[Federal Register Volume 74, Number 45 (Tuesday, March 10, 2009)]
[Notices]
[Pages 10305-10315]
From the Federal Register Online via the Government Publishing 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, http://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
[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 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 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
[[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 http://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 [email protected].
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
http://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, 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].
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, 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 email to [email protected].
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 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois.
Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania.
Date of application for amendments: February 28, 2008.
Brief description of amendments: The amendment incorporates
Technical Specification Task Force Change
[[Page 10314]]
Traveler No. 308, Rev. 1, ``Determination of Cumulative and Projected
Dose Contributions in the Radioactive Effluent Controls Program
(RECP),'' which clarified the existing wording in the RECP technical
specification to reflect the intent of Generic Letter 89-01,
``Implementation of Programmatic and Procedural Controls for
radiological Effluent Technical Specifications (RETS) in the
Administrative Controls Section of the Technical Specifications and the
Relocation of the Procedural Details of RETS to the Offsite Dose
Calculation Manual or to the Process Control Program,'' regarding the
periodicity of dose projections for the calendar quarter and year.
Date of issuance: February 23, 2009.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 156, 156, 161, 161, 184, 43, 230, 223, 190, 177,
197, 158, 272, 270, 274, 242, 237 and 268.
Facility Operating License Nos. NPF-72, NPF-77, NPF-37, NPF-66,
NPF-62, DPR-2, DPR-19, DPR-25, NPF-11, NPF-18, NPF-39, NPF-85, DPR-16,
DPR-44, DPR-56, DPR-29, DPR-30, and DPR-50: The amendments revised the
Technical Specifications/Licenses.
Date of initial notice in Federal Register: May 20, 2008 (73 FR
29162). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 23, 2009.
No significant hazards consideration comments received: No.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: February 8, 2008.
Description of amendment request: This amendment changes the
Technical Specifications to delete Surveillance Requirement 4.6.3.1,
which specifies post-maintenance testing requirements for containment
isolation valves.
Date of issuance: February 23, 2009.
Effective date: As of its date of issuance, and shall be
implemented within 90 days.
Amendment No.: 120.
Facility Operating License No. NPF-86: The amendment revised the
License and Technical Specifications.
Date of initial notice in Federal Register: August 26, 2008 (73 FR
50361). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 23, 2009.
No significant hazards consideration comments received: No comments
were received. However, a hearing was requested which included
contentions challenging the NRC staff's proposed no significant hazards
consideration determination. On October 14, 2008, the request for
hearing was denied by the Atomic Safety and Licensing Board. In
accordance with 10 CFR 50.91(a)(3), the NRC staff made a final
determination of no significant hazards consideration which is included
in the Safety Evaluation.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station, Unit No. 1 (NMP1), Oswego County, New York
Date of application for amendment: February 25, 2008.
Brief description of amendments: The amendment revises NMP1
Technical Specification (TS) Section 3/4.4.4, ``Emergency Ventilation
System,'' to remove the operability and surveillance requirements for
the 10,000 watt heater located in the common supply inlet air duct for
the Reactor Building Emergency Ventilation System. The amendment also
revises TS 3/4.4.5, ``Control Room Air Treatment System,'' to reduce
the 10-hour duration monthly system operational surveillance test
requirement to a 15-minute run surveillance test requirement.
Date of issuance: February 17, 2009.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 201.
Renewed Facility Operating License No. DPR-063: The amendment
revises the License and TSs.
Date of initial notice in Federal Register: April 8, 2008 (73 FR
19110). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 17, 2009.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit No. 2 (NMP2), Oswego County, New York
Date of application for amendment: August 14, 2008.
Brief description of amendment: The amendment revises the NMP1
Technical Specification (TS) Surveillance Requirement frequency in TS
3.1.3, ``Control Rod Operability,'' and Example 1.4-3 in TS Section
1.4, ``Frequency,'' to clarify the applicability of the 1.25 test
interval extension. The proposed changes are consistent with the
Nuclear Regulatory Commission (NRC)-approved Revision 1 to TS Task
Force (TSTF) Change Traveler, TSTF-475, ``Control Rod Notch Testing
Frequency and SRM Insert Control Rod Action,'' and NUREG-1433,
``Standard Technical Specifications General Electric Plants, BWR/4,''
Revision 3.0. A notice of availability for this TS improvement using
the consolidated line item improvement process was published in the
Federal Register on November 13, 2007 (72 FR 63935).
Date of issuance: February 23, 2009.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 130.
Renewed Facility Operating License No. NPF-69: Amendment revises
the License and Technical Specifications.
Date of initial notice in Federal Register: October 21, 2008 (73 FR
62567). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 23, 2009.
No significant hazards consideration comments received: No.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of application for amendment: February 8, 2008, as
supplemented by letter dated April 25, 2008, and email dated January 7,
2009.
Brief description of amendment: The amendment revises Technical
Specification 5.6.6, ``Reactor Coolant System (RCS) Pressure and
Temperature Limits Report (PTLR),'' to include a new methodology for
establishing reactor pressure vessel pressure-temperature limits in the
Ginna PTLR. The new PTLR methodology is documented in WCAP-14040-A,
Revision 4, ``Methodology Used to Develop Cold Overpressure Mitigating
System Setpoints and RCS Heatup and Cooldown Limit Curves.''
Date of issuance: February 23, 2009.
Effective date: As of the date of issuance to be implemented within
90 days.
Amendment No.: 106.
Renewed Facility Operating License No. DPR-18: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: April 8, 2008 (73 FR
19111). The supplemental letter dated April 25, 2008, and email dated
January 7, 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
[[Page 10315]]
Safety Evaluation dated February 23, 2009.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendment request: October 8, 2008.
Brief description of amendment request: The amendments revise the
TS for the diesel fuel oil testing program. The proposed changes are
based on NRC-approved Technical Specifications Task Force (TSTF)
Traveler TSTF-374, revision 0. Prior notice of such a proposed change
using the Consolidated Line Item Improvement Process was provided in
the Federal Register on April 21, 2006 (71 FR 20735).
Date of issuance: February 20, 2009.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 181 and 174.
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revised
the licenses and the technical specifications.
Date of initial notice in Federal Register: December 16, 2008 (73
FR 76413) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 20, 2009.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: December 1, 2008.
Brief description of amendment: On October 31, 2008, the NRC
approved Amendment No. 186 to allow a one-time extension to the
Completion Times for both essential service water (ESW) trains and the
emergency diesel generators from 72 hours to 14 days. Amendment No. 186
was effective on the date of issuance and approved implementation by
December 31, 2008, to permit replacement of ESW piping. The licensee
completed the replacement of ESW Train A piping, but deferred the
replacement of ESW Train B piping to early 2009. Amendment No. 191
authorized implementation of the ESW Train B piping prior to April 30,
2009.
Date of issuance: February 24, 2009.
Effective date: As of its date of issuance, and shall be
implemented prior to April 30, 2009.
Amendment No.: 191.
Facility Operating License No. NPF-30: The amendment revised the
Operating License and Technical Specifications.
Date of initial notice in Federal Register: December 23, 2008 (73
FR 78858).
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.
Dated at Rockville, Maryland, this 26th day of February 2009.
For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E9-4898 Filed 3-9-09; 8:45 am]
BILLING CODE 7590-01-P