[Federal Register Volume 71, Number 205 (Tuesday, October 24, 2006)]
[Notices]
[Pages 62306-62318]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E6-17546]
[[Page 62306]]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission to publish notice of any amendments
issued, or proposed to be issued, and grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a determination by the Commission that such
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from September 29, 2006, to October 12, 2006. The
last biweekly notice was published on October 10, 2006 ( 71 FR 59529).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rulemaking,
Directives and Editing Branch, Division of Administrative Services,
Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the Commission's
Public Document Room (PDR), located at One White Flint North, Public
File Area O1F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. The filing of requests for a hearing and petitions for leave
to intervene is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted 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
[[Page 62307]]
fact. Contentions shall be limited to matters within the scope of the
amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendments request: August 16, 2006.
Description of amendments request: The proposed amendments would
revise several Surveillance Requirements (SRs) in Technical
Specification (TS) 3.8.1, ``AC Sources--Operating,'' to allow these SRs
to be performed, or partially performed, in reactor modes that
currently are not allowed by the TSs. The proposed changes would also
require certain SRs to be performed at a power factor of <=0.9 if
performed with the emergency diesel generators synchronized to the
grid, unless grid conditions do not permit.
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 emergency diesel generators (DGs) and their associated
emergency loads are accident mitigating features, rather than
accident initiating equipment. Each DG is dedicated to a specific
vital bus and these buses and DGs are independent of each other.
There is no common mode failure provided by the testing changes
proposed in this license amendment request (LAR) that would cause
multiple bus failures. Therefore, there will be no significant
impact on any accident probabilities by the approval of the
requested amendment.
The design of plant equipment is not being modified by these
proposed changes.
The changes include an increase in the online time the DG will
be paralleled to the grid in Mode[s] 1, 2, 3, and 4. The overall
time that the DG is paralleled in all modes (outage/non-outage)
should remain unchanged. As such, the ability of the DGs to respond
to a design basis accident (DBA) can be adversely impacted by the
proposed changes. However, the impacts are not considered
significant based on the DG under test maintaining its ability to
respond to an auto-start signal were one to be received during
testing, along with the ability of the remaining DG to mitigate a
DBA or provide a safe shutdown, and data that shows that the DG
itself will not perturb the electrical system significantly.
Furthermore, the proposed amendments for surveillance requirements
(SR) 3.8.1.10 and SR 3.8.1.14 share the same electrical
configuration alignment to the current monthly 1-hour loaded
surveillance.
SR changes that are consistent with Industry/Technical
Specification Task Force (TSTF) Standard Technical Specification
(STS) change TSTF-283, Revision 3 and NUREG-1432, Revision 2 have
been approved by the NRC, and the on-line tests allowed by the TSTF
and the NUREG are only to be performed for the purpose of
establishing operability of the DG being tested. Performance of
these SRs during previously restricted modes will require an
assessment to assure plant safety is maintained or enhanced.
The proposed changes to SRs 3.8.1.10 and 3.8.1.14 to require
that these SRs be performed at a power factor of <=0.9 if performed
with the emergency diesel generators synchronized to the grid unless
grid conditions do not permit are consistent with NRC-approved
NUREG-1432, Standard Technical Specifications, Combustion
Engineering Plants, and NRC-approved TSTF-276, Revision 2. This
requirement ensures that the DG is tested under load conditions that
are as close to design basis conditions as possible. A power factor
of <=0.9 is representative of the actual inductive loading a DG
would see under design basis accident conditions. Under certain
conditions, however, the proposed change allows the surveillance to
be conducted at a power factor other than <=0.9. These conditions
occur when grid voltage is high, and the additional field excitation
needed to get the power factor to <=0.9 results in voltages on the
emergency busses that are too high. Under these conditions, the
power factor should be maintained as close as practicable to 0.9
while still maintaining acceptable voltage limits on the emergency
busses. In other circumstances, the grid voltage may be such that
the DG excitation levels needed to obtain a power factor of 0.9
[[Page 62308]]
may not cause unacceptable voltages on the emergency busses, but the
excitation levels are in excess of those recommended for the DG. In
such cases, the power factor shall be maintained as close as
practicable to 0.9 without exceeding DG excitation limits.
As stated above, a power factor <=0.9 should be able to be
achieved when performing this SR at power and synchronized with
offsite power by transferring house loads from the auxiliary
transformer to the startup transformer in order to lower the Class
1E bus voltage. Transferring house loads from the auxiliary
transformer to the startup transformer is routinely performed at
power, in accordance with procedure 40OP-9NA03. The circuit breakers
supplying the house loads (NAN-S01 and NAN-S02) from the auxiliary
and startup transformers are interlocked such that one supply
breaker does not open until the alternate supply breaker is closed.
This ensures that the bus remains energized during the transfer.
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 accident from any accident previously evaluated?
Response: No.
The proposed changes would create no new accidents since no
changes are being made to the plant that would introduce any new
accident causal mechanisms. Equipment will be operated in the same
configuration currently allowed by other DG SRs that allow testing
in plant Modes 1, 2, 3, and 4. This license amendment request does
not impact any plant systems that are accident initiators or
adversely impact any accident mitigating systems.
Therefore, the proposed change does not create the possibility
of a new or different 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 do not involve a significant reduction in a
margin of safety. The margin of safety is related to the ability of
the fission product barriers to perform their design safety
functions during and following an accident situation. These barriers
include the fuel cladding, the reactor coolant system, and the
containment system. The proposed changes to the testing requirements
for the plant DGs do not affect the operability requirements for the
DGs, as verification of such operability will continue to be
performed as required (except during different allowed modes).
Continued verification of operability supports the capability of the
DGs to perform their required function of providing emergency power
to plant equipment that supports or constitutes the fission product
barriers. Only one DG is tested at a time and the remaining DG will
be available to safely shut down the plant or respond to a DBA, if
required. Consequently, the performance of these fission product
barriers will not be impacted by implementation of the proposed
amendment.
In addition, the proposed changes involve no changes to safety
setpoints or limits established or assumed by the accident analysis.
On this and the above basis, no safety margins will be impacted.
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
that review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: Michael G. Green, Senior Regulatory Counsel,
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695,
Phoenix, Arizona 85072-2034.
NRC Branch Chief: David Terao.
Dominion Energy Kewaunee, Inc., Docket No. 50-305, Kewaunee Power
Station, Kewaunee County, Wisconsin
Date of amendment request: September 25, 2006.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 4.2.a, ``ASME Code Class 1, 2, 3,
and MC Components and Supports.'' The revised TS 4.2.a, Item 2, would
reference the American Society of Mechanical Engineers Code for
Operation and Maintenance.
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 change revises the Kewaunee Power Station
(Kewaunee) Technical Specification (TS) TS 4.2.a.2 regarding in-
service testing of ASME Code Class 1, Class 2 and Class 3 pumps and
valves. The proposed change revises the TS to be consistent with the
requirements of 10 CFR [Title 10, Code of Federal Regulations]
50.55a(f)(4) for pumps and valves which are classified as American
Society of Mechanical Engineers (ASME) Code Class 1, Class 2 and
Class 3. The proposed change incorporates revisions to the ASME Code
that result in a net improvement in the measures for in-service
testing of pumps and valves.
As a net improvement in the in-service testing of pumps and
valves, the proposed change does not negatively impact any accident
initiators, analyzed events, or assumed mitigation of accident or
transient events. It does not involve the addition or removal of any
equipment, or any design changes to the facility. Therefore, this
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?
No. The proposed change revises Kewaunee TS 4.2.a.2 regarding
in-service testing of ASME Code Class 1, Class 2 and Class 3 pumps
and valves, for consistency with the requirements of 10 CFR
50.55a(f)(4). The proposed change incorporates revisions to the ASME
Code that result in a net improvement in the measures for testing
pumps and valves.
The proposed change does not involve a modification to the
physical configuration of the plant (i.e., no new equipment will be
installed) or adversely affect methods governing normal plant
operation. The proposed change will not impose any new or different
requirements or introduce a new accident initiator, accident
precursor, or malfunction mechanism. The proposed change does not
alter existing test criteria or frequencies. Additionally, there is
no change in the types or increases in the amounts of any effluent
that may be released off-site and there is no increase in individual
or cumulative occupational exposure. Therefore, this proposed change
does not create the possibility of an accident of a different kind
than previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The proposed change revises TS 4.2.a.2 regarding in-service
testing of ASME Code Class 1, Class 2, and Class 3 pumps and valves,
for consistency with the requirements of 10 CFR 50.55a(f)(4). The
proposed change incorporates revisions to the ASME Code that result
in a net improvement in the measures for testing pumps and valves.
The safety function of the affected pumps and valves will continue
to be confirmed through testing. Therefore, this 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: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
NRC Branch Chief: M. Murphy (Acting).
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 amendment request: September 1, 2006.
Description of amendment request: The proposed amendment would
revise the Millstone Power Station, Unit Nos. 2 and 3 (MPS2 and MPS3)
Technical Specifications (TSs) to replace the terms
[[Page 62309]]
``trash racks and screens'' with the term ``strainers''.
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:
Criterion 1:
Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Although the configurations of the existing sump screen and the
replacement strainer assemblies are different, they serve the same
fundamental purpose of passively removing debris from the sump's
suction supply of the supported system pumps. Replacing trash racks
with strainers does not adversely impact the adequacy of pump net
positive suction head assumed in the safety analyses. In fact, it
will improve it. Likewise, the proposed change does not reduce the
reliability of any supported systems or introduce any new system
interactions. A missile evaluation of the new strainer design
concluded that there is no credible missile that could damage the
strainer when needed during a loss-of-coolant accident [LOCA]. A jet
impingement evaluation of the new strainer design concluded that
there are no credible high energy line break jets that could damage
the strainer when needed during a LOCA. The greatly increased
surface area of the new strainer will reduce the approach velocity
of the strainer face significantly, further decreasing the risk of
impact from large debris entrained in the sump flow stream. The
proposed rewording of the SRs [surveillance requirements] will
continue to ensure that the ECCS [emergency core cooling system]
sump suction inlet strainers show no evidence of structural distress
or abnormal corrosion for MPS2 and [MPS]3 with or without the
strainer modification complete. As such, the proposed change does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Criterion 2:
Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
During the next refueling outage for each unit, DNC [Dominion
Nuclear Connecticut, Inc.] is replacing the ECCS trash racks and
screens with strainers in support of the response to Generic Letter
2004-02 on Millstone Units 2 and 3. The ECCS strainers are passive
components in standby safety systems used for accident mitigation.
As such, they are not accident initiators. Therefore, there is no
possibility that this change could create any accident of any kind.
A change to TS SRs 4.5.2.j for MPS2 and 4.5.2.d.2 for MPS3 addresses
differences in nomenclature between the existing and [Generic Safety
Issue] GSI-191 designs. These changes do not alter the nature of
events postulated in the Final 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.
Criterion 3:
Does the proposed amendment involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not adversely affect any plant safety
limits, set points, or design parameters. The changes also do not
adversely affect the fuel, fuel cladding, reactor coolant system
(RCS), or containment integrity. Therefore, the proposed TS change,
which revises the terminology associated with TS SRs, 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: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT
06385.
NRC Acting Branch Chief: Brooke D. Poole.
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One, Units 1 and 2 (ANO-1&2), Pope County, Arkansas
Date of amendment request: October 25, 2005.
Description of amendment request: The proposed change modifies
inventory and inspection requirements associated with the Emergency
Cooling Pond (ECP), which is a common cooling water source for ANO-1&2
during conditions that may render the normal cooling water source
(Dardanelle Reservoir) unavailable.
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 indicated ECP level is an operator aid for routine
verification that the required ECP inventory of 70 acre-feet is
maintained. Relocation of this indication to the TS [technical
specification] Bases does not change the design basis and,
therefore, has no impact on any accident described in the SAR
[safety analysis report]. The relocation of excessive SR
[surveillance requirement] details to the TS Bases does not reduce
the level of testing required with regard to ECP operability
verifications. Actual ECP inspection is more detailed than that
currently described in the TSs. The relocation of this excessive
detail to the TS Bases, therefore, has no impact on any accident
described in the SAR. Finally, the inclusion of a new Action
associated with the discovery of degradation of the ECP structure is
more restrictive in that the proposed engineering evaluation must be
performed within 7 days. Previously, the TS Bases did not require a
completion time for this action. Actions associated with TS Limiting
Conditions for Operation (LCO) or SRs are below the level of detail
described in the SAR and, therefore, have no impact on any accident
currently described in the SAR.
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 aforementioned proposed change to the TSs does not require
any physical alteration to the plant or alter plant design. The ECP
is not an accident initiator. The proposed change does not adversely
impact the function of the ECP as credited in any safety analyses
for the prevention or mitigation of any accident.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not adversely impact a margin of safety
analysis for any accident previously evaluated. Relocation of the
indicated ECP level that corresponds to the required ECP volume of
70 acre-feet and the relocation of excessive SR details to the TS
Bases will not result in a credible increase in nuclear safety risk.
In addition, the TS Bases is part of the SAR and controlled under 10
CFR 50.59. The inclusion of a new action relocated from the TS Bases
to the TS with completion time constraint is more conservative than
currently described in the TS Bases. The proposed change acts to
correct current TS deficiencies and, therefore, is considered risk
neutral.
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: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: David Terao.
[[Page 62310]]
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: September 19, 2006.
Description of amendment request: The proposed change will revise
River Bend Station, Unit 1, (RBS) Technical Specifications (TS)
Surveillance Requirement 3.6.1.3.5 to replace the currently specified
frequency for leak testing containment purge supply and exhaust
isolation valves with resilient seal materials with a requirement to
test these valves in accordance with the Containment Leakage Rate
Testing Program. The RBS Containment Leakage Rate Testing Program is
implemented in accordance with the Code of Federal Regulations, Part
50, Appendix J, Option B, and Regulatory Guide (RG) 1.163,
``Performance-Based Containment Leak Test Program,'' dated September
1995. RG 1.163 allows a nominal test interval of 30 months for
containment purge and vent valves.
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.
This change deletes the augmented testing requirement for these
containment isolation valves and allows the surveillance intervals
to be set in accordance with the Containment Leakage Rate Testing
Programs. This change does not affect the system function or design.
The purge valves are not an initiator of any previously analyzed
accident. Leakage rates do not affect the probability of the
occurrence of any accident. Operating history has demonstrated that
the valves do not degrade and cause leakage as previously
anticipated. Because these valves have been demonstrated to be
reliable, these valves can be expected to perform the containment
isolation function as assumed in the accident analyses. Therefore,
there is no significant increase in the consequences of any
previously evaluated accident.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Extending the test intervals has no influence on, nor does it
contribute in any way to, the possibility of a new or different kind
of accident or malfunction from those previously analyzed. No change
has been made to the design, function or method of performing
leakage testing. Leakage acceptance criteria have not changed. No
new accident modes are created by extending the testing intervals.
No safety-related equipment or safety functions are altered as a
result of this 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 change involve a significant reduction in a
margin of safety?
Response: No.
The only margin of safety that has the potential of being
impacted by the proposed changes involves the offsite dose
consequences of postulated accidents which are directly related to
the containment leakage rate. The proposed change does not alter the
method of performing the tests nor does it change the leakage
acceptance criteria. Sufficient data has been collected to
demonstrate these resilient seals do not degrade at an accelerated
rate.
Because of this demonstrated reliability, this change will
provide sufficient surveillance to determine an increase in the
unfiltered leakage prior to the leakage exceeding that assumed in
the accident analysis.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn LLP,
1700 K Street, NW., Washington, DC 20006.
NRC Branch Chief: David Terao.
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Nuclear
Plant (PNP), Van Buren County, Michigan
Date of amendment request: May 30, 2006.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS), Section 5.5.8, ``Steam Generator
Program,'' to modify the steam generator (SG) provisions for tube
inspections, as contained in the PNP TS Surveillance Requirements,
Section 5.5.8.d. The purpose of these changes is to define the depth of
the required tube inspections. WCAP-16208-P, ``NDE Inspection Length
for CE [Combustion Engineering] Steam Generator Tubesheet Region
Explosive Expansions,'' Revision 1, provided recommended tubesheet
region inspection lengths for plants with CE-supplied steam generators
with explosive expansions. This inspection length is referred to as C*
(``C-Star''). Nuclear Management Company (NMC) intends to implement the
C* inspection methodology for PNP.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment does not involve a significant increase
in the probability of an accident previously evaluated because the
modification to TS Section 5.5.8.d maintains the existing design
limits and would not increase the probability or consequences of an
accident involving tube burst or primary to secondary accident-
induced leakage, as previously analyzed in the UFSAR [Updated Final
Safety Analysis Report]. Also, the tube burst and collapse criteria
of NRC [Nuclear Regulatory Commission] Regulatory Guide 1.121,
``Basis for Plugging Degraded PWR Steam Generator Tubes,'' would
continue to be satisfied.
Tube burst is precluded for a tube with defects within the
tubesheet region because of the constraint provided by the
tubesheet. As such, tube pullout resulting from the axial forces
induced by primary to secondary differential pressures would be a
prerequisite for tube burst to occur. A joint industry test program,
WCAP-16208-P, has defined the nondegraded tube to tubesheet joint
length required to preclude tube pullout C [deg]) and maintain
acceptable primary to secondary accident-induced leakage, assuming a
360[deg] circumferential through wall crack existed immediately
below this length. For PNP, C [deg] is 12.5 inches. Any degradation
below C [deg] is shown by empirical test results and analyses to be
acceptable, thereby precluding an event with consequences similar to
a postulated tube rupture event.
WCAP-1 6208-P incorporates an assumed primary to secondary
accident-induced leakage value of 0.1 gpm/SG. The NMC TSTF
[Technical Specifications Task Force]-449 submittal to the NRC
provided the PNP SG tube integrity related TS. LCO [Limiting
Condition for Operation] 3.4.13, item d., ``PCS Operational
Leakage,'' states that operational leakage through any one SG shall
be limited to 150 gallons per day. The UFSAR Chapter 14.14-6
accident-induced leakage limit assumption based on MSLB [main steam-
line break] is 0.3 gallons per minute (432 gallons per day).
Therefore, the LCO leakage limit is conservatively less than the
design basis accident induced leakage limit.
In summary, the proposed modifications to the PNP Technical
Specifications maintain existing design limits and do not involve a
significant increase in the probability or consequences of an
accident previously evaluated in the UFSAR.
Therefore, operation of the facility in accordance with the
proposed amendment
[[Page 62311]]
would not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment does not create the possibility of a new
or different kind of accident from any accident previously evaluated
because SG tube leakage and structural integrity will continue to be
maintained during all plant conditions upon implementation of the
proposed inspection scope to the PNP TSs. The revised inspection
scope does not introduce any new mechanisms that might result in a
different kind of accident from those previously evaluated. Even
with the limiting circumstances of a complete circumferential
separation (360-degree through wall crack) of a tube below the C*
length, tube pullout is precluded and leakage is predicted to be
maintained within the TS limits during all plant conditions.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment does not involve a significant reduction
in a margin of safety. The requirements for the inspection of SG
tubes are intended to ensure that this portion of the primary
coolant system maintains its integrity. Tube integrity means that
the tubes are capable of performing these functions in accordance
with the plant design and licensing basis. Tube integrity includes
both structural and leakage integrity. The proposed tubesheet
inspection depth of 12.5 inches will ensure tube integrity is
maintained because any degradation below C* is shown by empirical
test results and analyses to be acceptable. In addition, operation
with potential tube degradation below the C* inspection length
continues to meet the margin of safety as defined by RG [Regulatory
Guide] 1.121, ``Basis for Plugging Degraded PWR Steam Generator
Tubes,'' and RG 1.83, ``Inservice Inspection of Pressurized Water
Reactor Steam Generator Tubes.'' Therefore, the proposed
modifications 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: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Branch Chief: Martin C. Murphy, Acting Branch Chief.
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].
Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power
Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: March 28, 2006.
Brief description of amendment: The amendment revised Facility
Operating License No. NPF-49 by deleting Section 2.F, which specifies
reporting of violations of the requirements of Section 2.C of the
renewed operating license.
Date of issuance: October 4, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 234.
Facility Operating License No. NPF-49: The amendment revised the
License.
Date of initial notice in Federal Register: May 9, 2006 (71 FR
26997).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 4, 2006.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: September 19, 2005, as
supplemented by letters dated February 28, May 31, and September 26,
2006.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3.6.2.1, ``Containment Spray System.'' Specifically,
the change revised the allowable outage time (AOT) for TS 3.6.2.1 from
72 hours to 7 days during fuel cycles 19 and 20. Per the license
amendment request, the AOT extension may only be invoked twice (i.e.,
once for each train or twice for one train). The requested changes are
sought to provide needed flexibility in the performance of selected
corrective and preventative maintenance activities during power
operations. Currently, the licensee's maintenance activities on
containment spray system components are performed during the refueling
outages; taking several days of ``around the clock'' effort.
Date of issuance: September 28, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 268.
Renewed Facility Operating License No. NPF-6: The amendment revised
the Technical Specifications.
Date of initial notice in Federal Register: January 3, 2006 (71 FR
148). The supplements dated February 28, May 31, and September 26,
2006,
[[Page 62312]]
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 28, 2006.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of application for amendment: August 17, 2005, as supplemented
by letter dated May 19, 2006.
Brief description of amendment: The proposed changes revised the
Operating License Condition (OLC) 2.C.(41) to add reference to a
Nuclear Regulatory Commission (NRC) Safety Evaluation that allows the
application of certain risk-informed, performance-based fire protection
methods and tools.
Date of issuance: September 29, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment No: 170.
Facility Operating License No. NPF-29: The amendment revised the
OLC 2.C.(41).
Date of initial notice in Federal Register: October 25, 2005 (70 FR
61658). The supplement dated May 19, 2006, provided additional
information that clarified the change the staff's original proposed no
significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 29, 2006
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: November 5, 2004.
Brief description of amendment: The amendment modified Waterford 3
Technical Specification (TS) 3.7.4, ``Ultimate Heat Sink,'' to provide
clarification that the ambient temperature monitoring requirement that
is specified in TS 3.7.4.d only applies when the affected ultimate heat
sink train is considered to be operable. The NRC is not approving the
request to delete TS 3.7.4.c, which would allow the plant to take
credit for the dry cooling tower fans that are not protected from
tornado missiles when a tornado warning is in effect.
Date of issuance: September 28, 2006.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 208.
Facility Operating License No. NPF-38: The amendment revised the
Operating License and the Technical Specifications.
Date of initial notice in Federal Register: December 7, 2004 (69 FR
70717).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 28, 2006.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: October 27, 2005.
Brief description of amendment: The amendment modified Surveillance
Requirement (SR) 4.5.2e of Technical Specification (TS) 3.5.2, ``ECCS
[Emergency Core Cooling Systems] Subsystems--Modes 1, 2 and 3,'' SR
4.6.2.1d of TS 3.6.2, ``Containment Spray System,'' and SR 4.7.3b of TS
3.7.3, ``Component Cooling Water and Auxiliary Component Cooling Water
Systems,'' to remove the words ``during shutdown.'' This will provide
flexibility allowing components required to be tested by these SRs to
be tested online. Additionally, a revision to delete SR 4.7.12.1c of TS
3.7.12, ``Essential Services Chilled Water system,'' is approved. A
modification permanently separating the safety and non-safety portions
of the Essential Services Chilled Water system has eliminated the need
for automatic isolation valves and thus this SR.
Date of issuance: October 6, 2006.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 209.
Facility Operating License No. NPF-38: The amendment revised the
Technical Specifications and the Facility Operating License.
Date of initial notice in Federal Register: December 20, 2005 (70
FR 75491).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 6, 2006.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: October 25, 2005.
Brief description of amendment: The amendment modifies Waterford 3
Technical Specification 6.9.1.11, ``Core Operating Limits Report
COLR,'' to add a methodology that will allow the use of zirconium
diboride burnable absorber coating on fuel pellets.
Date of issuance: October 6, 2006.
Effective date: As of the date of issuance and shall be implemented
30 days from the date of issuance.
Amendment No.: 210.
Facility Operating License No. NPF-38: The amendment revised the
Operating License and the Technical Specifications.
Date of initial notice in Federal Register: December 6, 2005 (70 FR
72673).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 6, 2006.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: June 11, 2004, as supplemented
by letters dated December 12, 2005, April 4, 2006, and July 28, 2006.
Brief description of amendments: This amendment incorporated a
revision to the Technical Specifications (TSs) and licensing and design
bases that relocates surveillance test intervals of various TS
surveillance requirements to a new program, the Surveillance Frequency
Control Program, which will be located in the Administrative Controls
Section of the TSs. These amendments are pilot submittals in support of
the Boiling Water Reactor Owners' Group Risk-Informed Initiative 5b,
``Relocate Surveillance Test Intervals to Licensee Control.''
Date of issuance: September 28, 2006.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment Nos. 186, 147.
Facility Operating License Nos. NPF-39 and NPF-85. This amendment
revised the facility operating licenses and the TSs.
Date of initial notice in Federal Register: May 24, 2005 (70 FR
29793). The supplements provided clarifying information that did not
expand the scope of the application as originally
[[Page 62313]]
noticed, and did not change the NRC staff's original proposed no
significant hazards consideration determination as originally published
in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 28, 2006.
No significant hazards consideration comments received: No.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: September 29, 2005, as supplemented on
August 8, September 18, and September 28, 2006.
Description of amendment request: The amendment revised the
Seabrook Station, Unit No. 1 Technical Specifications (TSs) to permit a
one-time change in the steam generator tube inspection requirements to
include a sampling of the bulges and over-expansions for portions of
the steam generator tubes within the hot-leg tubesheet region.
Date of issuance: September 29, 2006.
Effective date: As of its date of issuance, and shall be
implemented within 90 days.
Amendment No.: 112.
Facility Operating License No. NPF-86: The amendment revised the
License and the Tss.
Date of initial notice in Federal Register: November 8, 2005 (70 FR
67749). The licensee's August 8 and September 28, 2006, supplements
provided clarifying information that did not change the scope of the
proposed amendment as described in the original notice of proposed
action published in the Federal Register, and did not change the
initial proposed no significant hazards consideration determination.
The supplement dated September 18, 2006, modified the requested
amendment to request a one-time change in lieu of a permanent one. This
narrowing of scope did not alter the validity of the NRC staff's
proposed no significant hazards consideration determination. The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated September 29, 2006.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook
Nuclear Plant, Unit 1, Berrien County, Michigan
Date of application for amendment: May 31, 2006.
Brief description of amendment: The amendment approved elimination
of the resistance temperature detector (RTD) bypass piping and
installing fast response thermowell-mounted RTDs in the reactor coolant
system loop piping. The amendment also revised Surveillance Requirement
3.3.1.15 of the Technical Specifications, deleting the requirement to
perform surveillance on the reactor coolant system RTD bypass loop flow
rate.
Date of issuance: October 6, 2006.
Effective date: As of the date of issuance and shall be implemented
prior to entry into Mode 2 from the fall 2006 refueling outage.
Amendment No.: 296.
Facility Operating License No. DPR-58: Amendment revise the
Technical Specifications.
Date of initial notice in Federal Register: July 5, 2006 (71 FR
38182). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 6, 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: January 30, 2006, as supplement by May
17 and August 29, 2006.
Brief description of amendment: The amendment revised the Cooper
Nuclear Station Technical Specification Section 5.5.12, ``Primary
Containment Leakage Rate Testing Program,'' to allow a one-time
extension of no more than 5 years for the Type A, Integrated Leakage
Rate Test (ILRT) interval. This revision is a one-time exception to the
10-year frequency of the performance-based leakage rate testing program
for Type A tests as defined in Nuclear Energy Institute (NEI) document,
NEI 94-01, Revision 0, ``Industry Guideline for Implementing
Performance-Based Option of 10 CFR Part 50, Appendix J,'' pursuant to
10 CFR Part 50, Appendix J, Option B. The requested exception is to
allow the ILRT to be performed within 15 years from the last ILRT, last
performed on December 7, 1998.
Date of issuance: October 3, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 224.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 25, 2006 (71 FR
23957). The supplement dated May 17 and August 29, 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 as published in the Federal Register. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated October 3, 2006.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station, Unit No. 1, Oswego County, New York
Date of application for amendment: January 18, 2006.
Brief description of amendment: The amendment deletes the reference
to the hydrogen monitors in Technical Specification 3.6.11, ``Accident
Monitoring Instrumentation.''
Date of issuance: October 2, 2006.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 191.
Facility Operating License No. DPR-63: Amendment revises the
Technical Specifications and License.
Date of initial notice in Federal Register: July 18, 2006 (71 FR
40749) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 2, 2006.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of application for amendments: December 13, 2005, supplemented
by letters dated June 7, and July 21, 2006.
Brief description of amendments: The amendments revise technical
specification 5.5.14 ``Containment Leakage Rate Testing Program'' for
Prairie Island Nuclear Generating Plant Units 1 and 2, to allow a one-
time interval extension of no more than 5 years for the Appendix J Type
A, Integrated Leakage Rate Test.
Date of issuance: October 2, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 174 and 164.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 31, 2006 (71 FR
5081) The supplemental information provided by letters dated June 7,
and July 21, 2006, did not change the no significant hazards
determination.
[[Page 62314]]
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 2, 2006.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of application for amendments: January 19, 2006, as
supplemented by letter dated June 20, 2006.
Brief description of amendments: The amendments deleted the
antitrust conditions from the facility operating licenses.
Date of issuance: October 2, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: Unit 1-189; Unit 2-191.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Facility Operating Licenses.
Date of initial notice in Federal Register: April 14, 2006 (71 FR
19551) The supplemental letter dated June 20, 2006, provided additional
information that clarified the application, and did not expand the
scope of the application as originally noticed. The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated October 2, 2006.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket No. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of application for amendments: November 9, 2004, as
supplemented on December 15, 2005, June 30, 2006, August 18, 2006, and
September 28, 2006.
Brief description of amendments: The amendments revise the SSES 1
and 2 Technical Specifications (TSs) 3.8.4, ``DC Sources--Operating,''
3.8.5, ``DC Sources-Shutdown,'' 3.8.6, ``Battery Cell Parameters,'' and
add a new TS Section, 5.5.13, ``Battery Monitoring and Maintenance
Program.'' These changes are consistent with TS Task Force (TSTF) 360,
Revision 1.
Date of issuance: September 28, 2006.
Effective date: As of the date of issuance and to be implemented
within 60 days.
Amendment Nos.: 238 and 215.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the TSs and license.
Date of initial notice in Federal Register: January 17, 2006 (71 FR
2596). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 28, 2006.
The supplements dated December 15, 2005, June 30, 2006, August 18,
2006, and September 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.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-311, Salem Nuclear Generating Station,
Unit No. 2, Salem County, New Jersey
Date of application for amendment: September 21, 2005, as
supplemented by letters dated June 28, 2006, and August 4, 2006.
Brief description of amendment: The amendment revises the extent of
steam generator tube inspections in the hot-leg side of the tubesheet.
Date of issuance: September 28, 2006.
Effective date: As of the date of issuance, to be implemented
within 60 days from date of issuance.
Amendment No.: 256.
Facility Operating License No. DPR-75: This amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: January 7, 2006 (71 FR
2594). The supplements did not expand the scope of the request, or
change the original proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 28 2006.
No significant hazards consideration comments received: No.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of application for amendments: April 17, 2006.
No significant hazards consideration comments received: No.
Brief description of amendments: The proposed amendments deleted
Section 2.G of the Facility Operating Licenses, which required
reporting of violations of the requirements in Sections 2.C(1), 2.C(3),
and 2.F of the Facility Operating Licenses.
Date of issuance: October 3, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: Unit 2-205; Unit 3-197.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
deleted Section 2.G of the Facility Operating Licenses.
Date of initial notice in Federal Register: May 9, 2006 (71 FR
27003) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 3, 2006.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear
Plant, Unit 1, Limestone County, Alabama
Date of application for amendment: August 16, 2004 (TS-433) as
supplemented by letter dated September 30, 2005.
Brief description of amendment: The proposed amendment extends the
frequency of ``once-per cycle'' from 18 to 24 months in several
Technical Specification (TS) Surveillance Requirements. This change
will allow the adoption of a 24-month refueling cycle.
Date of issuance: September 28, 2006.
Effective date: Date of issuance, to be implemented within 60 days.
Amendment No.: 263.
Renewed Facility Operating License No. DPR-33: Amendment revised
the TSs.
Date of initial notice in Federal Register: March 29, 2005 (70 FR
15947). The supplement dated September 30, 2005, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the NRC
staff's original proposed no significant hazards determination as
published in the Federal Register. The Commission's related evaluation
of the amendment is contained in a Safety Evaluation dated September
28, 2006.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear
Plant, Unit 1, Limestone County, Alabama
Date of amendment request: October 12, 2004, as supplemented April
27 and June 27, 2005 (TS-438).
Description of amendment request: The amendment revised the
frequency requirement for Technical Specification (TS) Surveillance
Requirement (SR) 3.6.1.3.8 by allowing a representative sample
(approximately 20 percent) of excess flow check valves (EFCVs) to be
tested every 24 months, so that each EFCV is tested once every 120
months. The current SR requires testing of each EFCV every 24 months.
Date of issuance: September 29, 2006.
[[Page 62315]]
Effective date: Date of issuance, to be implemented within 30 days.
Amendment No.: 264.
Facility Operating License Nos. DPR-33: Amendment revised the TSs.
Date of initial notice in Federal Register: March 29, 2005 (70 FR
15948). The supplemental letters provided clarifying information that
did not expand the scope of the original application or change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated September 29, 2006.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: February 24, 2006, as
supplemented by letter dated May 8, 2006 (TS-06-02).
Brief description of amendment: The amendment revises the Updated
Final Safety Analysis Report (UFSAR) by modifying the design and
licensing basis to incorporate revised dose analysis inputs and results
for the steam generator tube rupture accident. The analysis was revised
as a result of an error in the computer model used to calculate the
dose consequences to the Main Control Room subsequent to an accident.
Date of issuance: October 4, 2006.
Effective date: As of the date of issuance and shall be implemented
as part of the next UFSAR update made in accordance with 10 CFR
50.71(e).
Amendment No.: 64.
Facility Operating License No. NPF-90: Amendment authorizes
revision of the UFSAR.
Date of initial notice in the Federal Register: April 25, 2006 (71
FR 23962). 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 October 4, 2006.
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: December 16, 2005, as supplemented by
letters dated June 23 and August 25, 2006.
Brief description of amendments: The change revised Technical
Specifications (TSs) 3.3.2, ``ESFAS [Engineered Safety Features
Actuation System] Instrumentation''; and 3.5.2, ``ECCS [Emergency Core
Cooling System]--Operating.''
Date of issuance: October 5, 2006.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance for TS 3.5.2 revisions, and
within 120 days from the completion of the 12th refueling outage of
Unit 1, for TS 3.3.2 revisions.
Amendment Nos.: 129 and 129.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 14, 2006 (71 FR
13179). The supplements dated June 23 and August 25, 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 as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 5, 2006.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: July 23, 2004, as supplemented by
letters dated August 11 and September 22, 2006.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3.6.3, ``Containment Isolation Valves,'' by (1)
adding the abbreviation ``(CIV)'' for containment isolation valve in
Condition A of the Actions for the Limiting Condition for Operation;
(2) deleting the note and revising Condition A to be for only one
penetration flow path with one CIV inoperable; (3) revising the
completion time for Required Condition A.1 from 4 hours to as much as 7
days depending on the category of the inoperable CIV; and (4) revising
Condition C to be for two or more penetration flow paths with one CIV
inoperable. The amendment also added two conditions to the license.
Date of issuance: September 28, 2006.
Effective date: Effective as of its date of issuance and shall be
implemented prior to the start of Refueling Outage 18, which is
scheduled to start in spring 2008.
Amendment No.: 167.
Facility Operating License No. NPF-42. The amendment revised
Appendix A, ``Technical Specifications,'' and Appendix D, ``Additional
Conditions,'' of the license.
Date of initial notice in Federal Register: December 7, 2004 (69 FR
70724). The supplemental letters dated August 11 and September 22,
2006, provided additional information that clarified the application,
did not expand the scope of the application as originally noticed, and
did not change the NRC staff's original proposed no significant hazards
consideration determination published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 28, 2006.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: June 2, 2006.
Brief description of amendment: The amendment revised Surveillance
Requirement 3.5.2.8 by replacing the phrase ``trash racks and screens''
with the word ``strainers.'' The amendment reflects the replacement of
the containment sump suction inlet trash racks and screens with a
complex strainer design with significantly larger effective area in the
upcoming Refueling Outage 15.
Date of issuance: October 5, 2006.
Effective date: As of its date of issuance and shall be implemented
prior to the entry into Mode 4 in the restart from the fall 2006
refueling outage.
Amendment No.: 168.
Facility Operating License No. NPF-42: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 18, 2006 (71 FR
40756) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 5, 2006.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: June 30, 2006.
Brief description of amendment: The amendment revised Technical
Specification (TS) 5.5.9, ``Steam Generator (SG) Program,'' by changing
the ``Refueling Outage 14'' to ``Refueling Outage 15'' in two places.
This change extended the provisions for SG tube
[[Page 62316]]
repair criteria and inspections that were approved for Refueling Outage
14, and the subsequent operating cycle, in Amendment No. 162 issued
April 28, 2005, to the upcoming Refueling Outage 15, and the subsequent
operating cycle.
Date of issuance: October 10, 2006.
Effective date: As of its date of issuance and shall be implemented
prior to entry into Mode 4 during the startup from Refueling Outage 15,
scheduled to begin in October 2006.
Amendment No.: 169.
Facility Operating License No. NPF-42: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 24, 2006 (71 FR
41845) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 10, 2006.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
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 for the
amendment 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.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective 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
application for amendment, (2) the amendment to Facility Operating
License, 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 System's (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].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. 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, and electronically on the Internet at the NRC Web site,
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR Reference staff at
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. If a
request for a hearing or petition for leave to intervene is filed by
the above date, 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
[[Page 62317]]
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 identify 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
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
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.\1\ 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 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.
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\1\ To the extent that the applications contain attachments and
supporting documents that are not publicly available because they
are asserted to contain safeguards or proprietary information,
petitioners desiring access to this information should contact the
applicant or applicant's counsel and discuss the need for a
protective order.
---------------------------------------------------------------------------
Each contention shall be given a separate numeric or alpha
designation within one of the following groups:
1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the
applications.
2. Environmental--primarily concerns/issues relating to matters
discussed or referenced in the environmental analysis for the
applications.
3. Miscellaneous--does not fall into one of the categories outlined
above.
As specified in 10 CFR 2.309, if two or more petitioners/requestors
seek to co-sponsor a contention, the petitioners/requestors shall
jointly designate a representative who shall have the authority to act
for the petitioners/requestors with respect to that contention. If a
petitioner/requestor seeks to adopt the contention of another
sponsoring petitioner/requestor, the petitioner/requestor who seeks to
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the
petitioners/requestors with respect to that contention.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
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 or 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).
Duke Power Company LLC, 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: September 27, 2006, as supplemented
October 2 and 3, 2006.
The supplement dated October 2 and 3, 2006, provided additional
information that claried the application, did not expand the scope of
the original proposed no significant hazards consideration (NSHC)
determination, and did not change the NRC staff's original proposed
NSHC determination.
Description of amendment request: The amendments extend the
Completion Time of Technical Specification 3.8.1, ``AC Sources--
Operating,'' Required Action C.2.2.5 for one time only from 45 days to
75 days to allow time for repairs of Keowee Hydro Unit 2.
Date of issuance: October 3, 2006.
Effective date: As of the date of issuance and shall be implemented
on or before October 3, 2006.
Amendment Nos.: 354, 356, 355.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the technical specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. Public notice of the proposed amendments was
published in the Greenville News on September 29 and 30, and October 1,
2006, and in the Anderson Independent on September 29 and October 1,
2006. The notice issued a proposed NSHC and provided an opportunity to
submit comments to the NRC staff on the Commission's proposed NSHC
determination by close of business on October 3, 2006. No comments have
been received.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, consultation with the State of South Carolina,
and final NSHC determination are contained in a safety evaluation dated
October 3, 2006.
Attorney for licensee: Ms. Lisa F. Vaughn, Duke Power Company LLC,
526 South Church Street, Charlotte, North Carolina, 28201-1006.
NRC Branch Chief: Evangelos C. Marinos.
[[Page 62318]]
Florida Power and Light Company, Docket No. 50-250, Turkey Point
Nuclear Plant, Unit 3, Miami-Dade County, Florida
Date of amendment request: September 8, 2006.
Description of amendment request: The amendment allows the use of
an alternate method for determining the position of Control Rod M-6,
which has an inoperable analog rod position indicator (ARPI), until the
ARPI is repaired, but no later than the Cycle 23 refueling outage
scheduled for the fall of 2007.
Date of issuance: October 5, 2006.
Effective date: As of the date of issuance.
Amendment No.: 230.
Facility Operating License Nos. DPR-31: Amendment revises the
technical specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes (71 FR 54691, dated September 18, 2006). The
notice provided an opportunity to submit comments on the Commission's
proposed NSHC determination. No comments have been received. The notice
also provided an opportunity to request a hearing by November 17, 2006,
but indicated that if the Commission makes a final NSHC determination,
any such hearing would take place after issuance of the amendment.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, state consultation, and final NSHC determination
are contained in a safety evaluation dated October 5, 2006.
Attorney for licensee: M.S. Ross, Managing Attorney, Florida power
and Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: L. Raghavan.
Dated at Rockville, Maryland, this 13th day of October 2006.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E6-17546 Filed 10-23-06; 8:45 am]
BILLING CODE 7590-01-P