Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 10305-10315 [E9-4898]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 74, Number 45 (Tuesday, March 10, 2009)]
[Notices]
[Pages 10305-10315]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E9-4898]


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NUCLEAR REGULATORY COMMISSION

[NRC-2009-0100]


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 staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from February 12, 2009, to February 25, 2009. The 
last biweekly notice was published on February 24, 2009 (74 FR 8281).

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 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example, in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it

[[Page 10306]]

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, 
Directives and Editing Branch, 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. Copies of written comments 
received may be examined at the Commission'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 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, https://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, 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 set forth 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 participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, 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 on 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 a waiver in accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
five (5) days prior to the filing deadline, the petitioner/requestor 
must contact the Office of the Secretary by e-mail at 
hearingdocket@nrc.gov, 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 Viewer\TM\ to 
access the Electronic Information Exchange (EIE), a component of the E-
Filing system. The Workplace Forms Viewer\TM\ is free and is available 
at https://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 https://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 https://www.nrc.gov/
site-help/e-submittals.html. A filing is considered

[[Page 10307]]

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 may seek assistance through the 
``Contact Us'' link located on the NRC Web site at https://www.nrc.gov/
site-help/e-submittals.html or by calling the NRC electronic filing 
Help Desk, which is available between 8 a.m. and 8 p.m., Eastern Time, 
Monday through Friday, excluding government holidays. The electronic 
filing Help Desk can be contacted by telephone at 1-866-672-7640 or by 
e-mail at MSHD.Resource@nrc.gov.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file a motion, 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 petition 
and/or request 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 
https://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant 
to an order of the Commission, the 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 submission.
    For further details with respect to this amendment action, 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 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

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

    Date of amendment request: January 15, 2009.
    Description of amendment request: The amendments would modify 
Technical Specifications (TSs) 3.3.10, 3.6.7, and 5.6.6 to delete the 
requirements related to hydrogen recombiners and hydrogen monitors. The 
proposed TS changes would support implementation of the revisions to 10 
CFR 50.44, ``Standards for Combustible Gas Control System in Light-
Water-Cooled Power Reactors,'' that became effective on October 16, 
2003. The proposed changes are consistent with Revision 1 of the NRC-
approved Industry/Technical Specification Task Force (TSTF) Standard 
Technical Specification Change Traveler, TSTF-447, ``Elimination of 
Hydrogen Recombiners and Change to Hydrogen and Oxygen Monitors.''
    The NRC staff issued a notice of opportunity for public comments on 
TSTF-447, Revision 1, published in the Federal Register on August 2, 
2002 (67 FR 50374), soliciting comments on a model safety evaluation 
(SE) and a model no significant hazards consideration (NSHC) 
determination for the elimination of requirements for hydrogen 
recombiners, and hydrogen and oxygen monitors from TS. Based on its 
evaluation of the public comments received, the NRC staff made 
appropriate changes to the models and included final versions in a 
notice of availability published in the Federal Register on September 
25, 2003 (68 FR 55416), regarding the adoption of TSTF-447, Revision 1, 
as part of the NRC's consolidated line item improvement process 
(CLIIP).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC adopted by the licensee is presented below:
    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated
    The revised 10 CFR 50.44 no longer defines a design-basis loss-of-
coolant accident (LOCA) hydrogen release, and eliminates requirements 
for hydrogen control systems to mitigate such a release. The 
installation of hydrogen recombiners and/or vent and purge systems 
required by 10 CFR 50.44(b)(3) was intended to address the limited 
quantity and rate of hydrogen generation that was postulated from a 
design-basis LOCA. The Commission has found that this hydrogen release 
is not risk-significant because the design-basis LOCA hydrogen release 
does not contribute to the conditional probability of a large release 
up to approximately 24 hours after the onset of core damage. In 
addition, these systems were ineffective at mitigating hydrogen 
releases from risk-significant accident sequences that could threaten 
containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen monitors are no longer required to mitigate design-basis 
accidents and, therefore, the hydrogen monitors do not meet the 
definition of a safety-related component as defined in 10 CFR 50.2. RG 
[Regulatory Guide] 1.97 Category 1 is intended for key variables that 
most directly indicate the accomplishment of

[[Page 10308]]

a safety function for design-basis accident events. The hydrogen 
monitors no longer meet the definition of Category 1 in RG 1.97. As 
part of the rulemaking to revise 10 CFR 50.44 the Commission found that 
Category 3, as defined in RG 1.97, is an appropriate categorization for 
the hydrogen monitors because the monitors are required to diagnose the 
course of beyond design-basis accidents.
    The regulatory requirements for the hydrogen monitors can be 
relaxed without degrading the plant emergency response. The emergency 
response, in this sense, refers to the methodologies used in 
ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite releases 
of radioactivity, and establishing protective action recommendations to 
be communicated to offsite authorities. Classification of the hydrogen 
monitors as Category 3 and removal of the hydrogen monitors from TS 
will not prevent an accident management strategy through the use of the 
SAMGs [severe accident management guidelines], the emergency plan (EP), 
the emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of the hydrogen recombiner requirements 
and relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, does not involve a significant increase 
in the probability or the consequences of any 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 elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal of 
these requirements from TS, will not result in any failure mode not 
previously analyzed. The hydrogen recombiner and hydrogen monitor 
equipment was intended to mitigate a design-basis hydrogen release. The 
hydrogen recombiner and hydrogen monitor equipment are not considered 
accident precursors, nor does their existence or elimination have any 
adverse impact on the pre-accident state of the reactor core or post 
accident confinement of radionuclides within the containment building.
    Therefore, this change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety
    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal of 
these requirements from TS, in light of existing plant equipment, 
instrumentation, procedures, and programs that provide effective 
mitigation of and recovery from reactor accidents, results in a neutral 
impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that this hydrogen 
release is not risk-significant because the design-basis LOCA hydrogen 
release does not contribute to the conditional probability of a large 
release up to approximately 24 hours after the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the [Three Mile Island], Unit 2 
accident, can be adequately met without reliance on safety-related 
hydrogen monitors.
    Therefore, this change does not involve a significant reduction in 
the margin of safety. Removal of hydrogen monitoring from TS will not 
result in a significant reduction in their functionality, reliability, 
and availability.
    The NRC staff has reviewed the analysis adopted by the licensee 
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 request for amendments involves NSHC.
    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.

Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1, Dauphin County, Pennsylvania

    Date of amendment request: September 29, 2008.
    Description of amendment request: The proposed changes would revise 
the TMI-1 technical specifications (TSs) to reflect design changes 
resulting from the planned control rod drive control system (CRDCS) 
digital upgrade project. In addition, the proposed amendment would 
revise the TS to remove all references to the axial power shaping rods 
(APSRs) to reflect changes resulting from their proposed elimination 
from the TMI-1 reactor.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below, with U.S. Nuclear Regulatory 
Commission (NRC) staff edits in brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed license amendment modifies the Technical 
Specifications (TSs) to incorporate new TS requirements associated 
with the new Digital Control Rod Drive Control System (DCRDCS) and 
an evaluation to permanently remove the Axial Power Shaping Rods 
(APSRs) from the reactor core.
    The proposed license amendment will continue to ensure 
reliability and operability of the control rod drive Reactor Trip 
Breakers (RTBs) to perform their safety function of tripping the 
reactor. The existing channel independence, separation and 
performance requirements of the RTBs and the Reactor Protection 
System (RPS) response time are retained for the new configuration. 
The RTB design was reviewed for credible common mode failures and no 
credible common mode failures were identified that would prevent the 
breakers from performing the reactor trip function. Reliable RTBs 
and their associated support circuitry provide assurance that a 
reactor trip will occur when initiated. The planned DCRDCS 
modification upgrades the relay-based Control Rod Drive Control 
System (CRDCS) to a solid state programmable DCRDCS using single rod 
power supplies assigned to each of the 61 Control Rod Drives (CRDs). 
The new components will meet the same design requirements (i.e., 
seismic, environmental, quality, separation, single failure 
criteria) as the existing components in the CRDCS/RPS interface. The 
DCRDCS modification will improve the reliability of the system by 
resolving age-related degradation issues and replacing obsolete 
equipment.
    Malfunction of the CRD control system (or operator error) is an 
initiator of the startup and rod withdrawal accidents. The new 
DCRDCS meets the design requirements of the original system 
including redundancy of critical functions, isolation from safety 
related systems, reactivity rate limit, and single failure 
requirements. Electrical ratings, heat loading, structural and 
environmental aspects have been verified to be acceptable. 
Therefore, there is no increase in the frequency of occurrence or 
probability of a malfunction of equipment important to safety. The 
DCRDCS is not required for accident mitigation, post accident 
response

[[Page 10309]]

or offsite release mitigation. The action of the RPS to trip the 
RTBs, to remove power from the control rods, and drop the rods into 
the core, remains independent of the DCRDCS. Therefore, there is no 
increase in the consequences or probability of occurrence of an 
accident previously evaluated.
    The modified Diverse Scram System (DSS) design utilizes the same 
power sources as the existing DSS, which are independent of reactor 
trip (i.e., RPS) related power sources. There is no change to the 
DSS logic circuitry. The DSS sensors and trip setpoint remains 
unchanged. Updated Final Safety Analysis (UFSAR) Section 7.1.5.4 
indicates that: ``The DSS provides an independent method of 
automatically tripping the reactor in the event the RPS related 
reactor trip system fails. It is designed in accordance with the 
Anticipated Transient Without Scram (ATWS) rule and, as such, its 
critical features are independence and diversity from the reactor 
trip system and emphasis on not failing in a tripped state.'' 
However, DSS is not safety related and is not credited in any safety 
analysis in UFSAR Chapter 14, ``Safety Analysis.'' The assumed DSS 
response time increase from 1.0 second to 2.0 seconds has been 
evaluated and the results of the analysis concluded that the 
original acceptance criteria are maintained. Therefore, the proposed 
change to the DSS [is not adverse and] does not increase the 
consequence of an ATWS event.
    The proposed license amendment will continue to ensure the 
reliability and operation of the reactor core. Analyses have shown 
that the core designs employed at TMI-1 are stable with respect to 
axial oscillations and that xenon oscillations initiated during 
power transients are naturally damped or can be manually suppressed 
using regulating control rods (i.e., Control Rod Group 7 (CRG-7)). 
Actual operating experience at TMI-1 bears out the analysis 
conclusions that adequate axial imbalance control can be maintained 
using coordinated movements of CRG-7 [and] timed water additions. A 
review of the TMI-1 safety analyses found no mention or credit for 
APSRs in any of the events analyzed for TMI-1, and safety analysis 
assumptions are verified to bound key core parameters for each 
reload with explicit accounting for the presence of (or lack of) 
APSRs in the core. Therefore, there is no affect of APSRs on 
transient analyses, as APSR positions do not change in the event of 
a reactor trip.
    Therefore, the proposed changes do 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 systems affected by implementing the proposed changes to the 
TS are not assumed to initiate design basis accidents. Rather, the 
CRDCS/RPS interface (i.e., RTBs) is used to mitigate the 
consequences of an accident that has already occurred. The proposed 
TS changes do not affect the mitigating function of this system. The 
failure of any one RTB will not inhibit the reactor trip function. 
The modification interfaces with the DSS, which mitigates the ATWS 
event, but the interface function remains the same.
    A Failure Modes and Effects Analysis (FMEA) was performed on the 
DCRDCS design to determine if adverse effects (i.e., loss of reactor 
control, uncontrolled rod withdrawal, reactor trip, or prevention of 
reactor trip) could result from the credible failure of a single 
component. The FMEA concluded that no credible single component 
failure would cause a total loss of reactor control, an uncontrolled 
rod withdrawal, a reactor trip, or prevent a reactor trip. All 
operation critical to the safe and effective performance of the 
DCRDCS maintained sufficient redundancy such that no credible single 
failure could compromise the design functionality.
    The APSRs' original function was to control any reactor core 
tendency towards axial oscillations resulting from xenon 
instabilities that could occur for certain early reactor core 
designs (i.e., rodded core designs). More recent non-rodded feed-
and-bleed core designs have been shown to be self-dampened with 
respect to axial xenon oscillations such that APSRs have not been 
moved at TMI-1 for axial power control since 1994, and have been 
withdrawn from the reactor core since Fall 2005 with Core Operating 
Limits Report limits preventing insertion, consistent with AREVA 
reload methods.
    Use of [CRG-7] has been shown to adequately suppress axial xenon 
oscillations.
    The proposed changes to the CRDCS and APSRs and associated TS 
changes do not introduce any new accident initiators, nor do they 
reduce or adversely affect the capabilities of any plant structure, 
system, or component to perform their safety function.
    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?
    Response: No.
    The proposed TS changes do not adversely impact any plant safety 
limits, setpoints, response times, or design parameters. The changes 
do not negatively affect the fuel, fuel cladding, reactor coolant 
system, or containment integrity [under normal, transient or 
accident conditions].
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1, (TMI-1) Dauphin County, Pennsylvania

    Date of amendment request: November 6, 2008.
    Description of amendment request: The proposed amendment would 
modify the TMI-1 Technical Specifications (TS), to replace the current 
limits on primary coolant gross specific activity with limits on 
primary coolant noble gas activity. The noble gas activity would be 
based on dose equivalent Xenon-133 (DEX) and would take into account 
only the noble gas activity in the primary coolant. The completion time 
for DEX being out of specification would be increased to match the 
action time requirements for the dose equivalent Iodine-131 (DEI) 
specification. In addition, the current DEI definition would be revised 
to allow the use of additional options for determining thyroid dose 
conversion factors. This change was proposed by the industry's 
Technical Specification Task Force (TSTF) and is designated TSTF-490. 
The NRC staff issued a notice of opportunity for comment in the Federal 
Register on November 20, 2006 (71 FR 67170), on possible amendments 
concerning TSTF-490, including a model safety evaluation and model no 
significant hazards (NSHC) determination, using the consolidated line 
item improvement process (CLIIP). The NRC staff subsequently issued a 
notice of availability of the models for referencing in license 
amendment applications in the Federal Register on March 15, 2007 (72 FR 
12217). The licensee affirmed the applicability of the following NSHC 
determination in its application dated November 6, 2008.
    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:

    Criterion 1: The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated
    Reactor coolant specific activity is not an initiator for any 
accident previously evaluated. The Completion Time when primary 
coolant gross activity is not within limit is not an initiator for 
any accident previously evaluated. The current variable limit on 
primary coolant iodine concentration is not an initiator to any 
accident previously evaluated. As a result, the proposed change does 
not significantly increase the probability of an accident. The 
proposed change will limit primary coolant noble gases to 
concentrations consistent with the accident analyses. The proposed 
change to the Completion Time has no impact on the

[[Page 10310]]

consequences of any design basis accident since the consequences of 
an accident during the extended Completion Time are the same as the 
consequences of an accident during the Completion Time. As a result, 
the consequences of any accident previously evaluated are not 
significantly increased.
    Criterion 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 in specific activity limits does not alter 
any physical part of the plant nor does it affect any plant 
operating parameter. The change does not create the potential for a 
new or different kind of accident from any previously calculated.
    Criterion 3: The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety
    The proposed change revises the limits on noble gas 
radioactivity in the primary coolant. The proposed change is 
consistent with the assumptions in the safety analyses and will 
ensure the monitored values protect the initial assumptions in the 
safety analyses. Based upon the reasoning presented above, the 
requested change does not involve a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the 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: J. Bradley Fewell, Esquire, Associate 
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Harold K. Chernoff.

Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1, Dauphin County, Pennsylvania

    Date of amendment request: October 9, 2008.
    Description of amendment request: The proposed changes would revise 
the existing Three Mile Island (TMI), Unit 1, technical specifications 
(TSs) relating to the steam generator (SG) tube surveillance program. 
The proposed changes reflect the planned installation of replacement 
SGs and specifically address the new thermally treated Alloy 690 tubing 
design of the replacement SGs. Removal of sections of the TSs that are 
not applicable to the replacement SGs are proposed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below, with U.S. Nuclear Regulatory 
Commission (NRC) staff edits in brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to the Technical Specifications (TSs) for 
the TMI, Unit 1 Steam Generator (SG) Program recognize that the TMI, 
Unit 1 SGs are being replaced and the standard industry performance 
criteria documented in [Technical Specification Task Force (TSTF) 
Traveler,] TSTF-449[,] for Alloy 690-tubed SGs will apply. These 
changes eliminate criteria that were established to reflect the 
condition and materials of the current TMI, Unit 1 SGs, and add the 
requirements for inspection of Alloy 690-tubed SGs from TSTF-449.
    With these proposed TS changes, the operational primary-to-
secondary leakage rate limit established for the original TMI, Unit 
1 SGs is replaced with the standard industry primary-to-secondary 
leakage rate limit. The standard industry limit is that limit 
provided in TSTF-449. The current, reduced limit in the TMI, Unit 1 
TS was implemented in response to upper tubesheet tube expansion 
degradation, and repairs, in the original TMI, Unit 1 SGs. A reduced 
limit is not required for the replacement SGs since they are 
fabricated from advanced materials and [will not be] subjected to 
the degradation mechanisms that influenced the original TMI, Unit 1 
SGs. Thus, reverting to the standard industry limit is appropriate. 
The slightly higher, industry standard, leak rate limit is still low 
enough to provide assurance that the probability of tube ruptures, 
or of rapidly propagating tube leaks, remains acceptably low. Thus, 
the probability of a previously evaluated accident is not increased.
    The installation of the new SGs, with improved materials, will 
decrease the consequences of SG related accidents. The removal of 
accident-induced leakage attributable to the current degradation 
mechanisms from TS 6.19.c.1.b [provides a reduction in the] accident 
induced leakage limit to 1 gpm per SG. SG accident-induced leakage 
is proportional to dose; a lower accident-induced leakage limit will 
result in a lower dose than previously evaluated accident 
consequences.
    The proposed change to replace the 90-day report with a report 
required within 180 days is a change to an administrative 
requirement and does not affect the probability or consequences of 
an accident. The 180-day period is now industry ``standard'' 
practice per TSTF-449.
    These changes continue to provide reasonable assurance that the 
SG tubing will retain integrity over the full range of operating 
conditions (including startup, operation in the power range, hot 
standby, cooldown and all anticipated transients included in the 
design specification). With the proposed changes, the SG performance 
criteria (based on tube structural integrity, accident-induced 
leakage, and operational leakage) and SG Program are updated to 
reflect the replacement SGs while remaining consistent with TSTF-
449.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident that was 
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 TS changes recognize an improvement in SG design as 
a result of SG replacement. The replacement SGs contain a number of 
design improvements with respect to the plant's original SGs. 
However, even with the design improvements, the replacement SGs are 
very similar to the original SGs and new types of accidents are not 
created. There are no new design functions for the Alloy 690 tubing 
in the replacement SGs. The proposed new leakage and inspection 
requirements are the standard industry requirements for Alloy 690 
tubing.
    Primary-to-secondary leakage monitoring equipment is not 
affected by the proposed changes, and primary-to-secondary leakage 
will continue to be monitored to ensure it remains within current 
accident analysis assumptions and limits. The proposed changes 
implement the industry ``standard'' TSTF-449 primary-to-secondary 
leak limits for the plant's Alloy 690-tubed replacement SGs. No new 
types of primary-to-secondary leak accidents are created.
    The proposed change to replace the 90-day report with a report 
required within 180 days is a change to an administrative 
requirement and does not create a new or different kind of accident. 
The 180-day period is now industry ``standard'' practice per TSTF-
449.
    Therefore, the proposed changes do not create the possibility of 
a new or different type of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The SG tubes in pressurized water reactors [PWRS] are an 
integral part of the reactor coolant pressure boundary and, as such, 
are relied upon to maintain the primary system's pressure and 
inventory. As part of the reactor coolant pressure boundary, the SG 
tubes are unique in that they are also relied upon as a heat 
transfer surface between the primary and secondary systems such that 
residual heat can be removed from the primary system. The SG tubes 
also isolate the radioactive fission products in the primary coolant 
from the secondary system. In summary, the safety function of a SG 
is maintained by ensuring the integrity of its tubes.
    SG tube integrity is a function of the design, environment, and 
physical condition of the tubing. The proposed changes do not affect 
the operating environment but do recognize the improved tube 
material as a result of replacing the SGs. The proposed TS changes 
for inspection, repair, and leakage requirements are consistent with 
industry codes and standards for replacement SGs with Alloy 690 
tubing material. The requirements established by the SG Program are 
consistent with those in the applicable design codes and standards. 
The proposed changes update the requirements in the current TSs to 
reflect SG replacement.
    The proposed TS changes include a change to the current TS limit 
on primary-to-

[[Page 10311]]

secondary leakage of 144 GPD [gallons per day] that was established 
in the 1980s due to SG tube degradation. The basis for this limit 
will no longer be applicable with the installation of replacement 
SGs. The proposed limit of 150 gallons per day of primary-to-
secondary leakage through any one SG is ``standard'' for the U.S. 
PWR industry. This limit is based on operating experience with SG 
tube degradation mechanisms that result in leakage and provides 
reasonable assurance that the SG tubing will remain capable of 
fulfilling its specific safety function of maintaining reactor 
coolant pressure boundary integrity throughout each operating cycle 
and in the unlikely event of a design basis accident. Further, if it 
is not practical to assign the leakage to an individual SG, all the 
primary-to-secondary leakage is conservatively assumed to be from 
one SG. This operational leakage rate criterion, in conjunction with 
the implementation of the SG Program, is an effective measure for 
minimizing the frequency of SG tube ruptures. [Additionally, this TS 
requirement is significantly less than the conditions assumed in the 
safety analysis.]
    The proposed change to replace the 90-day report with a report 
required within 180 days is a change to an administrative 
requirement and does not affect the margin of safety. The 180-day 
period is now industry ``standard'' practice per TSTF-449.
    For the above reasons, the margin of safety is not reduced.

    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: J. Bradley Fewell, Esquire, Associate 
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Harold K. Chernoff.

R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of amendment request: December 4, 2008.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.9.3, ``Containment 
Penetrations,'' to permit refueling operations with both personnel 
airlock doors open under administrative control. Nuclear Regulatory 
Commission (NRC) review and approval of a revised non loss-of-coolant 
accident (LOCA) gas gap fractions and fuel-handling accident (FHA) 
using the revised gap fractions and a shorter decay time of 72 hours 
will be necessary to support this license amendment.
    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 has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below.
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    There are three separate items requiring NRC approval in the 
licensee's application. The licensee has submitted a plant-specific 
analysis to revise the non-LOCA gas gap fractions. Regulatory Guide 
1.183, ``Alternative Radiological Source Terms for Evaluating Design 
Basis Accidents at Nuclear Power Reactors,'' includes Table 3, ``Non-
LOCA Fraction of Fission Product Inventory in Gap.'' The Ginna licensee 
has determined that a small number of fuel rods may exceed the peak 
power and burnup criteria of Table 3 thus necessitating the plant-
specific analysis. The new non-LOCA gap fractions are considered a 
methodology change thus requiring NRC review and approval.
    The Ginna FHA currently assumes that fuel movement will not occur 
prior to 100 hours following reactor shutdown. The licensee has 
submitted a revised FHA that assumes both the new gas gap fractions 
discussed above and only 72 hours of decay time prior to fuel movement. 
The revised FHA must also be reviewed and approved by NRC.
    The proposed change to TS 3.9.3, which would permit refueling 
operations with both personnel airlock doors open under administrative 
control, impacts the release pathway for the FHA. The proposed TS 
change requires NRC review and approval.
    The proposed changes to the gas gap fractions and the FHA represent 
analytical changes and do not increase the probability of an accident 
previously evaluated. The change to TS 3.9.3 introduces a new release 
pathway for the FHA and does not increase the probability of an FHA or 
any other accident previously evaluated.
    The change in analyzed decay time and the non-LOCA gap fractions 
result in an increase in the estimated dose to the control room and 
off-site receptors and, upon approval, will become the analyses of 
record. However, the increase in dose is within regulatory limits so 
that the changes do not represent a significant increase in the 
consequences of the FHA or any other accident previously evaluated. The 
proposed change to TS 3.9.3 introduces a new release pathway for the 
FHA. However, control room and offsite dose calculations are bounded by 
the release pathway from the equipment hatch. As a result, the proposed 
change to TS 3.9.3 does not involve a significant increase in the 
consequences of an accident previously evaluated.
    Therefore, the probability or consequences of an accident 
previously evaluated will not be significantly increased.
    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 in analyzed decay time and the non-LOCA gap 
fractions only impact design inputs to the FHA. The proposed change to 
TS 3.9.3 only impacts isolation requirements during refueling 
operations within the containment. The only accident which could result 
in a significant release of radioactivity in the plant mode where 
refueling is possible is the FHA. No other initiators or accident 
precursors are created by this change.
    Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident not previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The change in analyzed decay time and the non-LOCA gap fractions 
result in an increase in estimated dose to the control room and off 
site receptors. However, the dose remains within regulatory guidelines 
and limits with adequate margin. The proposed change to TS 3.9.3 
introduces a new release pathway for the FHA which is bounded by the 
release pathway through the equipment hatch.
    Therefore, the proposed change does not involve a significant 
reduction in the 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: Carey Fleming, Sr. Counsel--Nuclear 
Generation, Constellation Group, LLC, 750 East Pratt Street, 17th 
Floor, Baltimore, MD 21202.
    NRC Branch Chief: Mark G. Kowal.

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

[[Page 10312]]

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 Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by email to pdr@nrc.gov.

AmerGen Energy Company, LLC, Docket No. 50-219, Oyster Creek Nuclear 
Generating Station, Ocean County, New Jersey

    Date of amendment request: November 13, 2007, as supplemented by 
letter dated February 18, 2009.
    Description of amendment request: The amendment deletes Technical 
Specification (TS) Section 6.5 and its associated subsections relating 
to the Review and Audit function, as well as correcting several 
administrative items. Additionally, the amendment implements changes to 
correct minor errors in TS Tables 3.1.1, 4.1.1, and 4.1.2.
    Date of issuance: February 24, 2009.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 273.
    Facility Operating License No. DPR-16: The amendment revised the 
License and Technical Specifications.
    Date of initial notice in Federal Register: April 8, 2008 (73 FR 
19108). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 24, 2009.
    No significant hazards consideration comments received: No.

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

    Date of application for amendment: November 13, 2007, supplemented 
by letters dated September 29, 2008, and February 18, 2009.
    Brief description of amendment: The amendment deletes Technical 
Specification (TS) Section 6.5 and its associated subsections relating 
to the Review and Audit function, as well as correcting several 
administrative items. The administrative items involve: correcting 
typographical errors, providing improved TS figure legibility, updating 
the description of the installed spent fuel pool storage locations, 
removing references to deleted TS sections, and correcting an error in 
the labeling of outfalls on the TMI site drawing.
    Date of issuance: February 24, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 269.
    Facility Operating License No. DPR-50. Amendment revised the 
license and the technical specifications.
    Date of initial notice in Federal Register: April 8, 2008 (73 FR 
19109). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 24, 2009.
    No significant hazards consideration comments received: No.

Duke Energy Carolinas, LLC, et al., Docket No. 50-414, Catawba Nuclear 
Station, Unit 2, York County, South Carolina

    Date of application for amendments: January 20, 2009.
    Brief description of amendments: The amendment revised Technical 
Specification Surveillance Requirement (SR) 3.3.1.4 frequency. SR 
3.3.1.4 is a Trip Actuating Device Operational Test of the reactor trip 
breakers and reactor trip bypass breakers.
    Date of issuance: February 13, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 242.
    Facility Operating License No. NPF-52: The amendment revised the 
license and the technical specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): The notice provided an opportunity to submit 
comments on the Commission's proposed NSHC determination by February 
28, 2009. No comments have been received to date. However, the notice 
also provided an opportunity to request a hearing by March 30, 2009, 
but indicated that if the Commission make a final NSHC determination, 
any such hearing would take place after issuance of the amendment.
    Date of initial notice in Federal Register: January 28, 2009 (74 FR 
4986). The supplement dated February 5, 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 February 13, 2009.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 25, 2008.
    Brief description of amendment: The amendment would revise Appendix 
A of Technical Specifications (TSs), as they apply to the spent fuel 
pool storage requirements in TS Section 3.7.16 and the criticality 
requirements for the Region I spent fuel pool and north tilt pit fuel 
storage racks, in TS Section 4.3.1.1.
    Date of issuance: February 6, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 236.
    Facility Operating License No. DPR-20: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 2, 2009 (74 FR 
123).

[[Page 10313]]

 The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated February 6, 2009.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 21, 2008.
    Brief description of amendment: The amendment supports a proposed 
change to the in-service inspection program that is based on topical 
report WCAP-16168-NP-A, Revision 2, ``Risk-Informed Extension of the 
Reactor Vessel In-Service Inspection Interval.'' In the referenced 
safety evaluation of the topical report, the NRC required licensees to 
amend their licenses to require that the information and analyses 
requested in Section (e) of the final 10 CFR 50.61a (or the proposed 10 
CFR 50.61a, given in 72 FR 56275 prior to issuance of the final 10 CFR 
50.61a) be submitted for NRC staff review and approval within one year 
of completing the required reactor vessel weld inspection. Entergy 
Nuclear Operations, Inc., added a new license condition to provide this 
information.
    Date of issuance: February 11, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 237.
    Facility Operating License No. DPR-20: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 4, 2008 (73 FR 
65690). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 11, 2009.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 1, 2007, as supplemented 
by letters dated September 5 and September 21, 2007, February 14, 2008, 
and January 19 and February 20, 2009.
    Brief description of amendment: The changes revised the allowable 
values in the Grand Gulf Nuclear Station, Unit 1, Technical 
Specification Tables 3.3.5.1-1 and 3.3.5.2-1 for the Condensate Storage 
Tank (CST) low level setpoints for the High Pressure Core Spray and 
Reactor Core Isolation Cooling suction swap from the CST to the 
Suppression Pool.
    Date of issuance: February 25, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No: 181.
    Facility Operating License No. NPF-29: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: May 8, 2007 (72 FR 
26176). The supplements dated September 5 and September 21, 2007, 
February 14, 2008, and January 19 and February 20, 2009, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 25, 2009.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 16, 2007, as supplemented by 
letter dated January 8, 2009.
    Brief description of amendment: The amendment added a new license 
condition on the control room envelope (CRE) habitability program; 
revised the TS requirements related to the CRE habitability in TS 
3.7.6, ``Control Room Emergency Air Filtration System--Operating,'' TS 
3.7.6.2, ``Control Room Emergency Air Filtration System--Shutdown,'' 
and TS 3.7.6.5, ``Control Room Isolation and Pressurization''; and 
established a CRE habitability program in TS Section 6.5, 
``Administrative Controls--Programs.'' These changes are consistent 
with the NRC-approved Industry/TS Task Force (TSTF) Traveler TSTF-448, 
Revision 3, ``Control Room Habitability.'' The availability of this TS 
improvement was published in the Federal Register on January 17, 2007 
(72 FR 2022), as part of the Consolidated Line Item Improvement 
Process.
    Date of issuance: February 20, 2009.
    Effective date: As of the date of issuance and shall be implemented 
120 days from the date of issuance.
    Amendment No.: 218.
    Facility Operating License No. NPF-38: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: September 25, 2007 (72 
FR 54473).
    The supplemental letter dated January 8, 2009, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 20, 2009.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois.

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

Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power 
Station, Unit No. 1, DeWitt County, Illinois.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 1, 2 and 3, Grundy County, Illinois.

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

Exelon Generation Company, LLC, Docket No. 50-352 and No. 50-353, 
Limerick Generating Station, Unit 1 and 2, Montgomery County, 
Pennsylvania.

Exelon Generation Company, LLC, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey.

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station,Units 2 and 3,York 
and Lancaster Counties, Pennsylvania.

Exelon Generation Company, LLC, Docket Nos. 50-254 a