[Federal Register Volume 75, Number 124 (Tuesday, June 29, 2010)]
[Notices]
[Pages 37471-37478]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2010-15439]


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NUCLEAR REGULATORY COMMISSION

[NRC-2010-0232]


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 June 3, 2010 to June 16, 2010. The last 
biweekly notice was published on June 15, 2010 (75 FR 33839).

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 Cindy Bladey, 
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 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, 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/

[[Page 37472]]

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

[[Page 37473]]

document on all other participants. Filing is considered complete by 
first-class mail as of the time of deposit in the mail, or by courier, 
express mail, or expedited delivery service upon depositing the 
document with the provider of the service. A presiding officer, having 
granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant 
to an order of the Commission, or the presiding officer. Participants 
are requested not to include personal privacy information, such as 
social security numbers, home addresses, or home phone numbers in their 
filings, unless an NRC regulation or other law requires submission of 
such information. With respect to copyrighted works, except for limited 
excerpts that serve the purpose of the adjudicatory filings and would 
constitute a Fair Use application, participants are requested not to 
include copyrighted materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Non-timely filings 
will not be entertained absent a determination by the presiding officer 
that the petition or request should be granted or the contentions 
should be admitted, based on a balancing of the factors specified in 10 
CFR 2.309(c)(1)(i)-(viii).
    For further details with respect to this license amendment 
application, see the application for amendment which is available for 
public inspection at the Commission's PDR, located at One White Flint 
North, Public File Area O1F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the ADAMS Public Electronic Reading Room on the Internet at the NRC Web 
site, http://www.nrc.gov/reading-rm/adams.html. Persons who do not have 
access to ADAMS or who encounter problems in accessing the documents 
located in ADAMS, should contact the NRC PDR Reference staff at 1-800-
397-4209, 301-415-4737, or by e-mail to [email protected].

Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power 
Station, Kewaunee County, Wisconsin

    Date of amendment request: April 13, 2010.
    Description of amendment request: The licensee proposed to revise 
Section 3.1.a.1.C, ``Reactor Coolant Pumps,'' Section 3.1.a.3, 
``Pressurizer Safety Valves,'' and Section 3.1.b, ``Heatup and Normal 
Cooldown Limit Curves for Normal Operation,'' of the Technical 
Specifications (TS). Specifically, the proposed amendment would replace 
the heatup and cooldown pressure-temperature (P-T) limit curves with 
new ones, and specifying a higher low temperature overpressure 
protection (LTOP) enabling temperature.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration (NSHC) analysis. The NRC staff reviewed the licensee's 
NSHC analysis and has prepared its own as follows:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The purpose of the P-T limits curves and LTOP is to ensure that 
the reactor vessel is operated within its material design limits. As 
such, the subject specifications specify the pressure limits inside 
the reactor vessel under different temperature conditions for normal 
operation. No conditions of operation within the approved P-T limits 
were postulated to be initiators of accidents previously analyzed in 
the Kewaunee Final Safety Analysis Report. Furthermore, the 
consequences of the analyzed accidents were not postulated to be 
exacerbated by normal operation within approved P-T limits. 
Accordingly, the probability of occurrence and the consequences of 
the previously analyzed accidents would not be affected in any way 
by the proposed P-T limits changes.
    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 any physical alteration of 
the plant (no new or different type of equipment will be installed) 
nor does it change methods and procedures governing plant operation. 
The proposed change will not impose any new or eliminate any old 
requirements. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change will not reduce a margin of safety because 
it has no effect on any safety analysis methods, scenarios, or 
assumptions. 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 proposed amendment involves no significant hazards 
consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., Counsel for Dominion Energy Kewaunee, Inc., 
120 Tredegar Street, Richmond, VA 23219.
    NRC Branch Chief: Robert J. Pascarelli.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of amendment request: April 28, 2010.
    Description of amendment request: The proposed change revises the 
Final Safety Analysis Report and Emergency Plan to support U.S. 
Department of Energy non-intrusive surveillance and characterization 
activities within the 618-11 Waste Burial Ground.
    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.
    Normal and postulated activities at the 618-11 site do not serve 
as initiators of any Columbia [Generating Station] accident 
previously evaluated, nor do they require reassessment of the 
previously evaluated accidents. The accident probabilities are 
unaffected and the outcomes remain unchanged.
    Therefore there is no 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 analyzed?
    Response: No.
    The only hazard postulated beyond the 618-11 site and onto the 
Columbia facility is a release of 44.5 mrem [millirem] at 100 m 
[meters]. This level of exposure does not impact the design function 
or operation of any Columbia SSCs [structures, systems, or 
components]. The protected area of the facility that encloses the 
safety related SSCs is greater than 300 m from the postulated 
release point. The calculated dose at 300 m is 3 mrem. This level of 
exposure does not cause any 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.

[[Page 37474]]

    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The only hazard postulated beyond the 618-11 site and onto the 
Columbia facility is a release of 44.5 mrem at 100 m. This level of 
exposure does not impact the design function or operation of any 
Columbia SSCs. The protected area of the facility that encloses the 
safety related SSCs is greater than 300 m from the postulated 
release point. The calculated dose at 300 m is 3 mrem. This level of 
exposure does not impact the equipment qualification of SSCs and is 
well within the mild environment range for SSCs. It does not exceed 
or alter a design safety limit.
    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 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William A. Horin, Esq., Winston & Strawn, 
1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Michael T. Markley.

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: April 13, 2010.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) to institute a requirement to 
perform a Logic System Functional Test of the Control Rod Block 
actuation instrumentation trip functions once every Operating Cycle.
    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 change does not impact the function of any structure, system 
or component that affects the probability of an accident or that 
supports mitigation or consequences of an accident previously 
evaluated. The proposed change adds a requirement to perform 
additional testing of the control rod block instrumentation. The 
proposed change does not affect reactor operations or accident 
analysis and there is no change to the radiological consequences of 
a previously analyzed accident. The operability requirements for 
accident mitigation systems remain consistent with the licensing and 
design basis.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve any physical alteration of 
plant equipment and does not change the method by which any safety-
related system performs its function. The proposed change involves 
the addition of a requirement to perform a logic system functional 
test of plant instrumentation. This test is within the design 
capability of the system and does not create the possibility of a 
different kind of accident. No new or different types of equipment 
will be permanently installed. Operation of existing installed 
equipment is unchanged. The methods governing plant operation and 
testing remain consistent with current safety analysis assumptions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    These changes do not change any existing design or operational 
requirements and do not adversely affect existing plant safety 
margins or the reliability of the equipment assumed to operate in 
the safety analysis. The proposed change only affects the testing of 
the control rod block instrumentation. As such, there are no changes 
being made to safety analysis assumptions, safety limits or safety 
system settings that would adversely affect plant safety as a result 
of the proposed change.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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: April 13, 2010.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to update the Table of 
Contents and the Applicability and Objective portions of TS 4.12 as a 
result of changes made by License Amendments 230 and 239, and to revise 
wording in TS 3.7.A.8. The proposed changes are considered 
administrative in nature and do not materially change any technical 
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. 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 changes are administrative in nature and do not 
involve any physical changes to the plant. The changes do not revise 
the methods of plant operation which could increase the probability 
or consequences of accidents. No new modes of operation are 
introduced by the proposed changes such that a previously evaluated 
accident is more likely to occur or more adverse consequences would 
result.
    Therefore, the proposed changes do 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 changes are administrative in nature and do not 
affect the operation of any systems or equipment, nor do they 
involve any potential initiating events that would create any new or 
different kind of accident. There are no changes to the design 
assumptions, conditions, configuration of the facility, or manner in 
which the plant is operated and maintained. The changes do not 
affect assumptions contained in plant safety analyses or the 
physical design and/or modes of plant operation. Consequently, no 
new failure mode is introduced.
    Therefore, the proposed changes do 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.
    There are no changes being made to the Technical Specification 
(TS) safety limits or safety system settings. The operating limits 
and functional capabilities of systems, structures and components 
are unchanged as a result of these administrative changes. These 
changes do not affect any equipment involved in potential initiating 
events or plant response to accidents. There is no change to the 
basis for any TS related to the establishment, or maintenance of, a 
nuclear safety margin. The proposed changes do not impact any safety 
limits, safety settings or safety margins.

[[Page 37475]]

    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.
    NRC Branch Chief: Nancy Salgado.

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

    Date of amendment request: March 31, 2010.
    Description of amendment request: The proposed amendment would 
implement an alternative source term (AST) for Arkansas Nuclear One, 
Unit 2 (ANO-2). The proposed amendment would modify Technical 
Specification (TS) 3.4.8, ``Specific Activity,'' and 6.5.12, ``Control 
Room Habitability Program,'' and associated definitions as related to 
the use of an AST associated with accident offsite and control room 
dose consequences.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) Section 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 use of an AST is recognized in 10 CFR 50.67. RG [Regulatory 
Guide] 1.183 provides guidance for implementation of an AST. The AST 
involves quantities, isotopic composition, chemical and physical 
characteristics, and release timing of radioactive material for use 
as inputs to accident dose analyses. As such, the AST cannot affect 
the probability of occurrence of a previously evaluated accident. In 
addition, the increase in the DEX [Dose Equivalent Xenon-133] 
activity limit and the terminology/reference changes proposed for 
the ANO-2 TSs are unrelated to accident initiators. No facility 
equipment, procedure, or process changes are required in conjunction 
with implementing the AST that could increase the likelihood of a 
previously analyzed accident. The proposed changes in the source 
term and the methodology for the dose consequence analyses follow 
the guidance of RG 1.183. As a result, there is no increase in the 
likelihood of existing event initiators.
    Regarding accident consequences, the increase in the DEX 
activity limit acts to support the analysis results given the 
application of an AST. The proposed limit was utilized as an 
assumption in the AST analysis and determined to be acceptable. The 
results of accident dose analyses using the AST are compared to TEDE 
[Total Effective Dose Equivalent] acceptance criteria that account 
for the sum of deep dose equivalent (for external exposure) and 
committed effective dose equivalent (for internal exposure). Dose 
results were previously compared to separate limits on whole body, 
thyroid, and skin doses as appropriate for the particular accident 
analyzed. The results of the revised dose consequences analyses 
demonstrate that the regulatory acceptance criteria are met for each 
analyzed event. The proposed TS terminology/reference changes are 
consistent with the analysis and adoption of an AST. Implementing 
the AST involves no facility equipment, procedure, or process 
changes that could affect the radioactive material actually released 
during an event. Subsequently, no conditions have been created that 
could significantly increase the consequences of any of the events 
being evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any of the events 
being 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 AST involves quantities, isotopic composition, chemical and 
physical characteristics, and release timing of radioactive material 
for use as inputs to accident dose analyses. As such, the AST cannot 
create the possibility of a new or different kind of accident. In 
addition, the increased DEX activity limit and proposed terminology/
reference changes within the TSs are unrelated to accident 
initiators and are supported by AST adoption.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Implementing the AST is relevant only to calculated accident 
dose consequences. The results of the revised dose consequences 
analyses demonstrate that the regulatory acceptance criteria are met 
for each analyzed event. In addition, the increased DEX activity 
limit and proposed terminology/reference changes within the TSs 
support adoption of the AST methodologies, have been determined to 
result in acceptable dose consequence and do not result in a 
significant impact to any margin of safety. The AST does not affect 
the transient behavior of non-radiological parameters (e.g., RCS 
[Reactor Coolant System] pressure, Containment pressure) that are 
pertinent to 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: Joseph A. Aluise, Associate General 
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Michael T. Markley.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: April 19, 2010.
    Description of amendment request: The proposed amendments would 
revise Technical Specification 3.4.11, ``RCS Pressure and Temperature 
(P/T) Limits,'' to incorporate revised P/T curves that are valid for up 
to 32 effective full power years of operation.
    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 revises Technical Specification (TS) Section 
3.4.11 to replace the existing P/T curves with revised curves that 
are valid up to 32 EFPY. The revised curves were developed using the 
methodology of General Electric (GE) Topical Report NEDC-32983P, 
``General Electric Methodology for Reactor Pressure Vessel Fast 
Neutron Flux Evaluations.'' The NEDC-32983P methodology has been 
approved by the NRC for use by licensees. The P/T limits are not 
derived from design basis accident analyses. They are prescribed 
during normal operation to avoid encountering pressure, temperature, 
and temperature rate of change conditions that might cause 
undetected flaws to propagate and cause non-ductile failure of the 
reactor coolant pressure boundary, a condition that is unanalyzed. 
Since the P/T limits are not derived from any design basis accident, 
there are no acceptance limits related to the P/T limits. Rather, 
the P/T limits are acceptance limits themselves since they preclude 
operation in an unanalyzed condition.
    Thus, the proposed changes do not have any affect on the 
probability of an accident previously evaluated.

[[Page 37476]]

    The P/T curves are used as operational limits during heatup or 
cooldown maneuvering, when the pressure and temperature indications 
are monitored and compared to the applicable curve to determine that 
operation is within the allowable region. The P/T curves provide 
assurance that station operation is consistent with a previously 
evaluated accident. Thus, the radiological consequences of any 
accident previously evaluated are not increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not change the response of plant 
equipment to transient conditions. The proposed change does not 
introduce any new equipment, modes of system operation, or failure 
mechanisms.
    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?
    Response: No.
    The proposed change adopts P/T curves that have been developed 
using the methodology of GE Topical Report NEDC-32983P. The NEDC-
32983P methodology adheres to the guidance in NRC Regulatory Guide 
1.190, ``Calculation and Dosimetry methods for Determining Pressure 
Vessel Neutron Fluence,'' dated March 2001. In a letter dated 
September 14, 2001, the NRC approved NEDC-32983P for use by 
licensees. The proposed change does not alter the manner in which 
safety limits, limiting safety system settings, or limiting 
conditions for operation are determined. The setpoints at which 
protective actions are initiated are not altered by the proposed 
change.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of amendment request: March 29, 2010, as supplemented on May 
28, 2010.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to extend the allowed outage 
time (AOT) for the ``A'' and ``B'' emergency diesel generators (EDGs) 
from 72 hours to 14 days.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The emergency diesel generators are safety related components 
which provide backup electrical power supply to the onsite 
Safeguards Distribution System. The emergency diesel generators are 
not accident initiators; the EDGs are designed to mitigate the 
consequences of previously evaluated accidents including a loss of 
offsite power [LOOP]. Extending the AOT for a single EDG would not 
affect the previously evaluated accidents since the remaining EDGs 
supporting the redundant Engineered Safety Features (ESF) systems 
would continue to be available to perform the accident mitigation 
functions.
    Thus allowing an emergency diesel generator to be inoperable for 
an additional 11 days for performance of maintenance or testing does 
not increase the probability of a previously evaluated accident.
    Deterministic and probabilistic risk assessments evaluated the 
effect of the proposed Technical Specification changes on the 
availability of an electrical power supply to the plant emergency 
safeguards features systems. These assessments concluded that the 
proposed Technical Specification changes do not involve a 
significant increase in the risk of power supply unavailability.
    There is incremental risk associated with continued operation 
for an additional 11 days with one emergency diesel generator 
inoperable; however, the calculated impact on risk is very small and 
is consistent with the acceptance guidelines contained in Regulatory 
Guides 1.174 and 1.177. This risk is judged to be reasonably 
consistent with the risk associated with operations for 72 hours 
with one emergency diesel generator inoperable as allowed by the 
current Technical Specifications. Specifically, the remaining 
operable emergency diesel generators and paths are adequate to 
supply electrical power to the onsite Safeguards Distribution 
System. An emergency diesel generator is required to operate only if 
both offsite power sources fail and there is an event which requires 
operation of the plant emergency safeguards features such as a 
design basis accident. The probability of a design basis accident 
occurring during this period is low.
    The consequences of previously evaluated accidents will remain 
the same during the proposed 14 day AOT as during the current 72 
hour AOT. The ability of the remaining TS required EDG to mitigate 
the consequences of an accident will not be affected since no 
additional failures are postulated while equipment is inoperable 
within the TS AOT. The standby power supply for each of the four 
safety-related load groups consists of one EDG complete with its 
auxiliaries, which include the cooling water, starting air, 
lubrication, intake and exhaust, and fuel oil systems. The sizing of 
the EDGs and the loads assigned among them is such that any 
combination of three out of four of these EDGs is capable of 
shutting down the plant safely, maintaining the plant in a safe 
shutdown condition, and mitigating the consequences of accident 
conditions.
    Thus, this change does not involve a significant increase in the 
probability or consequences of a previously analyzed accident.
    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 Technical Specification changes do not involve a 
change in the plant design, system operation, or procedures involved 
with the emergency diesel generators. The proposed changes allow an 
emergency diesel generator to be inoperable for additional time. 
Equipment will be operated in the same configuration and manner that 
is currently allowed and designed for. There are no new failure 
modes or mechanisms created due to plant operation for an extended 
period to perform emergency diesel generator maintenance or testing. 
Extended operation with an inoperable emergency diesel generator 
does not involve any modification in the operational limits or 
physical design of plant systems. There are no new accident 
precursors generated due to the extended AOT.
    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.
    Currently, if an inoperable emergency diesel generator is not 
restored to operable status within 72 hours, Technical Specification 
3.8.1.1 ACTION b requires the unit be in at least HOT SHUTDOWN 
within the next 12 hours and in COLD SHUTDOWN within the following 
24 hours. The proposed Technical Specification changes will allow 
steady state plant operation at 100% power for an additional 11 
days.
    Deterministic and probabilistic risk assessments evaluated the 
effect of the proposed Technical Specification changes on the 
availability of an electrical power supply to the plant emergency 
safeguards features systems. These assessments concluded that the 
proposed Technical Specification changes do not involve a 
significant increase in the risk of power supply unavailability.
    The EDGs continue to meet their design requirements; there is no 
reduction in capability or change in design configuration. The EDG 
response to LOOP, LOCA [loss-of-

[[Page 37477]]

coolant accident], SBO [station blackout], or fire is not changed by 
this proposed amendment; there is no change to the EDG operating 
parameters. In the extended AOT, as in the existing AOT, the 
remaining operable emergency diesel generators and paths are 
adequate to supply electrical power to the onsite Safeguards 
Distribution System. The proposed change does not alter a design 
basis or safety limit; therefore it does not significantly reduce 
the margin of safety. The EDGs will continue to operate per the 
existing design and regulatory requirements.
    Therefore, based on the considerations given above, 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: 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: April 13, 2010, as supplemented by 
letter dated June 1, 2010.
    Description of amendment request: The proposed amendment to Renewed 
Facility Operating License No. NPF-42 would revise the approved fire 
protection program, as described in the Wolf Creek Generating Station 
Updated Safety Analysis Report, by removing the high/low pressure 
interface designation of the pressurizer power-operated relief valves 
(PORVs) and their associated block valves.
    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 are 
not impacted by the proposed change. This amendment classifies the 
pressurizer PORVs and their associated block valves based on the 
guidance in Regulatory Guide 1.189, ``Fire Protection for Nuclear 
Power Plants,'' Revision 2, and Nuclear Energy [Institute] (NEI) 00-
01, ``Guidance for Post-Fire Safe-Shutdown Circuit Analysis,'' 
Revision 2, Appendix C. The classification change only affects the 
post fire safe shutdown (PFSSD) analysis methodology for the PORVs 
and block valves. Reclassification of the PORVs and block valves 
will not impact the use of the valves to depressurize the Reactor 
Coolant System (RCS) to recover from certain transients if normal 
pressurizer spray is not available.
    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.
    There are no changes in the method by which any safety related 
plant system performs its safety function, and the normal manner of 
plant operation is unaffected. No new accident scenarios, transient 
precursors, failure mechanisms, or limiting single failures are 
introduced as a result of this change. There will be no adverse 
effect or challenges imposed on any safety related system as a 
result of this change. The classification change only affects the 
PFSSD analysis methodology for the PORVs and block valves.
    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 ensure the accomplishment 
of protection functions. There will be no impact on departure from 
nuclear boiling [ratio] (DNBR) limits, heat flux hot channel factor 
(FQ(Z)) limits, nuclear enthalpy rise hot channel factor 
(F\N\[Delta]H) limits, peak centerline temperature (PCT) 
limits, peak local power density or any other margin of safety.
    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.

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, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of application for amendment: August 17, 2009, as supplemented 
by letter dated January 21, 2010.
    Brief description of amendment: The amendment modified (1) 
Technical Specification (TS) 3.8.3, ``Diesel Fuel Oil, Lube Oil, and 
Starting Air,'' to relocate specific numerical values for fuel oil and 
lube oil storage volumes

[[Page 37478]]

from the TS to the TS Bases, (2) TS 3.8.1, ``AC [Alternating Current] 
Sources--Operating,'' to relocate specific values for the day tank fuel 
oil volumes from the TS to the TS Bases, and (3) TS 5.5.9, ``Diesel 
Fuel Oil Testing Program,'' to relocate the specific standard for 
particulate concentration testing of fuel oil from the TS to the TS 
Bases.
    Date of issuance: May 27, 2010.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 215.
    Facility Operating License No. NPF-21: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: November 3, 2009 (74 FR 
56884). The supplemental letter dated January 21, 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 amendment is contained 
in a Safety Evaluation dated May 27, 2010.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear 
Plant, Van Buren County, Michigan

    Date of application for amendment: March 31, 2010, supplemented by 
letter dated May 13, 2010.
    Brief description of amendment: The amendment adds a new license 
condition 2.C (4) to Palisades Nuclear Plant, renewed facility license 
No. DPR-20. This license condition would state that performance of 
Technical Specification (TS) surveillance requirement (SR) 3.1.4.3 is 
not required for control rod drive 22 through cycle 21 or until the 
next entry into Mode 3. The amendment consists of changes to TS by 
addition of a note in SR 3.1.4.3, stating:
    ``Not required to be performed or met for control rod 22 during 
cycle 21 provided control rod 22 is administratively declared 
immovable, but trippable and Condition D is entered for control rod 
22.''
    Date of issuance: June 2, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 15 days.
    Amendment No.: 239.
    Facility Operating License No. DPR-20: Amendment revised the 
Technical Specifications and license.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): The notice provided an opportunity to submit 
comments on the Commission's proposed NSHC determination. No comments 
have been received. The notice also provided an opportunity to request 
a hearing by June 13, 2010, which is within 60 days of the individual 
notice published on April 14; but indicated that if the Commission 
makes a final NSHC determination, any such hearing would take place 
after issuance of the amendment.
    Date of initial individual notice in Federal Register: April 14, 
2010 (75 FR 19428), followed by the repeat biweekly notice in the 
Federal Register on May 4, 2010 (75 FR 23818).
    The Commission's related evaluation of the amendment, state 
consultation, and final NSHC determination are contained in a Safety 
Evaluation dated June 2, 2010.
    Attorney for licensee: Mr. William Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White 
Plains, NY 10601.
    NRC Branch Chief: Robert J. Pascarelli.

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 application for amendments: September 14, 2009, as 
supplemented on April 12, 2010.
    Brief description of amendments: The amendments make miscellaneous 
administrative and editorial changes to the Technical Specifications 
(TSs) and the Facility Operating Licenses (FOLs) including correction 
of typographical and format errors, correction of administrative 
differences between units, and deletion of historical requirements that 
have expired.
    Date of issuance: June 15, 2010.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment Nos.: 295 and 278.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the TSs and the FOLs.
    Date of initial notice in Federal Register: November 17, 2009 (74 
FR 59262). The letter dated April 12, 2010, provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination or expand the application beyond 
the scope of the original Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 15, 2010.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 18th day of June 2010.

    For the Nuclear Regulatory Commission.
Robert A. Nelson,
Deputy Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2010-15439 Filed 6-28-10; 8:45 am]
BILLING CODE 7590-01-P