[Federal Register Volume 71, Number 195 (Tuesday, October 10, 2006)]
[Notices]
[Pages 59529-59538]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E6-16560]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a determination by the Commission that such
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from September 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, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the
[[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, [email protected]; or (4) facsimile
transmission addressed to the Office of the Secretary, U.S. Nuclear
Regulatory Commission, Washington, DC, Attention: Rulemakings and
Adjudications Staff at (301) 415-1101, verification number is (301)
415-1966. A copy of the request for hearing and petition for leave to
intervene should also be sent to the Office of the General Counsel,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and it
is requested that copies be transmitted either by means of facsimile
transmission to (301) 415-3725 or by e-mail to [email protected]. A
copy of the request for hearing and petition for leave to intervene
should also be sent to the attorney for the licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
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, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-219, 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 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
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