Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 59529-59538 [E6-16560]
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Federal Register / Vol. 71, No. 195 / Tuesday, October 10, 2006 / Notices
report to identify any non-radiological
hazards that may have impacted the
environment surrounding the Facility.
No such hazards or impacts to the
environment were identified. The NRC
has identified no other radiological or
non-radiological activities in the area
that could result in cumulative
environmental impacts.
The NRC staff finds that the proposed
release of the Facility for unrestricted
use is in compliance with 10 CFR
20.1402. Although the Licensee will
continue to perform licensed activities
at other areas of the South Charleston
site, the Licensee must ensure that this
decommissioned area does not become
recontaminated. Before the license can
be terminated, the Licensee will be
required to show that its entire site,
including previously-released areas,
complies with the radiological criteria
in 10 CFR 20.1402. Based on its review,
the staff considered the impact of the
residual radioactivity at the Facility and
concluded that the proposed action will
not have a significant effect on the
quality of the human environment.
Environmental Impacts of the
Alternatives to the Proposed Action
Due to the largely administrative
nature of the proposed action, its
environmental impacts are small.
Therefore, the only alternative the staff
considered is the no-action alternative,
under which the staff would leave
things as they are by simply denying the
amendment request. This no-action
alternative is not feasible because it
conflicts with 10 CFR 30.36(d),
requiring that decommissioning of
byproduct material facilities be
completed and approved by the NRC
after licensed activities cease. The
NRC’s analysis of the Licensee’s final
status survey data confirmed that the
Facility meets the requirements of 10
CFR 20.1402 for unrestricted release.
Additionally, denying the amendment
request would result in no change in
current environmental impacts. The
environmental impacts of the proposed
action and the no-action alternative are
therefore similar, and the no-action
alternative is accordingly not further
considered.
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Conclusion
The NRC staff has concluded that the
proposed action is consistent with the
NRC’s unrestricted release criteria
specified in 10 CFR 20.1402. Because
the proposed action will not
significantly impact the quality of the
human environment, the NRC staff
concludes that the proposed action is
the preferred alternative.
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Agencies and Persons Consulted
NRC provided a draft of this EA to the
State of West Virginia for review on
August 28, 2006. On September 15,
2006, the State of West Virginia
Radiological Health program responded
by electronic mail. The State agreed
with the conclusions of the EA, and
otherwise had no comments.
The NRC staff has determined that the
proposed action is of a procedural
nature, and will not affect listed species
or critical habitat. Therefore, no further
consultation is required under Section 7
of the Endangered Species Act. The
NRC staff has also determined that the
proposed action is not the type of
activity that has the potential to cause
effects on historic properties. Therefore,
no further consultation is required
under Section 106 of the National
Historic Preservation Act.
III. Finding of No Significant Impact
The NRC staff has prepared this EA in
support of the proposed action. On the
basis of this EA, the NRC finds that
there are no significant environmental
impacts from the proposed action, and
that preparation of an environmental
impact statement is not warranted.
Accordingly, the NRC has determined
that a Finding of No Significant Impact
is appropriate.
IV. Further Information
Documents related to this action,
including the application for license
amendment and supporting
documentation, are available
electronically at the NRC’s Electronic
Reading Room at https://www.nrc.gov/
reading-rm/adams.html. From this site,
you can access the NRC’s Agencywide
Document Access and Management
System (ADAMS), which provides text
and image files of NRC’s public
documents. The documents related to
this action are listed below, along with
their ADAMS accession numbers.
1. Letters dated August 3, 2006
[ML062220617], June 19, 2006
[ML061720331], and January 27, 2006
[ML060320507].
2. Facsimile dated January 31, 2006
[ML060320519].
3. NUREG–1757, ‘‘Consolidated
NMSS Decommissioning Guidance.’’
4. Title 10 Code of Federal
Regulations, Part 20, Subpart E,
‘‘Radiological Criteria for License
Termination.’’
5. Title 10, Code of Federal
Regulations, Part 51, ‘‘Environmental
Protection Regulations for Domestic
Licensing and Related Regulatory
Functions.’’
6. NUREG–1496, ‘‘Generic
Environmental Impact Statement in
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59529
Support of Rulemaking on Radiological
Criteria for License Termination of NRCLicensed Nuclear Facilities.’’
If you do not have access to ADAMS,
or if there are problems in accessing the
documents located in ADAMS, contact
the NRC Public Document Room (PDR)
Reference staff at 1–800–397–4209, 301–
415–4737, or by e-mail to pdr@nrc.gov.
These documents may also be viewed
electronically on the public computers
located at the NRC’s PDR, O 1 F21, One
White Flint North, 11555 Rockville
Pike, Rockville, MD 20852. The PDR
reproduction contractor will copy
documents for a fee.
Dated at 475 Allendale Road, King of
Prussia, Pennsylvania this 29th day of
September 2006.
For the Nuclear Regulatory Commission.
James P. Dwyer,
Chief, Commercial and R&D Branch, Division
of Nuclear Materials Safety, Region I.
[FR Doc. E6–16647 Filed 10–6–06; 8:45 am]
BILLING CODE 7590–01–P
NUCLEAR REGULATORY
COMMISSION
Biweekly Notice; Applications and
Amendments to Facility Operating
Licenses Involving No Significant
Hazards Considerations
I. Background
Pursuant to section 189a.(2) of the
Atomic Energy Act of 1954, as amended
(the Act), the U.S. Nuclear Regulatory
Commission (the Commission or NRC
staff) is publishing this regular biweekly
notice. The Act requires the
Commission publish notice of any
amendments issued, or proposed to be
issued and grants the Commission the
authority to issue and make
immediately effective any amendment
to an operating license upon a
determination by the Commission that
such amendment involves no significant
hazards consideration, notwithstanding
the pendency before the Commission of
a request for a hearing from any person.
This biweekly notice includes all
notices of amendments issued, or
proposed to be issued from September
15, 2006, to September 28, 2006. The
last biweekly notice was published on
September 26, 2006 (71 FR 56189).
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.
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Federal Register / Vol. 71, No. 195 / Tuesday, October 10, 2006 / Notices
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 60day 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
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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, https://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
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effect of any decision or order which
may be entered in the proceeding on the
requestor’s/petitioner’s interest. The
petition must also set forth the specific
contentions which the petitioner/
requestor seeks to have litigated at the
proceeding.
Each contention must consist of a
specific statement of the issue of law or
fact to be raised or controverted. In
addition, the petitioner/requestor shall
provide a brief explanation of the bases
for the contention and a concise
statement of the alleged facts or expert
opinion which support the contention
and on which the petitioner/requestor
intends to rely in proving the contention
at the hearing. The petitioner/requestor
must also provide references to those
specific sources and documents of
which the petitioner is aware and on
which the petitioner/requestor intends
to rely to establish those facts or expert
opinion. The petition must include
sufficient information to show that a
genuine dispute exists with the
applicant on a material issue of law or
fact. Contentions shall be limited to
matters within the scope of the
amendment under consideration. The
contention must be one which, if
proven, would entitle the petitioner/
requestor to relief. A petitioner/
requestor who fails to satisfy these
requirements with respect to at least one
contention will not be permitted to
participate as a party.
Those permitted to intervene become
parties to the proceeding, subject to any
limitations in the order granting leave to
intervene, and have the opportunity to
participate fully in the conduct of the
hearing.
If a hearing is requested, and the
Commission has not made a final
determination on the issue of no
significant hazards consideration, the
Commission will make a final
determination on the issue of no
significant hazards consideration. The
final determination will serve to decide
when the hearing is held. If the final
determination is that the amendment
request involves no significant hazards
consideration, the Commission may
issue the amendment and make it
immediately effective, notwithstanding
the request for a hearing. Any hearing
held would take place after issuance of
the amendment. If the final
determination is that the amendment
request involves a significant hazards
consideration, any hearing held would
take place before the issuance of any
amendment.
A request for a hearing or a petition
for leave to intervene must be filed by:
(1) First class mail addressed to the
Office of the Secretary of the
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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,
HearingDocket@nrc.gov; 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 email to OGCMailCenter@nrc.gov. 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, https://
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
pdr@nrc.gov.
FirstEnergy Nuclear Operating
Company, et al., Docket No. 50–346,
Davis-Besse Nuclear Power Station, Unit
1, Ottawa County, Ohio
Date of amendment request: May 30,
2006.
Description of amendment request:
The proposed amendment would revise
technical specification (TS) Definitions
1.14, ‘‘LEAKAGE’’, and 1.16,
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‘‘PRESSURE BOUNDARY LEAKAGE’’;
revise TS 3/4.6.2, ‘‘Reactor Coolant
System Operational Leakage’’; add a
new TS 3/4.4.5, ‘‘Steam Generator (SG)
Tube Integrity;’’ add a new TS 6.8.4.g,
‘‘Steam Generator (SG) Program;’’ and
add a new TS 6.9.1.12, ‘‘Steam
Generator Tube Inspection Report’’; as
well as administrative and editorial
changes. These changes are consistent
with the NRC-approved Revision 4 to
TS Task Force (TSTF) Standard TS
change traveler, TSTF–449, ‘‘Steam
Generator Tube Integrity.’’ The
proposed changes are necessary in order
to implement the guidance for the
industry initiative on Nuclear Energy
Institute (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
concerning 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 May 30, 2006.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), an
analysis of the issue of NSHC 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[n] SGTR [SG tube rupture] event is one
of the design basis accidents that are
analyzed as part of a plant’s licensing basis.
In the analysis of a[n] 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 doubleended rupture of a single tube is assumed.
For other design basis accidents such as
[an] MSLB [main steam line break], rod
ejection, and reactor coolant pump locked
rotor[,] the tubes are assumed to retain their
structural integrity (i.e., they are assumed not
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59531
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[s] to the TS[s] to 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[s] to the TS[s]. 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 1[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 1[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 with no more than [500 gallons per
day or 720 gallons per day] in any one SG,
and that the reactor coolant activity levels of
DOSE EQUIVALENT 1[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[n] 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
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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 [a]
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.
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The NRC staff proposes to determine
that the amendment request involves no
significant hazards consideration.
Attorney for licensee: David W.
Jenkins, Attorney, FirstEnergy
Corporation, Mail Stop A–GO–18, 76
South Main Street, Akron, OH 44308.
NRC Branch Chief: Daniel S. Collins.
Tennessee Valley Authority (TVA),
Docket No. 50–259 , Browns Ferry
Nuclear Plant (BFN), Unit 1, Limestone
County, Alabama
Date of amendment request:
September 22, 2006 (TS–431).
Description of amendment request:
The proposed amendment supplements
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a June 28, 2004, request to increase the
licensed thermal power from 3293
megawatt thermal (MWt) to 3952 MWt,
an approximate 20 percent increase in
thermal power. This supplement
requests interim approval of an increase
in licensed thermal power from 3293
MWt to 3458 MWt with an attendant 30psi increase in reactor pressure. This
represents an approximate 5 percent
increase above the original licensed
thermal power (OLTP) of 3293 MWt. An
interim approval would provide for
operation at 105 percent power until
such time as certain steam dryer
analyses can be completed. The NRC
staff’s review of the remainder of the
June 2004 application would resume
upon receipt of the satisfactorily
completed steam dryer analyses.
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 probability (frequency of occurrence)
of Design Basis Accidents occurring is not
affected by the increased power level,
because BFN Unit 1 continues to comply
with the regulatory and design basis criteria
established for plant equipment. An
evaluation of the Boiling Water Reactor
probabilistic risk assessments concludes that
the calculated core damage frequency does
not significantly change due to operation at
105% OLTP.
Scram setpoints (equipment settings that
initiate automatic plant shutdowns) are
established such that there is no significant
increase in scram frequency due to operation
at 105% OLTP. No new challenges to safetyrelated equipment result from operation at
105% OLTP.
The probability of Design Basis Accidents
occurring is not affected by taking credit for
containment overpressure in ensuring
adequate NPSH [Net Positive Suction Head]
for the BFN Unit 1 low pressure ECCS
pumps. NRC Bulletin 96–03 requested that
BWR [Boiling-Water Reactors] owners
implement appropriate measures to minimize
the potential clogging of the Emergency Core
Cooling System (ECCS) suppression chamber
strainers by potential debris generated by a
LOCA [loss-of-coolant accident]. TVA
installed new, high-capacity passive strainers
on BFN Unit 1 of the same design as BFN
Units 2 and 3. In addition, TVA’s proposed
resolution of NRC Bulletin 96–03 for BFN
Unit 1 takes credit for containment
overpressure to maintain adequate ECCS
pump Net Positive Suction Head (NPSH).
Containment pressure will increase following
a pipe break occurring inside containment.
Crediting containment overpressure in the
analysis of the consequences of the Loss of
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Coolant Accident (LOCA) does not affect the
precursors for the LOCA, nor does it affect
the precursors for any other accident or
transient analyzed in Chapter 14 of the BFN
Updated Final Safety Analysis Report
(UFSAR). Therefore, there is no increase in
the probability of any accident previously
evaluated.
The changes in consequences of
hypothetical accidents, which would occur
from 102% of the stretch power uprate
reactor thermal power compared to those
previously evaluated, are in all cases
insignificant. The stretch power uprate
accident evaluations do not exceed any of
their NRC-approved acceptance limits. The
spectrum of hypothetical accidents and
transients has been investigated, and are
shown to meet the plant’s currently licensed
regulatory criteria. In the area of core design,
for example, the fuel operating limits such as
Maximum Average Planar Linear Heat
Generation Rate (MAPLHGR) and Safety
Limit Minimum Critical Power Ratio
(SLMCPR) are still met, and fuel reload
analyses will show plant transients meet the
criteria accepted by the NRC. Challenges to
fuel (ECCS performance) are evaluated, and
shown to continue to meet the criteria of 10
CFR 50.46.
Challenges to the containment have been
evaluated at the increased power level, and
the containment and its associated cooling
systems continue to meet the design and
licensing criteria. Radiological release events
(accidents) have been evaluated at the
increased power level, and shown to be less
than the limits of 10 CFR 50.67.
The radiological consequences of the
design basis accident are not increased by
taking credit for the post-LOCA suppression
chamber airspace pressure. The containment
will continue to function as designed. This
proposed change only takes credit for
containment pressure that would exist
following a LOCA. Crediting this pressure in
ensuring adequate ECCS NPSH will not
result in an increase in containment leakage
assumed in any analysis.
Therefore, the proposed amendment does
not result in a significant increase in
consequences or a significant increase in the
probability or consequences of any accident
previously evaluated.
(2) Does the proposed change create the
possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
Equipment that could be affected by
operation at 105% OLTP has been evaluated.
No new operating mode, safety-related
equipment lineup, accident scenario or
equipment failure mode was identified. The
full spectrum of accident considerations has
been evaluated and no new or different kind
of accident has been identified. Operation at
105% OLTP uses developed technology, and
applies it within the capabilities of existing
plant safety related equipment in accordance
with the regulatory criteria, including NRC
approved codes, standards and methods. No
new power dependent accidents have been
identified.
The BFN Unit 1 TS [Technical
Specifications] require revision to implement
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operation at 105% OLTP. All revisions have
been assessed, and it has been determined
that the proposed change will not introduce
a different accident than that previously
evaluated.
The proposed use of the post-LOCA
suppression chamber airspace pressure in the
calculation of NPSH for the ECCS pumps
does not introduce any new modes of plant
operation or make physical changes to plant
systems. Rather, the post-LOCA suppression
chamber airspace pressure is a consequence
of the conditions that would exist in the
containment following a large pipe break
inside containment. 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.
Therefore, the change will 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 calculated loads on all affected
structures, systems and components will
remain within their design allowables for all
design basis event categories. No NRC
acceptance criterion is exceeded. Because the
BFN Unit 1 configuration and reactions to
transients and hypothetical accidents does
not result in exceeding the presently
approved NRC acceptance limits, operation
at 105% OLTP does not involve a significant
reduction in a margin of safety.
The post-LOCA suppression chamber
airspace pressure is a byproduct of the
conditions that will exist in the containment
after a line break inside containment.
Conservative analyses have been performed
that demonstrate that sufficient post-accident
suppression chamber airspace pressure will
be available to meet the NPSH requirements
for the low pressure ECCS pumps. By
enabling credit of these conditions for the
low pressure ECCS pumps, adequate NPSH
margin will be ensured, and accordingly, the
ECCS pumps will meet their performance
requirements. Therefore, the credit for
containment overpressure does not involve a
significant reduction in a margin of safety.
jlentini on PROD1PC65 with NOTICES
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: L. Raghavan.
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
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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, https://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
pdr@nrc.gov.
AmerGen Energy Company, LLC, Docket
No. 50–219, Oyster Creek Nuclear
Generating Station, Ocean County, New
Jersey
Date of application for amendment:
December 2, 2005.
Brief description of amendment: The
amendment revised the Oyster Creek
Nuclear Generating Station Technical
Specifications (TSs) to increase the
allowable as-found main steam safety
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59533
valve code safety function lift setpoint
tolerance from ±1% to ±3%.
Date of Issuance: September 13, 2006.
Effective date: As of the date of
Issuance to be implemented within 60
days.
Amendment No.: 261.
Facility Operating License No. DPR–
16: The amendment revised the TSs.
Date of initial notice in Federal
Register: January 17, 2006 (71 FR
2588).
The Commission’s related evaluation
of this amendment is contained in a
Safety Evaluation dated September 13,
2006.
No significant hazards consideration
comments received: No.
Calvert Cliffs Nuclear Power Plant, Inc.,
Docket Nos. 50–317 and 50–318, Calvert
Cliffs Nuclear Power Plant, Unit Nos. 1
and 2, Calvert County, Maryland
Date of application for amendments:
June 7, 2005, as supplemented on May
12, 2006.
Brief description of amendments: The
amendments revise the Technical
Specifications (TSs) to eliminate the use
of the defined term Core Alterations.
The amendments incorporate the
changes reflected in TS Task Force
(TSTF) Travelers 471-T (TSTF–471-T),
‘‘Eliminate use of term CORE
ALTERATIONS in ACTIONS and
Notes,’’ and TSTF–51-A, ‘‘Revise
containment requirements during
handling irradiated fuel and core
alterations.’’ In addition, the
amendments revise TS 3.9.2, ‘‘Nuclear
Instrumentation,’’ by replacing ‘‘Core
Alterations’’ with ‘‘positive reactivity
additions’’ in the Required Action for an
inoperable source range monitor during
refueling operations. The limiting
conditions for operation in TS 3.9.4,
‘‘Shutdown Cooling (SDC) and Coolant
Recirculation—High Water Level,’’ are
also revised by replacing ‘‘core
alterations’’ with ‘‘movement of fuel
assemblies within containment.’’
Date of issuance: September 21, 2006
Effective date: As of the date of
issuance to be implemented within 60
days.
Amendment Nos.: 279 and 256.
Renewed Facility Operating License
Nos. DPR–53 and DPR–69: Amendments
revised the Licenses and Technical
Specifications.
Date of initial notice in Federal
Register: July 5, 2005 (70 FR 38716).
The May 12, 2006, letter 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.
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The Commission’s related evaluation
of these amendments is contained in a
Safety Evaluation dated September 21,
2006.
No significant hazards consideration
comments received: No.
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated September 22,
2006.
No significant hazards consideration
comments received: No.
Carolina Power & Light Company,
Docket No. 50–261, H. B. Robinson
Steam Electric Plant, Unit No. 2,
Darlington County, South Carolina
Dominion Nuclear Connecticut, Inc., et
al., Docket No. 50–423, Millstone Power
Station, Unit No. 3, New London
County, Connecticut
Date of application for amendment:
March 3, 2005, as supplemented by
letter dated July 6, 2006.
Brief description of amendment: The
amendment revises the requirements of
Technical Specification (TS) 5.6.5,
‘‘Core Operating Limits Report (COLR).’’
Date of issuance: September 20, 2006.
Effective date: As of the date of
issuance and shall be implemented
within 60 days.
Amendment No. 209.
Renewed Facility Operating License
No. DPR–23. Amendment revises the
Technical Specifications.
Date of initial notice in Federal
Register: May 24, 2005 (70 FR 29787).
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 September 20,
2006.
No significant hazards consideration
comments received: No.
Date of application for amendment:
May 27, 2004, as supplemented
September 27, 2004, October 20, 2004,
March 23, 2005, January 30, 2006 and
May 25, 2006.
Brief description of amendment: The
amendment revised the Technical
Specifications (TSs) to incorporate a
full-scope application of an alternate
source term methodology in accordance
with 10 CFR 50.67.
Date of issuance: September 15, 2006.
Effective date: As of the date of
issuance and shall be implemented
within 90 days from the date of
issuance.
Amendment No.: 232.
Facility Operating License No. NPF–
49: The amendment revised the TSs.
Date of initial notice in Federal
Register: September 14, 2004 (69 FR
55468). The supplements contained
clarifying information only, and did not
change the initial no significant hazards
consideration determination or expand
the scope of the initial Federal Register
notice.
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated September 15,
2006.
No significant hazards consideration
comments received: No.
jlentini on PROD1PC65 with NOTICES
Carolina Power & Light Company,
Docket No. 50–261, H. B. Robinson
Steam Electric Plant, Unit No. 2,
Darlington County, South Carolina
Date of application for amendment:
August 20, 2004, as supplemented by
letters dated June 22, 2005, June 26,
2006, and September 18, 2006.
Brief description of amendment: The
amendment revises Table 3.3.1–1,
Functions 3, 14, 17.a., 20, and the
footnote related to Function 20.
Date of issuance: September 22, 2006.
Effective date: As of the date of
issuance and shall be implemented
within 60 days.
Amendment No. 210.
Renewed Facility Operating License
No. DPR–23. Amendment revises the
Technical Specifications.
Date of initial notice in Federal
Register: November 23, 2004 (69 FR
68182). The letters dated June 22, 2005,
June 26, 2006, and September 18, 2006,
provided clarifying information that did
not change the initial proposed no
significant hazards consideration
determination.
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16:42 Oct 06, 2006
Jkt 211001
Dominion Nuclear Connecticut, Inc., et
al., Docket No. 50–423, Millstone Power
Station, Unit No. 3, New London
County, Connecticut
Date of application for amendment:
September 13, 2005, as supplemented
by letters dated June 13 and August 14,
2006.
Brief description of amendment: The
amendment revised the Technical
Specification (TS) surveillance
requirements for the recirculation spray
system.
Date of issuance: September 20, 2006.
Effective date: As of the date of
issuance and shall be implemented
prior to entering Mode 1 following
refueling outage 3R11.
Amendment No.: 233.
Facility Operating License No. NPF–
49: The amendment revised the TSs.
Date of initial notice in Federal
Register: October 25, 2005 (70 FR
61657). The supplements dated June 13
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Sfmt 4703
and August 14, 2006, provided
clarifying information that did not
change the initial proposed no
significant hazards consideration
determination or expand the application
beyond the scope of the original Federal
Register notice.
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated September 20,
2006.
No significant hazards consideration
comments received: No.
Duke Power Company LLC, et al.,
Docket Nos. 50–413 and 50–414,
Catawba Nuclear Station, Units 1 and 2,
York County, South Carolina
Date of application for amendments:
June 29, 2005, as supplemented May 1,
2006, and September 8, 2006.
Brief description of amendments: The
amendments requested authorization to
revise the Updated Final Safety
Analysis Report and the emergency
operating procedures to allow an
additional operator action to manually
start one containment air return fan in
the air return system in response to
Nuclear Regulatory Commission
Bulletin 2003–01, ‘‘Potential Impact of
Debris Blockage on Emergency Sump
Recirculation at Pressurized-Water
Reactors,’’ June 6, 2003.
Date of issuance: September 25, 2006.
Effective date: As of the date of
issuance and shall be implemented
within 60 days from the date of
issuance.
Amendment Nos.: 231 and 227.
Renewed Facility Operating License
Nos. NPF–35 and NPF–52: Amendments
revised the licenses.
Date of initial notice in Federal
Register: October 25, 2005 (70 FR
61657).
The supplement dated May 1, 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 September 25,
2006.
No significant hazards consideration
comments received: No.
Duke Power Company LLC, et al.,
Docket Nos. 50–413 and 50–414,
Catawba Nuclear Station, Units 1 and 2,
York County, South Carolina
Date of application for amendments:
July 25, 2005, as supplemented July 28,
2005, and August 1, 2005.
Brief description of amendments: The
amendments revised the Technical
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Specifications temperature limit for the
standby nuclear service water pond
from 91.5 °F to 95 °F.
Date of issuance: September 25, 2006.
Effective date: As of the date of
issuance and shall be implemented
within 30 days from the date of issuance
September 25, 2006.
Amendment Nos.: 232 and 228.
Renewed Facility Operating License
Nos. NPF–35 and NPF–52: Amendments
revised the licenses and the technical
specifications.
Date of initial notice in Federal
Register: August 4, 2005 (70 FR 44946).
The supplements dated July 28, 2005,
and August 1, 2005, provided additional
information that clarified the
application, did not expand the scope of
the application as originally noticed,
and did not change the original
proposed no significant hazards
consideration determination.
The Commission’s related evaluation
of the amendments is contained in a
Safety Evaluation dated September 25,
2006.
No significant hazards consideration
comments received: No.
jlentini on PROD1PC65 with NOTICES
Duke Power Company LLC, et al.,
Docket Nos. 50–413 and 50–414,
Catawba Nuclear Station, Units 1 and 2,
York County, South Carolina
Date of application for amendments:
September 13, 2005, as supplemented
March 20, 2006.
Brief description of amendments: The
amendments revised the Technical
Specifications (TSs) to correct a
nonconservative TS associated with
spent fuel storage in the spent fuel pool.
The licensee identified the
nonconservative TS while comparing
results from spent fuel pool criticality
codes.
Date of issuance: September 27, 2006.
Effective date: As of the date of
issuance and shall be implemented
within 60 days from the date of issuance
September 27, 2006.
Amendment Nos.: 233 and 229.
Renewed Facility Operating License
Nos. NPF–35 and NPF–52: Amendments
revised the license and the technical
specifications.
Date of initial notice in Federal
Register: November 21, 2005 (70 FR
70104). The supplement dated March
20, 2006, provided additional
information that clarified the
application, did not expand the scope of
the application as originally noticed,
and did not change the Nuclear
Regulatory Commission (NRC) staff’s
original proposed no significant hazards
consideration determination.
The Commission’s related evaluation
of the amendments is contained in a
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16:42 Oct 06, 2006
Jkt 211001
Safety Evaluation dated September 27,
2006.
No significant hazards consideration
comments received: No.
Duke Power Company LLC, Docket Nos.
50–369 and 50–370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg
County, North Carolina
Date of application for amendments:
June 29, 2005, as supplemented May 1,
2006.
Brief description of amendments: The
amendments requested authorization to
revise the Updated Final Safety
Analysis Report and the emergency
operating procedures to allow an
additional operator action to manually
start one containment air return fan in
the air return system in response to
Nuclear Regulatory Commission
Bulletin 2003–01, ‘‘Potential Impact of
Debris Blockage on Emergency Sump
Recirculation at Pressurized-Water
Reactors,’’ June 6, 2003.
Date of issuance: September 25, 2006.
Effective date: As of the date of
issuance and shall be implemented
within 60 days from the date of
issuance.
Amendment Nos.: 234 and 216.
Renewed Facility Operating License
Nos. NPF–9 and NPF–17: Amendments
revised the licenses.
Date of initial notice in Federal
Register: October 25, 2005 (70 FR
61657).
The supplement dated May 1, 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 September 25,
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:
June 29, 2005, as supplemented by letter
dated May 18, 2006.
Brief description of amendment: By
letter dated June 29, 2005, Entergy
Operations, Inc., the licensee for
Arkansas Nuclear One, Unit 2 (ANO–2),
requested a license amendment to
relocate the shutdown cooling (SDC)
open permissive interlock (OPI) license
condition from the operating license to
the licensee’s technical requirements
manual. The license condition to
maintain OPI operability was previously
accepted by the NRC staff in a letter to
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Sfmt 4703
59535
the licensee, dated March 30, 2005, and
incorporated into ANO–2’s operating
license. The OPI prevents the two SDC
suction isolation valves from opening
above a selected set point to separate the
high-pressure reactor coolant system
from the low-pressure SDC system.
Date of issuance: September 13, 2006.
Effective date: As of the date of
issuance and shall be implemented
within 60 days from the date of
issuance.
Amendment No.: 267.
Renewed Facility Operating License
No. NPF–6: Amendment revised the
Facility Operating License.
Date of initial notice in Federal
Register: December 6, 2005 (70 FR
72671). The supplement dated May 18,
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 September 13,
2006.
No significant hazards consideration
comments received: No.
Entergy Nuclear Operations, Inc.,
Docket No. 50–293, Pilgrim Nuclear
Power Station, Plymouth County,
Massachusetts
Date of application for amendment:
December 14, 2004, revised by letter
dated August 30, 2006.
Brief description of amendment: The
Technical Specification amendment
relocates structural integrity
requirements to the Final Safety
Analysis Report.
Date of issuance: September 14, 2006.
Effective date: As of the date of
issuance, and shall be implemented
within 60 days.
Amendment No.: 224.
Facility Operating License No. DPR–
35: The amendment revised the License
and Technical Specifications.
Date of initial notice in Federal
Register: March 1, 2005 (70 FR 9991).
The licensee originally requested for
additional TS relocations in their
submittal dated December 14, 2004. The
NRC staff found these unacceptable.
Therefore, the licensee revised the
original application by letter dated
August 30, 2006, reducing the scope of
the application as originally noticed.
Hence, there is no change to the NRC
staff’s original proposed no significant
hazards consideration determination.
The Commission’s related evaluation
of the amendment is contained in a
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Safety Evaluation dated September 14,
2006.
No significant hazards consideration
comments received: No.
Entergy Nuclear Operations, Inc.,
Docket No. 50–293, Pilgrim Nuclear
Power Station, Plymouth County,
Massachusetts
Date of application for amendment:
June 2, 2005, supplemented by letter
dated June 14, 2006.
Brief description of amendment: The
amendment revised the Technical
Specification (TS) reactor coolant
system leakage detection
instrumentation requirements and
actions.
Date of issuance: September 20, 2006.
Effective date: As of the date of
issuance, and shall be implemented
within 60 days.
Amendment No.: 225.
Facility Operating License No. DPR–
35: The amendment revised the Facility
Operating License and Technical
Specifications.
Date of initial notice in Federal
Register: May 23, 2006 (71 FR 29676).
The supplement dated June 14, 2006,
provided additional information that
did not expand the scope of the
application as originally noticed, and
did not change the staff’s original
proposed no significant hazards
consideration determination.
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated September 20,
2006.
No significant hazards consideration
comments received: No.
jlentini on PROD1PC65 with NOTICES
FirstEnergy Nuclear Operating
Company, et al., Docket No. 50–412,
Beaver Valley Power Station, Unit No. 2,
Beaver County, Pennsylvania
Date of application for amendment:
April 11, 2005, as supplemented
December 2, 2005, and January 27, April
14, August 16, and September 1, 2006.
Brief description of amendment: The
amendment revised the scope of the
steam generator tubesheet inspections
and subsequent repair using the F*
inspection methodology.
Date of issuance: September 27, 2006.
Effective date: As of the date of
issuance to be implemented within 60
days.
Amendment No: 160.
Facility Operating License No. NPF–
73: The amendment revised the License
and Technical Specifications.
Date of initial notice in Federal
Register: June 7, 2005 (70 FR 33214).
The supplements dated December 2,
2005, and January 27, April 14, August
16, and September 1, 2006, provided
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16:42 Oct 06, 2006
Jkt 211001
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. The
Commission’s issuance of Amendment
No. 158 to Facility Operating License
NPF–73 for BVPS–2, regarding steam
generator tube integrity (Technical
Specification Task Force (TSTF) Item
449) on September 7, 2006, resulted in
renumbering and rewording the
requirements as originally proposed by
the licensee to fit the TSTF–449 format,
but did not change the scope of the
application.
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated September 27,
2006.
No significant hazards consideration
comments received: No.
FirstEnergy Nuclear Operating
Company, et al., Docket Nos. 50–334
and 50–412, Beaver Valley Power
Station, Unit Nos. 1 and 2 (BVPS–1 and
2), Beaver County, Pennsylvania
Date of application for amendments:
November 7, 2005, as supplemented
April 25, June 1, and August 3, 2006.
Brief description of amendments: The
amendments include changes to the
definition of leakage, changes to the
primary-to-secondary leakage
requirements, changes to the steam
generator (SG) tube surveillance
program (SG tube integrity), and
changes to the SG reporting
requirements.
Date of issuance: September 7, 2006.
Effective date: As of the date of
issuance to be implemented within 90
days for BVPS–1 and prior to entry into
Mode 4 following the fall 2006 refueling
outage for BVPS–2.
Amendment Nos.: 276 and 158.
Facility Operating License Nos. DPR–
66 and NPF–73: Amendments revised
the Technical Specifications and
Licenses.
Date of initial notice in Federal
Register: December 20, 2005 (70 FR
75491). The supplements dated April
25, June 1, and August 3, 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 September 7,
2006.
No significant hazards consideration
comments received: No.
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FirstEnergy Nuclear Operating
Company, et al., Docket Nos. 50–334
and 50–412, Beaver Valley Power
Station, Unit Nos. 1 and 2 (BVPS–1 and
2), Beaver County, Pennsylvania
Date of application for amendments:
February 17, 2005, as supplemented
May 12 and August 22, 2006.
Brief description of amendments: The
amendments revised the Technical
Specifications (TSs) 3.7.7.1 (BVPS–1),
‘‘Control Room Emergency Habitability
Systems,’’ and 3.7.7 (BVPS–2), ‘‘Control
Room Emergency Air Cleanup and
Pressurization System,’’ by dividing
these TSs into two specifications,
addressing control room emergency
ventilation and control room air cooling
functions separately. The amendments
also improved consistency with the
Standard TSs and improved consistency
between the units.
Date of issuance: September 25, 2006
Effective date: As of the date of
issuance, to be implemented within 60
days.
Amendment Nos.: 277 and 159
Facility Operating License Nos. DPR–
66 and NPF–73: Amendments revised
the License and Technical
Specifications.
Date of initial notice in Federal
Register: April 26, 2005 (70 FR 21458).
The supplements dated May 12 and
August 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 staff’s original
proposed no significant hazards
consideration determination as
published in the Federal Register.
The Commission’s related evaluation
of the amendments is contained in a
Safety Evaluation dated September 25,
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:
February 1, 2006.
Brief description of amendments: The
amendments revise the Technical
Specification (TS) testing frequency for
the Surveillance Requirements in TS
3.1.4, ‘‘Control Rod Scram Times,’’
based on the TS Task Force (TSTF)
change traveler TSTF–222, Revision 1.
Date of issuance: September 12, 2006.
Effective date: As of the date of
issuance and to be implemented within
60 days.
Amendment Nos.: 237 and 214.
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Facility Operating License Nos. NPF–
14 and NPF–22: The amendments
revised the License and TSs.
Date of initial notice in Federal
Register: May 9, 2006 (71 FR 27001).
The Commission’s related evaluation
of the amendments is contained in a
Safety Evaluation dated September 12,
2006.
No significant hazards consideration
comments received: No.
jlentini on PROD1PC65 with NOTICES
South Carolina Electric & Gas Company,
South Carolina Public Service
Authority, Docket 50–395, Virgil C.
Summer Nuclear Station, Unit No. 1,
Fairfield County, South Carolina
Date of application for amendment:
June 30, 2005, as supplemented July 21,
2006.
Brief description of amendment: The
amendment revises the Virgil C.
Summer Nuclear Station Technical
Specifications to permit the use of a best
estimate methodology in performing
loss-of-coolant accident analyses.
Date of issuance: September 7, 2006.
Effective date: As of the date of
issuance and shall be implemented
within 30 days.
Amendment No. 176.
Renewed Facility Operating License
No. NPF–12: Amendment revises the
Technical Specifications.
Date of initial notice in Federal
Register: October 11, 2005 (70 FR
59087). 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.
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated September 7,
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:
November 30, 2005, as supplemented by
letter dated May 30, 2006.
Brief description of amendments: The
proposed amendments revised the
Technical Specification (TS)
requirements related to steam generator
tube integrity, based on the NRCapproved Revision 4 to TS Task Force
(TSTF)-449, ‘‘Steam Generator Tube
Integrity.’’ Date of issuance: September
19, 2006.
Effective date: As of the date of
issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: Unit 2—204; Unit
3—196.
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16:42 Oct 06, 2006
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Facility Operating License Nos. NPF–
10 and NPF–15: The amendments
revised the TSs.
Date of initial notice in Federal
Register: February 14, 2006 (71 FR
7812). The May 30, 2006, supplemental
letter 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 September 19,
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:
March 9, 2004 (TS–434) as
supplemented on November 15, 2004,
and March 7, 2006.
Brief description of amendment: The
amendment reduced the Allowable
Value used for Reactor Vessel Water
Level—Low, Level 3, for several
instrument functions.
Date of issuance: September 18, 2006.
Effective date: September 18, 2006.
Amendment No.: 258.
Renewed Facility Operating License
No. DPR–33: Amendment revised the
Technical Specifications.
Date of initial notice in Federal
Register: April 13, 2004 (69 FR 19575).
The supplements dated November 15,
2004, and March 7, 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 September 18,
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 amendments:
November 3, 2003, as supplemented
May 6, 2004, and August 1, 2006.
Description of amendment request:
The amendment revised Technical
Specification (TS) Table 3.3.1.1 –1,
Reactor Protection system
Instrumentation, Function 7.b.
Date of issuance: September 19, 2006.
Effective date: Date of issuance, to be
issued within 60 days.
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59537
Amendment No.: 259.
Renewed Facility Operating License
No. DPR–33: Amendment revised the
TSs.
Date of initial notice in Federal
Register: April 13, 2004 (69 FR 19575).
The supplements dated May 6, 2004,
and August 1, 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 September 19,
2006.
No significant hazards consideration
comments received: No.
Tennessee Valley Authority, Docket
Nos. 50–259, 50–260, and 50–296,
Browns Ferry Nuclear Plant, Units 1, 2,
and 3, Limestone County, Alabama
Date of application for amendments:
August 16, 2004, as supplemented by
letters dated March 11, 2005, November
4, 2005, and April 14, 2006.
Description of amendment request: To
extend the channel calibration
frequency requirements for
instrumentation in the high-pressure
coolant injection, reactor core isolation
cooling, and reactor water core isolation
cooling systems.
Date of issuance: September 21, 2006.
Effective date: Date of issuance, to be
implemented within 60 days.
Amendment Nos.: 260, 297 and 255.
Renewed Facility Operating License
Nos. DPR–33, DPR–52, and DPR–68:
Amendments revised the Technical
Specifications.
Date of initial notice in Federal
Register: May 23, 2006 (71 FR 29680).
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 21,
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:
July 9, 2004 (TS 436).
Brief description of amendment: The
amendment revised Technical
Specification (TS) Surveillance
Requirement 3.6.1.3.10 to increase the
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jlentini on PROD1PC65 with NOTICES
allowed main steam isolation valve
(MSIV) leak rate from 11.5 standard
cubic feet per hour (scfh) per valve to
100 scfh for individual MSIVs with a
150 scfh combined leakage for all four
main steam lines.
Date of issuance: September 27, 2006.
Effective date: Date of issuance, to be
implemented within 60 days.
Amendment No.: 261.
Facility Operating License No. DPR–
33: Amendment revised the TSs.
Date of initial notice in Federal
Register: May 23, 2006 (71 FR 29680).
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated September 27,
2006.
No significant hazards consideration
comments received: No.
Tennessee Valley Authority, Docket
Nos. 50–327 and 50–328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton
County, Tennessee
Date of application for amendments:
August 14, 2004, as supplemented on
April 11, 2005, and July 11, 2006 (TS–
02–01).
Brief description of amendments: The
amendments revise Technical
Specifications (TSs) relating to the
reactor protection system and
engineered safety features
instrumentation. The Trip Setpoint
column of TS Tables 2.2–1 and 3.3–4
will be renamed Nominal Trip Setpoint;
inequality signs in TS Tables 2.2–1 and
3.3–4 will be removed; the trip setpoint
and allowable value for the Intermediate
Range Neutron Flux P–6 permissive will
be revised; Minimum Channels
Operable in TS Table 3.3–3 will be
revised; editorial changes will be made
to TS Table 3.3–4 to replace ± signs with
inequalities; and a correction will be
made to an alarm/trip setpoint in TS
Table 3.3–6.
Date of issuance: September 13, 2006.
Effective date: As of the date of
issuance and shall be implemented
within 45 days.
Amendment Nos. 310 and 299.
Facility Operating License Nos. DPR–
77 and DPR–79: Amendments revised
the technical specifications.
Date of initial notice in Federal
Register: October 12, 2004 (69 FR
60688). The supplemental letters
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 amendments is contained in a
Safety Evaluation dated September 13,
2006.
No significant hazards consideration
comments received: No.
VerDate Aug<31>2005
16:42 Oct 06, 2006
Jkt 211001
Tennessee Valley Authority, Docket
Nos. 50–327 and 50–328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton
County, Tennessee
Date of application for amendments:
September 30, 2004, as supplemented
on May 25, 2006.
Brief description of amendments: The
amendments revise the technical
specifications to relocate the
requirements for the emergency diesel
generator start loss of power
instrumentation and associated actions
in the engineering safety features tables
to a new limiting condition for
operation (LCO). In addition, an upper
allowable value limit has been added to
the voltage sensors for loss of voltage
and degraded voltage consistent with
Technical Specification Task Force
(TSTF) Item, TSTF–365, along with a
lower allowable value limit for the
degraded voltage diesel generator start
and load shed timer. The auxiliary
feedwater loss of power start setpoints
and allowable values have been
relocated to this new LCO.
Date of issuance: September 14, 2006.
Effective date: As of the date of
issuance and shall be implemented
within 45 days.
Amendment Nos. 311 and 300.
Facility Operating License Nos. DPR–
77 and DPR–79: Amendments revised
the technical specifications.
Date of initial notice in Federal
Register: January 18, 2005 (70 FR
2900). 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 amendments is contained in a
Safety Evaluation dated September 14,
2006.
No significant hazards consideration
comments received: No.
Union Electric Company, Docket No.
50–483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment:
June 29, 2006.
Brief description of amendment: The
amendment revises Technical
Specification (TS) 3.4.15, ‘‘RCS [Reactor
Coolant System] Leakage Detection
Instrumentation.’’ The TS changes
delete the containment atmosphere
gaseous radioactivity monitor from TS
3.4.15 and revise the existing
conditions, required actions, completion
times, and surveillance requirements in
TS 3.4.15 to account for the monitor
being deleted. The June 29, 2006, letter
superceded the license amendment
PO 00000
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Fmt 4703
Sfmt 4703
request in the August 26, 2005, letter to
authorize changes to the Final Safety
Analysis Report.
Date of issuance: September 26, 2006.
Effective date: As of its date of
issuance, and shall be implemented
within 90 days from the date of
issuance.
Amendment No.: 175.
Facility Operating License No. NPF–
30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal
Register: July 24, 2006 (71 FR 41843).
The Commission’s related evaluation of
the amendment is contained in a Safety
Evaluation dated September 26, 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 26,
2006.
Brief description of amendment: The
amendment revises Technical
Specification (TS) 3.4.15, ‘‘RCS [Reactor
Coolant System] Leakage Detection
Instrumentation.’’ The TS changes
delete the monitor from TS 3.4.15 and
revise the existing conditions, required
actions, completion times, and
surveillance requirements in TS 3.4.15
to account for the monitor being
deleted. The June 26, 2006, letter
superceded the license amendment
request in the August 26, 2005, letter to
authorize changes to the Updated Safety
Analysis Report.
Date of issuance: September 26, 2006.
Effective date: As of its date of
issuance and shall be implemented
within 90 days from the date of
issuance.
Amendment No.: 166.
Facility Operating License No. NPF–
42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal
Register: July 24, 2006 (71 FR 41848).
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated September 26,
2006.
No significant hazards consideration
comments received: No.
Dated at Rockville, Maryland, this 2nd day
of October 2006.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor
Licensing, Office of Nuclear Reactor
Regulation.
[FR Doc. E6–16560 Filed 10–6–06; 8:45 am]
BILLING CODE 7590–01–P
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[Federal Register Volume 71, Number 195 (Tuesday, October 10, 2006)]
[Notices]
[Pages 59529-59538]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E6-16560]
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a determination by the Commission that such
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from September 15, 2006, to September 28, 2006.
The last biweekly notice was published on September 26, 2006 (71 FR
56189).
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.
[[Page 59530]]
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, https://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the
[[Page 59531]]
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, HearingDocket@nrc.gov; 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 OGCMailCenter@nrc.gov. 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, https://
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 pdr@nrc.gov.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346,
Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: May 30, 2006.
Description of amendment request: The proposed amendment would
revise technical specification (TS) Definitions 1.14, ``LEAKAGE'', and
1.16, ``PRESSURE BOUNDARY LEAKAGE''; revise TS 3/4.6.2, ``Reactor
Coolant System Operational Leakage''; add a new TS 3/4.4.5, ``Steam
Generator (SG) Tube Integrity;'' add a new TS 6.8.4.g, ``Steam
Generator (SG) Program;'' and add a new TS 6.9.1.12, ``Steam Generator
Tube Inspection Report''; as well as administrative and editorial
changes. These changes are consistent with the NRC-approved Revision 4
to TS Task Force (TSTF) Standard TS change traveler, TSTF-449, ``Steam
Generator Tube Integrity.'' The proposed changes are necessary in order
to implement the guidance for the industry initiative on Nuclear Energy
Institute (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
concerning 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 May 30, 2006.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC 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[n] SGTR [SG tube rupture] event is one of the design basis
accidents that are analyzed as part of a plant's licensing basis. In
the analysis of a[n] 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 [an] MSLB [main steam
line break], 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[s] to the TS[s] to
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[s]
to the TS[s]. 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 1[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 1[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 with no more than [500 gallons per
day or 720 gallons per day] in any one SG, and that the reactor
coolant activity levels of DOSE EQUIVALENT 1[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[n] 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
[[Page 59532]]
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 [a] 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 amendment request
involves no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy
Corporation, Mail Stop A-GO-18, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: Daniel S. Collins.
Tennessee Valley Authority (TVA), Docket No. 50-259 , Browns Ferry
Nuclear Plant (BFN), Unit 1, Limestone County, Alabama
Date of amendment request: September 22, 2006 (TS-431).
Description of amendment request: The proposed amendment
supplements a June 28, 2004, request to increase the licensed thermal
power from 3293 megawatt thermal (MWt) to 3952 MWt, an approximate 20
percent increase in thermal power. This supplement requests interim
approval of an increase in licensed thermal power from 3293 MWt to 3458
MWt with an attendant 30-psi increase in reactor pressure. This
represents an approximate 5 percent increase above the original
licensed thermal power (OLTP) of 3293 MWt. An interim approval would
provide for operation at 105 percent power until such time as certain
steam dryer analyses can be completed. The NRC staff's review of the
remainder of the June 2004 application would resume upon receipt of the
satisfactorily completed steam dryer analyses.
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 probability (frequency of occurrence) of Design Basis
Accidents occurring is not affected by the increased power level,
because BFN Unit 1 continues to comply with the regulatory and
design basis criteria established for plant equipment. An evaluation
of the Boiling Water Reactor probabilistic risk assessments
concludes that the calculated core damage frequency does not
significantly change due to operation at 105% OLTP.
Scram setpoints (equipment settings that initiate automatic
plant shutdowns) are established such that there is no significant
increase in scram frequency due to operation at 105% OLTP. No new
challenges to safety-related equipment result from operation at 105%
OLTP.
The probability of Design Basis Accidents occurring is not
affected by taking credit for containment overpressure in ensuring
adequate NPSH [Net Positive Suction Head] for the BFN Unit 1 low
pressure ECCS pumps. NRC Bulletin 96-03 requested that BWR [Boiling-
Water Reactors] owners implement appropriate measures to minimize
the potential clogging of the Emergency Core Cooling System (ECCS)
suppression chamber strainers by potential debris generated by a
LOCA [loss-of-coolant accident]. TVA installed new, high-capacity
passive strainers on BFN Unit 1 of the same design as BFN Units 2
and 3. In addition, TVA's proposed resolution of NRC Bulletin 96-03
for BFN Unit 1 takes credit for containment overpressure to maintain
adequate ECCS pump Net Positive Suction Head (NPSH). Containment
pressure will increase following a pipe break occurring inside
containment. Crediting containment overpressure in the analysis of
the consequences of the Loss of Coolant Accident (LOCA) does not
affect the precursors for the LOCA, nor does it affect the
precursors for any other accident or transient analyzed in Chapter
14 of the BFN Updated Final Safety Analysis Report (UFSAR).
Therefore, there is no increase in the probability of any accident
previously evaluated.
The changes in consequences of hypothetical accidents, which
would occur from 102% of the stretch power uprate reactor thermal
power compared to those previously evaluated, are in all cases
insignificant. The stretch power uprate accident evaluations do not
exceed any of their NRC-approved acceptance limits. The spectrum of
hypothetical accidents and transients has been investigated, and are
shown to meet the plant's currently licensed regulatory criteria. In
the area of core design, for example, the fuel operating limits such
as Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) and
Safety Limit Minimum Critical Power Ratio (SLMCPR) are still met,
and fuel reload analyses will show plant transients meet the
criteria accepted by the NRC. Challenges to fuel (ECCS performance)
are evaluated, and shown to continue to meet the criteria of 10 CFR
50.46.
Challenges to the containment have been evaluated at the
increased power level, and the containment and its associated
cooling systems continue to meet the design and licensing criteria.
Radiological release events (accidents) have been evaluated at the
increased power level, and shown to be less than the limits of 10
CFR 50.67.
The radiological consequences of the design basis accident are
not increased by taking credit for the post-LOCA suppression chamber
airspace pressure. The containment will continue to function as
designed. This proposed change only takes credit for containment
pressure that would exist following a LOCA. Crediting this pressure
in ensuring adequate ECCS NPSH will not result in an increase in
containment leakage assumed in any analysis.
Therefore, the proposed amendment does not result in a
significant increase in consequences or a significant increase in
the probability or consequences of any accident previously
evaluated.
(2) Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Equipment that could be affected by operation at 105% OLTP has
been evaluated. No new operating mode, safety-related equipment
lineup, accident scenario or equipment failure mode was identified.
The full spectrum of accident considerations has been evaluated and
no new or different kind of accident has been identified. Operation
at 105% OLTP uses developed technology, and applies it within the
capabilities of existing plant safety related equipment in
accordance with the regulatory criteria, including NRC approved
codes, standards and methods. No new power dependent accidents have
been identified.
The BFN Unit 1 TS [Technical Specifications] require revision to
implement
[[Page 59533]]
operation at 105% OLTP. All revisions have been assessed, and it has
been determined that the proposed change will not introduce a
different accident than that previously evaluated.
The proposed use of the post-LOCA suppression chamber airspace
pressure in the calculation of NPSH for the ECCS pumps does not
introduce any new modes of plant operation or make physical changes
to plant systems. Rather, the post-LOCA suppression chamber airspace
pressure is a consequence of the conditions that would exist in the
containment following a large pipe break inside containment. 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.
Therefore, the change will 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 calculated loads on all affected structures, systems and
components will remain within their design allowables for all design
basis event categories. No NRC acceptance criterion is exceeded.
Because the BFN Unit 1 configuration and reactions to transients and
hypothetical accidents does not result in exceeding the presently
approved NRC acceptance limits, operation at 105% OLTP does not
involve a significant reduction in a margin of safety.
The post-LOCA suppression chamber airspace pressure is a
byproduct of the conditions that will exist in the containment after
a line break inside containment. Conservative analyses have been
performed that demonstrate that sufficient post-accident suppression
chamber airspace pressure will be available to meet the NPSH
requirements for the low pressure ECCS pumps. By enabling credit of
these conditions for the low pressure ECCS pumps, adequate NPSH
margin will be ensured, and accordingly, the ECCS pumps will meet
their performance requirements. Therefore, the credit for
containment overpressure 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: L. Raghavan.
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, https://
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 pdr@nrc.gov.
AmerGen Energy Company, LLC, Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of application for amendment: December 2, 2005.
Brief description of amendment: The amendment revised the Oyster
Creek Nuclear Generating Station Technical Specifications (TSs) to
increase the allowable as-found main steam safety valve code safety
function lift setpoint tolerance from 1% to 3%.
Date of Issuance: September 13, 2006.
Effective date: As of the date of Issuance to be implemented within
60 days.
Amendment No.: 261.
Facility Operating License No. DPR-16: The amendment revised the
TSs.
Date of initial notice in Federal Register: January 17, 2006 (71 FR
2588).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated September 13, 2006.
No significant hazards consideration comments received: No.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: June 7, 2005, as supplemented
on May 12, 2006.
Brief description of amendments: The amendments revise the
Technical Specifications (TSs) to eliminate the use of the defined term
Core Alterations. The amendments incorporate the changes reflected in
TS Task Force (TSTF) Travelers 471-T (TSTF-471-T), ``Eliminate use of
term CORE ALTERATIONS in ACTIONS and Notes,'' and TSTF-51-A, ``Revise
containment requirements during handling irradiated fuel and core
alterations.'' In addition, the amendments revise TS 3.9.2, ``Nuclear
Instrumentation,'' by replacing ``Core Alterations'' with ``positive
reactivity additions'' in the Required Action for an inoperable source
range monitor during refueling operations. The limiting conditions for
operation in TS 3.9.4, ``Shutdown Cooling (SDC) and Coolant
Recirculation--High Water Level,'' are also revised by replacing ``core
alterations'' with ``movement of fuel assemblies within containment.''
Date of issuance: September 21, 2006
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment Nos.: 279 and 256.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the Licenses and Technical Specifications.
Date of initial notice in Federal Register: July 5, 2005 (70 FR
38716).
The May 12, 2006, letter 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.
[[Page 59534]]
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated September 21, 2006.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: March 3, 2005, as supplemented
by letter dated July 6, 2006.
Brief description of amendment: The amendment revises the
requirements of Technical Specification (TS) 5.6.5, ``Core Operating
Limits Report (COLR).''
Date of issuance: September 20, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No. 209.
Renewed Facility Operating License No. DPR-23. Amendment revises
the Technical Specifications.
Date of initial notice in Federal Register: May 24, 2005 (70 FR
29787). 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 September 20, 2006.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: August 20, 2004, as supplemented
by letters dated June 22, 2005, June 26, 2006, and September 18, 2006.
Brief description of amendment: The amendment revises Table 3.3.1-
1, Functions 3, 14, 17.a., 20, and the footnote related to Function 20.
Date of issuance: September 22, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No. 210.
Renewed Facility Operating License No. DPR-23. Amendment revises
the Technical Specifications.
Date of initial notice in Federal Register: November 23, 2004 (69
FR 68182). The letters dated June 22, 2005, June 26, 2006, and
September 18, 2006, provided clarifying information that 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 September 22, 2006.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423,
Millstone Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: May 27, 2004, as supplemented
September 27, 2004, October 20, 2004, March 23, 2005, January 30, 2006
and May 25, 2006.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) to incorporate a full-scope application of an
alternate source term methodology in accordance with 10 CFR 50.67.
Date of issuance: September 15, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 232.
Facility Operating License No. NPF-49: The amendment revised the
TSs.
Date of initial notice in Federal Register: September 14, 2004 (69
FR 55468). The supplements contained clarifying information only, and
did not change the initial no significant hazards consideration
determination or expand the scope of the initial Federal Register
notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 15, 2006.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423,
Millstone Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: September 13, 2005, as
supplemented by letters dated June 13 and August 14, 2006.
Brief description of amendment: The amendment revised the Technical
Specification (TS) surveillance requirements for the recirculation
spray system.
Date of issuance: September 20, 2006.
Effective date: As of the date of issuance and shall be implemented
prior to entering Mode 1 following refueling outage 3R11.
Amendment No.: 233.
Facility Operating License No. NPF-49: The amendment revised the
TSs.
Date of initial notice in Federal Register: October 25, 2005 (70 FR
61657). The supplements dated June 13 and August 14, 2006, provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination or expand the
application beyond the scope of the original Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 20, 2006.
No significant hazards consideration comments received: No.
Duke Power Company LLC, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: June 29, 2005, as supplemented
May 1, 2006, and September 8, 2006.
Brief description of amendments: The amendments requested
authorization to revise the Updated Final Safety Analysis Report and
the emergency operating procedures to allow an additional operator
action to manually start one containment air return fan in the air
return system in response to Nuclear Regulatory Commission Bulletin
2003-01, ``Potential Impact of Debris Blockage on Emergency Sump
Recirculation at Pressurized-Water Reactors,'' June 6, 2003.
Date of issuance: September 25, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 231 and 227.
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the licenses.
Date of initial notice in Federal Register: October 25, 2005 (70 FR
61657).
The supplement dated May 1, 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 September 25, 2006.
No significant hazards consideration comments received: No.
Duke Power Company LLC, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: July 25, 2005, as supplemented
July 28, 2005, and August 1, 2005.
Brief description of amendments: The amendments revised the
Technical
[[Page 59535]]
Specifications temperature limit for the standby nuclear service water
pond from 91.5 [deg]F to 95 [deg]F.
Date of issuance: September 25, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance September 25, 2006.
Amendment Nos.: 232 and 228.
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: August 4, 2005 (70 FR
44946).
The supplements dated July 28, 2005, and August 1, 2005, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the original proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 25, 2006.
No significant hazards consideration comments received: No.
Duke Power Company LLC, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: September 13, 2005, as
supplemented March 20, 2006.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) to correct a nonconservative TS
associated with spent fuel storage in the spent fuel pool. The licensee
identified the nonconservative TS while comparing results from spent
fuel pool criticality codes.
Date of issuance: September 27, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance September 27, 2006.
Amendment Nos.: 233 and 229.
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the license and the technical specifications.
Date of initial notice in Federal Register: November 21, 2005 (70
FR 70104). The supplement dated March 20, 2006, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the Nuclear
Regulatory Commission (NRC) staff's original proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 27, 2006.
No significant hazards consideration comments received: No.
Duke Power Company LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: June 29, 2005, as supplemented
May 1, 2006.
Brief description of amendments: The amendments requested
authorization to revise the Updated Final Safety Analysis Report and
the emergency operating procedures to allow an additional operator
action to manually start one containment air return fan in the air
return system in response to Nuclear Regulatory Commission Bulletin
2003-01, ``Potential Impact of Debris Blockage on Emergency Sump
Recirculation at Pressurized-Water Reactors,'' June 6, 2003.
Date of issuance: September 25, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 234 and 216.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the licenses.
Date of initial notice in Federal Register: October 25, 2005 (70 FR
61657).
The supplement dated May 1, 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 September 25, 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: June 29, 2005, as supplemented
by letter dated May 18, 2006.
Brief description of amendment: By letter dated June 29, 2005,
Entergy Operations, Inc., the licensee for Arkansas Nuclear One, Unit 2
(ANO-2), requested a license amendment to relocate the shutdown cooling
(SDC) open permissive interlock (OPI) license condition from the
operating license to the licensee's technical requirements manual. The
license condition to maintain OPI operability was previously accepted
by the NRC staff in a letter to the licensee, dated March 30, 2005, and
incorporated into ANO-2's operating license. The OPI prevents the two
SDC suction isolation valves from opening above a selected set point to
separate the high-pressure reactor coolant system from the low-pressure
SDC system.
Date of issuance: September 13, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 267.
Renewed Facility Operating License No. NPF-6: Amendment revised the
Facility Operating License.
Date of initial notice in Federal Register: December 6, 2005 (70 FR
72671). The supplement dated May 18, 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 September 13, 2006.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: December 14, 2004, revised by
letter dated August 30, 2006.
Brief description of amendment: The Technical Specification
amendment relocates structural integrity requirements to the Final
Safety Analysis Report.
Date of issuance: September 14, 2006.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 224.
Facility Operating License No. DPR-35: The amendment revised the
License and Technical Specifications.
Date of initial notice in Federal Register: March 1, 2005 (70 FR
9991).
The licensee originally requested for additional TS relocations in
their submittal dated December 14, 2004. The NRC staff found these
unacceptable. Therefore, the licensee revised the original application
by letter dated August 30, 2006, reducing the scope of the application
as originally noticed. Hence, there is no change to the NRC staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a
[[Page 59536]]
Safety Evaluation dated September 14, 2006.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: June 2, 2005, supplemented by
letter dated June 14, 2006.
Brief description of amendment: The amendment revised the Technical
Specification (TS) reactor coolant system leakage detection
instrumentation requirements and actions.
Date of issuance: September 20, 2006.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 225.
Facility Operating License No. DPR-35: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: May 23, 2006 (71 FR
29676). The supplement dated June 14, 2006, provided additional
information that did not expand the scope of the application as
originally noticed, and did not change the staff's original proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 20, 2006.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412,
Beaver Valley Power Station, Unit No. 2, Beaver County, Pennsylvania
Date of application for amendment: April 11, 2005, as supplemented
December 2, 2005, and January 27, April 14, August 16, and September 1,
2006.
Brief description of amendment: The amendment revised the scope of
the steam generator tubesheet inspections and subsequent repair using
the F* inspection methodology.
Date of issuance: September 27, 2006.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No: 160.
Facility Operating License No. NPF-73: The amendment revised the
License and Technical Specifications.
Date of initial notice in Federal Register: June 7, 2005 (70 FR
33214). The supplements dated December 2, 2005, and January 27, April
14, August 16, and September 1, 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.
The Commission's issuance of Amendment No. 158 to Facility Operating
License NPF-73 for BVPS-2, regarding steam generator tube integrity
(Technical Specification Task Force (TSTF) Item 449) on September 7,
2006, resulted in renumbering and rewording the requirements as
originally proposed by the licensee to fit the TSTF-449 format, but did
not change the scope of the application.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 27, 2006.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2),
Beaver County, Pennsylvania
Date of application for amendments: November 7, 2005, as
supplemented April 25, June 1, and August 3, 2006.
Brief description of amendments: The amendments include changes to
the definition of leakage, changes to the primary-to-secondary leakage
requirements, changes to the steam generator (SG) tube surveillance
program (SG tube integrity), and changes to the SG reporting
requirements.
Date of issuance: September 7, 2006.
Effective date: As of the date of issuance to be implemented within
90 days for BVPS-1 and prior to entry into Mode 4 following the fall
2006 refueling outage for BVPS-2.
Amendment Nos.: 276 and 158.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications and Licenses.
Date of initial notice in Federal Register: December 20, 2005 (70
FR 75491). The supplements dated April 25, June 1, and August 3, 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 September 7, 2006.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2),
Beaver County, Pennsylvania
Date of application for amendments: February 17, 2005, as
supplemented May 12 and August 22, 2006.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) 3.7.7.1 (BVPS-1), ``Control Room
Emergency Habitability Systems,'' and 3.7.7 (BVPS-2), ``Control Room
Emergency Air Cleanup and Pressurization System,'' by dividing these
TSs into two specifications, addressing control room emergency
ventilation and control room air cooling functions separately. The
amendments also improved consistency with the Standard TSs and improved
consistency between the units.
Date of issuance: September 25, 2006
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment Nos.: 277 and 159
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the License and Technical Specifications.
Date of initial notice in Federal Register: April 26, 2005 (70 FR
21458). The supplements dated May 12 and August 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 staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 25, 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: February 1, 2006.
Brief description of amendments: The amendments revise the
Technical Specification (TS) testing frequency for the Surveillance
Requirements in TS 3.1.4, ``Control Rod Scram Times,'' based on the TS
Task Force (TSTF) change traveler TSTF-222, Revision 1.
Date of issuance: September 12, 2006.
Effective date: As of the date of issuance and to be implemented
within 60 days.
Amendment Nos.: 237 and 214.
[[Page 59537]]
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the License and TSs.
Date of initial notice in Federal Register: May 9, 2006 (71 FR
27001).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 12, 2006.
No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket 50-395, Virgil C. Summer Nuclear Station, Unit No. 1,
Fairfield County, South Carolina
Date of application for amendment: June 30, 2005, as supplemented
July 21, 2006.
Brief description of amendment: The amendment revises the Virgil C.
Summer Nuclear Station Technical Specifications to permit the use of a
best estimate methodology in performing loss-of-coolant accident
analyses.
Date of issuance: September 7, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No. 176.
Renewed Facility Operating License No. NPF-12: Amendment revises
the Technical Specifications.
Date of initial notice in Federal Register: October 11, 2005 (70 FR
59087). 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.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 7, 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: November 30, 2005, as
supplemented by letter dated May 30, 2006.
Brief description of amendments: The proposed amendments revised
the Technical Specification (TS) requirements related to steam
generator tube integrity, based on the NRC-approved Revision 4 to TS
Task Force (TSTF)-449, ``Steam Generator Tube Integrity.'' Date of
issuance: September 19, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: Unit 2--204; Unit 3--196.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the TSs.
Date of initial notice in Federal Register: February 14, 2006 (71
FR 7812). The May 30, 2006, supplemental letter 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 September 19, 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: March 9, 2004 (TS-434) as
supplemented on November 15, 2004, and March 7, 2006.
Brief description of amendment: The amendment reduced the Allowable
Value used for Reactor Vessel Water Level--Low, Level 3, for several
instrument functions.
Date of issuance: September 18, 2006.
Effective date: September 18, 2006.
Amendment No.: 258.
Renewed Facility Operating License No. DPR-33: Amendment revised
the Technical Specifications.
Date of initial notice in Federal Register: April 13, 2004 (69 FR
19575). The supplements dated November 15, 2004, and March 7, 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 September 18, 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 amendments: November 3, 2003, as
supplemented May 6, 2004, and August 1, 2006.
Description of amendment request: The amendment revised Technical
Specification (TS) Table 3.3.1.1 -1, Reactor Protection system
Instrumentation, Function 7.b.
Date of issuance: September 19, 2006.
Effective date: Date of issuance, to be issued within 60 days.
Amendment No.: 259.
Renewed Facility Operating License No. DPR-33: Amendment revised
the TSs.
Date of initial notice in Federal Register: April 13, 2004 (69 FR
19575). The supplements dated May 6, 2004, and August 1, 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 September 19, 2006.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama
Date of application for amendments: August 16, 2004, as
supplemented by letters dated March 11, 2005, November 4, 2005, and
April 14, 2006.
Description of amendment request: To extend the channel calibration
frequency requirements for instrumentation in the high-pressure coolant
injection, reactor core isolation cooling, and reactor water core
isolation cooling systems.
Date of issuance: September 21, 2006.
Effective date: Date of issuance, to be implemented within 60 days.
Amendment Nos.: 260, 297 and 255.
Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: May 23, 2006 (71 FR
29680). 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 21, 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: July 9, 2004 (TS 436).
Brief description of amendment: The amendment revised Technical
Specification (TS) Surveillance Requirement 3.6.1.3.10 to increase the
[[Page 59538]]
allowed main steam isolation valve (MSIV) leak rate from 11.5 standard
cubic feet per hour (scfh) per valve to 100 scfh for individual MSIVs
with a 150 scfh combined leakage for all four main steam lines.
Date of issuance: September 27, 2006.
Effective date: Date of issuance, to be implemented within 60 days.
Amendment No.: 261.
Facility Operating License No. DPR-33: Amendment revised the TSs.
Date of initial notice in Federal Register: May 23, 2006 (71 FR
29680).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 27, 2006.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: August 14, 2004, as
supplemented on April 11, 2005, and July 11, 2006 (TS-02-01).
Brief description of amendments: The amendments revise Technical
Specifications (TSs) relating to the reactor protection system and
engineered safety features instrumentation. The Trip Setpoint column of
TS Tables 2.2-1 and 3.3-4 will be renamed Nominal Trip Setpoint;
inequality signs in TS Tables 2.2-1 and 3.3-4 will be removed; the trip
setpoint and allowable value for the Intermediate Range Neutron Flux P-
6 permissive will be revised; Minimum Channels Operable in TS Table
3.3-3 will be revised; editorial changes will be made to TS Table 3.3-4
to replace signs with inequalities; and a correction will
be made to an alarm/trip setpoint in TS Table 3.3-6.
Date of issuance: September 13, 2006.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos. 310 and 299.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revised the technical specifications.
Date of initial notice in Federal Register: October 12, 2004 (69 FR
60688). The supplemental letters 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 amendments is contained
in a Safety Evaluation dated September 13, 2006.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: September 30, 2004, as
supplemented on May 25, 2006.
Brief description of amendments: The amendments revise the
technical specifications to relocate the requirements for the emergency
diesel generator start loss of power instrumentation and associated
actions in the engineering safety features tables to a new limiting
condition for operation (LCO). In addition, an upper allowable value
limit has been added to the voltage sensors for loss of voltage and
degraded voltage consistent with Technical Specification Task Force
(TSTF) Item, TSTF-365, along with a lower allowable value limit for the
degraded voltage diesel generator start and load shed timer. The
auxiliary feedwater loss of power start setpoints and allowable values
have been relocated to this new LCO.
Date of issuance: September 14, 2006.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos. 311 and 300.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revised the technical specifications.
Date of initial notice in Federal Register: January 18, 2005 (70 FR
2900). The supplemental letter provided clarifying information that was
within the scope of the