[Federal Register Volume 71, Number 59 (Tuesday, March 28, 2006)]
[Notices]
[Pages 15479-15494]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 06-2908]
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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
[[Page 15480]]
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 March 3, 2006 to March 16, 2006. The last
biweekly notice was published on March 14, 2006 (71 FR 13169).
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, Rules and
Directives 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
[[Page 15481]]
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 1, DeWitt County, Illinois
Date of amendment request: December 1, 2006.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.6.4.1, ``Secondary Containment.''
Specifically, the change would modify Surveillance Requirements (SRs)
3.6.4.1.4 and 3.6.4.1.5 to clarify their intent with respect to
secondary containment boundary integrity. The change is submitted in
accordance with the TS Task Force Traveler 322-A, Revision 2,
``Secondary Containment and Shield Building Boundary Integrity SRs.''
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. This change involves an administrative clarification
to reflect the original intent of the Technical Specifications. There
is no impact on the availability or capability of the secondary
containment or Standby Gas Treatment (SGT) system as a result of the
proposed change. Both the secondary containment and SGT system are
considered accident-mitigating equipment and are not initiators of any
previously evaluated accidents. Therefore, the proposed change does not
involve an increase in the probability of an accident previously
evaluated. Additionally, the proposed change does not alter the
secondary containment or SGT systems' performance measures or their
ability to perform their accident mitigation functions.
Therefore, the proposed change 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 accident previously evaluated?
Response: No. The proposed changes to the wording of TS SRs
3.6.4.1.4 and 3.6.4.1.5 clarify that only one SGT subsystem is required
to ensure the requirements of TS 3.6.4.1 are met. The proposed change
does not alter the parameters within which the plant is operated. There
are no new system operating conditions or performance measures
introduced by this proposed change that will affect the secondary
containment and SGT systems' protective or mitigative functions. The
proposed changes will not alter the methods in which equipment is
operated or tested. No new accident scenarios or assumptions, failure
mechanisms, or limiting single failures are introduced as a result of
the proposed change.
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 proposed change does not impact any of these margins of
safety parameters. This change involves an administrative clarification
to reflect the original intent of the TS. There is no adverse effect on
the operability or design requirements of the secondary containment or
SGT system. The equipment will continue to be tested in a manner and at
a frequency necessary to provide confidence that the equipment can
perform its intended safety function. There is no impact on the plant
safety analyses.
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, Assistant General
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett
Square, PA 19348.
NRC Branch Chief: Daniel S. Collins.
[[Page 15482]]
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 amendments request: February 27, 2006.
Description of amendments request: The amendment would revise
Technical Specification 4.2.1, ``Fuel Assemblies,'' to allow fuel with
advanced cladding material to be installed in the core for Cycle 19
only at Unit No. 1 or Cycle 17 only at Unit No. 2. Advanced cladding
material from Framatome-ANP may be used in up to 2 lead test
assemblies, and advanced cladding material from Westinghouse may be
used in up to 2 lead test assemblies.
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. Would not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Calvert Cliffs Technical Specification 4.2.1, Fuel Assemblies,
states that fuel rods are clad with either Zircaloy or
ZIRLOTM. Calvert Cliffs Nuclear Power Plant, Inc. proposes
to re-insert up to four fuel assemblies into Calvert Cliffs Unit 1 or
Unit 2 that have some fuel rods clad in zirconium alloys that do not
meet the definition of Zircaloy or ZIRLOTM. A temporary
exemption to the regulations has also been requested to allow these
fuel assemblies to be re-inserted into Unit 1 or Unit 2. The proposed
change to the Calvert Cliffs Technical Specifications will allow the
use of cladding materials that are not Zircaloy or ZIRLOTM
for one fuel cycle once the temporary exemption is approved. The
proposed change to the Technical Specification is effective only as
long as the temporary exemption is effective. The addition of what will
be an approved temporary exemption for Unit 1 or Unit 2 to Technical
Specification 4.2.1 does not change the probability or consequences of
an accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Would not create the possibility of a new or different [kind] of
accident from any accident previously evaluated.
The proposed change does not add any new equipment, modify any
interfaces with existing equipment, change the equipment's function, or
change the method of operating the equipment. The proposed change does
not affect normal plant operations or configuration. Since the proposed
change does not change the design, configuration, or operation, it
could not become an accident initiator.
Therefore, the proposed change does not create the possibility of a
new or different [kind] of accident from any previously evaluated.
3. Would not involve a significant reduction in [a] margin of
safety.
The proposed change will add an approved temporary exemption to the
Calvert Cliffs Technical Specifications allowing the installation of up
to four lead fuel assemblies. The assemblies use advanced cladding
materials that are not specifically permitted by existing regulations
or Calvert Cliffs' Technical Specifications. A temporary exemption to
allow the installation of these assemblies has been requested. The
addition of an approved temporary exemption to Technical Specification
4.2.1 is an administrative change to allow the installation of the lead
fuel assemblies under the provisions of the temporary exemption. The
license amendment is effective only as long as the exemption is
effective. This amendment does not change the margin of safety since it
only adds a reference to an approved, temporary exemption to the
Technical Specifications.
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
amendments request involves no significant hazards consideration.
Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear
Generation, Constellation Generation Group, LLC, 750 East Pratt Street,
17th floor, Baltimore, MD 21202.
NRC Branch Chief: Richard J. Laufer.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: January 5, 2005, supplemented November
21, 2005.
Description of amendment request: The proposed amendments would
revise the Technical Specification (TS) 5.5.19.b, TS 5.5.19.c, and TS
Surveillance Requirement (SR) 3.8.1.9. TS 5.5.19.b currently requires
verification that a Lee Combustion Turbine (LCT) can supply the
equivalent of one Unit's maximum safeguard loads, plus two Units' Mode
3 loads, when connected to the system grid every 12 months. In the
proposed amendments, this requirement would be more clearly specified
as, ``Verify an LCT can supply equivalent of one Unit's Loss of Coolant
Accident (LOCA) loads plus two Unit's Loss of Offsite Power (LOOP)
loads when connected to system grid every 12 months.'' TS 5.5.19.b and
SR 3.8.1.9 would be revised for consistency.
This notice supersedes the notice published in the Federal Register
on February 15, 2005 (70 FR 7764).
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) Involve a significant increase in the probability or
consequences of an accident previously evaluated:
Duke proposes to revise TS 5.5.19.b to clarify the Lee Combustion
Turbine (LCT) testing requirements. Duke proposes to revise TS 5.5.19.c
and TS 3.8.1 Surveillance Requirement (SR) 3.8.1.19 to be consistent
with the proposed change to TS 5.5.19.b. The proposed change makes the
wording of the test requirement consistent with the UFSAR [Updated
Final Safety Analysis Report]. LCT testing has no impact on the
probability of an accident analyzed in the UFSAR. The LCT can be
credited to mitigate the consequences of an accident analyzed in the
UFSAR. However, this clarification of LCT testing requirements has no
impact on its ability to mitigate the consequences of an accident. As
such, the proposed LAR [license amendment request] does not involve a
significant increase in the probability or consequences of an accident
previously evaluated.
(2) Create the possibility of a new or different kind of accident
from any kind of accident previously evaluated:
Duke proposes to revise TS 5.5.19.b to clarify the Lee Combustion
Turbine (LCT) testing requirements. Duke proposes to revise TS 5.5.19.c
and TS 3.8.1 SR 3.8.1.9 to be consistent with the proposed change to TS
5.5.19.b. The proposed change makes the wording of the test requirement
consistent with the UFSAR. These changes do not alter the nature of
events postulated in the Safety Analysis Report nor do they introduce
any unique precursor mechanisms. Therefore, the proposed amendment will
not create the possibility of a new or different kind of accident from
any accident previously evaluated.
(3) Involve a significant reduction in a margin of safety:
[[Page 15483]]
The proposed TS change does not unfavorably affect any plant safety
limits, set points, or design parameters. The changes also do not
unfavorably affect the fuel, fuel cladding, RCS [reactor coolant
system], or containment integrity. Therefore, the proposed TS change,
which clarifies TS requirements associated with the LCT testing
program, does not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Branch Chief: Evangelos C. Marinos.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: March 13, 2006.
Description of amendment request: The proposed amendments would
make changes to the technical specifications (TS) for LaSalle County
Station (LSCS), Units 1 and 2. Surveillance Requirement (SR) 3.7.3.1
verifies the cooling water temperature supplied to the plant from the
core standby cooling system (CSCS) pond (i.e., the ultimate heat sink
(UHS)) is <= 100 [deg]F. Currently, if the temperature of the cooling
water supplied to the plant from the CSCS pond is > 100 [deg]F, the UHS
must be declared inoperable in accordance with TS 3.7.3. TS 3.7.3,
Required Action B.1, requires that both units be placed in Mode 3
within 12 hours and Required Action B.2 requires that both units be
placed in Mode 4 within 36 hours.
Prolonged hot weather in the area during the summer months, in
conjunction with high humidity during the daytime, minimal cooling at
night and little precipitation, has resulted in sustained elevated
cooling water temperature supplied to the plant from the CSCS pond.
This license amendment is being requested to increase the temperature
limit of the cooling water supplied to the plant from the CSCS pond to
<= 101.5 [deg]F by reducing the temperature measurement uncertainty by
replacing the existing thermocouples with higher precision temperature
measuring equipment. Should the UHS indicated temperature exceed 101.5
[deg]F, Required Action B.1 would be entered and both units would be
placed in Mode 3 within 12 hours and Mode 4 within 36 hours.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed change will allow the indicated temperature of the cooling
water supplied to the plant from the CSCS pond to be increased to <=
101.5 [deg]F based on reducing the temperature measurement uncertainty
by replacing the existing thermocouples with higher precision
temperature measuring equipment.
Analyzed accidents are assumed to be initiated by the failure of
plant structures, systems, or components. An inoperable UHS is not
considered as an initiator of any analyzed events. As such, there is
not a significant increase in the probability of a previously evaluated
accident. Allowing the UHS to operate at a higher allowable indicated
temperature, but still within the design limits of the equipment it
supplies, will not affect the failure probability of that equipment.
The current heat analyses calculations of record for LSCS, Units 1 and
2, assume a UHS temperature of 100 [deg]F and post-accident peak inlet
temperature of 104 [deg]F. The proposed temperature increase is based
solely on a reduction of the existing instrument loop uncertainty
value. The current analysis bounds the proposed change. This higher
allowable indicated temperature does not impact the LOCA [loss-of-
coolant accident] Peak Clad Temperature Analysis, LOCA Containment
Analysis or the non-LOCA analyses; therefore, continued operation with
a UHS temperature > 100 [deg]F but <= 101.5 [deg]F will not increase
the consequences of an accident previously evaluated in the UFSAR.
Based on the above information, the increase in the allowable
indicated temperature of the cooling water supplied to the plant from
the UHS to <= 101.5 [deg]F by reducing the existing instrument loop
uncertainty value has no effect on the result of the design basis event
and will continue to allow each required heat exchanger to perform its
safety function. The heat exchangers will continue to provide
sufficient cooling for the heat loads during the most severe 30-day
period.
Based on the above information, increasing the allowable indicated
temperature of the cooling water supplied to the plant from the CSCS
pond from <= 100 [deg]F to <= 101.5 [deg]F by reducing the instrument
uncertainty value has no impact on any analyzed accident; therefore,
the proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed change create the possibility of a new or different
kind of accident from any previously evaluated?
The proposed change involves replacing the presently installed
thermocouples with higher accuracy temperature measurement equipment.
This proposed action will not alter the manner in which equipment is
operated, nor will the functional demands on credited equipment be
changed. No alteration in the procedures that ensure the units remain
within analyzed limits is proposed, and no change is being made to
procedures relied upon to respond to an off-normal event. Raising the
UHS temperature limit does not introduce any new or different modes of
plant operation, nor does it affect the operational characteristics of
any safety-related equipment or systems; as such, no new failure modes
are being introduced. The proposed action reduces the instrument
uncertainty value but does not alter assumptions made in the safety
analysis.
Increasing the allowable indicated temperature of the cooling water
supplied to the plant from the CSCS pond from <= 100 [deg]F to <= 101.5
[deg]F has no impact on safety related systems. The plant is designed
such that the RHR [residual heat removal] pumps on the unit undergoing
the LOCA/LOOP [loss of offsite power] conditions would start upon the
receipt of a signal, and would load onto their respective Emergency
Diesel Generators emergency bus during the LOOP event. The increase in
the allowable indicated temperature of the cooling water supplied to
the plant from the CSCS pond will not require operation of additional
RHR pumps; therefore, system operation is unaffected by the proposed
change in the UHS temperature limit.
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. The proposed change does not involve a significant reduction in
a margin of safety.
The proposed change allows an increase in the allowable indicated
temperature of the cooling water supplied to the plant from the CSCS
[[Page 15484]]
pond to <= 101.5 [deg]F. The margin of safety is determined by the
design and qualification of the plant equipment, the operation of the
plant within analyzed limits, and the point at which protective or
mitigative actions are initiated. The proposed action does not impact
these factors as the analyzed peak inlet temperature of the UHS is
unaffected based on the improved instrument uncertainty of the new high
precision temperature measurement instrumentation. No setpoints are
affected, and no other change is being proposed in the plant
operational limits as a result of this change. All accident analysis
assumptions and conditions will continue to be met. Adequate design
margin is available to ensure that the required margin of safety is not
significantly reduced.
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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Assistant General
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett
Square, PA 19348.
NRC Branch Chief: Daniel S. Collins.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry
Nuclear Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: February 14, 2006.
Description of amendment request: The proposed amendment would
revise the frequency of the Mode 5 Intermediate Range Monitoring (IRM)
Instrumentation CHANNEL FUNCTIONAL TEST contained in Technical
Specification (TS) 3.3.1.1 from 7 days to 31 days. The methodology used
for the IRM drift analysis is based upon guidance contained in Generic
Letter 91-04, ``Changes in Technical Specification Surveillance
Intervals to Accommodate a 24-month Fuel Cycle,'' and Electric Power
Institute Report TI-103335, ``Guidance for Instrument Calibration
Extension/Reduction Programs.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed Technical Specifications (TS) change involves an
increase in the Mode 5 CHANNEL FUNCTIONAL TEST interval for RPS
[Reactor Protection System] IRM channels from 7 days to 31 days. The
IRM system is used for event mitigation. The failure of an IRM does not
initiate an accident or transient event. The proposed TS change does
not alter the design or function of the IRM system for no physical
changes are being made to the plant. Evaluation of the proposed testing
interval change demonstrated that the availability of IRMs to mitigate
the consequences of a control rod withdrawal event at low power levels
are not significantly affected based on the effectiveness of other,
required TS surveillance testing that is performed, the availability of
redundant systems and equipment, and the high reliability of the IRM
equipment.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed TS change involves an increase in the Mode 5 IRM
CHANNEL FUNCTIONAL TEST interval from 7 clays [days] to 31 days.
Existing TS testing requirements ensure the operability of the IRMs.
The proposed TS change does not introduce any failure mechanisms of a
different type than those previously evaluated, since no physical
changes to the plant are being made. No new or different equipment is
being installed, and no installed equipment is being operated in a
different manner. As a result, no new failure modes are introduced. In
addition, the manner in which surveillance tests are performed remains
unchanged.
Therefore, the proposed TS change does not create the possibility
of a new or different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction in
a margin of safety.
The proposed TS change involves an increase in the Mode 5 CHANNEL
FUNCTIONAL TEST interval for RPS IRM channels from 7 days to 31 days.
There is expected to be no impact on system operability, based upon the
performance of the more frequent Channel Checks, Control Room
monitoring when the IRMs are in use, and the overall IRM reliability.
Furthermore, a historical review of surveillance test results and
associated maintenance records did not indicate evidence of any failure
that would invalidate the above conclusions.
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: David W. Jenkins, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: Mindy S. Landau, Acting.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York
Date of amendment request: December 16, 2005.
Description of amendment request: The proposed change to Technical
Specification (TS) Surveillance Requirement (SR) 4.1.4d relocates the
SR for testing the core spray header differential pressure ([Delta]P)
instrumentation to licensee-controlled documents. TS SR 4.1.4d
currently requires that the core spray header [Delta]P instrumentation
be periodically tested such that a check of each sensor is performed at
least once each day and each channel is calibrated and tested at least
once every 3 months. The proposed change will allow these SRs to be
placed in licensee-controlled documents where future changes will be
made pursuant to Title 10 of the Code of Federal Regulations (10 CFR),
Section 50.59. The functional description of the core spray header
[Delta]P instrumentation will also be relocated from the TS Bases to
licensee-controlled documents consistent with the proposed TS change.
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 are limited to the relocation of selected
instrumentation requirements. The proposed relocated requirements were
[[Page 15485]]
determined not to meet the 10 CFR 50.36 screening criteria for
retention in the TSs and will be maintained in licensee-controlled
documents in accordance with the provisions of 10 CFR 50.59. The
proposed changes do not introduce any new modes of plant operation,
make any physical changes to the plant, or alter any operational
setpoints which could degrade the performance of any safety system
assumed to function in the accident analysis. Therefore, the proposed
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No. The proposed changes do not introduce any new modes
of plant operation, make any physical changes to the plant, or alter
any operational setpoints which could create new accident initiators or
failure mechanisms. The proposed changes are limited to the relocation
of selected instrumentation requirements, and will have no impact on
the accident assumptions and initial conditions as previously analyzed
in the UFSAR [Updated Final Safety Analysis Report]. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No. The proposed changes are consistent with the Improved
Standard TSs (NUREG-1433, Rev. 3) and will have no impact on the
instrumentation setpoints, logic, or functional requirements as
described in the TSs, TS Bases, and UFSAR. The proposed relocated
requirements were determined to not meet the 10 CFR 50.36 screening
criteria for retention in the TSs. Thus, the relocated requirements
will be maintained in accordance with 10 CFR 50.59 as required.
Accordingly, the proposed relocated requirements will not degrade the
quality or performance of any safety system assumed to mitigate an
accident or assure operation within the safety limits. Therefore, the
proposed changes do not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1700 K Street, NW., Washington, DC 20006.
NRC Branch Chief: Richard J. Laufer.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of amendment request: February 28, 2006.
Description of amendment request: The proposed amendments would
change the SSES 1 and 2 Technical Specification (TS) Surveillance
Requirements (SRs) 3.8.4.7 and 3.8.4.8 to clarify that diesel generator
``E'' (DG E) electrical power subsystem testing does not require a mode
restriction when the DG E diesel is not required to be OPERABLE.
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. Performance of TS required SRs are not initiators to
any accident sequences analyzed in the Final Safety Analysis Report
(FSAR). The changes do not involve any physical change to structures,
systems, or components, (SSCs) and do not alter the method of operation
or control of SSCs. The current assumptions in the safety analysis
regarding accident initiators and mitigation of accidents are
unaffected by these changes. No additional failure modes or mechanisms
are being introduced and the likelihood of previously analyzed failures
remains unchanged.
Operation in accordance with the proposed Technical Specification
(TS) ensures that the DC [direct current] distribution system and
supported equipment functions remain capable of performing the function
as described in the FSAR. Therefore, the mitigative functions supported
by the system will continue to provide the protection assumed by the
analysis.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No. The proposed change does not involve a physical
alteration of the plant. No new equipment is being introduced, and
installed equipment is not being operated in a new or different manner.
There are no setpoints, at which protective or mitigative actions are
initiated, affected by this change. This change will not alter the
manner in which equipment operation is initiated, nor will the function
demands on credited equipment be changed. No alterations in the
procedures that ensure the plant remains within analyzed limits are
being proposed, and no changes are being made to the procedures relied
upon to respond to an off-normal event as described in the FSAR. As
such, no new failure modes are being introduced. The change does not
alter assumptions made in the safety analysis and licensing basis.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No. The margin of safety is established through equipment
design, operating parameters, and the setpoints at which automatic
actions are initiated. The proposed change is acceptable because
performance of SRs on equipment not require[d] to be OPERABLE and
isolated from the OPERABLE plant equipment cannot affect any margin of
safety. Therefore, the plant response to analyzed events will continue
to provide the margin of safety assumed by the analysis.
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: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Branch Chief: Richard J. Laufer
Southern California Edison Company (SCE), et al., Docket Nos. 50-361
and 50-362, San Onofre Nuclear Generating Station, Units 2 and 3 (SONGS
2 and 3), San Diego County, California
Date of amendment requests: March 10, 2006.
Description of amendment requests: The licensee requests the
Nuclear Regulatory Commission consent to the transfer of the City of
Anaheim's 3.16 percent undivided ownership interest in SONGS 2 and 3 to
Southern California Edison, excluding Anaheim's interest in its spent
fuel and the SONGS 2 and 3 independent spent fuel storage installation.
[[Page 15486]]
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 amendments do not involve any change in the
design, configuration, or operation of the nuclear plant. All Limiting
Conditions for Operation, Limiting Safety System Settings, and Safety
Limits specified in the Technical Specifications remain unchanged. SCE
will continue to be the licensed operator of the units.
The technical qualifications of SCE to carry out its exclusive
responsibilities under the operating licenses, as amended, will remain
unchanged. Personnel engaged in operation, maintenance, engineering,
assessment, training, and other related services are not changed. The
SCE officers and executives currently responsible for the overall safe
operation of the nuclear plants will continue in the same capacity.
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 amendments do not involve any change in the
design, configuration, or operation of the nuclear plant. The current
plant design and design bases will remain the same. The current plant
safety analyses, therefore, remain complete and accurate in addressing
the design basis events and in analyzing plant response and
consequences.
The Limiting Conditions for Operation, Limiting Safety System
Settings, and Safety Limits specified in the Technical Specifications
are not affected by the change. As such, the plant conditions for which
the design basis accident analyses were performed remain valid.
The amendments do not introduce a new mode of plant operation or
new accident precursors, do not involve any physical alterations to
plant configurations, or make changes to system set points that could
initiate a new or different kind 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 amendments do not involve a change in the design,
configuration, or operation of the nuclear plants. The change does not
affect either the way in which the plant structures, systems, and
components perform their safety function, or their design and licensing
basis.
Plant safety margins are established through Limiting Conditions
for Operation, Limiting Safety System Settings, and Safety Limits
specified in the Technical Specifications. Because there is no change
to the physical design of the plant, there is no change to any of these
margins.
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: David Terao.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of amendment request: September 19, 2005.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Limiting Conditions for Operation
(LCO) 3.3.1, ``Reactor Trip system (RTS) Instrumentation'' and TS
Surveillance Requirements (SR) 3.2.4.2, ``Quadrant Power Tilt Ration
(QPTR)'' to avoid confusion as to when a flux map for QPTR is required.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed changes do not adversely affect accident
initiators or precursors nor alter the design assumptions, conditions,
or configuration of the facility or the manner in which the plant is
operated and maintained. The proposed changes do not alter or prevent
the ability of structures, systems, and components (SSCs) from
performing their intended function to mitigate the consequences of an
initiating event within the assumed acceptance limits. The proposed
changes do not affect the source term, containment isolation, or
radiological release assumptions used in evaluating the radiological
consequences of an accident previously evaluated. Further, the proposed
changes do not increase the types or amounts of radioactive effluent
that may be release offsite, nor significantly increase individual or
cumulative occupational/public radiation exposures. The proposed
changes are consistent with safety analysis assumptions and resultant
consequences.
Therefore, the proposed changes do not increase 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 previously evaluated?
No. The proposed changes do not result in a change in the manner in
which the RTS and ESFAS provide plant protection. The RTS and ESFAS
will continue to have the same set points after the proposed changes
are implemented. There are no design changes associated with the
license amendment.
The changes do not involve a physical alteration of the plant
(i.e., no new or different type of equipment will be installed) or a
change in the methods governing normal plant operation. In addition,
the changes do not impose any new or different requirements or
eliminate any existing requirements. The changes do not alter
assumptions made in the safety analysis. The proposed changes are
consistent with the safety analysis assumptions and current plant
operating practice.
Therefore, the proposed changes do 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?
No. The proposed changes do not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The safety analysis acceptance criteria are
not impacted by these changes. Redundant RTS and ESFAS trains are
maintained, and diversity with regard to the signals that provide
reactor trip and engineered safety features actuation is also
[[Page 15487]]
maintained. All signals credited as primary or secondary, and all
operator actions credited in the accident analyses will remain the
same. The proposed changes will not result in plant operation in a
configuration outside the design basis.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Branch Chief: Evangelos C. Marinos.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama
Date of amendment request: January 10, 2006 (TS-453).
Description of amendment request: The proposed amendment would
specify the methodology used for determining, setting, and evaluating
as-found setpoints for those drift susceptible instruments, which are
either necessary to ensure compliance with a Safety Limit or critical
in ensuring the fuel peak cladding temperature acceptance criteria of
10 CFR 50.46 are met.
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. Including references to TVA's methodology for
determining, setting, and evaluating as-found instrument setpoints in
the TS is an administrative change. There will be no change to the
manner in which Safety Limits, Analytical Limits, or Allowable Values
are determined. No changes are proposed in the manner in which the
Reactor Protection System (RPS), Emergency Core Cooling System (ECCS),
Reactor Core Isolation Cooling (RCIC), or Primary Containment Isolation
systems provide plant protection or which create new modes of plant
operation.
The proposed request will not affect the probability of any event
initiators. There will be no degradation in the performance of, or an
increase in the number of challenges imposed on, safety-related
equipment assumed to function during an accident situation. There will
be no change to normal plant operating parameters or accident
mitigation performance.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No. There are no hardware changes nor are there any
changes in the method by which any plant system performs a safety
function. This request does not affect the normal method of plant
operation. The proposed amendment does not introduce new equipment,
which could create a new or different kind of accident.
No new external threats, release pathways, or equipment failure
modes are created. No new accident scenarios, transient precursors,
failure mechanisms, or limiting single failures are introduced as a
result of this request. Therefore, the implementation of the proposed
amendment will not create a possibility for an accident of a new or
different type than those previously evaluated.
3. Does the proposed amendment involve a significant reduction in a
margin of safety?
Response: No. Including references to TVA's methodology for
determining, setting, and evaluating as-found instrument setpoints in
the TS is an administrative change. No changes are proposed in the
manner in which the RPS, ECCS, RCIC, or Primary Containment Isolation
systems satisfy the Updated Final Safety Analysis Report requirements
for accident mitigation or unit safe shutdown. There will be no change
to Safety Limits, Analytical Limits, Allowable Values, or post-Loss Of
Coolant Accident peak clad temperatures. For these reasons, 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: Michael L. Marshall, Jr.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: February 6, 2006.
Description of amendment request: The proposed amendment would
modify technical specification (TS) requirements for inoperable
snubbers by adding Limiting Condition for Operation 3.0.7. The changes
are consistent with Nuclear Regulatory Commission approved Industry/
Technical Specification Task Force (TSTF) standard TS change TSTF-373,
Revision 4. The availability of this TS improvement was published in
the Federal Register on May 4, 2005 (70 FR 23252), 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), 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 allows a delay time before
declaring supported TS systems inoperable when the associated
snubber(s) cannot perform its required safety function. Entrance into
Actions or delaying entrance into Actions is not an initiator of any
accident previously evaluated. Consequently, the probability of an
accident previously evaluated is not significantly increased. The
consequences of an accident while relying on the delay time allowed
before declaring a TS supported system inoperable and taking its
Conditions and Required Actions are no different than the consequences
of an accident under the same plant conditions while relying on the
existing TS supported system Conditions and Required Actions.
Therefore, the consequences of an accident previously evaluated are not
significantly increased by this change. Therefore, this change does not
involve a significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind
[[Page 15488]]
of accident from any accident previously evaluated?
Response: No. The proposed change allows a delay time before
declaring supported TS systems inoperable when the associated
snubber(s) cannot perform its required safety function. The proposed
change does not involve a physical alteration of the plant (no new or
different type of equipment will be installed) or a change in the
methods governing normal plant operations. Thus, 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 allows a delay time before
declaring supported TS systems inoperable when the associated
snubber(s) cannot perform its required safety function. The proposed
change restores an allowance in the pre-ISTS conversion TS that was
unintentionally eliminated by the conversion. The pre-ISTS TS were
considered to provide an adequate margin of safety for plant operation,
as does the post-ISTS conversion TS. Therefore, this 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 Section Chief: Michael L. Marshall, Jr.
Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant,
Unit 2, Hamilton County, Tennessee
Date of amendment request: February 15, 2006.
Description of amendment request: The amendment would revise the
Technical Specifications (TS) to adopt NRC-approved Revision 4 to
Technical Specification Task Force (TSTF) Standard Technical
Specification Change Traveler, TSTF-449, ``Steam Generator Tube
Integrity.'' The proposed amendment includes changes to the TS
definition of Leakage, TS 3.4.6.2, ``Reactor Coolant System,
Operational Leakage,'' TS 3.4.5, ``Steam Generator (SG) Tube
Integrity,'' and adds TS 6.8.4.k, ``Steam Generator (SG) Program,'' and
TS 6.9.1.16, ``Steam Generator Tube Inspection Report.'' The proposed
changes are necessary in order to implement the guidance for the
industry initiative on NEI 97-06, ``Steam Generator Program
Guidelines.''
The amendment would also delete License Condition 2.C.8 Item b.
This License Condition references the licensee's letters from 1997 that
contain commitments associated with NRC Generic Letter 95-05,
``Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes
Affected by Outside Diameter Stress Corrosion Cracking,'' and the
application of voltage-based alternate repair criteria to the steam
generators. The licensee has concluded that the provisions and
requirements of the proposed TS changes bound the commitments
identified in the existing License Condition.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on March 2, 2005 (70 FR 10298), on possible amendments
adopting TSTF-449, 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 May 6, 2005
(70 FR 24126). The licensee affirmed the applicability of the following
NSHC determination in its application dated August 31, 2005.
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 requires a SG Program that includes performance
criteria that will provide reasonable assurance that the SG tubing will
retain integrity over the full range of operating conditions (including
startup, operation in the power range, hot standby, cooldown and all
anticipated transients included in the design specification). The SG
performance criteria are based on tube structural integrity, accident
induced leakage, and operational LEAKAGE.
A steam generator tube rupture (SGTR) event is one of the design
basis accidents that are analyzed as part of a plant's licensing basis.
In the analysis of a SGTR event, a bounding primary to secondary
LEAKAGE rate equal to the operational LEAKAGE rate limits in the
licensing basis plus the LEAKAGE rate associated with a double-ended
rupture of a single tube is assumed.
For other design basis accidents such as a main steamline break
(MSLB), rod ejection, and reactor coolant pump locked rotor the tubes
are assumed to retain their structural integrity (i.e., they are
assumed not to rupture). These analyses typically assume that primary
to secondary LEAKAGE for all SGs is 1 gallon per minute or increases to
1 gallon per minute as a result of accident induced stresses. The
accident induced leakage criterion introduced by the proposed changes
accounts for tubes that may leak during design basis accidents. The
accident induced leakage criterion limits this leakage to no more than
the value assumed in the accident analysis.
The SG performance criteria proposed change to the TS identify the
standards against which tube integrity is to be measured. Meeting the
performance criteria provides reasonable assurance that the SG tubing
will remain capable of fulfilling its specific safety function of
maintaining reactor coolant pressure boundary integrity throughout each
operating cycle and in the unlikely event of a design basis accident.
The performance criteria are only a part of the SG Program required by
the proposed change to the TS. The program, defined by NEI 97-06, Steam
Generator Program Guidelines, includes a framework that incorporates a
balance of prevention, inspection, evaluation, repair, and leakage
monitoring. The proposed changes do not, therefore, significantly
increase the probability of an accident previously evaluated.
The consequences of design basis accidents are, in part, functions
of the DOSE EQUIVALENT I-131 in the primary coolant and the primary to
secondary LEAKAGE rates resulting from an accident. Therefore, limits
are included in the plant technical specifications for operational
leakage and for DOSE EQUIVALENT I-131 in primary coolant to ensure the
plant is operated within its analyzed condition. The typical analysis
of the limiting design basis accident assumes that primary to secondary
leak rate after the accident is 1 gallon per minute and that the
reactor coolant activity levels of DOSE EQUIVALENT I-131 are at the TS
values before the accident.
The proposed change does not affect the design of the SGs, their
method of operation, or primary coolant chemistry controls. The
proposed approach updates the current TSs and enhances the requirements
for SG inspections. The proposed change does not adversely impact any
other previously evaluated
[[Page 15489]]
design basis accident and is an improvement over the current TSs.
Therefore, the proposed change does not affect the consequences of
a SGTR accident and the probability of such an accident is reduced. In
addition, the proposed changes do not affect the consequences of an
MSLB, rod ejection, or a reactor coolant pump locked rotor event, or
other previously evaluated accident.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed performance based requirements are an improvement over
the requirements imposed by the current technical specifications.
Implementation of the proposed SG Program will not introduce any
adverse changes to the plant design basis or postulated accidents
resulting from potential tube degradation. The result of the
implementation of the SG Program will be an enhancement of SG tube
performance. Primary to secondary LEAKAGE that may be experienced
during all plant conditions will be monitored to ensure it remains
within current accident analysis assumptions.
The proposed change does not affect the design of the SGs, their
method of operation, or primary or secondary coolant chemistry
controls. In addition, the proposed change does not impact any other
plant system or component. The change enhances SG inspection
requirements.
Therefore, the proposed change does not create the possibility of a
new or different type of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The SG tubes in pressurized water reactors are an integral part of
the reactor coolant pressure boundary and, as such, are relied upon to
maintain the primary system's pressure and inventory. As part of the
reactor coolant pressure boundary, the SG tubes are unique in that they
are also relied upon as a heat transfer surface between the primary and
secondary systems such that residual heat can be removed from the
primary system. In addition, the SG tubes isolate the radioactive
fission products in the primary coolant from the secondary system. In
summary, the safety function of an SG is maintained by ensuring the
integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change does not affect tube design or operating environment. The
proposed change is expected to result in an improvement in the tube
integrity by implementing the SG Program to manage SG tube inspection,
assessment, repair, and plugging. The requirements established by the
SG Program are consistent with those in the applicable design codes and
standards and are an improvement over the requirements in the current
TSs.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed change to the TS.
The NRC staff proposes to determine that the amendments 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: Michael L. Marshall, Jr.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: December 15, 2005.
Description of amendment request: The amendment would revise the
Technical Specifications (TS) to adopt NRC-approved Revision 4 to
Technical Specification Task Force (TSTF) Standard Technical
Specification Change Traveler, TSTF-449, ``Steam Generator Tube
Integrity.'' The proposed amendment includes:
--Revised TS definition of Leakage,
--Revised TS 3.4.13, ``RCS [Reactor Coolant System] Operational
Leakage,''
--Added new TS 3.4.17, ``Steam Generator Tube Integrity,''
--Revised TS 5.7.2.12, ``Steam Generator (SG) Tube Surveillance
Program,'' and
--Revised TS 5.9.9, ``SG Tube Inspection Report.''
The proposed changes are necessary in order to implement the guidance
for the industry initiative on NEI 97-06, ``Steam Generator Program
Guidelines.''
The NRC staff issued a notice of opportunity for comment in the
Federal Register on March 2, 2005 (70 FR 10298), on possible amendments
adopting TSTF-449, 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 May 6, 2005
(70 FR 24126). The licensee affirmed the applicability of the following
NSHC determination in its application dated December 15, 2005.
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 requires a SG Program that includes performance
criteria that will provide reasonable assurance that the SG tubing will
retain integrity over the full range of operating conditions (including
startup, operation in the power range, hot standby, cooldown and all
anticipated transients included in the design specification). The SG
performance criteria are based on tube structural integrity, accident
induced leakage, and operational LEAKAGE.
A steam generator tube rupture (SGTR) event is one of the design
basis accidents that are analyzed as part of a plant's licensing basis.
In the analysis of a SGTR event, a bounding primary to secondary
LEAKAGE rate equal to the operational LEAKAGE rate limits in the
licensing basis plus the LEAKAGE rate associated with a double-ended
rupture of a single tube is assumed.
For other design basis accidents such as a main steamline break
(MSLB), rod ejection, and reactor coolant pump locked rotor the tubes
are assumed to retain their structural integrity (i.e., they are
assumed not to rupture). These analyses typically assume that primary
to secondary LEAKAGE for all SGs is 1 gallon per minute or increases to
1 gallon per minute as a result of accident induced stresses. The
accident induced leakage criterion introduced by the proposed changes
accounts for tubes that may leak during design basis accidents. The
accident induced leakage criterion limits this leakage to no more than
the value assumed in the accident analysis.
The SG performance criteria proposed change to the TS identify the
standards against which tube integrity is to be measured. Meeting the
performance criteria provides reasonable assurance that the SG tubing
will remain capable of fulfilling its specific safety function of
maintaining reactor coolant pressure boundary integrity throughout each
operating cycle and in the unlikely event of a design basis accident.
The performance criteria are only a part of the SG Program required by
the
[[Page 15490]]
proposed change to the TS. The program, defined by NEI 97-06, Steam
Generator Program Guidelines, includes a framework that incorporates a
balance of prevention, inspection, evaluation, repair, and leakage
monitoring. The proposed changes do not, therefore, significantly
increase the probability of an accident previously evaluated.
The consequences of design basis accidents are, in part, functions
of the DOSE EQUIVALENT I-131 in the primary coolant and the primary to
secondary LEAKAGE rates resulting from an accident. Therefore, limits
are included in the plant technical specifications for operational
leakage and for DOSE EQUIVALENT I-131 in primary coolant to ensure the
plant is operated within its analyzed condition. The typical analysis
of the limiting design basis accident assumes that primary to secondary
leak rate after the accident is 1 gallon per minute and that the
reactor coolant activity levels of DOSE EQUIVALENT I-131 are at the TS
values before the accident.
The proposed change does not affect the design of the SGs, their
method of operation, or primary coolant chemistry controls. The
proposed approach updates the current TSs and enhances the requirements
for SG inspections. The proposed change does not adversely impact any
other previously evaluated design basis accident and is an improvement
over the current TSs.
Therefore, the proposed change does not affect the consequences of
a SGTR accident and the probability of such an accident is reduced. In
addition, the proposed changes do not affect the consequences of an
MSLB, rod ejection, or a reactor coolant pump locked rotor event, or
other previously evaluated accident.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed performance based requirements are an improvement over
the requirements imposed by the current technical specifications.
Implementation of the proposed SG Program will not introduce any
adverse changes to the plant design basis or postulated accidents
resulting from potential tube degradation. The result of the
implementation of the SG Program will be an enhancement of SG tube
performance. Primary to secondary LEAKAGE that may be experienced
during all plant conditions will be monitored to ensure it remains
within current accident analysis assumptions.
The proposed change does not affect the design of the SGs, their
method of operation, or primary or secondary coolant chemistry
controls. In addition, the proposed change does not impact any other
plant system or component. The change enhances SG inspection
requirements.
Therefore, the proposed change does not create the possibility of a
new or different type of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The SG tubes in pressurized water reactors are an integral part of
the reactor coolant pressure boundary and, as such, are relied upon to
maintain the primary system's pressure and inventory. As part of the
reactor coolant pressure boundary, the SG tubes are unique in that they
are also relied upon as a heat transfer surface between the primary and
secondary systems such that residual heat can be removed from the
primary system. In addition, the SG tubes isolate the radioactive
fission products in the primary coolant from the secondary system. In
summary, the safety function of an SG is maintained by ensuring the
integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change does not affect tube design or operating environment. The
proposed change is expected to result in an improvement in the tube
integrity by implementing the SG Program to manage SG tube inspection,
assessment, repair, and plugging. The requirements established by the
SG Program are consistent with those in the applicable design codes and
standards and are an improvement over the requirements in the current
TSs.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed change to the TS.
The NRC staff proposes to determine that the amendments 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: Michael L. Marshall, Jr.
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 12, 2005.
Brief description of amendments: The amendments requested would
revise Technical Specification (TS) 3.3.1, ``RTS [Reactor Trip System]
Instrumentation,'' Surveillance Requirements (SRs) 3.3.1.2 and SR
3.3.1.3.
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. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No. Overall protection system performance will remain
within the bounds of the previously performed accident analyses since
there are no hardware changes. The Reactor Trip System (RTS)
Instrumentation will be unaffected. Protection systems will continue to
function in a manner consistent with the plant design basis. All
design, material, and construction standards that were applicable prior
to the request are maintained.
The probability and consequences of accidents previously evaluated
in the Final Safety Analysis Report (FSAR) are not adversely affected
because the change to the daily surveillance for the normalization of
the Nuclear Instrumentation System (NIS) Power Range and Nitrogen-16
(N-16) Power Monitor indications assures the conservative response of
the channel even at reduced power levels.
The proposed changes will not affect the probability of any event
initiators. There will be no degradation in the performance of, or an
increase in the number of challenges imposed on, safety-related
equipment assumed to function during an accident situation. There will
be no change to normal plant operating parameters or accident
mitigation performance.
The proposed changes will not alter any assumptions or change any
mitigation actions in the radiological consequence evaluations in the
FSAR.
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. There are no hardware changes nor are there any
changes in the method by which any safety-related plant system performs
its safety function. This amendment will not affect the normal method
of plant operation or change any operating parameters. No performance
requirements or response time limits will be affected.
[[Page 15491]]
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result of
this amendment. There will be no adverse effect or challenges imposed
on any safety-related system as a result of this amendment.
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. The proposed changes require a revision to the
criteria for implementation of NIS Power Range and N-16 Power Monitor
indication adjustments based on secondary power calorimetric
calculations; however, the changes do not eliminate any RTS
surveillances or alter the frequency of surveillances required by the
TS. The revision to the criteria for implementation of the daily
surveillance will remove a requirement for normalization of the NIS
Power Range and N-16 Power Monitor indications at reduced power
conditions that could result in safety performance outside the bounds
of the safety analyses. Therefore, the Nominal Trip Setpoints and
Allowable Values for the Reactor Trip System functions, as specified in
the TS and related Bases, as well as the safety analysis limits assumed
in the transient and accident analyses, are unchanged. None of the
acceptance criteria for any accident analysis is changed.
There will be no effect on the manner in which safety limits or
limiting safety systems settings are determined nor will there be any
effect on those plant systems necessary to assure the accomplishment of
protection functions. There will be no impact on the overpower limit,
departure from nucleate boiling ratio (DNBR) limits, heat flux hot
channel factor (FQ), nuclear enthalpy rise hot channel
factor (F[Delta]H), loss of coolant accident peak cladding temperature
(LOCA PCT), peak local power density, or any other margin of safety.
The radiological dose consequences are unaffected by this proposed
change.
The imposition of appropriate surveillance testing requirements
will not reduce any margin of safety since the changes will assure that
safety analysis assumptions on equipment operability are verified on a
periodic frequency.
Therefore the proposed change does not involve a 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: David Terao.
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.
Florida Power and Light Company, Docket No. 50-389, St. Lucie Plant,
Unit No. 2, St. Lucie County, Florida
Date of amendment request: February 14, 2006.
Description of amendment request: Revise the Technical
Specifications regarding the Containment Ventilation System to allow
additional corrective actions for inoperable containment purge supply
and exhaust valves.
Date of publication of individual notice in the Federal Register:
March 1, 2006 (71 FR 10566).
Expiration date of individual notice: March 15, 2006.
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 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site,
http://www.nrc.gov/reading-rm/adams.html.
If you do not have access
to ADAMS or if there are problems in accessing the documents
located in ADAMS, contact the PDR Reference staff at
1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
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 application for amendment: June 20, 2005, as supplemented
by letter dated November 2, 2005.
Brief description of amendment: This amendment revises the
footnotes in Tables 3.4-2 and 4.4-3 of Technical Specification (TS) 3/
4.4.7 by increasing the temperature limit above which (1) reactor
coolant sampling and analysis for dissolved oxygen is required, and (2)
when limit for dissolved oxygen, specified in TS 4.4.7, applies. This
temperature limit will be increased from 180 [deg]F to 250 [deg]F.
Date of issuance: March 8, 2006.
Effective date: March 8, 2006.
Amendment No. 120.
[[Page 15492]]
Facility Operating License No. NPF-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: October 11, 2005 (70 FR
59084). The supplemental letter provided clarifying information that
was within the scope of the initial notice and did not change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 8, 2006.
No significant hazards consideration comments received: No.
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 application for amendment: September 1, 2005, as
supplemented by letters dated December 22, 2005, and January 23, 2006.
Brief description of amendment: This amendment revises the
technical specification (TS) requirements for pressurized-water reactor
Boraflex fuel storage racks and adds TS requirements for fuel storage
pool boron concentration. Specifically, the amendment (1) adds a new TS
3/4.7.14, ``Fuel Storage Pool Boron Concentration,'' with a Limiting
Condition for Operation that requires a fuel pool boron concentration
of at least 2000 ppm at all times, (2) revises and reformats TS 5.6.1
to specify the design features and fuel storage limitations in
accordance with the categorization of spent fuel storage racks in
various spent fuel pools, and (3) revises TS 5.3.1 to remove the cross-
reference to TS 5.6.1.b.
Date of issuance: March 10, 2006.
Effective date: March 10, 2006.
Amendment No. 121.
Facility Operating License No. NPF-63: Amendment revises the TS.
Date of initial notice in Federal Register: November 8, 2005 (70 FR
67745). The supplemental letters provided additional information that
was within the scope of the initial notice and did not change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 10, 2006.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket Nos. 50-336 and 50-423,
Millstone Power Station, Unit Nos. 2 and 3, New London County,
Connecticut
Date of application for amendments: February 25, 2005.
Brief description of amendments: The amendments made various
administrative changes to the Technical Specifications (TSs).
Date of issuance: March 16, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 291 and 229.
Facility Operating License Nos. DPR-65 and NPF-49: The amendments
revised the TSs.
Date of initial notice in Federal Register: March 29, 2005 (70 FR
15942).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 16, 2006.
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: December 17, 2004.
Brief description of amendments: The amendments revised Appendix B,
Environmental Protection Plan (non-radiological), of the LaSalle County
Facility Operating Licenses.
Date of issuance: March 8, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 176/162.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Environmental Protection Plan.
Date of initial notice in Federal Register: April 12, 2005 (70 FR
19115).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 8, 2006.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of application for amendments: June 15, 2005.
Brief description of amendments: The amendments revised Technical
Specification (TS) 3.3.2.2, ``Feedwater System and Main Turbine High
Water Level Trip Instrumentation,'' to reflect a design change in the
instrumentation logic that trips the three feedwater pumps and main
turbine.
Date of issuance: March 9, 2006.
Effective date: As of the date of issuance and shall be implemented
prior to start-up from the spring 2006 refueling outage for Unit 2 and
prior to start-up from the spring 2007 refueling outage for Unit 1.
Amendment Nos.: 330/225.
Facility Operating License Nos. DPR-29 and DPR-30: The amendments
revised the Technical Specifications and Surveillance Requirements.
Date of initial notice in Federal Register: August 30, 2005 (70 FR
51381).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 9, 2006.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al. (FENOC), Docket No. 50-
346, Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: August, 20, 2004, as
supplemented by letters dated June 16 and December 6, 2005.
Brief description of amendment: The amendment revised TS 3/
4.8.1.1,``A.C. Sources--Operating,'' by deleting Surveillance
Requirement (SR) 4.8.1.1.2.d.4, which requires verification that the
emergency diesel generator auto-connected loads do not exceed the 2000-
hour load limit. In addition, the amendment revised TS 4/3.8.1.2,
``A.C. Sources--Shutdown,'' to add exceptions to SR 4.8.1.2 when
performed in Modes 5 and 6. As a result of discussions held on October
20, 2005, FENOC decided to withdraw the portion of the amendment
request (LAR 01-0009) that requested clarification of SR 4.8.1.1.b.
Date of issuance: March 2, 2006
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 273.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 28, 2004 (69
FR 57989).
The June 16 and December 6, 2005, supplements, contained clarifying
information and did not change the NRC staff's initial proposed finding
of no significant hazards consideration or expand the scope of the
original Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 2, 2006.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: May 2, 2005, as supplemented by
letters
[[Page 15493]]
dated August 28, September 15, 2005, and January 12, 2006, and January
13, February 9, and February 28, 2006.
Brief description of amendment: This amendment revised the
Technical Specifications (TSs) Section 2.1.1, ``Safety Limits--Reactor
Core,'' and TS Section 2.2.1, ``Limiting Safety Settings--Reactor
Protection System Setpoints.'' The amendment supports the use of the
Framatome Mark B-HTP fuel design for Cycle 15, which is scheduled to
begin following the refueling outage in March 2006.
Date of issuance: March 2, 2006
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 274.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 24, 2005 (70 FR
29796).
The August 28, September 15, 2005, and January 12, January 13,
February 9, and February 28, 2006, 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 March 2, 2006.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: August 25, 2005.
Brief description of amendment: The amendment revised the
definitions of Channel Calibration, Channel Function Test, and Logic
System Functional Test in accordance with the Technical Specification
Task Force Traveler (TSTF)-205-A.
Date of issuance: March 10, 2006
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 217.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 11, 2005 (70 FR
59086).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 10, 2006
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: August 11, 2005.
Brief description of amendment: The amendment deleted the
surveillance requirement (SR) of TS 2.10.2(9)b(iii) to verify shutdown
margin every 8-hour shift during low power physics testing. This change
made TS 2.10.2(9)b more consistent with SR 3.1.7 of NUREG-1432,
``Standard Technical Specifications Combustion Engineering Plants,
Revision 3.'' In addition, the Containment Structural Tests Report has
been deleted from TS 5.9.3c and several administrative and editorial
changes were made.
Date of issuance: February 1, 2006.
Effective date: February 1, 2006 and shall be implemented within 60
days from the date of issuance.
Amendment No.: 237.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: September 27, 2005 (70
FR 56503)
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated February 1, 2006.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of application for amendments: June 27, 2005, as supplemented
on December 1, 2005, and February 28, 2006.
Brief description of amendments: These amendments change the SSES 1
and 2 technical specifications for reactor protection system and
control rod block instrumentation, oscillation power range monitor
instrumentation, recirculation loops operating, shutdown margin test--
refueling, and the core operating limits report. The amendments modify
the power range neutron monitor system (PRNMS) by installation of the
General Electric Nuclear Measurement Analysis and Control PRNMS. The
modification of the PRNMS replaces analog technology with a digital
upgrade.
Date of issuance: March 3, 2006
Effective date: As of the date of issuance and to be implemented
prior to startup following the Cycle 14 refueling outage for Unit 1 and
the Cycle 13 refueling outage for Unit 2.
Amendment Nos.: 230 and 207.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 13, 2005 (70
FR 54088).
The supplements dated December 1, 2005, and February 28, 2006,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the staff's original proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 3, 2006.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric
Station, Unit 2 (SSES 2), Luzerne County, Pennsylvania
Date of application for amendment: March 18, 2005, as supplemented
on February 28, 2006.
Brief description of amendment: The amendment revises the SSES 2
Technical Specification 3.3.8.1, ``Loss of Power (LOP)
Instrumentation,'' to (1) clarify that Condition A applies to the LOP
instrumentation associated with both the Unit 1 and Unit 2 4.16
Kilovolt (kV) Engineered Safeguards System (ESS) buses since both the
Unit 1 and Unit 2 buses are required to support Unit 2 operation, (2)
add a new Condition B to allow the LOP instrumentation for two Unit 1
4.16kV ESS buses in the same division to be inoperable for up to 8
hours for the performance of Surveillance Requirement 3.8.1.19 on Unit
1. In addition, the amendment revises the SSES 2 TS 3.8.7,
``Distribution Systems--Operating,'' to (1) eliminate ``or more'' and
the plural to ``subsystems'' such that the condition will read ``one
Unit 1 AC [alternating current] electrical power distribution subsystem
inoperable,'' and (2) add a new Condition D for two Unit 1 AC
electrical power distribution subsystems inoperable.
Date of issuance: March 16, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 208.
Facility Operating License No. NPF-22: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 24, 2005 (70 FR
29800).
The supplement dated February 28, 2006, provided additional
information that clarified the application, did not expand the scope of
the application as
[[Page 15494]]
originally noticed, and did not change the staff's original proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 16, 2006.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: February 4, 2005.
Brief description of amendment: The amendment relocated the
Transversing In-Core Probe (TIP) system Technical Specification (TS) to
the Hope Creek Generating Station Updated Final Safety Analysis Report,
as well as removed the note on the TIP system from the Reactor
Protection System Instrumentation Surveillance Requirements table.
Date of issuance: March 8, 2006.
Effective date: As of the date of issuance, to be implemented
within 60 days from date of issuance.
Amendment No.: 164.
Facility Operating License No. NPF-57: This amendment revised the
TSs.
Date of initial notice in Federal Register: March 15, 2005 (70 FR
12750).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 8, 2006.
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: April 29, 2005, as supplemented
on July 1 and November 21, 2005.
Brief description of amendment: The amendment revised Technical
Specification 3.7.3, ``Main Feedwater Regulating Valves (MFRVs),
Associated Bypass Valves, and Main Feedwater Pump Discharge Valves
(MFPDVs),'' to allow the use of the main feedwater isolation valves in
lieu of the MFPDVs to provide isolation capability to the steam
generators in the event of a steam line break.
Date of issuance: March 16, 2006
Effective date: As of the date of issuance to be implemented prior
to startup from the fall 2006 refueling outage.
Amendment No.: 95.
Renewed Facility Operating License No. DPR-18: Amendment revised
the Technical Specifications.
Date of initial notice in Federal Register: June 7, 2005 (70 FR
33218).
The July 1 and November 21, 2005, letters 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 March 16, 2006.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant,
Unit 1, Hamilton County, Tennessee
Date of application for amendment: August 31, 2005.
Brief description of amendment: The amendment revises the Technical
Specifications associated with steam generator tube integrity
consistent with Revision 4 to Technical Specification Task Force (TSTF)
Standard Technical Specification Change Traveler, TSTF-449, ``Steam
Generator Tube Integrity.''
Date of issuance: February 23, 2006.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment No.: 306.
Facility Operating License No. DPR-77: Amendment revises the
technical specifications.
Date of initial notice in Federal Register: November 22, 2005 (70
FR 70643).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 23, 2006.
No significant hazards consideration comments received: No.
TXU Generation Company LP, Docket No. 50-446, Comanche Peak Steam
Electric Station, Unit No. 2, Somervell County, Texas
Date of amendment request: April 27, 2005, as supplemented by
letter dated July 20, 2005.
Description of amendment: The amendment revises the Technical
Specifications to add Topical Report WCAP-13060-P-A to the list of NRC
approved methodologies to be used at Comanche Peak Steam Electric
Station, Unit 2.
Date of issuance: March 15, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 123.
Facility Operating License No. NPF-89: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 8, 2005 (70 FR
67753).
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: October 6, 2004, as supplemented by
letters dated September 16 and November 22, 2005.
Brief description of amendments: The amendments revised the
Technical Specification 3.8.1, ``AC Sources--Operating,'' to remove
mode restrictions on surveillance requirements.
Date of issuance: March 15, 2006.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: 124.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 15, 2005 (70 FR
12751).
The supplements dated September 16 and November 22, 2005, 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 March 15, 2006.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 20th day of March 2006.
For the Nuclear Regulatory Commission.
Edwin M. Hackett,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 06-2908 Filed 3-27-06; 8:45 am]
BILLING CODE 7590-01-P