Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 13786-13798 [2010-6052]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 75, Number 55 (Tuesday, March 23, 2010)]
[Notices]
[Pages 13786-13798]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2010-6052]


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NUCLEAR REGULATORY COMMISSION

[NRC-2010-0106]


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses Involving No Significant Hazards Considerations

I. Background

    Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC) is publishing this regular biweekly notice. The Act 
requires the Commission publish notice of any amendments issued, or 
proposed to be issued and grants the Commission the authority to issue 
and make immediately effective any amendment to an operating license 
upon a determination by the Commission that such amendment involves no 
significant hazards consideration, notwithstanding the pendency before 
the Commission of a request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from February 25, 2010, to March 10, 2010. The 
last biweekly notice was published on March 9, 2010 (75 FR 10823).

Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in Title 10 of the Code of Federal 
Regulations (10 CFR), Section 50.92, this means that operation of the 
facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rulemaking 
and Directives Branch (RDB), TWB-05-B01M, Division of Administrative 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and should cite the publication date and 
page number of this Federal Register notice. Written comments may also 
be faxed to the RDB at 301-492-3446. Documents may be examined, and/or 
copied for a fee, at the NRC's Public Document Room (PDR), located at 
One White Flint North, Public File Area O1F21, 11555 Rockville Pike 
(first floor), Rockville, Maryland.
    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license. 
Requests for a hearing and a petition for leave to intervene shall be 
filed in accordance with the Commission's ``Rules of Practice for 
Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested person(s) 
should consult a current copy of 10 CFR 2.309, which is available at 
the Commission's PDR, located at One White Flint North, Public File 
Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed by the above date, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention

[[Page 13787]]

at the hearing. The requestor/petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the requestor/petitioner intends to rely to 
establish those facts or expert opinion. The petition must include 
sufficient information to show that a genuine dispute exists with the 
applicant on a material issue of law or fact. Contentions shall be 
limited to matters within the scope of the amendment under 
consideration. The contention must be one which, if proven, would 
entitle the requestor/petitioner to relief. A requestor/petitioner who 
fails to satisfy these requirements with respect to at least one 
contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held. If 
the final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards consideration, any hearing held 
would take place before the issuance of any amendment.
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC E-Filing rule (72 FR 49139, 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the Internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
ten (10) days prior to the filing deadline, the participant should 
contact the Office of the Secretary by e-mail at 
hearing.docket@nrc.gov, or by telephone at (301) 415-1677, to request 
(1) a digital ID certificate, which allows the participant (or its 
counsel or representative) to digitally sign documents and access the 
E-Submittal server for any proceeding in which it is participating; and 
(2) advise the Secretary that the participant will be submitting a 
request or petition for hearing (even in instances in which the 
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the 
Secretary will establish an electronic docket for the hearing in this 
proceeding if the Secretary has not already established an electronic 
docket.
    Information about applying for a digital ID certificate is 
available on NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing 
the E-Submittal server are detailed in NRC's ``Guidance for Electronic 
Submission,'' which is available on the agency's public Web site at 
https://www.nrc.gov/site-help/e-submittals.html. Participants may 
attempt to use other software not listed on the Web site, but should 
note that the NRC's E-Filing system does not support unlisted software, 
and the NRC Meta System Help Desk will not be able to offer assistance 
in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through EIE, users will be required to install a Web 
browser plug-in from the NRC Web site. Further information on the Web-
based submission form, including the installation of the Web browser 
plug-in, is available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with NRC guidance 
available on the NRC public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an e-mail notice confirming receipt of the document. The 
E-Filing system also distributes an e-mail notice that provides access 
to the document to the NRC Office of the General Counsel and any others 
who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC Web site 
at https://www.nrc.gov/site-help/e-submittals.html, by e-mail at 
MSHD.Resource@nrc.gov, or by a toll-free call at (866) 672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 
20852, Attention: Rulemaking and Adjudications Staff. Participants 
filing a document in this manner are responsible for serving the 
document on all other participants. Filing is considered complete by 
first-class mail as of the time of deposit in the mail, or by courier, 
express mail, or expedited delivery service upon depositing the 
document with the provider of the service. A presiding officer, having 
granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the

[[Page 13788]]

reason for granting the exemption from use of E-Filing no longer 
exists.
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
https://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant 
to an order of the Commission, or the presiding officer. Participants 
are requested not to include personal privacy information, such as 
social security numbers, home addresses, or home phone numbers in their 
filings, unless an NRC regulation or other law requires submission of 
such information. With respect to copyrighted works, except for limited 
excerpts that serve the purpose of the adjudicatory filings and would 
constitute a Fair Use application, participants are requested not to 
include copyrighted materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Non-timely filings 
will not be entertained absent a determination by the presiding officer 
that the petition or request should be granted or the contentions 
should be admitted, based on a balancing of the factors specified in 10 
CFR 2.309(c)(1)(i)-(viii).
    For further details with respect to this license amendment 
application, see the application for amendment which is available for 
public inspection at the Commission's PDR, located at One White Flint 
North, Public File Area O1F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the ADAMS Public Electronic Reading Room on the Internet at the NRC Web 
site, https://www.nrc.gov/reading-rm/adams.html. Persons who do not have 
access to ADAMS or who encounter problems in accessing the documents 
located in ADAMS, should contact the NRC PDR Reference staff at 1-800-
397-4209, 301-415-4737, or by e-mail to pdr.resource@nrc.gov.
Carolina Power and 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: January 27, 2010.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TS) Section 3.6.2.2.a to incorporate 
an expanded range of eductor flow rates for the containment spray 
additive system. These changes are supported by the use of a new 
chemical model and new boric acid equilibrium data, revised sump 
hydrogen-ion concentration (pH) limits, and changes to the containment 
spray additive tank concentration and volume limits. 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 provides revised requirements for an 
expanded range of eductor flow rates using a new chemical model and 
new boric acid equilibrium data, revised sump pH limits, and changes 
to CSAT concentration and volume limits. This ensures that the Spray 
Additive System remains operable within the TS requirements or 
appropriate actions be taken. The proposed changes do not affect the 
automatic shutdown capability of the reactor protection system and 
no accident analyses are impacted by the proposed changes.
    Expanding the range of acceptable values of eductor flow rate 
does not increase the probability of occurrence of any accident. 
Analyzed events are initiated by the failure of plant structures, 
systems or components. The containment spray additive system is not 
considered as an initiator of any analyzed accident. The proposed 
changes ensure that the spray additive system and the associated 
containment spray system can perform the accident mitigation 
functions required during a LOCA [loss-of-coolant accident] or MSLB 
[main steam line break] event.
    The proposed change does not have a detrimental impact on the 
integrity of any plant structure, system or component that initiates 
an analyzed event and will not alter the operation of, or otherwise 
increase the failure probability of any plant equipment that 
initiates an analyzed accident. Furthermore, this action does not 
affect the initiating frequency of a LOCA or MSLB event.
    Therefore, this amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    As described above, the proposed change provides revised 
requirements for an expanded range of eductor flow rates using a new 
chemical model and new boric acid equilibrium data, revised sump pH 
limits, and changes to CSAT concentration and volume limits. These 
proposed changes ensure that the spray additive system and the 
associated containment spray system can perform the required 
accident mitigation functions during a LOCA or MSLB event. There are 
no other types of accidents that can be postulated that would 
require the use of the spray additive system or the associated 
containment spray system for mitigation.
    The proposed changes do not introduce any new association 
between the spray additive system and any radioactive system, 
including the RCS [reactor coolant system].
    Emergency operation of the spray additive system, or postulated 
failures of the spray additive system, cannot initiate any type of 
accident. No new accident initiators are introduced by the proposed 
requirements and no new failure modes are created that would cause a 
new or different kind of accident from any accident previously 
evaluated.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The Bases of TS 3.6.2.2 state that the operability of the Spray 
Additive System ensures that sufficient NaOH [sodium hydroxide] is 
added to the containment spray in the event of a LOCA. The limits on 
NaOH volume and concentration ensure a pH value of between 7.0 and 
11.0 for the solution that is recirculated within containment after 
a LOCA. The spray additive system adds NaOH to the containment spray 
water being supplied from the refueling water storage tank (RWST) to 
adjust the pH of the containment spray and containment recirculation 
sump solutions. This pH range minimizes both the evolution of iodine 
and the effect of chloride and caustic stress corrosion on 
mechanical systems and components. The proposed range of flow rate 
from the RWST through each eductor ensures that the original margin 
of safety is maintained through acceptable pH control following a 
LOCA or MSLB event. The initial conditions of the accident analyses 
are preserved and the consequences of previously analyzed accidents 
are unaffected.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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, North Carolina 27602.
    NRC Acting Branch Chief: Douglas A. Broaddus (Acting).

[[Page 13789]]

Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, 
Oconee Nuclear Station, Units 1, 2, and 3 (Oconee 1, 2, and 3), Oconee 
County, South Carolina; Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2 (McGuire 1 and 2), Mecklenburg County, North 
Carolina; Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 
1 and 2 (Catawba 1 and 2), York County, South Carolina
    Date of amendment request: December 15, 2009.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications to replace the current limits on 
primary coolant gross specific activity with limits on primary coolant 
noble gas activity. The noble gas activity would be based on DOSE 
EQUIVALENT XE-133 and would take into account only the noble gas 
activity in the primary coolant. The changes are consistent with 
nuclear Regulatory Commission (NRC) approved Industry/Technical 
Specification Task Force (TSTF) Standard Technical Specification Change 
Traveler, TSTF-490.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of no significant hazards. The NRC staff has 
reviewed the licensee's analysis against the standards of 10 CFR 
50.92(c). The NRC staff's analysis of the no significant hazards 
consideration is presented below:
    Criterion 1: Does the proposed change 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 
licensee's current accident analyses for Catawba 1 and 2, McGuire 1 and 
2 and Oconee 1, 2, and 3. 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.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    Criterion 2: Does the proposed change 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.
    Therefore the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
calculated.
    Criterion 3: Does the proposed change involve a significant 
reduction in a 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 licensee's safety analysis and will ensure the 
monitored values protect the initial assumptions in the safety 
analysis.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Associate General 
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South 
Church Street, EC07H, Charlotte, NC 28202.
    NRC Branch Chief: Gloria Kulesa.
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: December 3, 2009.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) to incorporate Standard Technical 
Specification 3.1.8 ``Scram Discharge Volume (SDV) Vent and Drain 
Valves'' and associated Bases of NUREG-1433, Revision 3, ``Standard 
Technical Specifications General Electric Plants, BWR/4,'' modified to 
account for plant specific design details.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The operation of Vermont Yankee Nuclear Power Station (VY) in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed amendment does not impact the operability of any 
structure, system or component that affects the probability of an 
accident or that supports mitigation of an accident previously 
evaluated. The proposed amendment does not affect reactor operations 
or accident analysis and has no radiological consequences. The 
operability requirements for accident mitigation systems remain 
consistent with the licensing and design basis. Therefore, the 
proposed amendment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The operation of VY in accordance with the proposed amendment 
will 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 plant operation. Thus, this 
change does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. The operation of VY in accordance with the proposed amendment 
will not involve a significant reduction in a margin of safety.
    The proposed change ensures that the safety functions of the SDV 
vent and drain valves are fulfilled. The isolation function is 
maintained by valves in the vent and drain lines and by the required 
action to isolate the affected line. The ability to vent and drain 
the SDVs is maintained through administrative controls. In addition, 
the reactor protection system ensures that an SDV will not be filled 
to the point that it has insufficient volume to accept a full scram. 
Maintaining the safety functions related to isolation of the SDV and 
insertion of control rods ensures that the proposed change does not 
involve a significant reduction in the margin of safety. The 
proposed amendment does not change the design or function of any 
component or system. The proposed amendment does not impact any 
safety limits, safety settings or safety margins. Therefore, 
operation of VY in accordance with the proposed amendment will not 
involve a significant reduction in the margin to 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.

[[Page 13790]]

    NRC Branch Chief: Nancy Salgado.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California
    Date of amendment request: December 14, 2009.
    Description of amendment request: The proposed amendments would 
change the design basis and Final Safety Analysis Report Update (FSARU) 
to allow use of a damping value of 5 percent of critical damping for 
the structural dynamic qualification of the control rod drive mechanism 
(CRDM) pressure housings on the replacement reactor vessel head for the 
design earthquake (DE), double design earthquake (DDE), Hosgri 
earthquake (HE), and loss-of-coolant accident (LOCA) loading 
conditions.
    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?
    The proposed change revises the design basis and Final Safety 
Analysis Report Update (FSARU) to reflect a damping value of 5 
percent of critical damping for the structural dynamic qualification 
of the control rod drive mechanism (CRDM) pressure housings for the 
replacement reactor vessel head for the design earthquake (DE), 
double design earthquake (DDE), Hosgri earthquake (HE), and loss of 
coolant accident (LOCA). The 5 percent damping value has been 
accepted by the NRC staff at several other plants with equivalent 
CRDMs and seismic support structures.
    The damping value is an element of the structural dynamic 
analysis performed to confirm the CRDMs' ability to function under a 
postulated seismic disturbance or LOCA while maintaining resulting 
stresses under ASME Code [American Society of Mechanical Engineers 
Boiler and Pressure Vessel Code] Section III allowable values. 
Because the ASME Code requirements continue to be met, this proposed 
change to the damping value could not result in an increase in the 
probability or consequences of an accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change revises the design basis and FSARU to 
reflect a damping value of 5 percent of critical damping for the 
structural dynamic qualification of the CRDM pressure housings for 
the replacement reactor vessel head for the DE, DDE, HE, and LOCA. 
The 5 percent damping value has been accepted by the NRC staff at 
several other plants with equivalent CRDMs and seismic support 
structures and is a conservative value based on the testing 
performed by the OEM [original equipment manufacturer].
    The damping value is an element of the structural dynamic 
analysis performed to confirm the CRDMs' ability to function under a 
postulated seismic disturbance or LOCA while maintaining resulting 
stresses under ASME Code Section III allowable values. Because the 
ASME Code requirements continue to be met, this proposed change to 
the damping value could not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    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 change involve a significant reduction in a margin 
of safety?
    The proposed change revises the design basis and FSARU to 
reflect a damping value of 5 percent of critical damping for the 
structural dynamic qualification of the CRDM pressure housings for 
the replacement reactor vessel head for the DE, DDE, HE, and LOCA. 
The 5 percent damping value for CRDMs has been accepted by the NRC 
staff at several other plants with equivalent CRDMs and seismic 
support structures.
    The damping value is an element of a structural dynamic analysis 
performed to confirm the CRDMs' ability to function under a 
postulated seismic disturbance or LOCA while maintaining resulting 
stresses under ASME Code, Section III, allowable values. The margin 
of safety is maintained by meeting the ASME Code requirements.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Jennifer Post, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Branch Chief: Michael T. Markley.
Pacific Gas and Electric Company (PG&E), Docket Nos. 50-275 and 50-323, 
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo 
County, California
    Date of amendment request: December 29, 2009.
    Description of amendment request: The proposed amendments would 
revise the licensing basis as described in the Final Safety Analysis 
Report Update (FSARU) to discuss the conformance of the delayed access 
offsite power circuit (the 500-kV delayed access circuit) to the 
General Design Criterion 17 requirement that each of the offsite power 
circuits be designed to be available in sufficient time following a 
loss of all onsite alternating current power supplies and the other 
offsite electric power circuit, to assure that specified acceptable 
fuel design limits and design conditions of the reactor coolant 
pressure boundary are not exceeded. The proposed amendment will also 
add information related to reactor coolant pump seal performance during 
and after (1) a loss of seal injection (with continued thermal barrier 
cooling); (2) a loss of thermal barrier cooling (with continued seal 
injection); and (3) a loss of all seal cooling (both thermal barrier 
cooling and seal injection).
    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 amendments would revise the licensing basis as 
described in the Final Safety Analysis Report Update (FSARU) to 
discuss the conformance of the delayed access offsite alternating 
current (ac) power circuit (the 500-kV delayed access circuit) to 
the General Design Criterion (GDC) 17 requirement that ``each of the 
offsite power circuits be designed to be available in sufficient 
time following a loss of all onsite alternating current power 
supplies and the other offsite electric power circuit, to assure 
that specified acceptable fuel design limits and design conditions 
of the reactor coolant pressure boundary are not exceeded.'' It 
would also add information related to reactor coolant pump (RCP) 
seal performance during and after (1) a loss of seal injection (with 
continued thermal barrier cooling); (2) a loss of thermal barrier 
cooling (with continued seal injection); and (3) a loss of all seal 
cooling (both thermal barrier cooling and seal injection).
    PG&E Calculation STA-274 demonstrates that specified acceptable 
fuel design limits and design conditions of the reactor coolant 
pressure boundary are not exceeded following a loss of the 230-kV 
immediate access offsite power circuit and all onsite emergency ac 
power supplies until the 500-kV delayed access circuit can be 
aligned for backfeed. Alignment of the 500-kV delayed offsite 
circuit to backfeed, implementing RCP seal coping strategy actions 
to limit maximum RCP seal leakage to 21 gpm [gallons per minute] per 
pump, and restoring reactor coolant system (RCS) makeup flow to 
stabilize the plant can be completed within approximately 54 minutes 
to assure that specified acceptable fuel design limits and

[[Page 13791]]

design conditions of the reactor coolant pressure boundary are not 
exceeded.
    The proposed changes will not add any accident initiators, or 
adversely affect how the plant safety-related structures, systems, 
or components (SSCs) are operated, maintained, modified, tested, or 
inspected. There is no increase in the probability of a GDC 17 loss 
of all ac event occurring, and since the same applicable GDC 17 
acceptance criteria continue to be met with the increased RCP seal 
leakage, there is no change in the consequences associated with this 
event.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different accident from any accident previously evaluated?
    Response: No.
    The RCP Seal coping strategy implemented in response to 
Westinghouse Technical Bulletin TB-04-22, Revision 1, ensures that 
RCP seal integrity is maintained following a loss of all seal 
cooling associated with the GDC 17 loss of all ac event. PG&E 
Calculation STA-274 demonstrates that the GDC 17 requirements for a 
delayed offsite ac power source are met for up to a one-hour time 
period for the operators to complete the necessary actions 
associated with establishing the 500-kV backfeed, implementing the 
RCP seal coping strategy to limit maximum RCS seal leakage to 21 gpm 
per pump, and restoring RCS makeup flow. This proposed change 
provides assurance that specified acceptable fuel design limits and 
design conditions of the reactor coolant pressure boundary are not 
exceeded. The proposed change does not introduce new equipment that 
could create a new or different kind of accident, and no new 
equipment failure modes are created. As a result, no new accident 
scenarios, failure mechanisms, or limiting single failures are 
introduced as a result of this proposed amendment.
    Therefore, the proposed changes do not create the possibility of 
a new or different accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The implementation of the RCP seal coping strategy ensures that 
RCP seal leakage is limited to 21 gpm per pump following a loss of 
all seal cooling such that there is no impact or reduction in the 
margin of safety associated with the GDC 17 loss of all ac event. 
The analysis associated with the change supports the ability to 
align the 500-kV delayed access circuit, implement the RCP seal 
coping strategy actions, and restore RCS makeup flow in sufficient 
time following a loss of all onsite ac power supplies and the other 
offsite electric power circuit, to assure that specified acceptable 
fuel design limits and design conditions of the reactor coolant 
pressure boundary are not exceeded. The proposed amendment would not 
alter the way any safety-related SSC functions and would not alter 
the way the plant is operated. The amendment demonstrates that the 
500-kV backfeed, isolation of RCP seal cooling, and restoration of 
RCS makeup flow can be reliably completed within 54 minutes, and 
that there is considerable margin to the GDC 17 acceptance criteria 
for the 500-kV backfeed as a delayed offsite ac power source. The 
proposed amendment would not introduce any new uncertainties or 
change any existing uncertainties associated with any safety limit. 
Since the proposed amendment would have no impact on the structural 
integrity of the fuel cladding or reactor coolant pressure boundary, 
and maintains the RCP seal leakage within controllable limits, there 
is no impact on the containment structure. Based on the above 
considerations, the proposed amendment would not degrade the ability 
to safely shut down the plant in the event of a loss of all ac 
power.
    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: Jennifer Post, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Branch Chief: Michael T. Markley.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
    Date of amendment request: January 26, 2010 (TS 09-05).
    Description of amendment request: The proposed amendments would 
revise the Technical Specification (TS) Table 3.3-1, ``Reactor Trip 
System Instrumentation,'' Functional Unit 5, ``Intermediate Range, 
Neutron Flux,'' to resolve an oversight regarding the operability 
requirements for the intermediate range neutron flux channels. The 
amendments would add an action to TS Table 3.3-1 to define that the 
provisions of Specification 3.0.3 are not applicable above 10 percent 
of thermal rated power with the number of operable intermediate range 
neutron flux channels two less than the minimum channels operable 
requirement.
    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 intermediate range neutron flux trip must be operable in 
Mode 1 below the P-10 setpoint and in Mode 2 when there is a 
potential for an uncontrolled rod withdrawal accident during reactor 
startup. Above the P-10 setpoint, the power range neutron flux high 
setpoint trip and the power range neutron flux high positive rate 
trip provide core protection for a rod withdrawal accident. The 
intermediate range channels have no protection function above the P-
10 setpoint. The proposed change does not affect the design of 
structures, systems, or components (SSCs) credited in accident or 
transient analyses, the operational characteristics or function of 
SSCs, the interfaces between credited SSCs and other plant systems, 
or the reliability of SSCs. The proposed change does not impact the 
initiating frequency of any UFSAR accident or transient previously 
evaluated. In addition, the proposed change does not impact the 
capability of credited SSCs to perform their required safety 
functions. Thus, eliminating the requirement to apply Specification 
3.0.3 provisions when two intermediate range channels are inoperable 
in Mode 1 with the thermal power above the P-10 setpoint 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 intermediate range neutron flux trip must be operable in 
Mode 1 below the P-10 setpoint and in Mode 2 when there is a 
potential for an uncontrolled rod withdrawal accident during reactor 
startup. Above the P-10 setpoint, the power range neutron flux high 
setpoint trip and the power range neutron flux high positive rate 
trip provide core protection for a rod withdrawal accident. The 
intermediate range channels have no protection function above the P-
10 setpoint. The proposed change does not involve a change in 
design, configuration, or method of operation of the plant. The 
proposed change does not alter the manner in which equipment 
operation is initiated, nor will the functional demands on credited 
equipment be changed. The capability of credited SSCs to perform 
their required function will not be affected by the proposed change. 
In addition, the proposed change does not affect the interaction of 
plant SSCs with other plant SSCs whose failure or malfunction can 
initiate an accident or transient. As such, no new failure modes are 
being introduced. Thus, eliminating the requirement to apply 
Specification 3.0.3 provisions when two intermediate range channels 
are inoperable in Mode 1 with the thermal power above the P-10 
setpoint does not create the possibility of a new or different kind 
of accident.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change resolves an oversight regarding the 
operability requirements for the

[[Page 13792]]

intermediate range neutron flux channels. Currently, Specification 
3.0.3 provisions apply when two intermediate range neutron flux 
channels are declared inoperable in Mode 1 when thermal power is 
above the P-10 setpoint. Above the P-10 setpoint, the power range 
neutron flux trip and the power range neutron flux high positive 
rate trip provide core protection for a rod withdrawal accident. The 
intermediate range channels have no protection function above the P-
10 setpoint. The proposed change does not change the conditions, 
operating configurations, or minimum amount of operating equipment 
assumed in the safety analyses for accident or transient mitigation. 
The proposed change does not alter the plant design, including 
instrument setpoints, nor does it alter the assumptions contained in 
the safety analyses. 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 impact the redundancy or availability of SSCs required to 
accident or transient mitigation, or the ability of the plant to 
cope with design basis events. In addition, no changes are proposed 
in the manner in which the credited SSCs provide plant protection or 
which create new modes of plant operation. Thus, eliminating the 
requirement to apply Specification 3.0.3 provisions when two 
intermediate range channels are inoperable in Mode 1 with thermal 
power above the P-10 setpoint 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Douglas A. Broaddus (Acting).
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas
    Date of amendment request: January 28, 2010.
    Description of amendment request: The proposed amendment will 
revise the Limiting Condition for Operation (LCO) of Technical 
Specification (TS) 3.6.3, ``Containment Isolation Valves,'' for Wolf 
Creek Generating Station. A note will be added to LCO 3.6.3 to allow 
the reactor coolant pump (RCP) seal injection valves to be considered 
OPERABLE with the valves open and power removed.
    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 change affects the RCP seal cooling and the containment 
isolation system. The change allows the removal of power to the four 
RCP seal injection valves such that they will not close in response 
to a spurious signal. A spurious closure of one or more of the seal 
injection valves could lead to a loss of coolant from the RCP seal. 
Allowance for removal of power to the valve reduces the probability 
of this event. The RCP seal performance depends on the design, flow 
rates, pressures and temperatures. There are no changes to the RCP 
seal design, nor to the seal cooling flow rates, pressures or 
temperatures.
    Therefore, the consequences of a loss of coolant from the RCP 
seal are not impacted.
    The seal injection valves are containment isolation valves. The 
system design for RCP seal cooling does not require automatic 
closure of the seal injection valves or closure of the valve within 
a specified time frame. The design of the system is such that the 
cooling water pressure passing through these valves is higher than 
the operating pressure of the reactor coolant system. The cooling 
water is needed to prevent a loss of coolant from the pump seals and 
the cooling water is assured because it is provided by the safety 
related charging pumps. In addition, a check valve is installed 
inside the containment on each seal injection line to provide a 
second containment isolation valve on the line. The seal injection 
valves fail as-is upon loss of electrical power and are not designed 
to change position following an accident. The seal injection valves 
are remote manual valves that can be operated from the control room 
based on plant procedures. These valves are not modeled as 
containment isolation valves in any accident analysis. A failure in 
the open position has no consequence due to the normal inflow of the 
seal injection water.
    Therefore, this change will not 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 previously evaluated?
    Response: No.
    The proposed amendment does not change the method by which any 
safety related plant system, subsystem, or component performs its 
specified safety function. The proposed changes will not affect the 
normal method of plant operation or change any operating parameters. 
No equipment performance requirements will be affected. Plant 
procedures will still provide for the appropriate closure of the 
seal injection valves when restoring seal injection. The proposed 
changes will not alter any assumptions made in the safety analyses 
regarding limits on RCP seal injection flow.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures will be introduced as a 
result of this amendment. There will be no adverse effect or 
challenges imposed on any safety related system as a result of this 
amendment. The proposed amendment will not alter the design or 
performance of the 7300 Process Protection System, Nuclear 
Instrumentation System, or Solid State Protection System used in the 
plant protection systems.
    Therefore, this change will 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 affect the acceptance criteria for 
any analyzed event. There will be no effect on the manner in which 
safety limits or limiting safety system settings are determined nor 
will there be any effect on those plant systems necessary to assure 
the accomplishment of protection function. Removing power from the 
RCP seal injection valves during normal operation does not impact 
the assumed ECCS [emergency core cooling system] flow that would be 
available for injection into the RCS following an accident.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Jay Silberg, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: Michael T. Markley.

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.

[[Page 13793]]

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, 
Davis-Besse Nuclear Power Station (DBNPS), Unit No. 1, Ottawa County, 
Ohio
    Date of amendment request: September 28, 2009.
    Brief description of amendment request: The proposed amendment 
would support application of optimized weld overlays or full structural 
weld overlays. Applying these weld overlays on the reactor coolant pump 
suction and discharge nozzle dissimilar metal welds requires an update 
to the DBNPS leak-before-break evaluation.
    Date of publication of individual notice in Federal Register: 
February 22, 2010 (75 FR 7628)
    Expiration date of individual notice: March 24, 2010 (Public 
comments) and April 22, 2010 (Hearing requests).
FPL Energy, Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin
    Date of amendment request: April 17, 2009, as supplemented by 
letter dated January 19, 2010.
    Description of amendment request: On July 14, 2009, the Nuclear 
Regulatory Commission published a Notice of Consideration of Issuance, 
Proposed No Significant Hazards Consideration Determination, and 
Opportunity for Hearing in the Federal Register (74 FR 34048) for a 
proposed amendment that would change the legal name of the licensee and 
owner from ``FPL Energy Point Beach, LLC'' to ``NextEra Energy Point 
Beach, LLC.''
    On January 19, 2010, the licensee submitted a supplement which 
expanded the original scope of work. The proposed revisions would 
correct an administrative error within a License Condition contained in 
Appendix C of the Renewed Facility Operating Licenses. The correction 
changes ``FPLE Group Capital'' to the appropriately titled ``FPL Group 
Capital.''
    Date of publication of individual notice in Federal Register: March 
3, 2010 (75 FR 9616)
    Expiration date of individual notice: May 3, 2010, 60 days from 
publication of the individual notice.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management System (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, https://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr.resource@nrc.gov.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos. 
1, 2, and 3, Maricopa County, Arizona
    Date of application for amendment: February 19, 2009, as 
supplemented by letters dated December 22, 2009, and February 23, 2010.
    Brief description of amendment: The amendments revised the 
Technical Specifications (TSs) to relocate the reactor coolant system 
pressure and temperature (P/T) limits and the low temperature 
overpressure protection (LTOP) enable temperatures to a licensee-
controlled document outside of the TSs. The P/T limits and LTOP enable 
temperatures will be specified in a Pressure and Temperature Limits 
Report (PTLR) that will be located in the PVNGS Technical Requirements 
Manual and administratively controlled by a new TS 5.6.9. The proposed 
changes are in accordance with the guidance in NRC Generic Letter 96-
03, ``Relocation of the Pressure Temperature Limit Curves and Low 
Temperature Overpressure Protection System Limits,'' dated January 31, 
1996.
    Date of issuance: February 25, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 150 days from the date of issuance.
    Amendment No.: Unit 1-178; Unit 2-178; Unit 3-178.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendment revised the Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: May 19, 2009 (74 FR 
23442). The supplemental letters dated December 22, 2009, and February 
23, 2010, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 25, 2010.
    No significant hazards consideration comments received: No.
Carolina Power and Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina
    Date of application for amendments: August 18, 2009, as 
supplemented on December 7, 2009.
    Brief Description of amendments: The proposed license amendments 
revised Technical Specification 3.3.1.1, ``Reactor Protection System 
(RPS) Instrumentation,'' Surveillance Requirement 3.3.1.1.8, to 
increase the frequency interval between local power range monitor 
calibrations from 1100 megawatt-days per metric ton average core 
exposure (i.e., equivalent to approximately 907 effective full-power 
hours (EFPH)) to 2000 EFPH.
    Date of issuance: February 24, 2010.

[[Page 13794]]

    Effective date: Date of issuance, to be implemented prior to start-
up from the 2010 refueling outage (RFO) for Unit 1, and prior to start-
up from the 2011 RFO for Unit 2.
    Amendment Nos.: 254 and 282.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 1, 2009 (74 FR 
62833). The supplement letter dated December 7, 2009, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 24, 2010.
    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 amendment request: October 27, 2009.
    Description of amendment request: This amendment request would 
change the Technical Specifications to provide revised values for the 
Safety Limit Minimum Critical Power Ratio for both single and dual 
recirculation loop operation.
    Date of Issuance: March 8, 2010.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 243.
    Facility Operating License No. DPR-28: Amendment revised the 
License and Technical Specifications.
    Date of initial notice in Federal Register: January 5, 2010 (75 FR 
461).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated March 8, 2010.
    No significant hazards consideration comments received: No.
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 application for amendments: February 16, 2009.
    Brief description of amendments: To remove the structural integrity 
requirements contained in TS 3/4.4.10, and its associated Bases from 
the Technical Specifications. Also relocate the reactor coolant pump 
(RCP) motor flywheel inspection requirem