are superseded by the 
worker fatigue requirements in 10 CFR Part 26. Working hours will 
continue to be controlled in accordance with NRC requirements. The 
new rule allows for deviations from controls to mitigate or prevent 
a condition adverse to safety or as necessary to maintain the 
security of the facility. This ensures that the new rule will not 
unnecessarily restrict working hours and thereby create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed change does not alter the plant configuration, 
require new plant equipment to be installed, alter accident analysis 
assumptions, add any initiators, or affect the function of plant 
systems or the manner in which systems are operated, maintained, 
modified, tested, or inspected.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.

Criterion 3: The Proposed Change Does Not Involve a Significant 
Reduction in a Margin of Safety

    The proposed change removes Technical Specification restrictions 
on working hours for personnel who perform safety related functions. 
The Technical Specification restrictions are superseded by the 
worker fatigue requirements in 10 CFR Part 26. The proposed change 
does not involve any physical changes to the plant or alter the 
manner in which plant systems are operated, maintained, modified, 
tested, or inspected. The proposed change does not alter the manner 
in which safety limits, limiting safety system settings or limiting 
conditions for operati