[Federal Register Volume 75, Number 248 (Tuesday, December 28, 2010)]
[Notices]
[Pages 81667-81675]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2010-32668]


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NUCLEAR REGULATORY COMMISSION

[NRC-2010-0393]


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 December 2, 2010, to December 15, 2010. The 
last biweekly notice was published on December 14, 2010 (75 FR 77906).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in Title 10 of the Code of Federal 
Regulations (10 CFR), 50.92, this means that operation of the facility 
in accordance with the proposed amendment would not (1) Involve a 
significant increase in the probability or consequences of an accident 
previously evaluated; or (2) create the possibility of a new or 
different kind of accident from any accident previously evaluated; or 
(3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules, 
Announcements and Directives Branch (RADB), TWB-05-B01M, Division of 
Administrative Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be faxed to the RADB at 301-492-3446. 
Documents may be examined, and/or copied for a fee, at the NRC's Public 
Document Room (PDR), located at One White Flint North, 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 O1 F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed by the above date, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to

[[Page 81668]]

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 
<a href="/cdn-cgi/l/email-protection#741c1115061d1a135a101b171f1100341a06175a131b02"><span class="__cf_email__" data-cfemail="d6beb3b7a4bfb8b1f8b2b9b5bdb3a296b8a4b5f8b1b9a0">[email protected]</span></a>, or by telephone at 301-415-1677, to request (1) 
a digital ID certificate, which allows the participant (or its counsel 
or representative) to digitally sign documents and access the E-
Submittal server for any proceeding in which it is participating; and 
(2) advise the Secretary that the participant will be submitting a 
request or petition for hearing (even in instances in which the 
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the 
Secretary will establish an electronic docket for the hearing in this 
proceeding if the Secretary has not already established an electronic 
docket.
    Information about applying for a digital ID certificate is 
available on NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing 
the E-Submittal server are detailed in NRC's ``Guidance for Electronic 
Submission,'' which is available on the agency's public Web site at 
http://www.nrc.gov/site-help/e-submittals.html. Participants may 
attempt to use other software not listed on the Web site, but should 
note that the NRC's E-Filing system does not support unlisted software, 
and the NRC Meta System Help Desk will not be able to offer assistance 
in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through 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 http://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with NRC guidance 
available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an e-mail notice confirming receipt of the document. The 
E-Filing system also distributes an e-mail notice that provides access 
to the document to the NRC Office of the General Counsel and any others 
who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC Web site 
at http://www.nrc.gov/site-help/e-submittals.html, by e-mail at 
<a href="/cdn-cgi/l/email-protection#28657b606c067a4d5b475d5a4b4d68465a4b064f475e"><span class="__cf_email__" data-cfemail="7d302e3539532f180e12080f1e183d130f1e531a120b">[email protected]</span></a>, 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 reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant 
to an order of the Commission, or the presiding officer. Participants 
are requested not to include personal privacy information, such as 
social security numbers, home

[[Page 81669]]

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 O1-F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the ADAMS Public Electronic Reading Room on the Internet at the NRC Web 
site, http://www.nrc.gov/reading-rm/adams.html. Persons who do not have 
access to ADAMS or who encounter problems in accessing the documents 
located in ADAMS, should contact the NRC PDR Reference staff at 1-800-
397-4209, 301-415-4737, or by e-mail to <a href="/cdn-cgi/l/email-protection#93e3f7e1bde1f6e0fce6e1f0f6d3fde1f0bdf4fce5"><span class="__cf_email__" data-cfemail="037367712d7166706c76716066436d71602d646c75">[email protected]</span></a>.

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

    Date of amendment request: July 22, 2010.
    Description of amendment request: The amendments would revise an 
element of the methodology used in evaluating the radiological 
consequences of design basis steam generator tube rupture (SGTR) 
accidents. Specifically, the changes will revise the Palo Verde Nuclear 
Generating Station (PVNGS) Updated Final Safety Analysis Report 
(UFSAR), Section 15.6.6, ``Steam Generator Tube Rupture,'' to reflect a 
lower iodine spiking factor assumed for the coincident event Generated 
Iodine Spike (GIS) and the resulting reduction in the radiological 
consequences provided in UFSAR Table 15.6.3-5, ``Radiological 
Consequences for the Limiting SGTRLOPSF [Steam Generator Tube Rupture 
with Loss of Offsite Power and Single Failure] Event.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment changes an element of the methodology 
used in evaluating the radiological consequences of design basis 
SGTR accidents. This change will revise the iodine spiking factor 
used for a GIS from a value of 500 to a value of 335. The proposed 
change in the methodology element does not involve any design or 
physical changes to the facility or any component of that facility. 
The proposed change creates no new failure modes or initiating 
occurrences that could result in a design basis transient or 
accident evaluated in the Palo Verde Nuclear Generating Station 
(PVNGS) Updated Final Safety Analysis Report (UFSAR). Therefore the 
proposed change does not involve a significant increase in the 
probability of an accident previously evaluated.
    The proposed change in the methodology element does change the 
design basis analyses results for PVNGS. However, the results remain 
bounded by the previous analyzed values and remain within the 
acceptance criteria for PVNGS of 100% of the 10 CFR [Part] 100 
maximum thyroid dose limit of 300 rem [roentgen equivalent man].
    Therefore, the proposed change does not involve a significant 
increase in the consequences of an accident previously analyzed.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment changes an element of the methodology 
used in evaluating the radiological consequences of design basis 
SGTR accidents. This change will revise the iodine spiking factor 
used for a GIS from a value of 500 to a value of 335. The proposed 
change in the methodology element does not involve any design or 
physical changes to the facility or any component of that facility. 
The proposed change in the methodology element does change the 
design basis analyses results for PVNGS; however, these results 
remain bounded by the previous analyzed values and remain within the 
acceptance criteria for PVNGS of 100% of the 10 CFR [Part] 100 
maximum thyroid dose limit of 300 rem.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment changes an element of the methodology 
used in evaluating the radiological consequences of design basis 
SGTR accidents. This change will revise the iodine spiking factor 
used for a GIS from a value of 500 to a value of 335. The proposed 
change in the methodology element does not involve any design or 
physical changes to the facility or any component of that facility. 
The proposed methodology element change for a postulated SGTR, with 
a coincident loss of offsite power, GIS, and a failed open 
atmospheric dump valve (ADV), results in lower maximum dose 
consequences at the Exclusion Area Boundary (EAB) and Low Population 
Zone (LPZ) [than] previously analyzed for this event combination. 
The methodology element change results in the 2-hour maximum thyroid 
dose value of 182 rem at the EAB being reduced to 124 rem. In 
addition, the 8-hour maximum thyroid dose of 125 rem at the LPZ, 
would be reduced to 84 rem.
    Previously for PVNGS, the GIS 8-hour maximum thyroid dose was 
bounding at the LPZ and the pre-Accident Iodine Spike (PIS) 2-hour 
maximum thyroid dose was bounding at the EAB. The methodology 
element change reduces the GIS calculated dose at both the EAB and 
LPZ for SGTR events, but it does not affect the PIS dose values. 
Since the GIS calculated dose at the LPZ drops below the PIS 8-hour 
LPZ maximum thyroid dose (91 rem), the PIS 8-hour LPZ dose will 
become bounding for PVNGS. The PIS 2-hour EAB maximum thyroid dose 
(294 rem), remains the bounding dose at the EAB.
    The revised dose consequences remain bounded by the previous 
analyzed values and remain within the 10 CFR Part 100 guideline 
values which are the acceptance criteria for PVNGS Units 1, 2, and 
3. In addition, the proposed change has no effect on previously 
reported dose consequences for control room personnel following any 
postulated SGTR event.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Michael G. Green, Senior Regulatory Counsel, 
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695, 
Phoenix, Arizona 85072-2034.
    NRC Branch Chief: Michael T. Markley.

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point 
Nuclear Generating Unit No. 3, Westchester County, New York

    Date of amendment request: October 6, 2010.
    Description of amendment request: The proposed change will revise 
the note in Surveillance Requirement (SR) 3.5.4.1 in the Refueling 
Water Storage

[[Page 81670]]

Tank (RWST) Technical Specification (TS). Specifically, the proposed 
change will not require monitoring of the RWST temperature every 24 
hours when the RWST heating steam supply isolation valves are locked 
closed.
    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?
    No. The proposed change revises the existing Indian Point 
[Nuclear Generating Unit No.] 3 [(IP3)] Refueling Water Storage Tank 
(RWST) Technical Specification (TS) Surveillance Requirement (SR) 
3.5.4.1 to revise the note that eliminates the requirement to 
perform SR 3.5.4.1 when ambient air temperatures are within the 
operating limits of the RWST. The revision to the note adds a 
requirement that the steam heating supply isolation valves be locked 
closed when not performing the surveillance. The additional 
requirement does not increase the probability of an accident 
occurring since it is not an accident initiator and does not 
increase the consequences of an accident since it is providing 
additional assurance that the RWST is within the temperature limits 
assumed for accident analyses. The change increases observation of 
the RWST temperature when the steam supply isolation valves are not 
locked closed and does not otherwise affect [* * *] the performance 
capability of the structures, systems, and components relied upon to 
mitigate the consequences of postulated accidents.
    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?
    No. The proposed change revises the note that eliminates the 
requirement to perform SR 3.5.4.1 when ambient air temperatures are 
within the operating limits of the RWST. The revision adds the 
additional requirement of locking closed the steam supply isolation 
valves. The proposed change does not involve installation of new 
equipment or modification of existing equipment, so that no new 
equipment failure modes are introduced. Also, the proposed change 
does not result in a change to the way that the equipment or 
facility is operated so that no new accident initiators are created.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The proposed change revises the note that eliminates the 
requirement to perform SR 3.5.4.1 when ambient air temperatures are 
within the operating limits of the RWST. The revision adds the 
additional requirement of locking closed the steam supply isolation 
valves. The change does not reduce margin since it increases the 
temperature surveillance frequency for the RWST to provide further 
assurance that the required water temperature is maintained at all 
times.
    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., 440 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Nancy L. Salgado.

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

    Date of amendment request: November 8, 2010.
    Description of amendment request: The proposed license amendment 
request will make changes related to the final resolution of an 
unresolved issue associated with Technical Specification (TS) Amendment 
No. 181 dated February 25, 2009. This issue was resolved with the 
approval of Revision 4 of Technical Specification Task Force (TSTF) 
Change Traveler TSTF-493, ``Clarify Application of Setpoint Methodology 
for LSSS [Limiting Safety System Setting] Functions,'' which included 
the instrument function (i.e., Condensate Storage Tank (CST) Level-Low) 
that was the subject of Amendment No. 181. Specifically, the proposed 
change will add the appropriate notes as specified in TSTF-493 to the 
surveillance requirements associated with TS Table 3.3.5.1-1, 
``Emergency Core Cooling System Instrumentation,'' Function 3.d, 
Condensate Storage Tank Level--Low, and to TS Table 3.3.5.2-1, 
``Reactor Core Isolation Cooling System Instrumentation,'' Function 3, 
Condensate Storage Tank Level--Low. The supporting TS Bases will also 
be revised.
    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 adds test requirements to the CST Level-Low 
function to ensure the CST Level-low instruments will function as 
required. Surveillance tests are not an initiator of any accident 
previously evaluated. The CST components, for which the additional 
requirements were added, continue to be operable and capable of 
performing their intended function.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed 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 a physical change to the 
plant, i.e., no new or different type of equipment will be 
installed. The proposed change does not alter assumptions made in 
the safety analysis but ensures that the CST Level-low instruments 
perform as assumed in the [Updated Final Safety Analysis Report].
    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 change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change adds test requirements that will assure that 
(1) the CST Level-low instrumentation for the setpoint allowable 
value will be the limiting setting for assessing instrumentation 
channel operability and (2) will be conservatively determined so 
that the evaluation of CST instrument performance history and the 
requirements of the calibration procedures will not have an adverse 
effect on equipment operability. The testing methods and acceptance 
criteria for the CST Level-low instrumentation will continue to be 
met. There is no impact to the safety analysis acceptance criteria 
as described in the plant licensing basis because no change is made 
to the accident analysis assumptions.
    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: Joseph A. Aluise, Associate General 
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.

[[Page 81671]]

    NRC Branch Chief: Michael T. Markley.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: July 12, 2010.
    Description of amendment request: The proposed amendment would 
modify Item 1 of Table 2-5, ``Instrumentation Operating Requirements 
for Other Safety Feature Functions,'' of Technical Specification (TS) 
2.15, ``Instrumentation and Control Systems,'' to provide new Note (e), 
and Surveillance Requirement (SR) Items 1 and 2 of Table 3-3, ``Minimum 
Frequencies for Checks, Calibrations and Testing of Miscellaneous 
Instrumentation and Controls,'' of TS 3.1, ``Instrumentation and 
Control,'' which pertain to operability of the primary and secondary 
control element assembly (CEA) position indication system (CEAPIS) 
channels. A new SR is proposed for Item 4 of Table 3-3 of TS 3.1, which 
will verify the position of CEAs each shift. The proposed amendment 
will ensure that CEA alignment is maintained during power operations so 
that the power distribution and reactivity limits defined by the design 
power peaking and shutdown margin (SDM) limits are preserved.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment will allow plant operation to continue 
when a CEAPIS channel is inoperable by requiring prompt verification 
of CEA positions following CEA movement. CEAs are most likely to 
become misaligned during movement and therefore, this change will 
cause CEA alignment errors to be promptly detected and corrected. It 
is appropriate to clarify that CEAPIS channels are not subject to 
the requirements of TS 2.15(1), (2), and (3) as they are not 
designed to be placed in trip or bypass, nor are they engineered 
safety feature (ESF) or isolation logic subsystems.
    The proposed amendment does not alter the requirements of TS 
2.15(4) regarding the rod block function of the secondary CEAPIS 
channel. Should the secondary CEAPIS channel or its rod block 
function be inoperable, several additional CEA deviation events are 
possible. However, this situation is already addressed by TS 
2.15(4), which requires the CEAs (rods) to be maintained fully 
withdrawn with the control rod drive system mode switch in the off 
position except when manual motion of CEA Group 4 is required to 
control axial power distribution. This is the same position that the 
CEAs must be in (fully withdrawn) when the plant is at power (Mode 
1) in order to utilize distributed control system (DCS) core mimic 
to CHANNEL CHECK the CEAPIS channels.
    If it was not possible to use DCS core mimic to verify the 
primary CEAPIS channel as would be the case if CEA Group 4 was 
inserted to control axial power distribution, then the primary 
CEAPIS channel would be declared inoperable when the CHANNEL CHECK 
could not be accomplished. The plant would then be placed in hot 
shutdown (Mode 3) within 12 hours in accordance with TS 2.15(4). 
Therefore, although the proposed amendment will allow a CEAPIS 
channel to be inoperable indefinitely, there is no significant 
increase in the probability or consequences of an accident as the 
requirements of TS 2.15(4) will continue to be met. This serves to 
prevent the type of CEA deviation events that the rod block function 
was designed for.
    Replacing the current method of verifying CEAPIS data with the 
defined term CHANNEL CHECK is an improvement that provides 
additional flexibility without weakening the intent of the 
surveillance. As a result, when it is feasible to obtain CEA 
position indication from DCS core mimic (i.e., when the CEAs are 
either fully inserted or fully withdrawn), the primary and secondary 
CEAPIS channels will be compared with DCS core mimic indication as 
well as each other.
    As an additional means of verifying CEA positions, DCS core 
mimic indication provides added confidence that the CEAs are in the 
indicated positions. Should the primary or secondary CEAPIS channel 
become inoperable, the accuracy and reliability of DCS core mimic 
indication is assured by its previous comparison with both OPERABLE 
channels. Comparison of the OPERABLE CEAPIS channel with DCS core 
mimic will satisfy the required CHANNEL CHECK and allow continued 
operation while the inoperable channel is repaired. The proposed 
amendment ensures that the CEA alignment required by TS 2.10.2(4) is 
met each shift by requiring all full length (shutdown and 
regulating) CEAs to be positioned within 12 inches of all other CEAs 
in the group.
    The change proposed for TS 2.10.2(7)c incorporates more 
conservative wording to ensure that the regulating CEA groups are 
maintained within the Long Term Insertion Limit. The proposed change 
will ensure that corrective actions are taken if either time 
interval is exceeded and makes TS 2.10.2(7)c more consistent with CE 
STS.
    The proposed amendment 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. As an additional means of verifying 
primary and secondary CEAPIS data, DCS core mimic indication 
increases confidence in the reliability of CEAPIS data.
    The proposed amendment will help minimize unplanned shutdowns 
that can cause plant transients yet continues to ensure that power 
distribution and reactivity limits are maintained.
    Therefore, it is concluded that this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment does not change the design function or 
operation of the primary or secondary CEAPIS channels. If one CEAPIS 
channel should become inoperable, the position of CEAs will be 
verified within 15 minutes of any CEA movement to quickly detect and 
correct CEA alignment errors. Data from each CEAPIS channel will 
continue to be compared to the other channel each shift as before. 
However, a CHANNEL CHECK will require that CEAPIS channel data also 
be compared with DCS core mimic indication when it is available. 
Thus, when the CEAPIS channels are required to be OPERABLE, there 
will be at least two means of verifying the position of CEAs or else 
appropriate actions must be taken. The CEA alignment required by TS 
2.10.2(4) is assured by requiring verification each shift that all 
full length (shutdown and regulating) CEAs are positioned within 12 
inches of all other CEAs in the group.
    No changes are proposed to testing and calibration of the CEAPIS 
channels and these requirements will continue to ensure that they 
are capable of performing their design function. Use of the defined 
term CHANNEL CHECK is an appropriate surveillance method as it 
requires that the channel be compared with other independent 
channels measuring the same variable where feasible. DCS core mimic 
is a diverse, accurate and reliable means of verifying CEA positions 
when the CEAs are fully inserted or fully withdrawn. The change 
proposed for TS 2.10.2(7)c ensures that appropriate corrective 
actions are taken when the regulating CEA groups are below the Long 
Term Insertion Limit in excess of either of the specified time 
intervals.
    No new structures, systems, or components (SSCs) are being 
installed, and no credible new failure mechanisms, malfunctions, or 
accident initiators are created.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    When a CEAPIS channel is inoperable, the proposed amendment 
allows plant operation to continue but requires more frequent 
verification of CEA positions following any CEA movement, which is 
when CEAs are most likely to become misaligned. This will enable CEA 
alignment errors to be detected and corrected more promptly. As 
CEAPIS channels are not designed to be placed in trip

[[Page 81672]]

or bypass, nor are they engineered safety feature (ESF) or isolation 
logic subsystems, it is appropriate to clarify that TS 2.15(1), (2), 
and (3) do not apply. FCS normally operates with the CEAs fully 
withdrawn and maintains reactivity control by adjusting reactor 
coolant system (RCS) boric acid concentration. When the CEAs are 
fully withdrawn (or fully inserted), DCS core mimic indication 
provides accurate and reliable indication of CEA positions suitable 
for comparison with the primary and secondary CEAPIS channels. Thus, 
even with one CEAPIS channel inoperable, a diverse means of 
verifying the accuracy of the OPERABLE CEAPIS channel will be 
available. The accuracy and reliability of DCS core mimic is assured 
by testing conducted each refueling outage with continued assurance 
provided by comparison with primary and secondary CEAPIS each shift.
    The change also ensures that the CEA alignment required by TS 
2.10.2(4) is met each shift by requiring all full length (shutdown 
and regulating) CEAs to be positioned within 12 inches of all other 
CEAs in the group. The proposed amendment does not alter the TS 
2.15(4) requirement to place the reactor in hot shutdown in the 
event that both CEAPIS channels are inoperable. The change proposed 
for TS 2.10.2(7)c incorporates more conservative wording to ensure 
that the regulating CEA groups are maintained within the Long Term 
Insertion Limit.
    The proposed amendment will help minimize unplanned shutdowns 
that can cause plant transients yet continues to ensure that power 
distribution and reactivity limits are maintained. The proposed 
amendment 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 amendment does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700 
K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Michael T. Markley.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

    Date of amendment request: October 4, 2010.
    Description of amendment request: The proposed amendment would 
modify the Technical Specification (TS) requirements for snubbers in TS 
3/4.7.9 due to planned revisions to the inservice inspection (ISI) 
program.
    For the current third 10-year ISI intervals, at Salem Nuclear 
Generating Station (Salem), Units 1 and 2, snubber testing and 
examination are performed in accordance with the specific requirements 
of TS 3/4.7.9 in lieu of the requirements contained in American Society 
of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code), 
Section XI, Article IWF-5000, as previously authorized by the U.S. 
Nuclear Regulatory Commission (NRC or the Commission).
    Section 50.55a(g)(4)(ii) of Title 10 of the Code of Federal 
Regulations (10 CFR) requires that inservice examination of components 
conducted during successive 120-month inspection intervals must comply 
with the requirements of the latest edition and addenda of the ASME 
Code incorporated by reference in 10 CFR 50.55a(b), 12 months before 
the start of the inspection interval. For the Salem Unit 1 fourth 10-
year ISI interval beginning on May 20, 2011, the licensee intends to 
adopt Subsection ISTD of the ASME Code for Operation and Maintenance of 
Nuclear Power Plants (OM Code), 2004 Edition, in place of the 
requirements for snubbers in ASME Code, Section XI, Articles IWF-
5200(a) and (b) and IWF-5300(a) and (b), as permitted by 10 CFR 
50.55a(b)(3)(v). The licensee also intends to adopt Subsection ISTD of 
the ASME OM Code for the remainder of the Salem Unit 2 third 10-year 
ISI interval which ends on November 27, 2013.
    In accordance with 10 CFR 50.55a(g)(5)(ii), if a revised ISI 
program for a facility conflicts with the TSs for the facility, 
licensees are required to apply to the Commission for amendment of the 
TSs to conform the TSs to the revised program. Due to the planned 
changes to the ISI program, the proposed amendment would replace the 
specific TS requirements for snubbers, currently contained in 
surveillance requirement (SR) 4.7.9, with reference to the program for 
examination, testing and service life monitoring for snubbers. In 
addition, the current reference to SR 4.7.9c in TS ACTION 3.7.9 would 
be replaced with reference to the program for snubbers.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below, with Nuclear Regulatory 
Commission (NRC) staff edits in square brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes revise [TS 3/4.7.9 due to planned changes 
to the ISI program for snubbers. Specifically, the proposed 
amendment would replace the TS SRs for snubbers with reference to 
the program for examination, testing and service life monitoring for 
snubbers. Following implementation of the proposed amendment, in 
lieu of the TS SRs, snubber examination, testing and service life 
monitoring would be governed by the requirements in Section XI of 
the ASME Code or the OM Code as required by 10 CFR 50.55a(g) or 10 
CFR .55a(b)(3)(v), except where the NRC has granted specific written 
relief, pursuant to 10 CFR 50.55a(g)(6)(i), or authorized 
alternatives pursuant to 10 CFR 50.55a(a)(3).]
    Snubber examination, testing and service life monitoring is not 
an initiator of any accident previously evaluated. Therefore, the 
probability of an accident previously evaluated is not significantly 
increased.
    Snubbers will continue to be demonstrated OPERABLE by 
performance of a program for examination, testing and service life 
monitoring in compliance with 10 CFR 50.55a or authorized 
alternatives. The proposed change to TS ACTION 3.7.9 for inoperable 
snubbers is administrative in nature and is required for consistency 
with the proposed change to SR 4.7.9. Therefore the proposed change 
does not adversely affect plant operations, design functions or 
analyses that verify the capability of systems, structures, and 
components to perform their design functions. The consequences of 
accidents 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.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not involve any physical alteration of 
plant equipment. The proposed change does not change the method by 
which any safety-related system performs its function. As such, no 
new or different types of equipment will be installed, and the basic 
operation of 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 previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes ensure snubber examination, testing and 
service life monitoring will continue to meet the requirements of 10 
CFR 50.55a(g) except where the NRC has granted specific written 
relief, pursuant to 10 CFR 50.55a(g)(6)(i), or authorized 
alternatives pursuant to 10 CFR

[[Page 81673]]

50.55a(a)(3). Snubbers will continue to be demonstrated OPERABLE by 
performance of a program for examination, testing and service life 
monitoring in compliance with 10 CFR 50.55a or authorized 
alternatives. The proposed change to TS ACTION 3.7.9 for inoperable 
snubbers is administrative in nature and is required for consistency 
with the proposed change to SR 4.7.9.
    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, and with the changes noted above in square brackets, 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: Vincent Zabielski, PSEG Nuclear LLC-N21, 
P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Harold K. Chernoff.

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

    Date of amendment request: September 22, 2010, as supplemented by 
letter dated November 22, 2010.
    Description of amendment request: The proposed amendment consists 
of changes to the approved fire protection program as described in the 
Wolf Creek Generating Station (WCGS) Updated Safety Analysis Report 
(USAR). Specifically, amendment proposes a deviation from a commitment 
to certain technical requirements of 10 CFR, Part 50, Appendix R, 
Section III.L.1, as described in Appendix 9.5E of the WCGS USAR. The 
licensee has proposed to revise USAR Table 9.5E-1 to include 
information on Reactor Coolant System process variables not maintained 
within those predicted for a loss of normal ac [alternating current] 
power.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The design function of structures, systems and components (SSCs) 
are not impacted by the proposed change. Evaluation SA-08-006 Rev. 1 
[RETRAN-3D Post-Fire Safe Shutdown (PFSSD) Consequence Evaluation 
for a Postulated Control Room Fire] has demonstrated that the 
formation of voids in the reactor head for a short time following a 
fire in the control room and spurious temporary opening of the 
pressurizer power operated relief valve (PORV) does not result in 
damage to a fission product barrier and does not result in a loss of 
natural circulation cooldown. The proposed change does not alter or 
prevent the ability of SSCs from performing their intended function 
to mitigate the consequences of an initiating event within the 
assumed acceptance limits.
    Therefore, the probability of any accident previously evaluated 
is not increased. Equipment required to mitigate an accident remains 
capable of performing the assumed function.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change will not alter the requirements or function 
for systems required during accident conditions. The design function 
of structures, systems and components are not impacted by the 
proposed change. The thermal hydraulic analysis of the reactor 
coolant system identified that the process variables are not 
maintained within those predicted for a loss of normal ac power, 
however, the fission product boundary integrity is not affected.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    There 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 functions. There will be no impact on departure from 
nuclear boiling ratio (DNBR) limits, heat flux hot channel factor 
(F<INF>Q</INF>(Z)) limits, nuclear enthalpy rise hot channel factor 
(FN<INF>N</INF>[Delta]<INF>H</INF>) limits, peak centerline 
temperature (PCT) limits, peak local power density or any other 
margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: November 4, 2010.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 5.6.5, ``CORE OPERATING LIMITS 
REPORT (COLR),'' to replace the existing large break loss-of-coolant 
accident (LOCA) analysis methodology. Specifically, the proposed change 
adds a reference of Westinghouse Electric Company's topical report 
WCAP-16009-P-A, ``Realistic Large Break LOCA Evaluation Methodology 
Using Automated Statistical Treatment of Uncertainty Method (ASTRUM),'' 
to TS 5.6.5b.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises TS Section 5.6.5 to incorporate a 
new large break LOCA analysis methodology. Specifically, the 
proposed change adds WCAP-16009-P-A to TS 5.6.5b as a method used 
for establishing core operating limits.
    Accident analyses are not accident initiators; therefore, the 
proposed change does not involve a significant increase in the 
probability of an accident. The analyses using ASTRUM demonstrated 
that the acceptance criteria in 10 CFR 50.46, ``Acceptance criteria 
for emergency core cooling systems for lightwater nuclear power 
reactors,'' were met. Large break LOCA analyses performed consistent 
with the methodology in NRC-approved WCAP-16009-P-A, including 
applicable assumptions, limitations and conditions, demonstrate that 
10 CFR 50.46 acceptance criteria are met; thus, this change does not 
involve a significant increase in the consequences of an accident. 
No physical changes to the plant are associated with the proposed 
change.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any previously evaluated?
    Response: No.
    The proposed change revises TS Section 5.6.5 to incorporate a 
new large break LOCA analysis methodology. Specifically, the 
proposed change adds WCAP-16009-P-A to TS 5.6.5b as a method used 
for establishing core operating limits. There are no physical 
changes being made to the plant as a result

[[Page 81674]]

of using the Westinghouse ASTRUM analysis methodology in WCAP-16009-
P-A for performance of the large break LOCA analyses. Large break 
LOCA analyses performed consistent with the methodology in NRC-
approved WCAP-16009-P-A, including applicable assumptions, 
limitations and conditions, demonstrate that 10 CFR 50.46 acceptance 
criteria are met. No new modes of plant operation are being 
introduced. The configuration, operation, and accident response of 
the structures or components are unchanged by use of the new 
analysis methodology. Analyses of transient events have confirmed 
that no transient event results in a new sequence of events that 
could lead to a new accident scenario. The parameters assumed in the 
analyses are within the design limits of existing plant equipment.
    In addition, employing the Westinghouse ASTRUM large break LOCA 
analysis methodology does not create any new failure modes that 
could lead to a different kind of accident. The design of systems 
remains unchanged and no new equipment or systems have been 
installed which could potentially introduce new failure modes or 
accident sequences. No changes have been made to instrumentation 
actuation setpoints. Adding the reference to WCAP-16009-P-A in TS 
Section 5.6.5b is an administrative change that does not create the 
possibility of a new or different kind of accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change revises TS Section 5.6.5 to incorporate a 
new large break LOCA analysis methodology. Specifically, the 
proposed change adds WCAP-1 6009-P-A to TS 5.6.5b as a method used 
for establishing core operating limits. The analyses using ASTRUM 
demonstrated that the applicable acceptance criteria in 10 CFR 50.46 
are met. Margins of safety for large break LOCAs include 
quantitative limits for fuel performance established in 10 CFR 
50.46. These acceptance criteria are not being changed by this 
proposed new methodology. Large break LOCA analyses performed 
consistent with the methodology in NRC-approved WCAP-16009-P-A, 
including applicable assumptions, limitations and conditions, 
demonstrate that 10 CFR 50.46 acceptance criteria are met; thus, 
this change does not involve a significant reduction in a margin of 
safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: 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.

NextEra Energy Point Beach, Docket Nos. 50-266 and 50-301, Point Beach 
Nuclear Plant (PBNP), Units 1 and 2, Manitowoc County, Wisconsin

    Date of amendment request: April 7, 2009, as supplemented by 
letters dated June 17 (two letters) and December 8 of 2009; and April 
15, July 8, July 28, August 24, September 9, September 21, October 14, 
and November 1 of 2010.
    Brief description of amendment request: The proposed amendment 
would increase the licensed core power level for PBNP Units 1 and 2 
from 1540 to 1800 megawatts thermal. The increase in core thermal power 
will be approximately 17 percent over the current licensed thermal 
power level and is categorized as an Extended Power Uprate.
    Date of publication of individual notice in Federal Register: 
November 17, 2010 (75 FR 70305).
    Expiration date of individual notice: January 18, 2011.

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 01 F21, 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, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1-800-397-4209, 301-415-4737 or by 
e-mail to <a href="/cdn-cgi/l/email-protection#057561772b7760766a70776660456b77662b626a73"><span class="__cf_email__" data-cfemail="d5a5b1a7fba7b0a6baa0a7b6b095bba7b6fbb2baa3">[email protected]</span></a>.

Indiana Michigan Power Company (IandM), Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of application for amendment: September 8, 2010.
    Brief description of amendment: The amendments delete the Technical 
Specification requirements related to the containment hydrogen 
recombiners and the hydrogen monitors, in accordance with Nuclear 
Energy Institute Technical Specification Task Force (TSTF) initiative 
designated as TSTF-447.
    Date of issuance: December 14, 2010.
    Effective date: As of the date of issuance and shall be implemented

[[Page 81675]]

within 120 days from the date of issuance.
    Amendment Nos.: 313 (for Unit 1) and 296 (for Unit 2).
    Facility Operating License Nos. DPR-58 and DPR-74: Amendment 
revised the Renewed Operating License and Technical Specifications.
    Date of initial notice in Federal Register: October 14, 2010 (75 FR 
63209).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 14, 2010.
    No significant hazards consideration comments received: No.

Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue 
County, Minnesota

    Date of application for amendments: November 24, 2009, as 
supplemented by letter dated May 26, 2010.
    Brief description of amendments: These amendments revise Technical 
Specification (TS) 4.2.1, ``Fuel Assemblies,'' to add Optimized 
ZIRLO\TM\ as an acceptable fuel rod cladding material and add two 
Westinghouse topical reports to the analytical methods identified in TS 
5.6.5.b.
    Date of issuance: November 29, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 199, 187.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 4, 2010 (75 FR 
23816).
    The supplemental letter contained clarifying information and did 
not change the initial no significant hazards consideration 
determination, and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 29, 2010.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland this 16th day of December, 2010.

    For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2010-32668 Filed 12-27-10; 8:45 am]
BILLING CODE 7590-01-P


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