[Federal Register Volume 72, Number 98 (Tuesday, May 22, 2007)]
[Notices]
[Pages 28717-28728]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-9523]
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NUCLEAR REGULATORY COMMISSION
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 April 27, 2007, to May 10, 2007. The last
biweekly notice was published on May 8, 2007 (72 FR 26173).
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. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rulemaking,
Directives and Editing Branch, Division of Administrative Services,
Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays.
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. The filing of requests for a hearing and petitions for leave
to intervene is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. 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 persons 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
[[Page 28718]]
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.
A request for a hearing or a petition for leave to intervene must
be filed 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; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this 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].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit No. 1, DeWitt County, Illinois
Date of amendment request: December 12, 2006.
Description of amendment request: The proposed amendment would
revise Surveillance Requirement (SR) 3.3.1.1.8 and SR 3.3.1.3.2 to
increase the interval between local power range monitor (LPRM)
calibrations from 1000 megawatt-days per ton (MWD/T) average core
exposure to 2000 MWD/T average core exposure. The proposed increase in
the interval between required LPRM calibrations is acceptable due to
improvements in fuel analytical bases, core monitoring processes, and
nuclear instrumentation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment revises the surveillance interval for the
LPRM calibration from 1000 MWD/T average core exposure to 2000 MWD/T
average core exposure. Increasing the frequency interval between
required LPRM calibrations is acceptable due to improvements in fuel
analytical bases, core monitoring processes, and nuclear
instrumentation. Therefore, the revised surveillance interval
continues to ensure that the LPRM detector signal will continue to
be adequately calibrated.
This change will not alter the operation of process variables,
structures, systems, or components as described in the CPS [Clinton
Power Station] Updated Safety Analysis Report (USAR). The proposed
change does not alter the initiation conditions or operational
parameters for the LPRM subsystem and there is no new equipment
introduced by the extension of the LPRM calibration interval. The
performance of the Average Power Range Monitor (APRM) system,
Oscillation Power Range Monitor (OPRM) system, Rod Control and
Information System (RC&IS) and 3D MONICORE core monitoring system is
not significantly affected by the proposed surveillance interval
increase. The proposed LPRM calibration interval extension will have
no significant effect on the Reactor Protection System (RPS)
instrumentation accuracy during power maneuvers or transients and
will therefore not significantly affect the performance of the RPS.
As such, the probability of occurrence for a previously evaluated
accident is not increased.
The radiological consequences of an accident can be affected by
the thermal limits existing at the time of the postulated accident;
however, LPRM chamber exposure has no significant affect on the
calculated thermal limits since LPRM accuracy does not
[[Page 28719]]
significantly deviate with exposure. For the LPRM extended
calibration interval, the total nodal power uncertainty remains less
than the uncertainty assumed in the General Electric BWR [boiling
water reactor] Thermal Analysis Basis (GETAB) safety limit,
maintaining the accuracy of the thermal limit calculation.
Therefore, the thermal limit calculation is not significantly
affected by LPRM calibration frequency, and thus the radiological
consequences of any accident previously evaluated are not increased.
Based on the above information, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any previously evaluated?
Response: No.
The performance of the APRM, OPRM, RC&IS and 3D MONICORE systems
is not significantly affected by the proposed LPRM surveillance
interval increase. The proposed change does not affect the control
parameters governing unit operation or the response of plant
equipment to transient conditions. The proposed amendment does not
change or introduce any new equipment, modes of system operation or
failure mechanisms.
Therefore, based on the above information, the proposed change
does not create the possibility of a new or different kind of
accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change has no impact on equipment design or
fundamental operation, and there are no changes being made to safety
limits or safety system allowable values that would adversely affect
plant safety as a result of the proposed LPRM surveillance interval
increase. The performance of the APRM, OPRM, RC&IS and 3D MONICORE
systems is not significantly affected by the proposed change. The
proposed LPRM calibration interval extension will have no
significant effect on RPS instrumentation accuracy during power
maneuvers or transients and will therefore not significantly affect
the performance of the RPS. The margin of safety can be affected by
the thermal limits existing at the time of the postulated accident;
however, uncertainties associated with LPRM chamber exposure have no
significant effect on the calculated thermal limits. The thermal
limit calculation is not significantly affected since LPRM
sensitivity with exposure is well defined. LPRM accuracy remains
within the total nodal power uncertainty assumed in the GETAB,
therefore maintaining thermal limits and the safety margin. The
proposed change does not affect safety analysis assumptions or
initial conditions and therefore, the margin of safety in the
original safety analyses is maintained.
Based on the above information, the proposed change does not
involve a significant reduction in a margin of safety .
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Russell Gibbs.
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit No. 1, DeWitt County, Illinois
Date of amendment request: January 26, 2007.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.3.1.1, ``Reactor Protection
System (RPS) Instrumentation,'' Table 3.3.1.1-1, ``Reactor Protection
System Instrumentation,'' Function 8, ``Scram Discharge Volume Water
Level--High,'' item b, ``Float Switches,'' by replacing Surveillance
Requirement (SR) 3.3.1.1.9 with SR 3.3.1.1.12. This change will
effectively revise the surveillance frequency for the scram discharge
volume (SDV) level float switch from every 92 days to every 24 months.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed TS change involves a change in the surveillance
frequency for the SDV water level float switch channel functional
test. The proposed TS change does not physically impact the plant.
The proposed change does not affect the design of the SDV water
level instruments, the operational characteristics or function of
the instruments, the interfaces between the instruments and the RPS,
or the reliability of the SDV water level instruments. The proposed
TS change does not degrade the performance of, or increase the
challenges to, any safety systems assumed to function in the
accident analysis. As noted in the Bases to TS 3.3.1.1, even though
the two types of SDV Water Level--High Functions are an input to the
RPS logic, no credit is taken for a scram initiated from these
functions for any of the design basis accidents or transients
evaluated in the CPS [Clinton Power Station] Updated Safety Analysis
Report (USAR). An inoperable SDV water level instrument is not
considered as an initiator of any analyzed event. The proposed TS
change does not impact the usefulness of the SRs in evaluating the
operability of required systems and components, or the way in which
the surveillances are performed. In addition, the frequency of
surveillance testing is not considered an initiator of any analyzed
accident, nor does a revision to the frequency introduce any
accident initiators. Therefore, the proposed change does not involve
a significant increase in the probability of an accident previously
evaluated.
The consequences of a previously analyzed event are dependent on
the initial conditions assumed in the analysis, the availability and
successful functioning of equipment assumed to operate in response
to the analyzed event, and the setpoints at which these actions are
initiated. The consequences of a previously evaluated accident are
not significantly increased by the proposed change. The proposed
change does not affect the performance of any equipment credited to
mitigate the radiological consequences of an accident. The risk
assessment of the proposed changes has concluded that there is an
insignificant increase in the core damage frequency as well as the
total population dose rate. Historical review of surveillance test
results and associated maintenance records did not find evidence of
failures that would invalidate the above conclusions.
Therefore, the proposed change does not alter the ability to
detect and mitigate events and, as such, does not involve a
significant increase in the consequences of an accident previously
evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any previously evaluated?
Response: No.
The proposed TS change does not introduce any failure mechanisms
of a different type than those previously evaluated, since there are
no physical changes being made to the facility. No new or different
equipment is being installed. No installed equipment is being
operated in a different manner. There is no change being made to the
parameters within which CPS is operated. There are no setpoints at
which protective or mitigative actions are initiated that are
affected by this proposed action. The change does not alter
assumptions made in the safety analysis. This proposed action will
not alter the manner in which equipment operation is initiated, nor
will the function demands on credited equipment be changed. No
alteration in the procedures, which ensure the unit remains within
analyzed limits, is proposed, and no change is being made to
procedures relied upon to respond to an off-normal event. As a
result, no new failure modes are being introduced. The way
surveillance tests are performed remains unchanged. A historical
review of surveillance test results and associated maintenance
records indicated there was no evidence of any failures that would
invalidate the above conclusions.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Margins of safety are established in the design of components,
the configuration of components to meet certain performance
parameters, and in the establishment of setpoints to initiate alarms
or actions. The
[[Page 28720]]
proposed TS change involves a change in the surveillance frequency
for the SDV water level float switch channel functional test. There
is no change in the design of the affected systems, no alteration of
the setpoints at which alarms or actions are initiated, and no
change in plant configuration from original design. The proposed
change does not significantly impact the condition or performance of
structures, systems, and components relied upon for accident
mitigation. The proposed change does not result in any hardware
changes or in any changes to the analytical limits assumed in
accident analyses. Existing operating margin between plant
conditions and actual plant setpoints is not significantly reduced
due to these changes. The proposed change does not significantly
impact any safety analysis assumptions or results.
AmerGen has conducted a risk assessment to determine the impact
of a change to the SDV water level instrument surveillance frequency
from the current once every 92 days to once every 24 months for the
risk measures of Core Damage Frequency (CDF) and Large Early Release
Frequency (LERF). This assessment indicated that the proposed CPS
surveillance frequency extension has a very small change in risk to
the public and is an acceptable plant change from a risk
perspective.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Russell Gibbs.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: April 30, 2007.
Description of amendment request: The amendment will revise the
technical specifications to use other narrow range containment sump
water level instrumentation rather than the existing redundant
instruments to allow installation of new emergency core cooling system
recirculation sumps strainers as specified in the Nuclear Regulatory
Commission Generic Letter 2004-02, Potential Impact of Debris Blockage
on Emergency Recirculation during Design Basis Accidents at Pressurized
Water Reactors.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated, and it does not change an accident previously evaluated
in the Final Safety Analysis Report (FSAR). The use of other narrow
range containment sump water level instruments rather than the
existing narrow range containment recirculation sump water level
instruments, which have level elements located inside the emergency
core cooling system (ECCS) recirculation sumps, will continue to
ensure that acceptable narrow range containment sump water level
monitoring is maintained during post-accident conditions. Operation
of the containment spray and residual heat removal systems is
unchanged as a result of the proposed amendment. The level elements
associated with the existing narrow range containment recirculation
sump water level instruments are not accident initiators, and the
FSAR does not credit these level elements in the dose analyses for
loss-of-coolant accidents. The proposed amendment does not adversely
affect the ability of structures, systems, or components (SSCs) to
perform their design function. SSCs required for post-accident
recirculation remain capable of performing their design functions.
Therefore, this amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed amendment does not create the possibility of a new
or different kind of accident from any accident previously
evaluated, and it does not change an accident previously evaluated
in the Final Safety Analysis Report (FSAR). The use of other narrow
range containment sump water level instruments rather than the
existing narrow range containment recirculation sump water level
instruments supports the replacement of the existing containment
recirculation sump screens with new strainers in accordance with the
response to Generic Letter 2004-02, Potential Impact of Debris
Blockage on Emergency Recirculation during Design Basis Accidents at
Pressurized-Water Reactors. The proposed amendment does not change
the design function or the operation of the containment spray and
residual heat removal systems associated with the containment
recirculation sumps. The proposed amendment does not create new
failure mechanisms or malfunctions or accident initiators. The
proposed amendment will continue to ensure that acceptable narrow
range containment sump water level monitoring is maintained during
post-accident conditions, and that SSCs required for post-accident
recirculation remain capable of performing their design functions.
Therefore, this amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed amendment does not involve a significant reduction
in a margin of safety. The proposed amendment does not adversely
affect a plant safety limit or a limiting safety system setting, and
does not alter a design basis limit for a parameter evaluated in the
FSAR. The use of other narrow range containment sump water level
instruments, which meet the requirements of the FSAR, rather than
the existing narrow range containment recirculation sump water level
instruments, will continue to ensure that acceptable narrow range
containment sump water level monitoring is maintained during post-
accident conditions. The proposed amendment does not adversely
affect the ability of SSCs to perform their design functions or the
reliability of equipment to mitigate accidents evaluated in the
FSAR. The proposed amendment will continue to ensure that SSCs
required for post-accident recirculation remain capable of
performing their design functions.
Therefore, this amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602.
NRC Branch Chief: Thomas H. Boyce.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: April 24, 2007.
Description of amendment request: The proposed change will add
Optimized ZIRLOTM as an acceptable fuel rod cladding
material in the Waterford Steam Electric Station, Unit 3 (Waterford 3),
Technical Specification (TS) 5.3.1, ``Fuel Assemblies.'' TS 5.3.1
currently identifies, in part, Zircaloy or ZIRLOTM fuel rod
cladding as the allowable fuel rod cladding material.
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
[[Page 28721]]
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The NRC-approved topical report WCAP-12610-P-A and CENPD-404-P-
A, Addendum 1-A, ``Optimized ZIRLOTM,'' prepared by
Westinghouse Electric Company, LLC (Westinghouse), addresses
Optimized ZIRLOTM and demonstrates that Optimized
ZIRLOTM has essentially the same properties as currently
licensed ZIRLOTM. The fuel cladding itself is not an
accident initiator and does not affect accident probability. Use of
Optimized ZIRLOTM fuel cladding has been shown to meet
all 10 CFR 50.46 design criteria and, therefore, will not increase
the consequences of an accident.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Use of Optimized ZIRLOTM clad fuel will not result in
changes in the operation or configuration of the facility. Topical
report WCAP-12610-P-A and CENPD-404-P-A demonstrated that the
material properties of Optimized ZIRLOTM are similar to
those of standard ZIRLOTM. Therefore, Optimized
ZIRLOTM fuel rod cladding will perform similarly to those
fabricated from standard ZIRLOTM, thus precluding the
possibility of the fuel becoming an accident initiator and causing a
new or different type of accident.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will not involve a significant reduction in
the margin of safety because it has been demonstrated that the
material properties of the Optimized ZIRLOTM are not
significantly different from those of standard ZIRLOTM.
Optimized ZIRLOTM is expected to perform similarly to
standard ZIRLOTM for all normal operating and accident
scenarios, including both loss-of-coolant accident (LOCA) and non-
LOCA scenarios. For LOCA scenarios, where the slight difference in
Optimized ZIRLOTM material properties relative to
standard ZIRLOTM could have some impact on the overall
accident scenario, plant-specific LOCA analyses using Optimized
ZIRLOTM properties will be performed prior to the use of
fuel assemblies with fuel rods containing Optimized
ZIRLOTM. These LOCA analyses will demonstrate that the
acceptance criteria of 10 CFR 50.46 will be satisfied when Optimized
ZIRLOTM fuel rod cladding is implemented.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Thomas G. Hiltz.
Nine Mile Point Nuclear Station (NMPNS), LLC, Docket No. 50-410, Nine
Mile Point Nuclear Station Unit No. 2 (NMP2), Oswego County, New York
Date of amendment request: March 30, 2007.
Description of amendment request: The proposed amendment would
change the NMP2 Technical Specifications to reflect an expanded
operating domain resulting from implementation of Average Power Range
Monitor/Rod Block Monitor/Technical Specifications/Maximum Extended
Load Line Analysis (ARTS/MELLLA). The Average Power Range Monitor
(APRM) flow-biased simulated thermal power Allowable Value would be
revised to permit operation in the MELLLA region. The current flow-
biased Rod Block Monitor (RBM) would be replaced by a power dependent
RBM, which also would require new Allowable Values. The flow-biased
APRM simulated thermal power setdown requirement would be replaced by
more direct power and flow dependent thermal limits administration. The
Surveillance Requirement for the standby liquid control (SLC) system
would be revised to require each SLC pump to deliver required flow at a
discharge pressure >=1325 psig in lieu of >=1320 psig; the SLC relief
valve setpoint would be increased from 1394 psig to 1400 psig. Finally,
the proposed amendment employs a new model for performing the
anticipated transients without scram (ATWS) analysis for ARTS/MELLLA
conditions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates the APRM flow-biased simulated
thermal power setdown requirement and substitutes power and flow
dependent adjustments to the Minimum Critical Power Ratio (MCPR) and
Linear Heat Generation Rate (LHGR) thermal limits. Thermal limits
will be determined using NRC [Nuclear Regulatory Commission]
approved analytical methods. The proposed change will have no effect
upon any accident initiating mechanism. The power and flow dependent
adjustments will ensure that the MCPR safety limit will not be
violated as a result of any Anticipated Operational Occurrence
(AOO), and that the fuel thermal and mechanical design bases will be
maintained.
The proposed change also expands the power and flow operating
domain by relaxing the restrictions imposed by the formulation of
the APRM flow-biased simulated thermal power Allowable Value and the
replacement of the current flow-biased RBM with a new power
dependent RBM. The APRM and RBM are not involved in the initiation
of any accident, and the APRM flow-biased simulated thermal power
function is not credited in any NMP2 safety analyses. The proposed
change will not introduce any initial conditions that would result
in NRC approved criteria being exceeded and the APRM and RBM will
remain capable of performing their design functions.
The Standby Liquid Control (SLC) System is provided to mitigate
anticipated transients without scram (ATWS) events and, as such, is
not considered an initiator of an ATWS event or any other analyzed
accident. The revised SLC discharge pump test pressure neither
reduces the ability of the SLC system to respond to or mitigate an
ATWS event nor increases the likelihood of a system malfunction that
could increase the consequences of an accident.
Based on the above discussion, it is concluded that the proposed
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change eliminates the APRM flow-biased simulated
thermal power setdown requirement and substitutes power and flow
dependent adjustments to the MCPR and LHGR thermal limits. Because
the thermal limits will continue to be met, no analyzed transient
event will escalate into a new or different type of accident due to
the initial starting conditions permitted by the adjusted thermal
limits.
The proposed change also expands the power and flow operating
domain by relaxing the restrictions imposed by the formulation of
the APRM flow-biased simulated thermal power Allowable Value and the
replacement of the current flow-biased RBM with a new power
dependent RBM. Changing the formulation for the APRM flow-biased
simulated thermal power Allowable Value and changing from a flow-
biased RBM to a power dependent RBM does not change their respective
functions and manner of
[[Page 28722]]
operation. The change does not introduce a sequence of events or
introduce a new failure mode that would create a new or different
[kind] of accident. While not credited, the APRM flow-biased
simulated thermal power Allowable Value and associated scram trip
setpoint will continue to initiate a scram to protect the MCPR
safety limit. The power dependent RBM will prevent rod withdrawal
when the power dependent RBM rod block setpoint is reached. No new
failure mechanisms, malfunctions, or accident initiators are being
introduced by the proposed change. In addition, operating within the
expanded power flow map will not require any systems, structures or
components to function differently than previously evaluated and
will not create initial conditions that would result in a new or
different kind of accident from any accident previously evaluated.
The proposed change to the SLC pump test discharge pressure is
consistent with the functional requirements of the ATWS rule (10 CFR
50.62). This proposed change does not involve the installation of
any new or different type of equipment, does not introduce any new
modes of plant operation, and does not change any methods governing
normal plant operation.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change eliminates the APRM flow-biased simulated
thermal power setdown requirement and substitutes power and flow
dependent adjustments to the MCPR and LHGR thermal limits.
Replacement of the APRM setdown requirement with power and flow
dependent adjustments to the MCPR and LHGR thermal limits will
continue to ensure that margins to the fuel cladding Safety Limit
are preserved during operation at other than rated conditions.
Thermal limits will be determined using NRC approved analytical
methods. The power and flow dependent adjustments will ensure that
the MCPR safety limit will not be violated as a result of any AOO,
and that the fuel thermal and mechanical design bases will be
maintained.
The proposed change also expands the power and flow operating
domain by relaxing the restrictions imposed by the formulation of
the APRM flow-biased simulated thermal power Allowable Value and the
replacement of the current flow-biased RBM with a new power
dependent RBM. The APRM flow-biased simulated thermal power
Allowable Value and associated scram trip setpoint will continue to
initiate a scram to protect the MCPR safety limit. The RBM will
continue to prevent rod withdrawal when the power dependent RBM rod
block setpoint is reached. The MCPR and LHGR thermal limits will be
developed to ensure that fuel thermal mechanical design bases remain
within the licensing limits during a control rod withdrawal error
event and to ensure that the MCPR safety limit will not be violated
as a result of a control rod withdrawal error event. Operation in
the expanded operating domain will not alter the manner in which
safety limits, limiting safety system settings, or limiting
conditions for operation are determined. AOOs and postulated
accidents within the expanded operating domain will continue to be
evaluated using NRC approved methods. The 10 CFR 50.46 acceptance
criteria for the performance of the ECCS [emergency core cooling
system] following postulated LOCAs [loss-of-coolant accidents] will
continue to be met.
The proposed change to the SLC pump discharge test pressure does
not alter the results of any accident analyses. The proposed change
is consistent with the functional requirements of the ATWS rule (10
CFR 50.62). The ability of the SLCS to respond to and mitigate an
ATWS event is not affected.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1700 K Street, NW., Washington, DC 20006.
NRC Branch Chief: Mark G. Kowal.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment requests: April 17, 2007.
Description of amendment requests: A change is proposed to the
standard technical specifications (STS) (NUREGs 1430 through 1434) and
plant-specific technical specifications (TS), to strengthen TS
requirements regarding control room envelope (CRE) habitability by
changing the action and surveillance requirements associated with the
limiting condition for operation operability requirements for the CRE
emergency ventilation system, and by adding a new TS administrative
controls program on CRE habitability. Accompanying the proposed TS
change are appropriate conforming technical changes to the TS Bases.
The proposed revision to the Bases also includes editorial and
administrative changes to reflect applicable changes to the
corresponding STS Bases, which were made to improve clarity, conform
with the latest information and references, correct factual errors, and
achieve more consistency among the STS NUREGs. The proposed revision to
the TS and associated Bases is consistent with STS as revised by TS
Task Force (TSTF)-448, Revision 3, ``Control Room Envelope
Habilitability.''
The proposed amendment would revise the TS Improvement To Modify
Requirements Regarding CRE Habitability using the Consolidated Line
Item Improvement Process, based on the NRC-approved to TSTF-448,
Revision 3. The NRC staff issued a notice of opportunity for comment in
the Federal Register on October 17, 2006 (71 FR 61075), on possible
amendments adopting TSTF-448, including a model safety evaluation and
model no significant hazards consideration (NSHC) determination, using
the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on January 17, 2007 (72 FR 2022). The licensee affirmed the
applicability of the following NSHC determination in its application
dated April 17, 2007.
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
The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility. The proposed change
does not alter or prevent the ability of structures, systems, and
components (SSCs) to perform their intended function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change revises the TS for the CRE emergency
ventilation system, which is a mitigation system designed to
minimize unfiltered air leakage into the CRE and to filter the CRE
atmosphere to protect the CRE occupants in the event of accidents
previously analyzed. An important part of the CRE emergency
ventilation system is the CRE boundary. The CRE emergency
ventilation system is not an initiator or precursor to any accident
previously evaluated. Therefore, the probability of any accident
previously evaluated is not increased. Performing tests to verify
the operability of the CRE boundary and implementing a program to
assess and maintain CRE habitability ensure that the CRE emergency
ventilation system is capable of adequately mitigating radiological
consequences to CRE occupants during accident conditions, and that
the CRE emergency ventilation system will perform as assumed in the
consequence analyses of design basis accidents. Thus, the
consequences of any accident previously evaluated are not increased.
Therefore, the proposed change does not involve a
[[Page 28723]]
significant increase in the probability or consequences of an
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Accident Previously
Evaluated
The proposed change does not impact the accident analysis. The
proposed change does not alter the required mitigation capability of
the CRE emergency ventilation system, or its functioning during
accident conditions as assumed in the licensing basis analyses of
design basis accident radiological consequences to CRE occupants. No
new or different accidents result from performing the new
surveillance or following the new program. The proposed change does
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a significant
change in the methods governing normal plant operation. The proposed
change does not alter any safety analysis assumptions and is
consistent with current plant operating practice. Therefore, this
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The proposed change does not affect safety
analysis acceptance criteria. The proposed change will not result in
plant operation in a configuration outside the design basis for an
unacceptable period of time without compensatory measures. The
proposed change does not adversely affect systems that respond to
safely shut down the plant and to maintain the plant in a safe
shutdown condition. Therefore, the proposed change does not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Branch Chief: Thomas G. Hiltz.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: February 28, 2007.
Brief description of amendments: The proposed amendment request
would revise the language of Technical Specification (TS) 3.7.1.2,
``Auxiliary Feedwater System,'' Action b from ``MODE 3 may be entered
with an inoperable turbine-driven auxiliary feedwater pump for the
purposes of performing Surveillance Requirement 4.7.1.2.1a.2'' to
``MODE 3 may be entered with an inoperable turbine-driven auxiliary
feedwater pump.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed deletion of the existing words in TS 3.7.1.2 Action
b is an administrative change that will clarify the Licensing Basis
for the turbine-driven auxiliary feedwater pump. Since this change
does not change the Licensing Basis for TS 3.7.1.2, this change
cannot affect the probability or consequence of any accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed deletion of the existing words in TS 3.7.1.2 Action
b is an administrative change that will clarify the Licensing Basis
for the turbine-driven auxiliary feedwater pump. Since this change
does not change the Licensing Basis for TS 3.7.1.2, this change
cannot affect the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed deletion of the existing words in TS 3.7.1.2 Action
b is an administrative change that will clarify the Licensing Basis
for the turbine-driven auxiliary feedwater pump. Since this change
does not change the Licensing Basis for TS 3.7.1.2, this change
cannot 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: A. H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Branch Chief: Thomas G. Hiltz.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: April 5, 2007.
Description of amendment request: The proposed amendments would
revise technical specifications (TSs) to change the surveillance
frequency for the turbine trip functions of the reactor trip system
instrumentation. The current frequency is prior to each reactor startup
and the proposed change will revise this to be prior to exceeding the
Permissive P-9 interlock whenever the unit has been in hot standby. The
proposed change is consistent with NRC-approved Technical Specification
Task Force Traveler TSTF-311, as incorporated into the latest revision
of Standard TSs (NUREG-1431, Revision 3).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes revise the surveillance frequency for
reactor trip functions from a turbine trip event. These changes do
not alter these functions physically or how they are maintained.
Delaying the performance of the surveillance up to the P-9 interlock
will continue to ensure operability of the function before the plant
is in a condition that would benefit from the associated actuation.
The incorporation of a surveillance frequency that is consistent
with the applicability for the function eliminates potential
misapplication of the TS requirements. The frequency changes support
turbine trip operability during plant startup and are consistent
with their ability to perform the reactor trip functions. Since
these changes will not affect the ability of these trips to perform
the initiation of reactor trips when appropriate, the off-site dose
consequences for an accident will not be impacted. Equally, the
potential to cause an accident is not affected because no plant
system or component has been altered by the proposed changes.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes only affect the surveillance frequency
requirement for the turbine trip functions. This does not affect
[[Page 28724]]
any physical features of the plant or the manner in which these
functions are utilized. The proposed surveillance frequency will
require the functions to be verified operable before the turbine
trip functions are applicable and able to perform their trip
functions. Delaying the performance of the surveillance up to the P-
9 interlock will continue to ensure operability of the function
before the plant is in a condition that would benefit from the
associated actuation. Therefore, the proposed change does not create
the possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not alter any plant setpoints or
functions that are assumed to actuate in the event of postulated
accidents. In fact, the proposed changes do not alter any plant
feature and only alter the requirements for when the function must
be verified to be operable through surveillance testing. The
proposed changes ensure the functionality of the turbine trips when
assumed in the analysis for accident mitigation. Therefore, the
proposed change does not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: Thomas H. Boyce.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: December 19, 2006.
Brief description of amendments: The proposed amendment request
would revise the requirements in Technical Specification (TS) 5.5.8,
``Inservice Testing Program,'' to update references to the American
Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,
Section XI, as the source of requirements for the inservice testing of
ASME Code Class 1, 2, and 3 pumps and valves, and address the
applicability of Surveillance Requirement 3.0.2 to other normal and
accelerated frequencies specified as 2 years or less in the Inservice
Testing Program.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed [change] involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes revise TS 5.5.8, ``Inservice Testing
Program,'' for consistency with the requirements of 10 CFR
50.55a(f)(4) regarding the inservice testing of pumps and valves.
The proposed change incorporates revisions to the ASME Code that
result in a net improvement in the measures for testing pumps and
valves.
The proposed changes do not impact any accident initiators or
analyzed events or assumed mitigation of accident or transient
events. They do not involve the addition or removal of any
equipment, or any design changes to the facility. Therefore, the
proposed changes do not represent a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes revise TS 5.5.8, ``Inservice Testing
Program,'' for consistency with the requirements of 10 CFR 50.55a(f)
regarding the inservice testing of pumps and valves. The proposed
changes incorporate revisions to the ASME Code that result in a net
improvement in the measures for testing pumps and valves.
The proposed changes do not involve a modification to the
physical configuration of the plant (i.e., no new equipment will be
installed) or change in the methods governing normal plant
operation. The proposed changes will not impose any new or different
requirements or introduce a new accident initiator, accident
precursor, or malfunction mechanism. Additionally, there is no
change in the types or increases in the amounts of any effluent that
may be released off-site and there is no increase in individual or
cumulative occupational exposure. Therefore, these proposed changes
do not create the possibility of an accident of a different kind
than previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes revise TS 5.5.8, ``Inservice Testing
Program,'' for consistency with the requirements of 10 CFR
50.55a(f)(4) regarding the inservice testing of pumps and valves.
The proposed changes incorporate revisions to the ASME Code that
result in a net improvement in the measures for testing pumps and
valves. The safety function of the affected pumps and valves will be
maintained. Therefore, these 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: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Branch Chief: Thomas G. Hiltz.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management 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,
[[Page 28725]]
(301) 415-4737 or by e-mail to [email protected].
Consumers Energy Company, Entergy Nuclear Palisades, LLC, and Entergy
Nuclear Operations, Inc., Docket No. 50-155, Big Rock Point Facility,
Charlevoix County, Michigan
Date of application for amendment: October 31, 2006.
Brief description of amendment: The license amendment reflects the
changes in ownership and operating authority for the Big Rock Facility
and its Independent Spent Fuel Storage Installation.
Date of issuance: April 11, 2007.
Effective date: As of the date of issuance.
Amendment No.: 127.
Facility Operating License No. DPR-06: The license amendment
reflects the changes in ownership and operating authority for the Big
Rock Facility and its Independent Spent Fuel Storage Installation.
Date of initial notice in Federal Register: January 30, 2007 (72 FR
4302-4303). The Commission's related evaluation of the amendment is
contained in a safety evaluation report dated April 6, 2007, which is
accessible to members of the public through ADAMS (Accession Number
ML070920385).
No significant hazards consideration comments received: No.
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power
Station, Kewaunee County, Wisconsin
Date of application for amendment: January 10, 2007, as
supplemented by letters dated April 5 and 27, 2007.
Brief description of amendment: The amendment modifies the
emergency diesel generators short-time load testing requirements.
Date of issuance: May 1, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 191.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 5, 2007 ( 72
FR 5303). The supplemental letters provided clarifying information that
did not expand the scope of the original application or change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 1, 2007.
No significant hazards consideration comments received: No.
Duke Power Company LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: April 11, 2006.
Brief description of amendments: The amendments revised an
organizational description in the Technical Specification Section
5.2.1, ``Onsite and Offsite Organizations.'' The change revises the
title of Executive Vice President to Group Vice President to reflect
title changes made by the licensee following the indirect transfer of
the facility operating licenses. The indirect transfer was reviewed and
approved by the NRC. This change is solely administrative in nature.
Date of issuance: April 13, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 239, 221.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: March 13, 2007 (72 FR
11387). The Commission's related evaluation, final no significant
hazards consideration finding, and State consultation of the amendments
is contained in a Safety Evaluation dated April 13, 2007.
No significant hazards consideration comments received: No.
Duke Power Company LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: March 8, 2007, as supplemented
March 27, April 13, and May 3, 2007.
Brief description of amendments: The amendments revise the McGuire
Nuclear Station, Units 1 and 2, Technical Specification 3.5.2.8, and
the associated Bases and authorize changes to the Updated Final Safety
Analysis Report (USFAR) concerning modifications to the emergency core
cooling system sump.
Date of issuance: May 4, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 240, 222.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the licenses and the technical specifications and
authorize changes to the UFSAR.
Date of initial notice in Federal Register: March 19, 2007 (72 FR
12835).
The supplements dated March 27, April 13, and May 3, 2007, 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, final no
significant hazards finding, and state consultation of the amendments
are contained in a Safety Evaluation dated May 4, 2007.
No significant hazards consideration comments received: No.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: September 19, 2006, as supplemented by
letter dated February 28, 2007.
Brief description of amendment: The amendment revised River Bend
Station (RBS), Unit 1, Technical Specifications (TS) Surveillance
Requirement (SR) 3.6.1.3.5 to replace the currently specified frequency
for leak testing containment purge supply and exhaust isolation valves
with resilient seal materials with a requirement to test these valves
in accordance with the RBS's Primary Containment Leakage Rate Testing
Program. RBS's Primary Containment Leakage Rate Testing Program is
implemented in accordance with the Title 10 of the Code of Federal
Regulations, Part 50, Appendix J, Option B, and Regulatory Guide (RG)
1.163, ``Performance-Based Containment Leak Test Program,'' dated
September 1995. RG 1.163 allows a nominal test interval of 30 months
for containment purge and vent valves.
Date of issuance: May 3, 2007.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 152.
Facility Operating License No. NPF-47: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: October 24, 2006 (71 FR
62310). The supplement dated February 28, 2007, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 3, 2007.
[[Page 28726]]
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: May 31, 2005, as supplemented by
letters dated February 8, 2006, and January 5, February 13, February
22, and March 22, 2007.
Brief description of amendment: The amendment modifies Technical
Specification (TS) Sections 3.8.1, ``AC [Alternating Current] Sources--
Operating,'' 3.8.4, ``DC [Direct Current] Sources--Operating,'' 3.8.5,
``DC Sources--Shutdown,'' 3.8.6, ``Battery Cell Parameters,'' and 5.5,
``Programs and Manuals.'' The change incorporates clarifying
requirements in surveillance testing of diesel generators and new
actions for an inoperable battery charger. The change includes a
revision to the Administrative Program to be consistent with Institute
of Electrical and Electronics Engineers Standard 450-2002, and changes
consistent with TS Task Force (TSTF) Traveler TSTF-360, Revision 1,
``DC Electrical Rewrite,'' and TSTF-283, Revision 3, ``Modify Section
3.8 Mode Restriction Notes.''
Date of issuance: May 1, 2007.
Effective date: As of its date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment No.: 204.
Facility Operating License No. NPF-21: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: February 27, 2007 (72
FR 8803). The supplemental letters dated February 8, 2006, and January
5, February 13, February 22, and March 22, 2007, 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 amendment is contained in a
Safety Evaluation dated May 1, 2007.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts.
Date of amendment request: October 18, 2005, as supplemented by
letter dated February 23, 2007.
Description of amendment request: The proposed amendment revised
applicability requirements related to single control rod withdrawal
allowances in shutdown modes. The amendment also corrected a
typographical error and administratively relocated the existing TS 3/
4.10.D, ``Multiple Control Rod Removal,'' to TS 3/4.14.E to be
consistent with the intent and presentation of special operations.
Date of issuance: April 25, 2007.
Effective date: As of the date of issuance, and shall be
implemented within 90 days.
Amendment No.: 228.
Facility Operating License No. DPR-35: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: January 3, 2006 (71 FR
148). The February 23, 2007, supplemental letter provided additional
information that clarified the application, but 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 April 25, 2007.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-254, Quad Cities Nuclear
Power Station, Unit 1, Rock Island County, Illinois
Date of application for amendments: January 16, 2007, as
supplemented by letter dated April 10, 2007.
Brief description of amendment: The amendment revises the values of
the safety limit minimum critical power ratio (SLMCPR) in the Quad
Cities Nuclear Power Station (Quad Cities), Unit 1, Technical
Specification (TS) Section 2.1.1, ``Reactor Core SLs [Safety Limits].''
Specifically, the proposed change would require that for Unit 1, the
minimum critical power ratio shall be greater than or equal to 1.11 for
two recirculation loop operation, or greater than or equal to 1.13 for
single recirculation loop operation. This change is needed to support
the next cycle of operation for Quad Cities, Unit 1.
Date of issuance: May 2, 2007.
Effective date: As of the date of issuance and shall be implemented
prior to startup from Q1R19 Refueling Outage.
Amendment No.: 234.
Renewed Facility Operating License No. DPR-29: The amendments
revised the Technical Specifications and License.
Date of initial notice in Federal Register: March 13, 2007 (71 FR
11388). The supplements contained clarifying information and did not
change the NRC staff's initial proposed finding of no significant
hazards consideration.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 2, 2007.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: February 14, 2006, as
supplemented by letters dated October 17, 2006, and February 8, 2007.
Brief description of amendment: The amendment revised Perry Nuclear
Power Plant, Unit No. 1, Technical Specifications (TSs) to change the
frequency of the Mode 5 Intermediate Range Monitoring Instrumentation
CHANNEL FUNCTIONAL TEST contained in TS 3.3.1.1 from 7 days to 31 days.
Date of issuance: April 27, 2007.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 141.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: March 28, 2006 (71 FR
15484) The October 17, 2006 and February 8, 2007 supplements, contained
clarifying information and did not change the NRC staff's initial
proposed finding of no significant hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 27, 2007.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: November 21, 2005, as
supplemented by letter dated February 22, 2007.
Brief description of amendment: This amendment revised the
acceptance criteria of technical specification (TS) surveillance
requirements associated with TS 3.8.1, to modify the emergency diesel
generator start tests to provide minimum voltage and frequency limits
and clarified other limits as steady state parameters.
[[Page 28727]]
Date of issuance: April 30, 2007.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 142.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: January 17, 2006 (71 FR
2591) The February 22, 2007, supplement contained clarifying
information and did not change the NRC staff's initial proposed finding
of no significant hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 30, 2007.
No significant hazards consideration comments received: No.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of application for amendments: April 27, 2006, as supplemented
December 5, 2006 and March 1, 2007.
Brief description of amendments: These amendments revised the
existing steam generator tube surveillance program to be consistent
with the Technical Specification Task Force (TSTF) Standard TS Change
Traveler, TSTF-449, ``Steam Generator Tube Integrity.''
Date of issuance: April 27, 2007.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos: 233 and 228.
Renewed Facility Operating License Nos. DPR-31 and DPR-41:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: July 18, 2006 (71 FR
40748). The supplements dated December 5, 2006, and March 1, 2007,
provided additional information clarifying information only and did not
change the initial no significant hazards consideration determination
or expand the scope of the initial application.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 27, 2007.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: June 6, 2006.
Brief description of amendments: The amendments revise information
in the Final Safety Analysis Report (FSAR) regarding the reactor
pressure vessel Charpy upper shelf energy (USE) requirements of Title
10 of the Code of Federal Regulations Part 50, Appendix G, Section
IV.A.1.c. The change updates the analysis for satisfying the RPV Charpy
USE requirements through the end of the current operating licenses.
Date of issuance: May 10, 2007.
Effective date: As of the date of issuance and shall be
incorporated into the FSAR during the next update of the FSAR, as
required by 10 CFR 50.71(c).
Amendment Nos.: 227 and 232.
Renewed Facility Operating License Nos. DPR-24 and DPR-27:
Amendments revise the Final Safety Analysis Report and the Licenses.
Date of initial notice in Federal Register: July 18, 2006 (71 FR
40750).
The Commission's related evaluation of the amendments is contained
in a safety evaluation dated May 10, 2007.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket No. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of application for amendments: May 31, 2006.
Brief description of amendments: The amendments correct
administrative errors in the SSES 1 and 2 Technical Specifications
(TSs) by adding a logical ``AND'' connector in Condition B of TS 3.8.1
for SSES 1, ``AC Sources--Operating,'' and correct the routing of
Interstate Route 80 on Figure 4.1-2 of TSs 4.1.2, ``Low Population
Zone,'' for SSES 1 and 2.
Date of issuance: April 26, 2007.
Effective date: As of the date of issuance and to be implemented
within 30 days.
Amendment Nos.: 243 and 221.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the TSs and License.
Date of initial notice in Federal Register: December 19, 2006 (71
FR 75996).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 26, 2007.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama
Date of application for amendments: November 15, 2006 January 11,
2007, as supplemented by letters dated January 11, and April 24, 2007.
Description of amendment request: The amendments revised the Fire
Protection License Condition numbers (13), (14), and (7) for Units 1,
2, and 3, respectively, to accommodate operation.
Date of issuance: April 25, 2007.
Effective date: Date of issuance, to be implemented within 30 days.
Amendment Nos.: 271, 300, and 259.
Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68:
Amendments revised the Operating Licenses.
Date of initial notice in Federal Register: December 19, 2006 (71
FR 76000). The supplements dated January 11, and April 24, 2007,
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 amendments is contained
in a Safety Evaluation dated April 25, 2007.
No significant hazards consideration comments received: No.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: March 22, 2006, supplemented by letter
dated September 12, 2006.
Brief description of amendments: The amendments revised the
Technical Specification (TS) 3.8.1 entitled, ``AC Sources--Operating.''
Specifically, the proposed change would revise the completion time for
TS 3.8.1, Condition F, Required Action F.1 from 12 hours to 24 hours.
Currently, TS 3.8.1, Condition F requires that an inoperable safety
injection (SI) sequencer must be restored to operable status within 12
hours. If this completion time is not met, Condition G becomes
applicable and the plant must be shutdown to at least Mode 3 within the
following 6 hours. The proposed change to the completion time for TS
3.8.1, Condition F, Required Action F.1 provides more time to complete
necessary repairs and required post-work testing to restore an
inoperable SI sequencer to operable status prior to commencing a plant
shutdown to Mode 3.
Date of issuance: April 27, 2007.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: NPF-87--138, NPF-89--138.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
[[Page 28728]]
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: March 28, 2007 (72 FR
14623).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 27, 2007.
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: June 7, 2006.
Brief description of amendment: The amendment deleted Required
Action D.1.2 in Technical Specification (TS) 3.7.10, ``Control Room
Emergency Ventilation System (CREVS),'' and Required Action C.1.2 in TS
3.7.11, ``Control Room Air Conditioning System (CRACS).'' For TS
3.7.13, ``Emergency Exhaust System (EES),'' the amendment also deletes
the phrase ``in MODE 1, 2, 3, or 4'' from Condition A (one EES train
inoperable) and revised Condition D to state the following: ``Required
Action and associated Completion Time of Condition A not met during
movement of irradiated fuel assemblies in the fuel building.''
Date of issuance: May 9, 2007.
Effective date: As of its date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 184.
Facility Operating License No. NPF-30: The amendment revised the
Operating License and Technical Specifications.
Date of initial notice in Federal Register: August 1, 2006 (71 FR
43536)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 9, 2007.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of application for amendments: January 31, 2007.
Brief Description of amendments: These amendments revised the
Technical Specification surveillance requirements for addressing a
missed surveillance, and is consistent with the Nuclear Regulatory
Commission approved Revision 6 of Technical Specification Task Force
(TSTF) Standard Technical Specifications Change Traveler TSTF-358,
``Missed Surveillance Requirements.''
Date of issuance: May 3, 2007.
Effective date: As of date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 253, 252.
Renewed Facility Operating License Nos. DPR-32 and DPR-37:
Amendments changed the licenses and the technical specifications.
Date of initial notice in Federal Register: February 27, 2007 (72
FR 8806).
No significant hazards consideration comments received: No.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 3, 2007.
Dated at Rockville, Maryland, this 11th day of May, 2007.
For the Nuclear Regulatory Commission.
Timothy McGinty,
Acting Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. E7-9523 Filed 5-21-07; 8:45 am]
BILLING CODE 7590-01-P