[Federal Register Volume 74, Number 229 (Tuesday, December 1, 2009)]
[Notices]
[Pages 62831-62840]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E9-28630]



[[Page 62831]]

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NUCLEAR REGULATORY COMMISSION

[NRC-2009-0518]


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 November 5, 2009, to November 18, 2009. The 
last biweekly notice was published on November 17, 2009 (74 FR 59259).

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 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 petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to

[[Page 62832]]

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, which the NRC 
promulgated in August 28, 2007 (72 FR 49139). The E-Filing process 
requires participants to submit and serve all adjudicatory documents 
over the internet, or in some cases to mail copies on electronic 
storage media. Participants may not submit paper copies of their 
filings unless they seek 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 petitioner/requestor 
should contact the Office of the Secretary by e-mail at 
[email protected], or by calling (301) 415-1677, to request (1) a 
digital ID certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and/or (2) 
creation of an electronic docket for the proceeding (even in instances 
in which the petitioner/requestor (or its counsel or representative) 
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms ViewerTM 
to access the Electronic Information Exchange (EIE), a component of the 
E-Filing system. The Workplace Forms ViewerTM is free and is 
available at http://www.nrc.gov/site-help/e-submittals/install-viewer.html. Information about applying for a digital ID certificate is 
available on NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html.
    Once a petitioner/requestor has obtained a digital ID certificate, 
had a docket created, and downloaded the EIE viewer, it can then submit 
a request for hearing or petition for leave to intervene. Submissions 
should be in Portable Document Format (PDF) in accordance with NRC 
guidance available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the 
time the filer submits its documents through EIE. To be timely, an 
electronic filing must be submitted to the EIE system no later than 
11:59 p.m. Eastern Time on the due date. Upon receipt of a 
transmission, the E-Filing system time-stamps the document and sends 
the submitter an e-mail notice confirming receipt of the document. The 
EIE system also distributes an e-mail notice that provides access to 
the document to the NRC Office of the General Counsel and any others 
who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the agency's adjudicatory e-
filing system may seek assistance through the ``Contact Us'' link 
located on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html or by calling the NRC Meta-System Help Desk, which is 
available between 8 a.m. and 8 p.m., Eastern Time, Monday through 
Friday, excluding government holidays. The Meta-System Help Desk can be 
contacted by telephone at 1-866-672-7640 or by e-mail at 
[email protected].
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, 
Maryland, 20852, Attention: Rulemaking and Adjudications Staff. 
Participants filing a document in this manner are responsible for 
serving the document on all other participants. Filing is considered 
complete by first-class mail as of the time of deposit in the mail, or 
by courier, express mail, or expedited delivery service upon depositing 
the document with the provider of the service.
    Non-timely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission, the presiding 
officer, or the Atomic Safety and Licensing Board that the request and/
or petition should be granted and/or the contentions should be 
admitted, based on a balancing of the factors specified in 10 CFR 
2.309(c)(1)(i)-(viii).
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant 
to an order of the Commission, an Atomic Safety and Licensing Board, or 
a Presiding Officer. Participants are requested not to include personal 
privacy information, such as Social Security numbers, home addresses, 
or home phone numbers in their filings, unless an NRC regulation or 
other law requires submission of such information. With respect to 
copyrighted works, except for limited excerpts that serve the purpose 
of the adjudicatory filings and would constitute a Fair Use 
application, participants are requested not to include copyrighted 
materials in their submissions.
    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].

[[Page 62833]]

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: September 28, 2009.
    Description of amendment request: The amendments would revise 
Required Action A.1 of Technical Specification (TS) 3.8.7, 
``Inverters--Operating,'' for the Palo Verde Nuclear Generating Station 
(PVNGS), Units 1, 2, and 3, by extending the Completion Time for 
restoration of an inoperable vital alternating current (AC) inverter 
from 24 hours to 7 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 proposed TS amendment does not affect the design of the 
vital AC inverters, the operational characteristics or function of 
the inverters, the interfaces between the inverters and other plant 
systems, or the reliability of the inverters. An inoperable vital AC 
inverter is not considered an initiator of an analyzed event. In 
addition, Required Actions and the associated Completion Times are 
not initiators of previously evaluated accidents. Extending the 
Completion Time for an inoperable vital AC inverter would not have a 
significant impact on the frequency of occurrence of an accident 
previously evaluated. The proposed amendment will not result in 
modifications to plant activities associated with inverter 
maintenance, but rather, provides operational flexibility by 
allowing additional time to perform inverter troubleshooting, 
corrective maintenance, and post-maintenance testing on-line.
    The proposed extension of the Completion Time for an inoperable 
vital AC inverter will not significantly affect the capability of 
the inverters to perform their safety function, which is to ensure 
an uninterruptible supply of 120-volt AC electrical power to the 
associated power distribution subsystems. An evaluation, using PRA 
[probabilistic risk assessment] methods, confirmed that the increase 
in plant risk associated with implementation of the proposed 
Completion Time extension is consistent with the NRC's Safety Goal 
Policy Statement, as further described in [NRC Regulatory Guide] RG 
1.174 and RG 1.177. In addition, a deterministic evaluation 
concluded that plant defense-in-depth philosophy will be maintained 
with the proposed Completion Time extension. Based on the above, the 
proposed amendment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment does not involve physical alteration of 
the PVNGS. No new equipment is being introduced, and installed 
equipment is not being operated in a new or different manner. There 
is no change being made to the parameters within which the PVNGS is 
operated. There are no setpoints at which protective or mitigating 
actions are initiated that are affected by this proposed action. The 
use of the alternate Class 1E power source for the vital AC 
instrument bus is consistent with the PVNGS plant design. The change 
does not alter assumptions made in the safety analysis. This 
proposed action will not alter the manner in which equipment 
operation is initiated, nor will the functional demands on credited 
equipment be changed. No alteration is proposed to the procedures 
that ensure the PVNGS remains within analyzed limits, and no change 
is being made to procedures relied upon to respond to an off-normal 
event. As such, no new failure modes are being introduced.
    Based on the above, the proposed amendment 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.
    Margins of safety are established in the design of components, 
the configuration of components to meet certain performance 
parameters, and in the establishment of setpoints to initiate alarms 
or actions. The proposed amendment does not alter the design or 
configuration of the vital AC inverters or their associated 120-volt 
AC subsystems, and does not alter the setpoints at which alarms and 
associated actions are initiated. With one of the required 120-volt 
AC vital instrumentation buses being powered from the alternate 
safety-related Class 1E power supply, which is backed by the 
divisional diesel generator (DG), there is no significant reduction 
in the margin of safety. Testing of the DGs and associated 
electrical distribution equipment provides confidence that the DGs 
will start and provide power to the associated equipment in the 
unlikely event of a loss of offsite power during the extended 7-day 
Completion Time.
    Applicable regulatory requirements will continue to be met, 
adequate defense-in-depth will be maintained, sufficient safety 
margins will be maintained, and any increases in risk are consistent 
with the NRC Safety Goal Policy Statement. Furthermore, during the 
proposed extended inverter Completion Time, any increases in risk 
posed by potential combinations of equipment out of service will be 
managed in accordance with the PVNGS site Configuration Risk 
Management Program, consistent with Paragraph (a)(4) of 10 CFR 
50.65, ``Requirements for monitoring the effectiveness of 
maintenance at nuclear power plants.''
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
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: August 18, 2009.
    Description of amendments request: The proposed license amendments 
revise Technical Specification 3.3.1.1, ``Reactor Protection System 
(RPS) Instrumentation,'' Surveillance Requirement 3.3.1.1.8, to 
increase the frequency interval between local power range monitor 
calibrations from 1100 megawatt-days per metric ton average core 
exposure (i.e., equivalent to approximately 907 effective full-power 
hours (EFPH)) to 2000 EFPH.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendments revise the surveillance interval for the 
LPRM [local power range monitor] calibration from 1100 MWD/T 
[megawatt days per metric ton] average core exposure to 2000 
effective full power hours (EFPH). Increasing the frequency interval 
between required LPRM calibrations is acceptable due to improvements 
in fuel analytical bases, core monitoring processes, and nuclear 
instrumentation. The revised surveillance interval continues to 
ensure that the LPRM detector signal will continue to be adequately 
calibrated.
    This change will not alter the operation of process variables, 
structures, systems, or components as described in the Updated Final 
Safety Analysis Report. The probability of an evaluated accident is 
derived from the probabilities of the individual precursors to that 
accident. The proposed change does not alter the initiation 
conditions or operational parameters for the LPRM subsystem and 
there is no new equipment introduced by the

[[Page 62834]]

extension of the LPRM calibration interval. The performance of the 
Average Power Range Monitor (APRM), Rod Block Monitor (RBM), and 
Oscillation Power Range Monitor (OPRM) systems is not affected by 
the proposed surveillance interval increase. The proposed LPRM 
calibration interval extension will have no significant effect on 
the Reactor Protection System (RPS) instrumentation accuracy during 
power maneuvers or transients and will, therefore, not significantly 
affect the performance of the RPS. As such, no individual precursors 
of an accident are affected and the proposed amendments do not 
increase the probability of a previously analyzed event.
    The radiological consequences of an accident can be affected by 
the thermal limits existing at the time of the postulated accident; 
however, increasing the surveillance interval frequency will not 
increase the calculated thermal limits since all uncertainties 
associated with the increased interval are currently implemented and 
are currently used to calculate the existing safety limits. Plant 
specific evaluation of LPRM sensitivity to exposure has determined 
that the extended calibration frequency increases the LPRM signal 
uncertainty value used in the SLMCPR [safety limit for minimum 
critical power] analysis; however, the increase is bounded by the 
values currently used in the safety analysis. Therefore, the thermal 
limit calculation is not significantly affected by LPRM calibration 
frequency, and thus the radiological consequences of any accident 
previously evaluated are not increased.
    Based on the above, the proposed amendments do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Creation of the possibility of a new or different kind of 
accident requires creating one or more new accident precursors. New 
accident precursors may be created by modifications of plant 
configuration, including changes in allowable modes of operation. 
The performance of the APRM, RBM, and OPRM systems are not affected 
by the proposed LPRM surveillance interval increase. The proposed 
change does not affect the control parameters governing unit 
operation or the response of plant equipment to transient 
conditions. For the proposed LPRM extended calibration interval 
frequency, all uncertainties remain less than the uncertainties 
assumed in the existing thermal limit calculations. The proposed 
change does not change or introduce any new equipment, modes of 
system operation, or failure mechanisms; therefore, no new accident 
precursors are created. Based on the above information, the proposed 
amendments do 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 has no impact on equipment design or 
fundamental operation, and there are no changes being made to safety 
limits or safety system allowable values that would adversely affect 
plant safety as a result of the proposed LPRM surveillance interval 
increase. The performance of the APRM, RBM, and OPRM systems are not 
affected by the proposed change. The margin of safety can be 
affected by the thermal limits existing at the time of the 
postulated accident; however, uncertainties associated with LPRM 
chamber exposure have no significant effect on the calculated 
thermal limits. Plant-specific evaluation of LPRM sensitivity to 
exposure has determined that the extended calibration frequency 
increases the LPRM signal uncertainty value used in the SLMCPR 
analysis; however, the increase is bounded by the values currently 
used in the safety analysis. The thermal limit calculation is not 
significantly affected since LPRM sensitivity with exposure is well 
defined. LPRM accuracy remains within that used to determine the 
total power uncertainty assumed in the thermal analysis basis, 
therefore maintaining thermal limits and the safety margin. The 
proposed change does not affect uncertainties or initial conditions 
assumed in the thermal limit calculations and therefore the margin 
of safety in the safety analyses is maintained. Based on the above 
information, the proposed amendments do 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.

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

    Date of amendment request: October 19, 2009.
    Description of amendment request: The proposed amendment relocates 
the Waterford Steam Electric Station, Unit 3 Steam Generator Level--
High trip requirements from Technical Specification Sections 2.2 and 3/
4.3.1 to the Technical Requirements Manual.
    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 relocates the Steam Generator Level--High 
Trip to a licensee-controlled document. The Steam Generator (SG) 
Level--High trip function is not credited in any DBA [design-basis 
accident] or transient analysis and is not an initiator to any 
accident analysis. As a result, neither the probability nor the 
consequences of an accident previously evaluated are significantly 
increased by this change.
    Therefore, this 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 relocates the Steam Generator Level--High 
trip function to a licensee-controlled document. The proposed change 
does not involve a physical alteration of the plant (no new or 
different type of equipment will be installed) or a change in the 
methods governing normal plant operation.
    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 relocates the Steam Generator Level--High 
trip function to a licensee-controlled document. This will allow 
changes to the Steam Generator Level--High Trip requirements 
currently in the Technical Specifications to be performed in 
accordance with the requirements of 10 CFR 50.59. As the Steam 
Generator Level--High trip function has been determined to not meet 
the definition of Technical Specifications or the criteria in 10 CFR 
50.36 (c)(2)(ii), lack of NRC review and approval prior to 
implementation for changes that are not determined to be a 
significant hazard will not lead to a significant reduction in the 
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: Terence A. Burke, Associate General 
Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Michael T. Markley.

[[Page 62835]]

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

    Date of amendment request: September 24, 2009.
    Description of amendment request: The amendment request proposes a 
one-time extension of the Completion Time (CT) to restore a unit-
specific essential service water train to operable status associated 
with Technical Specification Limiting Condition for Operation (LCO) 
3.7.8, Essential Service Water (SX) System, from 72 hours to 144 hours. 
The proposed change will only be used one time during the Byron Station 
Unit 2 spring 2010 refueling outage. The licensee is requesting an 
extension of the CT to 144 hours to replace two of the four SX pump 
suction isolation valves; maintenance history has shown that 
replacement of the SX pump suction isolation valves cannot be assured 
within the existing 72 hour CT window.
    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 have been evaluated using the risk-informed 
processes described in Regulatory Guide (RG) 1.174, ``An Approach 
for Using Probabilistic Risk Assessment in Risk-Informed Decisions 
on Plant-Specific Changes to the Licensing Basis,'' dated July 1998 
and RG 1.177, ``An Approach for Plant-Specific, Risk-Informed 
Decisionmaking: Technical Specifications,'' dated August 1998. In 
addition, proposed revised guidance as described in Draft Regulatory 
Guide DG-1226, ``An Approach for Using Probabilistic Risk Assessment 
in Risk-Informed Decisions on Plant-Specific Changes to the 
Licensing Basis,'' and Draft Regulatory Guide DG-1227, ``An Approach 
for Plant-Specific, Risk-Informed Decisionmaking: Technical 
Specifications,'' was reviewed for insights. The risk associated 
with the proposed changes was shown to be acceptable.
    The previously analyzed accidents are initiated by the failure 
of plant structures, systems, or components. The SX system is not 
considered an initiator for any of these previously analyzed events. 
The proposed change does not have a detrimental impact on the 
integrity of any plant structure, system, or component that 
initiates an analyzed event. No active or passive failure mechanisms 
that could lead to an accident are affected. The proposed change 
will not alter the operation of, or otherwise increase the failure 
probability of any plant equipment that initiates an analyzed 
accident. Therefore, the proposed change does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    The unit-specific SX system consists of two separate, 
electrically independent, 100% capacity, safety related, cooling 
water trains. Each train consists of a 100% capacity pump, piping, 
valving, and instrumentation. Normally, the pumps and valves are 
remotely and manually aligned. However, the pumps are automatically 
started upon receipt of a safety injection signal or an undervoltage 
on the engineered safety features (ESF) bus, and all essential 
valves are aligned to their post accident positions. The SX system 
is also the backup water supply to the auxiliary feedwater system 
and fire protection system.
    The design basis of the SX system is for one SX train, in 
conjunction with the component cooling water (CC) system and a 100% 
capacity containment cooling system, to remove core decay heat 
following a design basis LOCA [loss-of-coolant accident] as 
discussed in the UFSAR [updated final safety analysis report], 
Section 6.2, ``Containment Systems.'' This prevents the containment 
sump fluid from increasing in temperature during the recirculation 
phase following a LOCA and provides for a gradual reduction in the 
temperature of this fluid as it is supplied to the reactor coolant 
system by the emergency core cooling system pumps. The SX system is 
designed to perform its function with a single failure of any active 
component, assuming the loss of offsite power. The proposed one-time 
increase in the CT is consistent with the philosophy of the current 
Technical Specification LCO which allows one train of SX to be 
inoperable for 72 hours. This change only extends the 72 hour 
Completion Time to 144 hours which has been shown to be acceptable 
from a risk perspective; therefore, the proposed change does not 
involve a significant increase in the 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 the use or installation of 
new equipment and the currently installed equipment will not be 
operated in a new or different manner. No new or different system 
interactions are created and no new processes are introduced. The 
proposed changes will not introduce any new failure mechanisms, 
malfunctions, or accident initiators not already considered in the 
design and licensing bases. Based on this evaluation, 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 does not alter any existing setpoints at 
which protective actions are initiated and no new setpoints or 
protective actions are introduced. The design and operation of the 
SX system remains unchanged. The risk associated with the proposed 
increase in the time an SX pump is allowed to be inoperable was 
evaluated using the risk-informed processes described in RG 1.174, 
``An Approach for Using Probabilistic Risk Assessment in Risk-
Informed Decisions on Plant-Specific Changes to the Licensing 
Basis,'' dated July 1998 and RG 1.177, ``An Approach for Plant-
Specific, Risk-Informed Decisionmaking: Technical Specifications,'' 
dated August 1998. The risk was shown to be acceptable. Based on 
this evaluation, the proposed change does not involve a significant 
reduction in a margin of safety.

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

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334, 
Beaver Valley Power Station, Unit No. 1 (BVPS-1), Beaver County, 
Pennsylvania

    Date of amendment request: July 6, 2009.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 5.6.3, ``Core Operating Limits Report,'' 
to allow the use of the generically approved Topical Report, WCAP-
16009-P-A, ``Realistic Large Break LOCA [Loss-of-Coolant Accident] 
Evaluation Methodology Using Automated Statistical Treatment of 
Uncertainty Method (ASTRUM),'' for BVPS-1.
    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. No physical changes are required as a result of implementing 
the ASTRUM best-estimate large break [LOCA] methodology and 
associated technical specification changes. The plant conditions 
assumed in the analysis are bounded by the design conditions for all 
equipment in the plant. Therefore, there will be no increase in the 
probability of a LOCA. The consequences of a LOCA are not being 
increased, since it is shown that the emergency core cooling system 
is designed so that its calculated cooling performance conforms to 
the criteria contained in 10 CFR 50.46, Paragraph (b). No

[[Page 62836]]

other accident is potentially affected by this change.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. There are no physical changes being made to the plant. No 
new modes of plant operation are being introduced. The parameters 
assumed in the analysis are within the design limits of the existing 
plant equipment. All plant systems will perform as designed during 
the response to a potential accident.
    Therefore, the proposed change does not involve an increase in 
the probability or consequences of an accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The methodology used in the analysis would more 
realistically describe the expected behavior of plant systems during 
a postulated loss of coolant accident. Uncertainties have been 
accounted for as required by 10 CFR 50.46. A sufficient number of 
loss of coolant accidents with different break sizes, different 
locations and other variations in properties are analyzed to provide 
assurance that the most severe postulated LOCAs are calculated. As 
described in Section 3.3, there is a high level of probability that 
all criteria contained in 10 CFR 50.46, Paragraph (b) are met.

    The Nuclear Regulatory Commission (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 W. Jenkins, FirstEnergy Nuclear 
Operating Company, FirstEnergy Corporation, 76 South Main Street, 
Akron, OH 44308.
    NRC Branch Chief: Nancy L. Salgado.

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

    Date of amendment request: September 9, 2009.
    Description of amendment request: The proposed amendment would 
change the frequency of control rod notch testing, as specified in 
Technical Specification (TS) surveillance requirement (SR) 4.1.3.1.2.a, 
from at least once per 7 days to at least once per 31 days. The purpose 
of this SR is to confirm control rod insertion capability which is 
demonstrated by inserting each partially or fully withdrawn control rod 
at least one notch and observing that the control rod moves. This 
ensures that the control rod is not stuck and is free to insert on a 
scram signal. The proposed amendment would also add the word ``fully'' 
to the Action for TS Limiting Condition for Operation (LCO) 3.9.2 to 
clarify the requirement to fully insert all insertable control rods 
when the required source range monitor (SRM) instrumentation is 
inoperable. The licensee stated that the proposed amendment is based on 
Nuclear Regulatory Commission (NRC)-approved TS Task Force (TSTF) 
change, TSTF-475, Revision 1, ``Control Rod Notch Testing Frequency and 
SRM Insert Control Rod Action.'' The availability of this change to the 
Standard Technical Specifications (STS) was announced in the Federal 
Register on November 13, 2007 (72 FR 63935) as part of the consolidated 
line item improvement process. The Federal Register notice included a 
model safety evaluation, a model application and a model proposed a no 
significant hazards consideration (NSHC) determination. In its 
application dated September 9, 2009, the licensee affirmed the 
applicability of the proposed NSHC determination for TSTF-475 and has 
incorporated it by reference to satisfy the requirements of 10 CFR 
50.91(a). Since Hope Creek Generating Station has not adopted the STS 
(e.g., NUREG-1433), the licensee has proposed minor variations from the 
TS changes described in TSTF-475.
    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. The NRC staff's review is presented below.
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to SR 4.1.3.1.2.a reduces the frequency of 
control rod notch testing. Changing the frequency of testing is not 
expected to have any significant impact on the reliability of the 
control rods to insert as required on a scram signal. The proposed 
change to the Action for LCO 3.9.2 merely clarifies the intent of the 
action. There are no physical plant modifications associated with this 
change. The proposed amendment would not alter the way any structure, 
system, or component (SSC) functions and would not alter the way the 
plant is operated. As such, the proposed amendment would have no impact 
on the ability of the affected SSCs to either preclude or mitigate 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 proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment would not change the design function or 
operation of the SSCs involved and would not impact the way the plant 
is operated. As such, the proposed change would not introduce any new 
failure mechanisms, malfunctions, or accident initiators not already 
considered in the design and licensing bases. 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 margin of safety is associated with the confidence in the 
ability of the fission product barriers (i.e., fuel cladding, reactor 
coolant pressure boundary, and containment structure) to limit the 
level of radiation to the public. There are no physical plant 
modifications associated with the proposed amendment. The proposed 
amendment would not alter the way any SSC functions and would not alter 
the way the plant is operated. The proposed amendment would not 
introduce any new uncertainties or change any existing uncertainties 
associated with any safety limit. The proposed amendment would have no 
impact on the structural integrity of the fuel cladding, reactor 
coolant pressure boundary, or containment structure. Based on the above 
considerations, the NRC staff concludes that the proposed amendment 
would not degrade the confidence in the ability of the fission product 
barriers to limit the level of radiation to the public. Therefore, the 
proposed change does not involve a significant reduction in a margin of 
safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Vincent Zabielski, PSEG Nuclear LLC--N21, 
P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Harold K. Chernoff.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama

    Date of amendment request: October 20, 2009.

[[Page 62837]]

    Description of amendment request: The proposed amendment would 
delete paragraph d of Technical Specification 5.2.2, ``Unit Staff,'' 
superseded by Title 10 of the Code of Federal Regulations Part 26, 
Subpart I.
    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 proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change removes Technical Specification (TS) 
restrictions on working hours for personnel who perform safety 
related functions. The TS restrictions are superseded by the worker 
fatigue requirements in 10 CFR Part 26. The proposed change does not 
impact the physical configuration or function of plant structures, 
systems, or components (SSCs) or the manner in which SSCs are 
operated, maintained, modified, tested, or inspected. Worker fatigue 
is not an initiator of any accident previously evaluated. Worker 
fatigue is not an assumption in the consequence mitigation of any 
accident previously evaluated.
    Therefore, it is concluded that this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change removes TS restrictions on working hours for 
personnel who perform safety related functions. The TS restrictions 
are superseded by the worker fatigue requirements in 10 CFR Part 26. 
Working hours will continue to be controlled in accordance with NRC 
requirements. The new rule allows for deviations from controls to 
mitigate or prevent a condition adverse to safety or as necessary to 
maintain the security of the facility. This ensures that the new 
rule will not unnecessarily restrict working hours and thereby 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    The proposed change does not alter the plant configuration, 
require new plant equipment to be installed, alter accident analysis 
assumptions, add any initiators, or affect the function of plant 
systems or the manner in which systems are operated, maintained, 
modified, tested, or inspected.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change removes TS restrictions on working hours for 
personnel who perform safety related functions. The TS restrictions 
are superseded by the worker fatigue requirements in 10 CFR Part 26. 
The proposed change does not involve any physical changes to plant 
or alter the manner in which plant systems are operated, maintained, 
modified, tested, or inspected. The proposed change does not alter 
the manner in which safety limits, limiting safety system settings 
or limiting conditions for operation are determined. The safety 
analysis acceptance criteria are not affected by this change. The 
proposed change will not result in plant operation in a 
configuration outside the design basis. The proposed change does not 
adversely affect systems that respond to safely shut down the plant 
and to maintain the plant in a safe shut down condition.
    Removal of plant-specific TS administrative requirements will 
not reduce a margin of safety because the requirements in 10 CFR 
Part 26 are adequate to ensure that worker fatigue is managed.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Thomas Boyce.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: October 20, 2009.
    Description of amendment request: The proposed amendment would 
delete paragraph g of Technical Specification 6.2.2, ``Facility 
Staff,'' which was superseded by Title 10 of the Code of Federal 
Regulations (10 CFR), Part 26, Subpart I. This change is consistent 
with Nuclear Regulatory Commission approved Technical Specification 
Task Force (TSTF) Improved Standard Technical Specification Change 
Traveler TSTF-511, Revision 0, ``Eliminate Working Hour Restrictions 
from TS 5.2.2 to Support Compliance with 10 CFR Part 26.''
    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 proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change removes Technical Specification (TS) 
restrictions on working hours for personnel who perform safety 
related functions. The TS restrictions are superseded by the worker 
fatigue requirements in 10 CFR Part 26. The proposed change does not 
impact the physical configuration or function of plant structures, 
systems, or components (SSCs) or the manner in which SSCs are 
operated, maintained, modified, tested, or inspected. Worker fatigue 
is not an initiator of any accident previously evaluated. Worker 
fatigue is not an assumption in the consequence mitigation of any 
accident previously evaluated.
    Therefore, it is concluded that this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change removes TS restrictions on working hours for 
personnel who perform safety related functions. The TS restrictions 
are superseded by the worker fatigue requirements in 10 CFR Part 26. 
Working hours will continue to be controlled in accordance with NRC 
requirements. The new rule allows for deviations from controls to 
mitigate or prevent a condition adverse to safety or as necessary to 
maintain the security of the facility. This ensures that the new 
rule will not unnecessarily restrict working hours and thereby 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    The proposed change does not alter the plant configuration, 
require new plant equipment to be installed, alter accident analysis 
assumptions, add any initiators, or affect the function of plant 
systems or the manner in which systems are operated, maintained, 
modified, tested, or inspected.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change removes TS restrictions on working hours for 
personnel who perform safety related functions. The TS restrictions 
are superseded by the worker fatigue requirements in 10 CFR Part 26. 
The proposed change does not involve any physical changes to plant 
or alter the manner in which plant systems are operated, maintained, 
modified, tested, or inspected. The proposed change does not alter 
the manner in which safety limits, limiting safety system settings 
or limiting conditions for operation are determined. The safety 
analysis acceptance criteria are not affected by this change. The 
proposed change will not result in plant operation in a 
configuration outside the design basis. The proposed change does not 
adversely affect systems that respond to safely shut down the plant 
and to maintain the plant in a safe shutdown condition.
    Removal of plant specific TS administrative requirements will 
not reduce a margin of safety because the requirements

[[Page 62838]]

in 10 CFR Part 26 are adequate to ensure that worker fatigue is 
managed.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Thomas H. Boyce.

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: September 28, 2009.
    Description of amendment request: The proposed changes would 
address the filtration function of the Emergency Core Cooling System 
(ECCS) Pump Room Exhaust Air Cleanup System (PREACS) and are consistent 
with the associated design and licensing basis accident analysis 
assumptions. The proposed changes will add new Conditions B and C with 
associated Action Statements and Completion Times to Technical 
Specification (TS) 3.7.12 and modify Conditions A and D.
    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 license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    The proposed changes do not adversely affect accident initiators 
or precursors and do not alter the design assumptions, conditions, 
or configuration of the facility. The new conditions only affect the 
filtration function of ECCS PREACS, which is an accident mitigation 
function, so accident initiation probability is not impacted. 
Regarding significance of the proposed changes relative to the 
accident consequences, the new conditions remain consistent with 
existing design assumptions (i.e., dose calculations show that the 
filtration function is not required when ECCS leakage is less than 
the maximum allowable unfiltered leakage) and filtration is required 
to be operable as required to support the design analysis 
assumptions.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    The addition of the new Conditions B and C with associated 
Action Statements and Completion Times to TS 3.7.12 and modification 
of Condition D to address the filtration function of ECCS PREACS 
does not impact the accident analysis or associated assumptions. The 
new conditions only address actions to be taken when portions of 
ECCS PREACS (an accident mitigation system) is out-of-service.
    Therefore, the proposed changes do 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?
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. The proposed new conditions recognize that 
there may be limited leakage situations when filtration is not 
required to meet the accident analysis assumptions. Allowing safety 
equipment to be inoperable while it is not required is not reducing 
the analyzed margin of safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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, Esq., Senior Counsel, 
Dominion Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, 
Virginia 23219.
    NRC Branch Chief: Gloria J. Kulesa.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: October 16, 2009.
    Description of amendment request: The license amendment request 
(LAR) adds two references to the list of NRC approved methodologies 
contained in the Technical Specifications (TSs). Specifically, 
Westinghouse document WCAP-8745-P-A, ``Design Bases for Thermal 
Overpower Delta-T and Thermal Overtemperature Delta-T Trip Function,'' 
and the Dominion Fleet Report DOM-NAF-2-A, ``Reactor Core Thermal-
Hydraulics Using the VIPRE-D Computer Code,'' including Appendix B, 
``Qualification of the Westinghouse WRB-1 CHF [Critical Heat Flux] 
Correlation in the Dominion VIPRE-D Computer Code,'' in TS 6.2.C as a 
referenced analytical methodology 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 change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    Approval of the proposed changes will allow Dominion to use the 
VIPRE-D/WRB-1 and VIPRE-D/W-3 code/correlation pairs to perform 
licensing calculations of Westinghouse 15x15 Upgrade fuel in Surry 
cores, using the DDLs [Deterministic Design Limits] documented in 
Appendix B of the DOM-NAF-2-A Fleet Report and the SDL [Statistical 
Design Limit]. Neither the code/correlation pair nor the Statistical 
Departure from Nucleate Boiling Ratio (DNBR) Evaluation Methodology 
make any contribution to the potential accident initiators and thus 
cannot increase the probability of any accident. Further, since both 
the deterministic and statistical DNBR limits meet the required 
design basis of avoiding Departure from Nucleate Boiling (DNB) with 
95% probability at a 95% confidence level, the use of the new code/
correlation and the Statistical DNBR Evaluation Methodology do not 
increase the potential consequences of any accident. Finally, the 
full core DNB design limit provides increased assurance that the 
consequences of a postulated accident which includes radioactive 
release would be minimized because the overall number of rods in DNB 
would not exceed the 0.1% level. The pertinent evaluations to be 
performed as part of the cycle specific reload safety analysis to 
confirm that the existing safety analyses remain applicable have 
been performed and determined to be acceptable. The use of a 
different code/correlation pair will not increase the probability of 
an accident because plant systems will not be operated in a 
different manner, and system interfaces will not change. The use of 
the VIPRE-D/WRB-1 and VIPRE-D/W-3 code/correlation pairs to perform 
licensing calculations of Westinghouse 15x15 Upgrade fuel in Surry 
cores will not result in a measurable impact on normal operating 
plant releases and will not increase the predicted radiological 
consequences of accidents postulated in the UFSAR [Updated Final 
Safety Analysis Report].
    The remaining proposed changes are being made to enhance the 
completeness of the Surry TS and to achieve consistency with NUREG-
1431 Rev. 3. The proposed changes do not add or modify any plant 
systems, structures or components (SSCs). The proposed changes to 
relocate TS parameters to the COLR [Core Operating Limits Report] 
are programmatic and administrative in nature. These changes do not 
physically alter safety-related systems nor affect the way in which 
safety-related systems perform their functions. Additional Safety 
Limits on the DNB design basis and peak fuel centerline temperature 
are being imposed in TS 2.1, ``Safety Limit, Reactor Core,'' and the 
Reactor

[[Page 62839]]

Core Safety Limits figure is being relocated to the COLR. The 
additional Safety Limits are consistent with the values stated in 
the UFSAR and those being proposed herein. The proposed changes do 
not, by themselves, alter any of the relocated parameter limits. The 
removal of the cycle-specific parameter limits from the TS does not 
eliminate existing requirements to comply with the parameter limits. 
TS 6.2.C continues to ensure that the analytical methods used to 
determine the core operating limits meet NRC reviewed and approved 
methodologies and that applicable limits of the safety analyses are 
met. Deletion of the obsolete limits associated with N-1 loop 
operation (TS 2.1.A.2, TS 2.1.A.3, TS Figure 2.1-2, TS Figure 2.1-3) 
and fuel densification (TS figure 2.1-4) is acceptable since these 
limits no longer represent limiting conditions for operation and are 
not required to be in the Technical Specifications.
    Thus, the proposed changes do not affect initiators of analyzed 
events or assumed mitigation of accident or transient events. 
Therefore, this 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 changes do not involve a physical alteration of the 
plant (no new or different type of equipment will be installed).
    The use of VIPRE-D and its applicable fuel design limits for 
DNBR does not impact any of the applicable design criteria and all 
pertinent licensing basis criteria will continue to be met. 
Demonstrated adherence to these standards and criteria precludes new 
challenges to components and systems that could introduce a new type 
of accident. Setpoint safety analysis evaluations have demonstrated 
that the use of VIPRE-D is acceptable. Design and performance 
criteria will continue to be met and no new single failure 
mechanisms will be created. The use of the VIPRE-D code/correlation 
or the Statistical DNBR Evaluation Methodology does not involve any 
alteration to plant equipment or procedures that would introduce any 
new or unique operational modes or accident precursors.
    The proposed change adds a new surveillance requirement of RCS 
[Reactor Coolant System] Total Flow Rate and requests the addition 
of an already approved method for determining plant operating 
limits. The proposed change does not adversely affect accident 
initiators or precursors, nor does it alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of SSCs to perform their 
intended function to mitigate the consequences of an initiating 
event within the assumed acceptance limits.
    Thus, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    Response: No.
    The proposed changes to relocate TS parameters to the COLR are 
programmatic and administrative in nature. Additional Safety Limits 
on the DNB design basis and peak fuel centerline temperature are 
being imposed in TS 2.1, ``Safety Limit, Reactor Core,'' and the 
Reactor Core Safety Limits figure is being relocated to the COLR. 
The additional Safety Limits are consistent with the values stated 
in the UFSAR and those being proposed herein.
    Approval of the proposed changes will allow Dominion to use the 
VIPRE-D/WRB-1 and VIPRE-D/W-3 code/correlation pairs to perform 
licensing calculations of Westinghouse 15x15 Upgrade fuel in Surry 
cores, using the DDLs documented in Appendix B of the DOM-NAF-2-A 
Fleet Report and the SDL documented herein. The SDL has been 
developed in accordance with the Statistical DNBR Evaluation 
Methodology. The DNBR limits meet the design basis of avoiding DNB 
with 95% probability at a 95% confidence level. The use of the 
VIPRE-D/WRB-1 code/correlation provides the same margin to safety as 
the current code/correlation COBRA/WRB-1 used at Surry.
    Therefore, the proposed TS 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: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219.
    NRC Branch Chief: Gloria Kulesa.

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].

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

    Date of application for amendment: April 17, 2009.
    Brief description of amendment: The amendment revises Operating 
License No. DPR-49 by changing ``FPL Energy Duane Arnold, LLC'' to 
``NextEra Energy Duane Arnold, LLC,'' where appropriate, to reflect the 
renaming of FPL Energy Duane Arnold, LLC to NextEra Energy Duane 
Arnold, LLC.
    Date of issuance: November 13, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 275.
    Facility Operating License No. DPR-49: The amendment revised the 
License and Appendix B--Additional Conditions.
    Date of initial notice in Federal Register: June 30, 2009 (74 FR 
31324).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 13, 2009.
    No significant hazards consideration comments received: No.

[[Page 62840]]

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

    Date of amendment request: June 2, 2009.
    Brief description of amendment: The amendment (1) deleted Technical 
Specification (TS) surveillance requirement (SR) 3.1.3.2 and revised SR 
3.1.3.3, (2) removed reference to SR 3.1.3.2 from Required Action A.3 
of TS 3.1.3, ``Control Rod OPERABILITY,'' and (3) revised Example 1.4-3 
in Section 1.4, ``Frequency,'' to clarify the applicability of the 1.25 
surveillance test interval extension. The changes are in accordance 
with NRC-approved TS Task Force (TSTF) traveler TSTF-475, Revision 1, 
``Control Rod Notch Testing Frequency and SRM [Source Range Monitor] 
Insert Control Rod Action.''
    Date of issuance: November 12, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 235.
    Facility Operating License No. DPR-46: Amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: June 30, 2009 (74 FR 
31325).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 12, 2009.
    No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2, (SSES Units 1 and 2) Luzerne County, 
Pennsylvania

    Date of application for amendments: March 24, 2009, as supplemented 
by letters dated April 24, and September 11, 2009.
    Brief description of amendments: The change revised the allowable 
value in the Technical Specification (TS) Table 3.3.5.1-1 (Function 
3.d) for the high-pressure coolant injection automatic pump suction 
transfer from the condensate storage tank (CST) to the suppression 
pool. The present allowable value for this transfer is greater than or 
equal to 36 inches above the CST bottom. The change is to increase the 
allowable value for this transfer to occur at greater than or equal to 
40.5 inches above the CST bottom.
    Additionally, the amendment also included an editorial/
administrative change which corrected a typographical error in the SSES 
Units 1 and 2 TS Section 3.10.8.f.
    Date of issuance: November 9, 2009.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 254 for Unit 1 and 234 for Unit 2.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the License and Technical Specifications.
    Date of initial notice in Federal Register: October 6, 2009, (74 FR 
51332).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 9, 2009.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia

    Date of application for amendments: July 28, 2009, supplemented by 
letters dated September 16 and 30, 2009.
    Brief Description of amendments: These amendments revise the 
Technical Specifications (TS) of Surry Power Station, Units 1 and 2. 
The request proposed changes to the inspection scope and repair 
requirements of TS Section 6.4.Q, ``Steam Generator (SG) Program,'' to 
the reporting requirements of TS Section 6.6.A.3, ``Steam Generator 
(SG) Tube Inspection Report,'' and to TS Sections 4.13 and 3.1.C, ``RCS 
[Reactor Coolant System] Operational Leakage.'' The proposed changes 
would establish alternate repair inspection and criteria for portions 
of the SG tubes within the tubesheet. The alternate inspection and 
repair criteria would be applicable to Unit 1 during Refueling Outage 
23 (fall 2010) and the subsequent operating cycle and to Unit 2 during 
Refueling Outage 22 (fall 2009) and the subsequent operating cycle.
    Date of issuance: November 5, 2009.
    Effective date: Unit 1 is effective as of its date of issuance and 
shall be implemented by the end of the fall 2010 refueling outage. Unit 
2 is effective as of its date of issuance and shall be implemented by 
the end of the fall 2009 refueling outage.
    Amendment Nos.: 267 and 266.
    Renewed Facility Operating License Nos. DPR-32 and DPR-37: 
Amendments change the licenses and the technical specifications.
    Date of initial notice in Federal Register: August 19, 2009 (74 FR 
41939).
    The supplements dated September 16, 2009 and September 30, 2009, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated November 5, 2009.
    No significant hazards consideration comments received: No.

    Dated at Rockville, MD, this 19th day of November 2009.

    For The Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. E9-28630 Filed 11-30-09; 8:45 am]
BILLING CODE 7590-01-P