[Federal Register Volume 75, Number 16 (Tuesday, January 26, 2010)]
[Notices]
[Pages 4111-4122]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2010-1315]


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NUCLEAR REGULATORY COMMISSION

[NRC-2010-0017]


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 31, 2009 to January 13, 2010. The 
last biweekly notice was published on January 12, 2010 (75 FR 1655).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in Title 10 of the Code of Federal 
Regulations (10 CFR), Section 50.92, this means that operation of the 
facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rulemaking 
and Directives Branch (RDB), TWB-05-B01M, Division of Administrative 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and should cite the publication date and 
page number of this Federal Register notice. Written comments may also 
be faxed to the RDB at 301-492-3446. Documents may be examined, and/or 
copied for a fee, at the NRC's Public Document Room (PDR), located at 
One White Flint North, Public File Area O1F21, 11555 Rockville Pike 
(first floor), Rockville, Maryland.
    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license. 
Requests for a hearing and a petition for leave to intervene shall be 
filed in accordance with the Commission's ``Rules of Practice for 
Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested person(s) 
should consult a current copy of 10 CFR 2.309, which is available at 
the Commission's PDR, located at One White Flint North, Public File 
Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, 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

[[Page 4112]]

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 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#19717c786b70777e377d767a727c6d59776b7a377e766f"><span class="__cf_email__" data-cfemail="5e363b3f2c373039703a313d353b2a1e302c3d70393128">[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#1b5648535f35497e68746e69787e5b756978357c746d"><span class="__cf_email__" data-cfemail="317c6279751f6354425e44435254715f43521f565e47">[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,

[[Page 4113]]

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 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 January 26, 2010. 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 <a href="/cdn-cgi/l/email-protection#cabaaeb8e4b8afb9a5bfb8a9af8aa4b8a9e4ada5bc"><span class="__cf_email__" data-cfemail="047460762a7661776b71766761446a76672a636b72">[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: October 30, 2009.
    Description of amendment request: The amendments would revise 
License Condition C.(1) for Units 1 and 3, and the Technical 
Specifications (TS) for all three units, to remove requirements no 
longer applicable due to the completion of power uprate, replacement of 
steam generators, removal of part-length control element assemblies 
(CEAs), and completion of a core protection calculator (CPC) upgrade, 
and to make a minor administrative change to the nomenclature of the 
containment sump trash racks and screens.
    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 includes the following changes that are 
considered to be administrative and/or editorial changes:
    A. Remove superseded references to 3876 megawatts thermal (MWt) 
and related information to this value from Unit 1 and Unit 3 
Operating Licenses and Unit 1, 2, and 3 Technical Specifications.
    This change is administrative. The change only removes the 
references to 3876 MWt and related information to this value and 
leaves the references to 3990 MWt.
    B. Remove references to Part Length Control Element Assemblies.
    This change is administrative because it only removes references 
to part length CEAs which have been replaced by part strength CEAs.
    C. Remove outdated pages and other references as a result of the 
CPC upgrade, and adjust the indentation of the logical connectors 
AND and OR in TS 3.2.4, between Required Actions B.1, B.2.1, and 
B.2.2.
    This change is administrative because it removes the redundant 
TS pages identified as ``(Before CPC Upgrade) or (Before CPCS 
Upgrade)'' and removes the reference to ``(After CPC Upgrade) or 
(After CPCS Upgrade)'' from various TS pages that will be renumbered 
and remain in place. The CPC upgrade has been completed. The 
adjustment of the indentation of the logical connectors AND and OR 
in TS 3.2.4 is consistent with the Action numbers and with TS 1.2.
    D. Change ``trash racks and screens'' to ``strainers.''
    This change is administrative. The change from ``trash racks and 
screens'' to ``strainers'' does not change the intent of the 
Surveillance Requirement 3.5.3.8 to verify, by visual inspection, 
that each [emergency core cooling system] ECCS train containment 
sump suction inlet is not restricted by debris and the suction inlet 
strainers show no evidence of structural distress or abnormal 
corrosion.
    E. Delete inspection requirements for Steam Generators (SG) with 
Alloy 600 MA tubes.
    This change is administrative because APS [Arizona Public 
Service Company] has completed the SG replacement project which 
removed all SGs containing Alloy 600 MA tubes.
    As discussed above, the proposed amendment involves 
administrative and/or editorial changes only. The proposed amendment 
does not impact any accident initiators, analyzed events, or assumed 
mitigation of accident or transient events. The proposed changes do 
not involve the addition or removal of any equipment or any design 
changes to the facility. The proposed changes do not affect any 
plant operations, design function, or analysis that verifies the 
capability of structures, systems, and components (SSCs) to perform 
a design function. The proposed changes do not change any of the 
accidents previously evaluated in the UFSAR [updated final safety 
analysis report]. The proposed changes do not affect SSCs, operating 
procedures, and administrative controls that have the function of 
preventing or mitigating any of these accidents.
    Therefore, the proposed changes do not represent a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    As stated in response to standard 1, the proposed amendment only 
involves administrative and/or editorial changes. No actual plant 
equipment or accident analyses will be affected by the proposed 
changes. The proposed changes will not change the design function or 
operation of any SSCs. The proposed changes will not result in any 
new failure mechanisms, malfunctions, or accident initiators not 
considered in the design and licensing bases. The proposed amendment 
does not impact any accident initiators, analyzed events, or assumed 
mitigation of accident or transient events. Therefore, this proposed 
change does not create the possibility of an accident of a new or 
different kind than previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    As stated in response to standard 1, the proposed amendment only 
involves administrative and/or editorial changes. The proposed 
change does not involve any physical changes to the plant or alter 
the manner in which plant systems are operated, maintained, 
modified, tested, or inspected. The proposed change does not alter 
the manner in which safety limits, limiting safety system settings 
or limiting conditions for operation are determined. The safety 
analysis

[[Page 4114]]

acceptance criteria are not affected by this change. The proposed 
change will not result in plant operation in a configuration outside 
the design basis. The proposed change does not adversely affect 
systems that respond to safely shutdown the plant and to maintain 
the plant in a safe shutdown condition. Therefore, the proposed 
change does not involve a significant reduction in a margin of 
safety.

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

Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of amendments request: October 27, 2009.
    Description of amendments request: The proposed amendments would 
modify technical specifications (TSs) requirements related to primary 
containment isolation instrumentation in accordance with the Nuclear 
Regulatory Commission-approved Technical Specification Task Force 
(TSTF), Improved Standard Technical Specifications change traveler, 
TSTF-306, Revision 2, ``Add action to LCO 3.3.6.1 to give option to 
isolate the penetration.'' The proposed amendment would revise TS 
Section 3.3.6.1, ``Primary Containment Isolation Instrumentation,'' by 
adding an ACTIONS note allowing intermittent opening, under 
administrative control, of penetration flow paths that are isolated. 
Additionally, the traversing in-core probe (TIP) system would be added 
as a separate isolation function with an associated Required Action to 
isolate the penetration within 24 hours rather than immediately 
initiating a unit shutdown.
    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 addition of the note that the penetration flow path may be 
unisolated under administrative control simply provides consistency 
with what is already allowed elsewhere in TSs. The isolation 
function of the TIP valves is mitigative, and does not create any 
increased possibility of an accident. Also, the operation of the 
manual shear valves is unaffected by this activity. The ability to 
manually isolate the TIP system by either the normal isolation ball 
valves or the shear valves would be unaffected by the inoperable 
instrumentation. The Required Actions and their associated 
Completion Times are not initiating conditions for any accident 
previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as result of the proposed changes. 
All systems, structures, and components previously required for the 
mitigation of a transient remain capable of fulfilling their 
intended design functions. The proposed changes have no adverse 
effects on any safety-related system or component and do not 
challenge the performance or integrity of any safety-related system. 
As a result no new failure modes are being introduced.
    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 will not affect the operation of plant 
equipment or the function of any equipment assumed in the accident 
analysis. The allowance to unisolate a penetration flow path will 
not have a significant effect on the margin of safety because the 
penetration flow path can be isolated manually, if needed. This 
change simply provides consistency with what is already allowed 
elsewhere in TSs. The option to isolate a TIP penetration will 
ensure the penetration will perform as designed in the accident 
analysis. The ability to manually isolate the TIP system is 
unaffected by the inoperable instrumentation. The proposed change 
does not impact any safety analysis assumptions or results.
    Therefore, the proposed change does not result in a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, NC 27602.
    NRC Branch Chief: Thomas H. Boyce.

Dominion Nuclear Connecticut Inc., et al., Docket No. 50-423, Millstone 
Power Station, Unit No. 3, New London County, Connecticut

    Date of amendment request: November 23, 2009.
    Description of amendment request: The proposed license amendment 
request would revise the Millstone Power Station, Unit 3 Technical 
Specification (TS) 6.8.4.g, ``Steam Generator Program,'' to exclude a 
portion of the tubes below the top of the steam generator tubesheet 
from periodic steam generator tube inspections. This request would also 
remove reference to the previous Cycle 13 interim alternate repair 
criteria.
    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 previously analyzed accidents are initiated by the failure 
of plant structures, systems, or components. The proposed change 
that alters the steam generator inspection criteria and the steam 
generator inspection reporting criteria does not have a detrimental 
impact on the integrity of any plant structure, system, or component 
that initiates an analyzed event. The proposed change will not alter 
the operation of, or otherwise increase the failure probability of 
any plant equipment that initiates an analyzed accident.
    Of the applicable accidents previously evaluated, the limiting 
transients with consideration to the proposed change to the steam 
generator tube inspection and repair criteria are the steam 
generator tube rupture (SGTR) event and the feedline break (FLB) 
postulated accidents.
    During the SGTR event, the required structural integrity margins 
of the steam generator tubes and the tube-to-tubesheet joint over 
the H* distance will be maintained. Tube rupture in tubes with 
cracks within the tubesheet is precluded by the constraint provided 
by the tube-to-tubesheet joint. This constraint results from the 
hydraulic expansion process, thermal expansion mismatch between the 
tube and tubesheet, and from the differential pressure between the 
primary and secondary side. Based on this design, the structural 
margins against burst, as discussed in Regulatory Guide (RG) 1.121, 
``Bases for Plugging

[[Page 4115]]

Degraded [pressurized-water reactor] PWR Steam Generator Tubes,'' 
are maintained for both normal and postulated accident conditions.
    The proposed change has no impact on the structural or leakage 
integrity of the portion of the tube outside of the tubesheet. The 
proposed change maintains structural integrity of the steam 
generator tubes and does not affect other systems, structures, 
components, or operational features. Therefore, the proposed change 
results in no significant increase in the probability of the 
occurrence of a SGTR accident.
    At normal operating pressures, leakage from primary water stress 
corrosion cracking below the proposed limited inspection depth is 
limited by both the tube-to-tubesheet crevice and the limited crack 
opening permitted by the tubesheet constraint. Consequently, 
negligible normal operating leakage is expected from cracks within 
the tubesheet region. The consequences of an SGTR event are affected 
by the primary-to secondary leakage flow during the event. However, 
primary-to-secondary leakage flow through a postulated broken tube 
is not affected by the proposed changes since the tubesheet enhances 
the tube integrity in the region of the hydraulic expansion by 
precluding tube deformation beyond its initial hydraulically 
expanded outside diameter. Therefore, the proposed changes do not 
result in a significant increase in the consequences of a SGTR.
    The consequences of a steam line break (SLB) are also not 
significantly affected by the proposed changes. During a SLB 
accident, the reduction in pressure above the tubesheet on the shell 
side of the steam generator creates an axially uniformly distributed 
load on the tubesheet due to the reactor coolant system pressure on 
the underside of the tubesheet. The resulting bending action 
constrains the tubes in the tubesheet thereby restricting primary-
to-secondary leakage below the midplane.
    Primary-to-secondary leakage from tube degradation in the 
tubesheet area during the limiting accident (i.e., a SLB) is limited 
by flow restrictions. These restrictions result from the crack and 
tube-to-tubesheet contact pressures that provide a restricted 
leakage path above the indications and also limit the degree of 
potential crack face opening as compared to free span indications.
    The leakage factor of 2.49 for Millstone Power Station Unit 3 
(MPS3), for a postulated SLB/FLB, has been calculated as shown in 
Table RA124-2 of Enclosure 5. The leakage factor of 2.49 is a 
bounding value for all steam generators, both hot and cold legs, in 
Table RA124-2. Specifically, for the condition monitoring (CM) 
assessment, the component of leakage from the prior cycle from below 
the H* distance will be multiplied by a factor of 2.49 and added to 
the total leakage from any other source and compared to the 
allowable accident induced leakage limit. For the operational 
assessment (OA), the difference in the leakage between the allowable 
accident induced leakage and the accident induced leakage from 
sources other than the tubesheet expansion region will be divided by 
2.49 and compared to the observed operational leakage.
    The probability of a SLB is unaffected by the potential failure 
of a steam generator tube as the failure of the tube is not an 
initiator for a SLB event. SLB leakage is limited by leakage flow 
restrictions resulting from the leakage path above potential cracks 
through the tube-to-tubesheet crevice. The leak rate during 
postulated accident conditions (including locked rotor) has been 
shown to remain within the accident analysis assumptions for all 
axial and or circumferentially orientated cracks occurring 13.1 
inches below the top of the tubesheet. The accident induced leak 
rate limit is 1.0 gpm. The technical specification (TS) operational 
leak rate is 150 gpd (0.1 gpm) through any one steam generator. 
Consequently, there is significant margin between accident leakage 
and allowable operational leakage. The SLB/FLB leak rate ratio is 
only 2.49 resulting in significant margin between the conservatively 
estimated accident leakage and the allowable accident leakage (1.0 
gpm).
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change that alters the steam generator inspection 
criteria and the steam generator inspection reporting criteria does 
not introduce any new equipment, create new failure modes for 
existing equipment, or create any new limiting single failures. 
Plant operation will not be altered, and all safety functions will 
continue to perform as previously assumed in accident analyses.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The proposed change that alters the steam generator inspection 
criteria and the steam generator inspection reporting criteria 
maintains the required structural margins of the steam generator 
tubes for both normal and accident conditions. Nuclear Energy 
Institute (NEI) 97-06, Revision 2, ``Steam Generator Program 
Guidelines'' and RG 1.121, are used as the bases in the development 
of the limited tubesheet inspection depth methodology for 
determining that steam generator tube integrity considerations are 
maintained within acceptable limits. RG 1.121 describes a method 
acceptable to the Nuclear Regulatory Commission (NRC) for meeting 
General Design Criteria (GDC) 14, ``Reactor Coolant Pressure 
Boundary,'' GDC 15, ``Reactor Coolant System Design,'' GDC 31, 
``Fracture Prevention of Reactor Coolant Pressure Boundary,'' and 
GDC 32, ``Inspection of Reactor Coolant Pressure Boundary,'' by 
reducing the probability and consequences of a SGTR. RG 1.121 
concludes that by determining the limiting safe conditions for tube 
wall degradation the probability and consequences of a SGTR are 
reduced. This RG uses safety factors on loads for tube burst that 
are consistent with the requirements of Section III of the American 
Society of Mechanical Engineers (ASME) Code.
    For axially oriented cracking located within the tubesheet, tube 
burst is precluded due to the presence of the tubesheet. For 
circumferentially oriented cracking, Westinghouse Electric Company, 
LLC (Westinghouse) report WCAP-1 7071 -P, ``H*: Alternate Repair 
Criteria for the Tubesheet Expansion Region in Steam Generators with 
Hydraulically Expanded Tubes (Model F),'' defines a length of 
degradation free expanded tubing that provides the necessary 
resistance to tube pullout due to the pressure induced forces, with 
applicable safety factors applied. Application of the limited hot 
and cold leg tubesheet inspection criteria will preclude 
unacceptable primary-to-secondary leakage during all plant 
conditions. The methodology for determining leakage provides for 
large margins between calculated and actual leakage values in the 
proposed limited tubesheet inspection depth criteria.
    Therefore, the proposed change does not involve a significant 
reduction in any 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: Lillian M. Cuoco, Senior Counsel, Dominion 
Resource Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219.
    NRC Branch Chief: Harold K. Chernoff.

Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point 
Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: November 19, 2009.
    Description of amendment request: The proposed change will correct 
identified non-conservatisms in Technical Specification 5.5.9 
``Ventilation Filter Testing Program'' by modifying the charcoal 
testing criteria to account for the 95% charcoal efficiency assumed for 
elemental iodine in the accident analyses for alternate source term.
    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 testing acceptance criteria for 
the existing Indian Point 2 Control Room filtration system in

[[Page 4116]]

Technical Specification (TS) 5.5.9 ``Ventilation Filter Testing 
Program'' to reflect current assumptions of iodine removal in 
accident dose calculations. The revised testing criteria does not 
add equipment or change the process for taking the test sample and 
only changes the test in the laboratory to be more restrictive. 
Therefore it cannot increase the probability of an accident 
occurring. The revised testing criteria is more stringent and 
therefore does not increase the consequences of an accident since it 
is more capable of mitigating control room doses and is consistent 
with existing analyses. 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 testing acceptance criteria 
for the existing Control Room filtration system. The proposed change 
does not involve installation of new equipment, modification of 
existing equipment, or result in a change to the way that the 
equipment or facility is operated so that no new equipment failure 
modes are introduced. 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 testing acceptance criteria 
for the existing Control Room filtration system. There is no change 
to the design requirements or the surveillance interval. The 
proposed change reflects the accident analysis dose calculation 
assumptions that assumed increased iodine removal. The factor of 
safety applied to the testing acceptance criteria remains the same. 
The new acceptance criterion is well within the system design 
capabilities. 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 Nuclear Operations, Inc., Docket No. 50-286, Indian Point 
Nuclear Generating Unit No. 3 (IP3), Westchester County, New York

    Date of amendment request: December 15, 2009, as supplemented on 
December 22, 2009, January 4, 2010, and January 11, 2010.
    Description of amendment request: The proposed amendment would 
allow a one-time extension of the 72-hour completion time of Technical 
Specification (TS) 3.7.5, Condition B, Action B.1 ``Restore AFW 
[auxiliary feedwater] train to OPERABLE status'' by 34 hours.
    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 allowed outage time (AOT) 
for the steam driven Auxiliary Boiler Feedwater Pump (ABFP) on a one 
time basis. Revising the AOT is not an accident initiator since an 
ABFP is a mitigating system. Therefore the proposed changes do not 
increase the probability of an accident occurring. The proposed AOT 
change is a one time increase that will allow repairs without the 
transient of shutdown. The plant is designed for single failure and 
recognizes that inoperability for short periods does not cause a 
significant increase in the consequences of an accident. The one 
time increase in this outage time is compensated with measures to 
reduce the potential need for the ABFP and the effects of events 
that could require the pump. Therefore the increase does not 
significantly increase the consequences of an accident. Therefore 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. The proposed change revises the allowed outage time for the 
ABFP on a one time basis. The proposed change does not involve 
installation of new equipment or modification of existing equipment, 
so no new equipment failure modes are introduced. The proposed 
revision is not a change to the way that the equipment or facility 
is operated or analyzed and 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 reduction in the margin of safety associated with 
continued IP3 operation with Auxiliary Boiler Feedwater (ABF) pump 
32 out of service during a 34 hour period beyond current allowed 
outage time is represented by an increase of approximately 50 
percent in the allowed outage time. This change in the margin of 
safety has been compensated for by specific compensatory measures to 
reduce the potential need for the pump and to address postulated 
events that could require the pump. The increase in core damage 
frequency (CDF) associated with continued IP3 operation with ABFP 32 
out of service for a duration of 106 hours which represents a 34 
hour period beyond the current allowed outage time is 3.9E-5 per 
reactor year (ry). This results in an incremental conditional core 
damage probability (ICCDP) of 4.8E-07, which is below the ICCDP 
guidance threshold of 5E-07 identified in NRC Inspection Manual Part 
9900. The ICCDP includes risk due to external events due to seismic, 
fire, and flood. The increase in large early release frequency 
(LERF) was estimated as 4.2E-7/ry (including external events), which 
results in an incremental conditional large early release 
probability (ICLERP) of 5.1E-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, 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 Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian 
Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New 
York

    Date of amendment request: November 17, 2009.
    Description of amendment request: The proposed change will correct 
identified non-conservatisms in the calculation of Emergency Diesel 
Generator (EDG) air receiver pressure requirements for Technical 
Specification (TS) 3.8.3. In addition, the proposed change will modify 
the number of normal EDG starts the air receiver is capable of 
providing as listed in the Final Safety Analysis Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed change revises the pressure at which the 
Emergency Diesel [G]enerator (EDG) air receiver is required to be 
kept to meet surveillance requirements, revises the minimum EDG air 
receiver pressure required for one start of the EDG, and changes the 
number of normal starts in the air receiver. Revising the air 
receiver upper and lower pressure limits and reducing the number of 
starts in the air

[[Page 4117]]

receiver are not accident initiators since an EDG is a mitigating 
system. Therefore the proposed changes do not increase the 
probability of an accident occurring. The proposed changes will 
assure that each EDG is capable of starting consistent with assumed 
accident analyses. These analyses assume that an EDG starts the 
first time and accident analyses do not credit subsequent starts. 
The proposed new TS limits on the EDG air receiver will assure that 
air pressure is adequate to assure one attempt to start the EDG is 
available at the lower limit and will provide additional normal 
starts at the upper pressure established in the surveillance. 
Establishing acceptance criteria that replace non conservative 
criteria and assure the design bases is met assures the capability 
of equipment to mitigate accident conditions. Therefore the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. The proposed change revises the pressure limit for the air 
receiver to initiate an alarm for low pressure, revises the lower 
pressure limit that must be maintained to assure that air is 
sufficient for at least one EDG start and revises the number of 
normal starts in the air receiver based on the revised calculations. 
The proposed change does not involve installation of new equipment 
or modification of existing equipment, so no new equipment failure 
modes are introduced. The proposed revision to the air receiver 
pressure limits and minimum air receiver EDG starts is also is [sic] 
not a change to the way that the equipment or facility is operated 
or analyzed and 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 conduct of surveillance tests, the conditions for 
failure of those tests and the number of EDG starts in the air 
receiver are means of assuring that the equipment is capable of 
maintaining the margin of safety established in the safety analyses 
for the facility. The proposed change in the EDG surveillance test 
acceptance criteria is consistent with values assumed in existing 
safety analyses which assume one start attempt for each EDG. The 
requirement for a minimum air pressure in the EDG air start receiver 
assures that there will be adequate air to allow at least one EDG 
start attempt which meets the intent of the existing TS. The 
reduction in the number of starts maintained in the air receiver 
does not affect the margins in accident analyses for this reason and 
because an EDG failure to start would reduce the air pressure below 
that required for one start before the overcrank timer would lock 
out a further start attempt. 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 Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: November 23, 2009.
    Description of amendment request: The proposed amendment would 
modify the Technical Specification (TS) 5.5.7, Inservice Testing 
Program, by replacing the references from the American Society of 
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code to the 
current code of record, the ASME Operation and Maintenance Nuclear 
Power Plants Code (ASME OM Code), the code of record for the James A. 
FitzPatrick Nuclear Power Plant (JAF) Inservice Testing Program for 
Inservice Testing Program. This is an administrative amendment to 
maintain the TS current with the NRC accepted code of record for JAF.
    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. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The proposed TS changes are non-technical, and are provided for 
consistency. There is no plant change involved, and thus, proposed 
TS changes do not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: No.
    The proposed TS changes are non-technical, i.e., there is no 
plant change involved, and thus, do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    The proposed TS changes are non-technical, i.e., there is no 
plant change involved. The changes are consistent with the 
regulations, and only update the TS to refer to the current code of 
reference. No design or safety margin is involved. Therefore, the 
proposed changes do not involve a significant reduction in any 
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.

Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, 
Comanche Peak Steam Electric Station (CPSES), Units 1 and 2, Somervell 
County, Texas

    Date of amendment request: October 26, 2009.
    Description of amendment request: The proposed change will revise 
Technical Specification (TS) 3.8.1 entitled ``AC Sources--Operating'' 
to extend, on a one-time basis, the allowable Completion Time (CT) of 
Required Action A.3 for one offsite circuit inoperable, from 72 hours 
to 14 days. This change is only applicable to startup transformer (ST) 
XST2 and will expire on March 1, 2011. This change is needed to allow 
sufficient time to make final terminations as part of a plant 
modification to facilitate connection of either ST XST2 or the spare ST 
to the Class 1E buses.
    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 will revise the CT for the loss of one 
offsite source from 72 hours to 14 days. The proposed one-time 
extension of the CT for the loss of one offsite power circuit does 
not significantly increase the probability of an accident previously 
evaluated. The startup transformers are not the initiator of any 
previously evaluated

[[Page 4118]]

accidents involving a loss of offsite power (LOOP).
    The TS will continue to require equipment that will power safety 
related equipment necessary to perform any required safety function. 
The one-time extension of the CT to 14 days does not affect the 
design of the STs, the interface of the STs with other plant 
systems, the operating characteristic of the STs, or the reliability 
of the STs.
    Per Regulatory Guide (RG) 1.177, the risk acceptance guideline 
presented in RG 1.174 shows that Unit 1 met all the risk acceptance 
guidelines for delta core damage frequency (CDF), delta large early 
release frequency (LERF), incremental conditional core damage 
probability (ICCDP), and incremental conditional large early release 
probability (ICLERP). [CPSES,] Unit 2 met the same risk acceptance 
guidelines of delta LERF and ICLERP; however, the delta CDF and 
ICCDP were above the acceptance value. Since the increase above the 
regulatory guidance is small, and the risk reduction measures 
quantitatively addressed, the values for Unit 2 delta CDF and ICCDP 
would fall below the regulatory guidance as well as decrease the 
other risk metrics for both Units.
    The consequence of a LOOP event has been evaluated in the CPNPP 
[Comanche Peak Steam Electric Station] Final Safety Analysis Report 
[ ] and the Station Blackout evaluation. Increasing the CT for one 
offsite power source on a one-time basis from 72 hours to 14 days 
does not increase the consequences of a LOOP event nor change the 
evaluation of LOOP events.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not result in a change in the manner in 
which the electrical distribution subsystems provide plant 
protection. The proposed change will only affect the time allowed to 
restore the operability of the offsite power source through a 
startup transformer. The proposed change does not affect the 
configuration or operation of the plant. The proposed change to the 
CT will facilitate installation of a plant modification which will 
improve plant design and will eliminate the necessity to shut down 
both Units if [ST] XST2 fails or requires maintenance that goes 
beyond the current TS CT of 72 hours. This change will improve the 
long-term reliability of the 345kV [kiloVolt] offsite circuit STs 
which are common to both CPNPP Units.
    There are no changes to the STs or the supporting systems 
operating characteristics or conditions. The change to the CT does 
not change any existing accident scenarios, nor create any new or 
different accident scenarios. In addition, the change does not 
impose any new or different requirements or eliminate any existing 
requirements. The change does not alter any of the assumptions made 
in the safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not affect the acceptance criteria for 
any analyzed event nor is there a change to any safety limit. The 
proposed change does not alter the manner in which safety limits, 
limiting safety system settings, or limiting conditions for 
operation are determined. Neither the safety analyses nor the safety 
analysis acceptance criteria are affected by this change. The 
proposed change will not result in plant operation in a 
configuration outside the current design basis. The proposed 
activity only increases, for a one-time pre-planned occurrence, the 
period when the plant may operate with one offsite power source. The 
margin of safety is maintained by maintaining the ability to safely 
shut down the plant and remove residual heat.
    Therefore, the proposed change does not involve a 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: Timothy P. Matthews, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: November 4, 2009.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to: (1) Delete TS 4.0.5, 
which pertains to surveillance requirements (SRs) for inservice 
inspection (ISI) and inservice testing (IST) of American Society of 
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) 
Class 1, 2 and 3 components; (2) add a new TS for the IST Program to 
Section 6.0, ``Administrative Controls,'' of the TSs; (3) change TSs 
that currently reference TS 4.0.5 to reference the IST Program or ISI 
Program, as applicable; and (4) revise TS 6.10.3.h to reflect the 
deletion of the ISI Program from the TSs. The new TS for the IST 
Program, TS 6.8.4.i, will indicate that the program will include 
testing frequencies applicable to the ASME Code for Operation and 
Maintenance of Nuclear Power Plants (OM Code), replacing the current 
reference to Section XI of the ASME Code specified in TS 4.0.5. In 
addition, TS 6.8.4.i would revise the requirements, currently contained 
in TS 4.0.5, regarding the applicability of the surveillance interval 
extension provisions of SR 4.0.2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes revise TS 4.0.5, Surveillance Requirements 
for Inservice Inspections and Testing of ASME Code Components, for 
consistency with 10 CFR 50.55a(f)(4) requirements regarding 
inservice testing of pumps and valves. The proposed change 
incorporates revisions to the ASME OM Code and clarifies testing 
frequency requirements for testing pumps and valves. The proposed 
change also relocates the ISI and IST Programs consistent with 
NUREG-1433. A commitment is made to maintain [Generic Letter (GL)] 
88-01 inspection requirements in the ISI Program.
    The proposed changes do not impact any accident initiators or 
analyzed events or assumed mitigation of accident or transient 
events. They do not involve the addition or removal of any 
equipment, or any design changes to the facility.
    Therefore, the proposed changes do not represent a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not involve a modification to the 
physical configuration of the plant (i.e., no new equipment will be 
installed) or change in the methods governing normal plant 
operation. The proposed change will not impose any new or different 
requirements or introduce a new accident initiator, accident 
precursor, or malfunction mechanism. Therefore, this proposed change 
does not create the possibility of an accident of a different kind 
than previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes revise and relocate TS 4.0.5, Surveillance 
Requirements for Inservice Inspections and Testing of ASME Code 
Components, for consistency with (1) the requirements of 10 CFR 
50.55a(f)(4) regarding the inservice testing of pumps and valves and 
(2) NUREG-1433. The proposed change updates references to the ASME 
OM Code, clarifies testing frequency requirements for testing pumps 
and valves, and relocates the IST Program to Section 6.0 of TS, and 
the ISI Program to a licensee controlled document. The safety 
function of the affected pumps and valves will be maintained; the 
programs will continue to be

[[Page 4119]]

implemented with the required regulations and codes. A commitment is 
made to maintain GL 88-01 inspection requirements in the ISI 
Program; there will be no change to these requirements.
    Therefore, this 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: Vincent Zabielski, PSEG Nuclear LLC-N21, 
P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Harold K. Chernoff.

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

    Date of amendment request: December 1, 2009.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to change the required 
frequency of testing control rod scram times from ``at least once per 
120 days of POWER OPERATION'' to ``at least once per 200 days of POWER 
OPERATION.'' This change is based on TS Task Force (TSTF) change 
traveler TSTF-460, Revision 0, ``Control Rod Scram Time Testing 
Frequency.'' TSTF-460 has been approved generically by the Nuclear 
Regulatory Commission (NRC) for incorporation into the boiling water 
reactor (BWR) Standard TS (STS); NUREG-1433 (BWR/4) and NUREG-1434 
(BWR/6). The NRC staff published a notice announcing the availability 
of this proposed TS change using the consolidated line item improvement 
process (CLIIP) in the Federal Register on August 23, 2004 (69 FR 
51864). Since Hope Creek Generating Station has not adopted the STS, 
the licensee has proposed variations from the CLIIP to ensure 
consistency with NUREG-1433, Revision 3, ``Standard Technical 
Specifications, General Electric Plants, BWR/4.'' The changes to align 
with NUREG-1433 involve the adoption of a revised control rod scram 
time test methodology and an establishment of a category of operable 
but ``slow'' control rods.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes extend the frequency and revise the 
evaluation methodology for control rod scram times, and identify a 
new category of ``slow'' control rods for assessing control rod 
operability. The frequency of control rod scram testing is not an 
initiator of any accident previously evaluated. The frequency of 
surveillance testing does not affect the ability to mitigate any 
accident previously evaluated, because the tested component is still 
required to be operable. The proposed evaluation methodology is 
consistent with industry approved methods and ensures control rod 
operability requirements for the number and distribution of 
operable, slow, and stuck control rods [and] continue[s] to satisfy 
scram reactivity rate assumptions used in plant safety analysis. 
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 changes do not involve any physical alteration of 
the plant (no new or different type of equipment is being installed) 
and do not involve a change in the design, normal configuration, or 
basic operation of the plant. The proposed changes do not introduce 
any new accident initiators. The proposed changes do not involve 
significant changes in the fundamental methods governing normal 
plant operation and do not require unusual or uncommon operator 
actions. The proposed changes provide assurance that the plant will 
not be operated in a mode or condition that violates the assumptions 
or initial conditions in the plant safety analyses and that 
[structures, systems and components] remain capable of performing 
their intended safety functions as assumed in the same analyses. 
Consequently, the response of the plant and the plant operator to 
postulated events will not be significantly different. Therefore, 
the proposed TS 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.
    Margin of safety is related to confidence in the ability of the 
fission product barriers to perform their design functions during 
and following an accident situation. The proposed changes address 
control rod scram test performance and acceptance criteria as well 
as control rod operability requirements. The scram test acceptance 
criteria and control rod operability restrictions are based on 
industry approved methodology and will continue to ensure control 
rod scram design functions and reactivity insertion assumptions used 
in plant safety analyses continue to be protected. The proposed 
changes also extend the frequency of testing control rod scram times 
while at-power from 120 days to 200 days. The proposed change 
continues to test the control rod scram time to ensure the 
assumptions in the plant safety analysis are protected. The 
demonstrated reliability of the control rod scram function justifies 
the extension of the surveillance frequency. Therefore, the proposed 
changes do 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: Vincent Zabielski, PSEG Nuclear LLC-N21, 
P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Harold K. Chernoff.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Unit Nos. 1 and 2, Louisa County, Virginia

    Date of amendment request: December 16, 2009.
    Description of amendment request: The amendment would revise the 
Technical Specifications (TS) to adopt Nuclear Regulatory Commission 
(NRC)-approved Revision 2 to Technical Specification Task Force (TSTF) 
Standard Technical Specification Change Traveler, TSTF-427, ``Allowance 
for Non Technical Specification Barrier Degradation on Support System 
Operability.'' The proposed amendment would modify the requirements for 
unavailable barriers by adding Limiting Condition for Operation 3.0.9.
    The NRC staff published a notice of opportunity for comment in the 
Federal Register on June 2, 2006 (71 FR 32145), on possible amendments 
adopting TSTF-427, including a model safety evaluation and model no 
significant hazards consideration (NSHC) Determination, using the 
consolidated line-item improvement process. The NRC staff subsequently 
issued a notice of availability of the models for referencing in 
license amendment applications in the Federal Register October 3, 2006 
(71 FR 58444). The licensee affirmed the applicability of the following 
NSHC determination in its application dated December 16, 2009.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:


[[Page 4120]]



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

    The proposed change allows a delay time for entering a supported 
system technical specification (TS) when the inoperability is due 
solely to an unavailable hazard barrier if risk is assessed and 
managed. The postulated initiating events which may require a 
functional barrier are limited to those with low frequencies of 
occurrence, and the overall TS system safety function would still be 
available for the majority of anticipated challenges. Therefore, the 
probability of an accident previously evaluated is not significantly 
increased, if at all. The consequences of an accident while relying 
on the allowance provided by proposed LCO 3.0.9 are no different 
than the consequences of an accident while relying on the TS 
required actions in effect without the allowance provided by 
proposed LCO 3.0.9. Therefore, the consequences of an accident 
previously evaluated are not significantly affected by this change. 
The addition of a requirement to assess and manage the risk 
introduced by this change will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

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

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Allowing delay times for entering supported system TS when 
inoperability is due solely to an unavailable hazard barrier, if 
risk is assessed and managed, will not introduce new failure modes 
or effects and will not, in the absence of other unrelated failures, 
lead to an accident whose consequences exceed the consequences of 
accidents previously evaluated. The addition of a requirement to 
assess and manage the risk introduced by this change will further 
minimize possible concerns. Thus, this change does not create the 
possibility of a new or different kind of accident from an accident 
previously evaluated.

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

    The proposed change allows a delay time for entering a supported 
system TS when the inoperability is due solely to an unavailable 
barrier, if risk is assessed and managed. The postulated initiating 
events which may require a functional barrier are limited to those 
with low frequencies of occurrence, and the overall TS system safety 
function would still be available for the majority of anticipated 
challenges. The risk impact of the proposed TS changes was assessed 
following the three-tiered approach recommended in RG 1.177. A 
bounding risk assessment was performed to justify the proposed TS 
changes. This application of LCO 3.0.9 is predicated upon the 
licensee's performance of a risk assessment and the management of 
plant risk. The net change to the margin of safety is insignificant 
as indicated by the anticipated low levels of associated risk (ICCDP 
[incremental conditional core damage probability] and ICLERP 
[incremental conditional large early release probability]) as shown 
in Table 1 of Section 3.1.1 in the Safety Evaluation [published in 
the Federal Register on October 3, 2006 (71 FR 58444)]. Therefore, 
this change does not involve a significant reduction in a margin of 
safety.

    Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 
23219.
    NRC Branch Chief: Gloria Kulesa.

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

    Date of amendment request: October 10, 2009.
    Description of amendment request: The proposed changes will revise 
Technical Specification (TS) 3.1.7, ``Rod Position Indication,'' TS 
3.2.1, ``Heat Flux Hot Channel Factor (F<INF>Q</INF>(Z)) (F<INF>Q</INF> 
Methodology),'' TS 3.2.2, ``Nuclear Enthalpy Rise Hot Channel Factor 
(F<SUP>N</SUP>[Delta]H), TS 3.2.4, ``Quadrant Power Tilt Ratio 
(QPTR),'' and TS 3.3.1, ``Reactor Trip System (RTS) Instrumentation,'' 
for use of the Best Estimate Analyzer for Core Operations--Nuclear 
(BEACON) Power Distribution Monitoring System (PDMS) described in WCAP-
12472-P-A, ``BEACON Core Monitoring and Operations Support System,'' to 
perform power distribution surveillances.
    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 PDMS performs continuous core power distribution monitoring 
with data input from existing plant instrumentation. This system 
utilizes an NRC [U.S. Nuclear Regulatory Commission] approved 
Westinghouse proprietary computer code, i.e., Best Estimate Analyzer 
for Core Operations--Nuclear (BEACON), to provide data reduction for 
incore flux maps, core parameter analysis, load follow operation 
simulation, and core prediction. The PDMS does not provide any 
protection or control system function. Fission product barriers are 
not impacted by these proposed changes. The proposed changes 
occurring with PDMS will not result in any additional challenges to 
plant equipment that could increase the probability of any 
previously evaluated accident. The changes associated with the PDMS 
do not affect plant systems such that their function in the control 
of radiological consequences is adversely affected. These proposed 
changes will therefore not affect the mitigation of the radiological 
consequences of any accident described in the Updated Safety 
Analysis Report (USAR).
    Use of the PDMS supports maintaining the core power distribution 
within required limits. Further continuous on-line monitoring 
through the use of PDMS provides significantly more information 
about the power distributions present in the core than is currently 
available. This results in more time (i.e., earlier determination of 
an adverse condition developing) for operator action prior to having 
an adverse condition develop that could lead to an accident 
condition or to unfavorable initial conditions for an accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Other than use of the PDMS to monitor core power distribution, 
implementation of the PDMS and associated Technical Specification 
changes has no impact on plant operations or safety, nor does it 
contribute in any way to the probability or consequences of an 
accident. No safety-related equipment, safety function, or plant 
operation will be altered as a result of this proposed change. The 
possibility for a new or different type of accident from any 
accident previously evaluated is not created since the changes 
associated with implementation of the PDMS do not result in a change 
to the design basis of any plant component or system. The evaluation 
of the effects of using the PDMS to monitor core power distribution 
parameters shows that all design standards and applicable safety 
criteria limits are met.
    The proposed changes do not result in any event previously 
deemed incredible being made credible. Implementation of the PDMS 
will not result in any additional adverse condition and will not 
result in any increase in the challenges to safety systems. The 
cycle-specific variables required by the PDMS are calculated using 
NRC-approved methods. The Technical Specifications will continue to 
require operation within the required core operating limits, and 
appropriate actions will continue to be taken when or if limits are 
exceeded.
    The proposed change, therefore, does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    No margin of safety is adversely affected by the implementation 
of the PDMS. The margins of safety provided by current Technical 
Specification requirements and limits remain unchanged, as the 
Technical

[[Page 4121]]

Specifications will continue to require operation within the core 
limits that are based on NRC-approved reload design methodologies. 
Appropriate measures exist to control the values of these cycle-
specific limits, and appropriate actions will continue to be 
specified and taken for when limits are violated. Such actions 
remain unchanged.
    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 20, 2009.
    Description of amendment request: The proposed changes will revise 
Technical Specification (TS) 3.8.1, ``AC [Alternating Current] 
Sources--Operating,'' by adding a Note to the Required Actions B.3.1 
and B.3.2 to indicate that the TS 3.8.1 Required Actions B.3.1 and 
B.3.2 are satisfied if the diesel generator (DG) became inoperable due 
to an inoperable support system, an independently testable component, 
or preplanned preventive maintenance or testing. The amendment also 
proposes to revise the Completion Times for Required Actions B.3.1 and 
B.3.2 to specify a Completion Time based on the discovery of an issue 
or failure of the DG.
    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.
    WCNOC [Wolf Creek Nuclear Operating Corporation] is proposing to 
add a Note to Required Actions B.3.1 and B.3.2 to indicate that the 
TS 3.8.1 Required Actions of B.3 are satisfied if the DG became 
inoperable due to an inoperable support system, an independently 
testable component or preplanned preventative maintenance or 
testing. The proposed change to the TS does not involve a change in 
the operational limits or physical design of the emergency power 
system. Diesel generator (DG) OPERABILITY and reliability will 
continue to be assured while minimizing the potential number of 
required DG starts. The DGs are not an initiator of any accident 
previously evaluated. As a result, the probability of any accident 
previously evaluated is 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 previously evaluated?
    Response: No.
    No new or different accidents result for implementing the 
proposed change. The change does not involve a physical alteration 
of the plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operations. The change does not alter assumptions made in the safety 
analysis for DG performance.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not impacted by this change. The proposed change will not result 
in operation in a configuration outside the design basis.
    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.

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 <a href="/cdn-cgi/l/email-protection#92e2f6e0bce0f7e1fde7e0f1f7d2fce0f1bcf5fde4"><span class="__cf_email__" data-cfemail="38485c4a164a5d4b574d4a5b5d78564a5b165f574e">[email protected]</span></a>.

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

    Date of application for amendment: August 26, 2009.
    Brief description of amendment: The proposed amendment would revise 
the Technical Specification (TS) Section 6.5 that governs 
administrative controls of High Radiation Areas (HRA) to incorporate 
the HRA administrative controls contained within the Standard Technical 
Specifications, NUREG-1433, Revision 3.
    Date of issuance: January 4, 2010.

[[Page 4122]]

    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 241.
    Facility Operating License No. DPR-28: Amendment revised the 
License and Technical Specifications.
    Date of initial notice in Federal Register: October 20, 2009 (74 FR 
53778).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated January 4, 2010.
    No significant hazards consideration comments received: No.

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: March 22, 2009.
    Brief description of amendments: The amendments revise the 
Technical Specification (TS) definition of the fully withdrawn position 
of the Rod Cluster Control Assemblies (RCCAs) to minimize localized 
RCCA wear. Previously, the fully withdrawn position for the RCCAs was 
defined in the TSs as being within the interval of 222 to 228 steps 
withdrawn (i.e., steps above rod bottom). The approved change allows 
the fully withdrawn position to be defined as being within the interval 
of 222 to 230 steps withdrawn.
    Date of issuance: January 12, 2010.
    Effective date: As of the date of issuance. The Salem Unit No. 1 
amendment shall be implemented prior to entering Mode 2 following 
refueling outage 1R20 (currently scheduled for spring 2010). The Salem 
Unit No. 2 amendment shall be implemented prior to entering Mode 2 
following refueling outage 2R18 (currently scheduled for spring 2011).
    Amendment Nos.: 292 and 276.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the TSs and the License.
    Date of initial notice in Federal Register: June 2, 2009 (74 FR 
26435).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 12, 2010.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 13th day of January 2010.

    For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2010-1315 Filed 1-25-10; 8:45 am]
BILLING CODE 7590-01-P


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