[Federal Register Volume 72, Number 204 (Tuesday, October 23, 2007)]
[Notices]
[Pages 60032-60041]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-20679]
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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 September 27, 2007, to October 10, 2007. The
last biweekly notice was published on October 9, 2007 (72 FR 57352).
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,
[[Page 60033]]
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 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
[[Page 60034]]
the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or by e-
mail to [email protected].
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power
Station, Kewaunee County, Wisconsin
Date of amendment request: September 24, 2007.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to add a reference to
Dominion Topical Report DOM-NAF-5, ``Application of Dominion Nuclear
Core Design and Safety Analysis Methods to the Kewaunee Power Station
(KPS),'' to the list of approved analytical methods. The proposed
changes would permit the application of the Dominion nuclear core
design and safety analysis methods, including the methodology to
perform core thermal-hydraulic analysis to predict critical heat flux
and departure from nucleate boiling ratio for the Westinghouse 422 V+
fuel design. The proposed amendment would also: (1) Accommodate the use
of the methodologies proposed in DOM-NAF-5, (2) delete one approved
analytical method that will no longer be used, and (3) delete date and
revision numbers from the current TS list of approved analytical
methods, consistent with TS Task Force (TSTF) Change Traveler TSTF-363-
A, Revision 0, ``Revise Topical Report References in ITS [improved TSs]
5.6.5, COLR [Core Operating Limits Report],'' dated August 4, 2003.
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 analysis methods of DOM-NAF-5 do not make any contribution
to the potential accident initiators and thus do not increase the
probability of any accident previously evaluated. The use of the
approved Dominion analysis methodologies will not increase the
probability of an accident because plant systems, structures, and
components (SSC) will not be affected or operated in a different
manner, and system interfaces will not change.
Since the applicable safety analysis and nuclear core design
acceptance criteria will be satisfied when the Dominion analysis
methods are applied to KPS, the use of the approved Dominion
analysis methods does not increase the potential consequences of any
accident previously evaluated. The use of the approved Dominion
methods will not result in a significant impact on normal operating
plant releases, and will not increase the predicted radiological
consequences of postulated accidents described in the USAR [updated
safety analysis report].
Therefore, the proposed amendment does not involve a significant
increase in the probability or the consequences of any accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different type of accident from any accident previously
evaluated?
Response: No.
The use of Dominion analysis methods and the Dominion
statistical design limit (SDL) for fuel departure from nucleate
boiling ratio (DNBR) and fuel critical heat flux (CHF) does not
impact any of the applicable core design criteria. All pertinent
licensing basis limits and acceptance criteria will continue to be
met. Demonstrated adherence to these limits and acceptance criteria
precludes new challenges to SSCs that might introduce a new type of
accident. All design and performance criteria will continue to be
met and no new single failure mechanisms will be created. The use of
the Dominion methods does not involve any alteration to plant
equipment or procedures that might introduce any new or unique
operational modes or accident precursors.
Therefore, the proposed amendment does not create 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.
Nuclear core design and safety analysis acceptance criteria will
continue to be satisfied with the application of Dominion methods.
Meeting the analysis acceptance criteria and limits ensures that the
margin of safety is not significantly reduced. Nuclear core design
and safety analysis acceptance criteria will continue to be
satisfied with the application of Dominion methods. In particular,
use of VIPRE-D with the proposed SDL provides at least a 95%
probability at a 95% confidence level that DNBR will not occur (the
95/95 DNBR criterion). The required DNBR margin of safety for KPS,
which is the margin between the 95/95 DNBR criterion and clad
failure, is therefore not reduced.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., 120 Tredegar Street, Richmond, VA 23219.
NRC Acting Branch Chief: Travis L. Tate.
Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446,
Comanche Peak Steam Electric Station, Units 1 and 2, Somervell County,
Texas
Date of amendment request: August 28, 2007.
Brief description of amendments: Revision to the Operating License
and Technical Specification (TS) 1.0, ``Use and Application, and TS
3.7.17'', ``Spent Fuel Assembly Storage,'' to Revise Rated Thermal
Power from 3458 megawatts thermal (MWt) to 3612 MWt.
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. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The impacts of the proposed Stretch Power Uprate (SPU) on plant
systems, structures, and components (SSCs) were reviewed with
respect to SSC design capability, and it was determined that
following completion of plant changes to support the SPU, no system,
structure, or component would exceed its design conditions or
limits. Evaluations supporting those conclusions were performed
consistent with proposed Technical Specification changes.
Consequently, equipment reliability and structural integrity will
not be adversely affected. Control system studies demonstrated that
plant response to operational transients under SPU conditions will
not significantly increase reactor trip frequency, so there will be
no significant increase in the frequency of SSC challenges caused by
reactor trip.
New systems are not needed to implement the SPU, and new
interactions among SSCs are not created. The SPU does not create new
failure modes for existing SSCs. Modified components do not
introduce new failure modes relative to those of the components in
their pre-modified condition. Consequently, new initiators of
previously analyzed accidents are not created.
The fission product barriers--fuel cladding, reactor coolant
pressure boundary, and the containment building--remain unchanged.
The spectrum of previously analyzed postulated accidents and
transients was evaluated, and effects on the fuel, the reactor
coolant pressure boundary, and the containment were determined.
These analyses were performed consistent with the proposed Technical
Specification changes. The results demonstrate that existing reactor
coolant pressure boundary and containment limits are met and that
effects on the fuel are such that dose consequences meet existing
criteria at SPU conditions.
There is no increase in the probability of an accident
concerning the potential insertion of a fuel assembly in an
incorrect location in the Spent Fuel Pool Region I/
[[Page 60035]]
Region II racks as a result of the specified storage patterns.
Luminant Power [Luminant Generation Company LLC] has used
administrative controls to move fuel assemblies from location to
location since the initial receipt of fuel on site. Fuel assembly
placement will continue to be controlled pursuant to approved fuel
handling procedures and in accordance with the Technical
Specification for spent fuel rack storage configuration limitations.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
New systems are not required to implement the SPU, and new
interactions among SSCs are not created. The SPU does not create new
failure modes for existing SSCs. Modified components do not
introduce failures different from those of the components in their
pre-modified condition. Consequently, no new or different accident
sequences arise from SSC interactions or failures.
Training will be provided to address SPU effects, and the
plant's simulator will be updated consistent with SPU conditions.
Operating procedure changes are minor and do not result in any
significant changes in operating philosophy. For these reasons, the
SPU does not introduce human performance issues that could create
new accidents or different accident sequences.
The increase in power level does not create new fission product
release paths. The fission product barriers (fuel cladding, reactor
coolant pressure boundary, and the containment building) remain
unchanged.
The potential for criticality in the spent fuel pool is not a
new or different type of accident. The potential criticality
accidents have been reanalyzed to demonstrate that the pool remains
subcritical.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
Structural evaluations performed at SPU conditions demonstrated
that calculated loads on affected SSCs remain within their design
for all design basis event categories. American Society of
Mechanical Engineers (ASME) Code fatigue limits continue to be met.
Fuel performance evaluations were performed using parameter
values appropriate for a reload core operating at SPU conditions.
Those evaluations demonstrate that fuel performance acceptance
criteria continue to be met. Loss of Coolant Accident (LOCA) and
non-LOCA safety analyses were performed assuming SPU conditions and
consistent with the proposed Technical Specification change.
Emergency core cooling system performance was shown to meet the
criteria of 10 CFR 50.46. The non-LOCA events identified in the
Final Safety Analysis Report (FSAR) Chapter 15 were shown to meet
existing acceptance criteria.
The containment building response to mass and energy releases
was evaluated assuming SPU conditions. The evaluations showed that
temperature and pressure limits were met.
No plant changes associated with the SPU reduce the degree of
component or system redundancy. Existing Technical Specification
operability and surveillance requirements are not reduced by the
proposed changes.
The proposed fuel storage requirements in Technical
Specification 3.7.17 will provide adequate margin to assure that the
fuel storage array (Region I and Region II) will always remain
subcritical by the 5% margin recommended by the Nuclear Regulatory
Commission (NRC).
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: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Branch Chief: Thomas G. Hiltz.
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: August 16, 2007.
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) requirements related to control
room envelope habitability in TS 3.7.9, ``Control Room Emergency Air
Treatment System (CREATS),'' and TS section 5.5, ``Programs and
Manuals.'' The changes are consistent with the Nuclear Regulatory
Commission approved Industry/Technical Specification Task Force (TSTF)-
448, Revision 3. The availability of this TS improvement was published
in the Federal Register on January 17, 2007, as part of the
consolidated line item improvement process.
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 adopted by the licensee 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 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 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 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.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The 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
[[Page 60036]]
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 analysis adopted by the licensee
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: Daniel F. Stenger, Ballard Spahr Andrews &
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC
20005.
NRC Branch Chief: Mark G. Kowal.
Southern California Edison Company, et al., Docket Nos. 50-362, San
Onofre Nuclear Generating Station, Unit 3, San Diego County, California
Date of amendment requests: September 24, 2007.
Description of amendment requests: Approval of the revision to the
San Onofre Nuclear Generating Station Unit 3 Technical Specification
5.5.2.15, ``Containment Leakage Rate Testing Program.'' The request is
for a one-time extension from the currently approved 15-year interval
since the last Integrated Leak Rate Test (ILRT) to a 16-year interval
since the last ILRT.
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. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed revision to Technical Specifications adds a one
time extension to the current interval for Type A testing (10 CFR
50, Appendix J, Option B, Integrated Leak Rate Testing). The current
test interval of 15 years, based on past performance, would be
extended on a one time basis to 16 years from the last Type A test.
The proposed extension to Type A testing does not involve a
significant increase in the probability or consequences of an
accident since research documented in NUREG-1493, ``Performance-
Based Containment System Leakage Testing Requirements,'' September
1995, has found that, generically, very few potential containment
leakage paths are not identified by Type B and C tests. The NUREG
concluded that reducing the Type A testing frequency to once per
twenty years was found to lead to an imperceptible increase in risk.
A high degree of assurance is provided through testing and
inspection that the containment will not degrade in a manner
detectable only by Type A testing. The most recent Type A test at
Unit 3 shows leakage to be below acceptance criteria, indicating a
leak tight containment. Inspections required by the American Society
of Mechanical Engineers (ASME) Code Section Xl (Subsections IWE and
IWL) and maintenance rule monitoring (10 CFR 50.65, ``Requirements
for Monitoring the Effectiveness of Maintenance at Nuclear Power
Plants) are performed in order to identify indications of
containment degradation that could affect leak tightness. Type B and
C testing required by Technical Specifications will identify any
containment opening such as valves that would otherwise be detected
by the Type A tests. These factors show that a Type A test extension
will not represent a significant increase in the consequences of an
accident.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed revision to Technical Specifications adds a one
time extension to the current interval for Type A testing (10 CFR
50, Appendix J, Option B, Integrated Leak Rate Testing). The current
test interval of 16 years, based on past performance, would be
extended on a one time basis to 16 years from the last Type A test.
The proposed extension to Type A testing cannot create the
possibility of a new or different type of accident since there are
no physical changes being made to the plant and there are no changes
to the operation of the plant that could introduce a new failure
mode creating an accident or affecting the mitigation of an
accident. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed revision to Technical Specifications adds a one
time extension to the current interval for Type A testing (10 CFR
50, Appendix J, Option B, Integrated Leak Rate Testing). The current
test interval of 15 years, based on past performance, would be
extended on a one time basis to 16 years from the last Type A test.
The proposed extension to Type A testing will not significantly
reduce the margin of safety. The NUREG 1493, ``Performance-Based
Containment System Leakage Testing Requirements,'' September 1995,
generic study of the effects of extending containment leakage
testing found that a 20 year extension in Type A leakage testing
resulted in an imperceptible increase in risk to the public. NUREG
1493 found that, generically, the design containment leakage rate
contributes about 0.1 percent to the individual risk and that the
decrease in Type A testing frequency would have a minimal [e]ffect
on this risk since 95% of the potential leakage paths are detected
by Type C testing. Regular inspections required by the American
Society of Mechanical Engineers (ASME) Code Section Xl (Subsections
IWE and IWL) and maintenance rule monitoring (10 CFR 50.65,
``Requirements for Monitoring the Effectiveness of Maintenance at
Nuclear Power Plants['']) will further reduce the risk of a
containment leakage path going undetected.
Therefore[,] the proposed change does not involve a significant
reduction in a margin of safety.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Branch Chief: Thomas G. Hiltz.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: September 19, 2007.
Description of amendment request: The proposed amendments would
revise various Technical Specification (TS) setting limits and the
overtemperature [Delta]T/overpower [Delta]T time constants in TS 2.3
and TS 3.7. The methodology for determining the revised setting limits
and time constants is in agreement with methods 1 and 2 in ISA-RP67.04,
Part II.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
No. The proposed change revises [Limited Safety System Settings]
LSSSs and setting limits to ensure that safety limits are not
exceeded as a result of normal and expected instrument drift between
calibration intervals. The new allowable values (LSSSs and setting
limits) were derived to meet the intent of RIS 2006-17, ``NRC Staff
Position on the Requirements of 10 CFR 50.36, `Technical
Specifications,' Regarding Limiting Safety System Settings During
Periodic Testing and Calibration of Instrument Channels,'' dated
August 24, 2006.
The proposed TS change does not change any of the previously
evaluated accidents in the Updated Final Safety Analysis Report
(UFSAR). Rather, the proposed change ensures that reactor trip
system and engineered safety function actuation system actuations
occur as designed and within safety limits. In addition, it
increases the probability that a malfunctioning instrument channel
will be identified.
This change is not considered to represent a significant
increase in the probability or
[[Page 60037]]
consequences of an accident, since it will decrease the probability
of the malfunction of a system, structure or component (SSC),
thereby decreasing the probability or consequences of an accident
previously evaluated. Specifically, the change is conservative in
nature since it will increase the likelihood that a malfunctioning
instrument channel will be identified prior to that channel
exceeding its safety limit.
2. Does the proposed license amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
No. The proposed change revises LSSSs and setting limits to
ensure that safety limits are not exceeded as a result of normal and
expected instrument drift between calibration intervals.
The change is conservative and is intended to ensure the safety
analysis is maintained. Specifically, the proposed change is
intended to identify a malfunctioning channel prior to its exceeding
the safety limit sooner than the current instrument setting
methodology. Therefore the proposed change will 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?
No. The proposed change revises LSSSs and setting limits to
ensure that safety limits are not exceeded as a result of normal and
expected instrument drift between calibration intervals. The new
allowable values (LSSS and setting limits) were derived to meet the
intent of RIS 2006-17, ``NRC Staff Position on the Requirements of
10 CFR 50.36, `Technical Specifications,' Regarding Limiting Safety
System Settings During Periodic Testing and Calibration of
Instrument Channels,'' dated August 24, 2006.
Channel statistical allowance (CSA) calculations have been
performed on channels with an associated safety analysis limit to
determine the instrument channel uncertainty. Channel operational
test (COT) errors are associated with those portions of the
instrument channel tested to verify channel operability. These COT
errors were extracted from the CSA to derive an allowable value for
the channel. The allowable value is set at a distance from the
actual (nominal) trip setpoint equal to the COT errors (with some
minimal additional margin on some channels). The overall result is a
reduction in the distance between the allowable value and the
nominal trip setpoint. Consequently, for a malfunctioning channel,
the allowable value will be exceeded with less drift and, therefore,
corrective action will be initiated sooner after implementation of
the proposed change. This will increase the likelihood that the
safety analysis limit for the channel is not exceeded.
The distance between the safety analysis limit and the nominal
trip setpoint has not been decreased; therefore, the safety margin
has [not been] reduced. The likelihood that a malfunctioning channel
is identified prior to exceeding its safety analysis limit has
increased. Therefore, the proposed amendment does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Branch Chief: Evangelos C. Marinos.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Duke Power Company LLC, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: March 29, 2007.
Brief description of amendment request: The proposed amendments
would revise the Catawba Nuclear Station, Units 1 and 2, Technical
Specification section 3.5.2.8, and the associated Bases and authorize
changes to the Updated Final Safety Analysis Report concerning
modifications to the emergency core cooling system sumps.
Date of publication of individual notice in Federal Register:
August 13, 2007, (72 FR 45274).
Expiration date of individual notice: October 15, 2007.
Duke Power Company LLC, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: April 30, 2007.
Brief description of amendment request: The proposed amendments
would revise the Catawba Nuclear Station, Unit 2, Technical
Specification Section 5.5.9 concerning modifications to the steam
generator tube repair criteria.
Date of publication of individual notice in Federal Register:
August 13, 2007, (72 FR 45272).
Expiration date of individual notice: October 15, 2007.
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/
[[Page 60038]]
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. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of application for amendment: May 15, 2006, as supplemented by
letters dated October 6, 2006, December 12, 2006, May 31, 2007, July
25, 2007, and September 4, 2007.
Brief description of amendment: The amendment consists of changes
to various technical specifications (TSs) regarding steam generator
tube integrity. It is based on Revision 4 to Technical Specification
Task Force (TSTF) Standard Technical Specification Change Traveler,
TSTF-449, ``Steam Generator Tube Integrity,'' and is adapted for the
custom TSs used at TMI-1.
Date of issuance: September 27, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 261.
Facility Operating License No. DPR-50: Amendment revised the
license and the technical specifications.
Date of initial notice in Federal Register: July 18, 2006 (71 FR
40744). The supplements dated October 6, 2006, December 12, 2006, May
31, 2007, July 25, 2007, and September 4, 2007, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the NRC
staff's original proposed no significant hazards determination. The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated September 27, 2007.
No significant hazards consideration comments received: No.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: May 2, 2007.
Brief description of amendments: Consistent with the Nuclear
Regulatory Commission approved Technical Specification Task Force-427,
Revision 2, the amendments add a new limiting condition for operation
(LCO) 3.0.9, to the TS. LCO 3.0.9 will allow the licensee to delay
declaring an LCO not met for equipment supported by barriers unable to
perform their associated support function for up to 30 days provided
that risk is assessed and managed.
Date of issuance: September 27, 2007.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment Nos.: 282 and 259.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the License and Technical Specifications.
Date of initial notice in Federal Register: June 19, 2007 (72 FR
33781). The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated September 27, 2007.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of application for amendments: September 28, 2006, as
supplemented by letter dated September 20, 2007.
Brief Description of amendments: The amendments changed Technical
Specification (TS) 3.8.3, ``Diesel Fuel Oil,'' to allow the main fuel
oil storage tank to be taken out of service for 14 days for inspection,
maintenance, and associated repairs on a one-time basis.
Date of issuance: September 27, 2007.
Effective date: Date of issuance to be implemented within 60 days.
Amendment Nos.: 242 and 270.
Renewed Facility Operating License Nos. DPR-71 and DPR-62:
Amendments changed the TSs.
Date of initial notice in Federal Register: January 3, 2007 (72 FR
148). The supplement dated September 20, 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 September
27, 2007.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: January 18, 2007.
Brief description of amendment: The amendment revises the
expiration time limit of the reactor coolant system Pressure/
Temperature limit graphs in Technical Specifications (TS); revises the
adjusted reference temperature for the reactor vessel; and revises the
Low Temperature Overpressure Protection (LTOP) arming temperature value
specified in TSs. It also makes editorial changes in the use of
inequality signs in TSs associated with the LTOP arming temperature in
order to make them consistent.
Date of issuance: October 4, 2007.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 235.
Facility Operating License No. DPR-64: The amendment revised the
License and the Technical Specifications.
Date of initial notice in Federal Register: April 10, 2007 (72 FR
17946). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 4, 2007.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear
Plant, Van Buren County, Michigan
Date of application for amendment: September 25, 2006, as
supplemented by letters dated June 15, September 7, September 20, and
September 21, 2007.
Brief description of amendment: The amendment provides the
Technical Specification (TS) changes and evaluations of the
radiological consequences of design-basis accidents for implementation
of a full-scope alternative source term methodology.
Date of issuance: September 28, 2007.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 226.
Renewed Facility Operating License No. DPR-20. Amendment revised
the TSs and the Operating License.
Date of initial notice in Federal Register: February 27, 2007 (72
FR 8804). The supplemental letters contained clarifying information and
did not change the initial no significant hazards consideration
determination, and did not expand the scope of the original Federal
Register notice. The Commission's related evaluation of the amendment
is contained in a Safety Evaluation dated September 28, 2007.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear
Plant, Van Buren County, Michigan
Date of application for amendment: June 29, 2007, as supplemented
by letter dated August 20, 2007.
[[Page 60039]]
Brief description of amendment: The amendment revises Technical
Specification (TS) 3.5.5, ``Trisodium Phosphate,'' and the associated
surveillance requirements by replacing the containment sump buffering
agent, trisodium phosphate, with sodium tetraborate decahydrate (STB).
In particular, the amendment revises the TS Limiting Condition for
Operation (LCO) 3.5.5, with a new weight requirement for STB. The title
of the TS section is also changed from ``Trisodium Phosphate'' to
``Containment Sump Buffering Agent and Weight Requirements.''
Date of issuance: October 2, 2007.
Effective date: As of the date of issuance and shall be implemented
during the 2007 refueling outage, prior to Mode 3 entry following
refueling.
Amendment No.: 227.
Renewed Facility Operating License No. DPR-20. Amendment revised
the TS and License.
Date of initial notice in Federal Register: July 10, 2007 (72 FR
37544). The supplemental letter contained clarifying information and
did not change the initial no significant hazards consideration
determination, and did not expand the scope of the original Federal
Register notice. The Commission's related evaluation of the amendment
is contained in a Safety Evaluation dated October 2, 2007.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear
Plant, Van Buren County, Michigan
Date of application for amendment: April 18, 2007, as supplemented
by letters dated July 16 and September 20, 2007.
Brief description of amendment: The amendment changes Technical
Specification (TS) Surveillance Requirement (SR) 3.5.2.9, to make the
surveillance consistent with the plant design following planned
modifications to the containment sump. Entergy Nuclear Operations'
(ENO) modification removes the existing emergency core cooling system
(ECCS) suction inlet screens. In lieu of the ECCS suction inlet
screens, ENO is installing passive strainer assemblies on the 590 foot
elevation of containment. The SR change was necessary to reflect the
change in equipment.
Date of issuance: October 4, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 228.
Renewed Facility Operating License No. DPR-20. Amendment revised
the TS and License.
Date of initial notice in Federal Register: June 19, 2007 (72 FR
33782). The supplemental letters contained clarifying information and
did not change the initial no significant hazards consideration
determination, and did not expand the scope of the original Federal
Register notice. The Commission's related evaluation of the amendment
is contained in a Safety Evaluation dated October 4, 2007.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2 (ANO-2), Pope County, Arkansas
Date of application for amendment: March 30, 2007, as supplemented
on June 13, 2007.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3.9.12, ``Fuel Storage,'' and its associated tables,
figures, and surveillance requirements, TS 5.3, ``Fuel Storage,'' and
adds TS 6.5.17, ``Metamic Coupon Sampling Program.'' The ANO-2 TS
3.9.12 is changed to: (1) Support higher fuel assembly uranium-235 (U-
235) enrichment; (2) apply the appropriate loading restrictions; and
(3) delete the dry cask loading restrictions. ANO-2 TS 5.3.1 b is
changed to reflect a different spent fuel pool boron concentration that
is needed to assure K-effective remains less than or equal to 0.95.
ANO-2 TS 5.3.2a is modified to reflect a higher fuel assembly U-235
enrichment. A new coupon sampling program is added as TS 6.5.17, and TS
4.9.12.d is added to direct performance of the coupon sampling program.
Date of issuance: September 28, 2007.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 273.
Renewed Facility Operating License No. NPF-6: Amendment revised the
Renewed Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: May 8, 2007 (72 FR
26175). The supplement dated June 13, 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 September
28, 2007.
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
Date of application for amendment: September 26, 2006, as
supplemented by letter dated August 8, 2007.
Brief description of amendment: The amendment revised the technical
specifications to allow the AREVA NP Inc. Advanced Mark-BW(A) fuel
assemblies to be loaded into the Braidwood Station, Unit 1 core for
operating Cycles 15, 16, and 17.
Date of issuance: October 4, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 145/145.
Facility Operating License Nos. NPF-72 and NPF-77: The amendment
revised the Technical Specifications and License.
Date of initial notice in Federal Register: (72 FR 152; January 3,
2007). The August 8, 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 amendments is contained in a Safety Evaluation dated October 4,
2007.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: October 18, 2006, as
supplemented by letter dated, March 26, 2007.
Brief description of amendments: The amendments would modify the
technical specifications (TS) to risk-inform requirements regarding
selected required action end states consistent with the Nuclear
Regulatory Commission (NRC)-approved industry and TS task force (TSTF-
423), Revision 0, ``Technical Specifications End States, NEDC-32988-
A.'' This TSTF was published in the Federal Register on March 23, 2006,
as part of the consolidated line item improvement.
Date of issuance: September 27, 2007.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment Nos.: 184/171.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications and License.
Date of initial notice in Federal Register: May 8, 2007 (72 FR
26177).
[[Page 60040]]
The March 26, 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
amendments is contained in a Safety Evaluation dated September 27,
2007.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334,
Beaver Valley Power Station, Unit No. 1, Beaver County, Pennsylvania
Date of application for amendment: February 9, 2007, as
supplemented by letters dated August 8, August 23, and September 13,
2007.
Brief description of amendment: The amendment will address Generic
Safety Issue 191 ``Assessment of Debris Accumulation on PWR Sump
Performance,'' by implementing Technical Specification (TS) changes
that reflect the use of a new recirculation spray system pump start
signal due to a modification to the containment sump screens and
replace the use of LOCTIC with the Modular Accident Analysis Program-
Design Basis Accident calculation methodology to calculate containment
pressure, temperature, and condensation rates for input to the SWNAUA
code, which ultimately changes the aerosol removal coefficients used in
dose consequence analysis.
Date of issuance: October 5, 2007.
Effective date: As of the date of issuance, and shall be
implemented prior to the first entry into Mode 4 coming out of 1R18,
which begins September 2007.
Amendment No: 280.
Facility Operating License No. DPR-66: The amendment revised the
License and TS.
Date of initial notice in Federal Register: April 24, 2007 (72 FR
20383). The supplements dated August 8, August 23, and September 13,
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 October 5, 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: January 26, 2007, as
supplemented by letter dated July 11, 2007.
Brief description of amendments: The amendment conforms the license
to reflect the direct transfer of Wisconsin Electric Power Company's
ownership interest and the Nuclear Management Company's operating
authority for the renewed Facility Operating License, Nos. DPR-24 and
DPR-27 for Point Beach Nuclear Plant, Units 1 and 2 (Point Beach) to
FPL Energy Point Beach, LLC, as approved by order of the Commission
order dated July 31, 2007. Transfer of the licenses will also authorize
FPL Energy Point Beach, LLC, pursuant to the general license
requirements in 10 CFR 72.210, to store spent fuel in the Independent
Spent Fuel Storage Installation at Point Beach.
Date of issuance: September 28, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 228, 233.
Renewed Facility Operating License Nos. DPR-24 and DPR-27:
Amendments revised the Technical Specifications/License.
Date of initial notice in Federal Register: February 28, 2007 (72
FR 9035). The July 11, 2007, supplement contained clarifying
information and did not change the 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
July 31, 2007.
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: October 12, 2006.
Brief description of amendment: The amendment revises the number of
fuel assemblies that are allowed to be stored in the spent fuel pool
(SFP) from 1879 to 1321 in Technical Specification (TS) 4.3.3 and
removes the reference to Type 4 SFP storage racks in TS limiting
condition for operation 3.7.13.
Date of issuance: October 1, 2007.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 103.
Renewed Facility Operating License No. DPR-18: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: November 7, 2006 (71 FR
65145). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 1, 2007.
No significant hazards consideration comments received: No.
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 application for amendments: April 28, 2006, and as
supplemented by letters dated November 13 and December 22, 2006, May 7,
June 15, July 27, and September 11, 2007.
Brief description of amendments: The change increased the minimum
allowed boron concentration of the spent fuel pool and allowed credit
for soluble boron, guide tube inserts made from borated stainless
steel, and fuel storage patterns in place of Boraflex.
Date of issuance: September 27, 2007.
Effective date: As of the date of issuance, and shall be
implemented within 180 days of issuance.
Amendment Nos.: Unit 2-213; Unit 3-205.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: June 6, 2006 (71 FR
32606). The supplemental letters dated November 13 and December 22,
2006, May 7, June 15, July 27, and September 11, 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 September 27, 2007.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of application for amendments: February 13, 2007.
Brief description of amendments: The amendments revised the
Technical Specifications for refueling interlocks.
Date of issuance: October 4, 2007.
[[Page 60041]]
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 253, 197.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: March 27, 2007 (72 FR
14308). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 4, 2007.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: July 12, 2006, as supplemented
on December 7, 2006, January 26, 2007, May 8, 2007, August 14, 2007,
and August 22, 2007.
Brief description of amendments: The amendments revise the
technical specifications to establish 674 feet as the minimum water
level of the ultimate heat sink and 87 [deg]F as the maximum supply
header temperature of the emergency raw water cooling system.
Date of issuance: September 28, 2007
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 317 and 307.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revised the technical specifications.
Date of initial notice in Federal Register: August 15, 2006 (71 FR
46939). The supplements dated December 7, 2006, January 26, 2007, May
8, 2007, August 14, 2007, and August 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 as
published in the Federal Register. The Commission's related evaluation
of the amendments is contained in a safety evaluation dated September
28, 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: April 18, 2007, as supplemented by
letters dated July 20, and October 2, 2007.
Brief description of amendments: Amendments revise the licenses to
reflect changes in legal name of TXU Generation Company LP to Luminant
Generation Company LLC.
Date of issuance: October 9, 2007.
Effective date: As of the date of issuance and shall be implemented
within 7 days from the date of issuance.
Amendment Nos.: Unit 1-139; Unit 2-139.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Facility Operating Licenses.
Date of initial notice in Federal Register: June 13, 2007 (72 FR
32685). The supplemental letters dated July 20 and October 2, 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 September 10, 2007.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 15th day of October, 2007.
For the Nuclear Regulatory Commission.
John P. Boska,
Acting Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. E7-20679 Filed 10-22-07; 8:45 am]
BILLING CODE 7590-01-P