[Federal Register Volume 71, Number 69 (Tuesday, April 11, 2006)]
[Notices]
[Pages 18371-18380]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E6-5086]
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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 March 17, 2006 to March 30, 2006. The last
biweekly notice was published on March 28, 2006 (71 FR 15479).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
Involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: 1) The name, address, and telephone
number of the requestor or petitioner; 2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; 3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and 4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall
[[Page 18372]]
provide a brief explanation of the bases for the contention and a
concise statement of the alleged facts or expert opinion which support
the contention and on which the petitioner/requestor intends to rely in
proving the contention at the hearing. The petitioner/requestor must
also provide references to those specific sources and documents of
which the petitioner is aware and on which the petitioner/requestor
intends to rely to establish those facts or expert opinion. The
petition must include sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or fact.
Contentions shall be limited to matters within the scope of the
amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and
Adjudications Staff; (3) e-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Dominion Nuclear Connecticut, Inc., Docket Nos. 50-336 and 50-423,
Millstone Power Station, Unit Nos. 2 and 3, New London County,
Connecticut
Date of amendment request: February 7, 2006.
Description of amendment request: The proposed amendments would
increase the allowed outage time from 72 hours to 7 days for the
inoperability of the steam supply to the turbine-driven auxiliary
feedwater (AFW) pump or the inoperability of the turbine-driven AFW
pump under certain operating mode restrictions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1: Does the Proposed Amendment Involve a Significant Increase
in the Probability or Consequences of an Accident Previously Evaluated?
Response: No.
The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed amendment to MPS 2 and 3 [Millstone Power Station,
Unit Nos. 2 and 3] TS [Technical Specification] 3.7.1.2 permits a 7
day allowed outage time for the inoperability of the necessary steam
supply to the turbine-driven AFW pump in Modes 1, 2, and 3, or for
the inoperability of the turbine-driven AFW pump if the
inoperability occurs in Mode 3 following a refueling outage, if Mode
2 had not been entered. Extending the allowed outage time does not
involve a significant increase in the probability or consequences of
an accident previously evaluated because: 1) The proposed amendment
does not represent a change to the system design, 2) the proposed
amendment does not prevent the safety function of the AFW [system]
from being performed since the redundant trains are required to be
operable, 3) the proposed amendment does not alter, degrade, or
prevent action described or assumed in any accident described in the
MPS 2 and 3 FSARs [final safety analysis reports] from being
performed since the other trains of AFW are required to be operable,
4) the proposed amendment does not alter any assumptions previously
made in evaluating radiological consequences, and 5) the proposed
amendment does not affect the integrity of any fission product
barrier. No other safety related equipment is affected by the
proposed change. Therefore, this proposed amendment does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
Criterion 2: Does the Proposed Amendment Create the Possibility of a
New or Different Kind of Accident From Any Accident Previously
Evaluated?
Response: No.
The proposed amendment does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed amendment to MPS 2 and 3 TS 3.7.1.2 would allow a 7
day allowed outage time for the inoperability of the necessary steam
supply to the turbine-driven AFW pump in Modes 1, 2, and 3, or for
the inoperability of the turbine-driven AFW pump if the
inoperability occurs in Mode 3 following a refueling outage, if Mode
2 had not been entered. Extending the allowed action time does not
create the possibility of a new or different kind of accident from
any accident previously evaluated because: 1) the proposed amendment
does not represent a change to the system design, 2) the proposed
amendment does not alter how equipment is operated or the ability of
the system to deliver the required AFW flow, and 3) the
[[Page 18373]]
proposed amendment does not affect any other safety related
equipment. Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Criterion 3: Does the Proposed Amendment Involve a Significant
Reduction in a Margin of Safety?
Response: No.
The proposed amendment does not involve a significant reduction
in a margin of safety.
The proposed amendment to MPS 2 and 3 TS 3.7.1.2 would allow a 7
day allowed action time for the inoperability of the necessary steam
supply to the turbine-driven AFW pump in Modes 1, 2, and 3.
Extending the allowed action time does not involve a significant
reduction in a margin of safety because: 1) There is a redundant
steam supply to the turbine driven AFW pump, 2) the motor-driven AFW
pumps are required to be operable when Mode 3 is entered, 3) the
motor-driven AFW pumps can provide sufficient flow to remove decay
heat and cool the unit to shutdown cooling system entry conditions
from power operations, 4) the motor-driven AFW pumps are designed to
supply sufficient water to remove decay heat with steam generator
pressure at no load conditions to cool the unit to shutdown cooling
entry conditions, 5) the proposed change does not change or
introduce any new setpoints at which mitigating functions are
initiated, 6) no changes to the design parameters of the AFW
[system] are being proposed, and 7) no changes in system operation
that would impact an established safety margin are being proposed by
this change.
The proposed amendment to MPS 2 and 3 TS 3.7.1.2 would also
allow a 7 day allowed action time for the inoperability of the
turbine-driven AFW pump if the inoperability occurs in Mode 3
following a refueling outage, if Mode 2 had not been entered.
Extending the allowed action time does not involve a significant
reduction in a margin of safety because: (1) During a return to
power operations following a refueling outage, decay heat is at its
lowest levels, (2) the motor-driven AFW pumps are required to be
operable when Mode 3 is entered, (3) the motor-driven AFW pumps can
provide sufficient flow to remove decay heat and cool the unit to
shutdown cooling system entry conditions from power operations, (4)
the motor-driven AFW pumps are designed to supply sufficient water
to remove decay heat with steam generator pressure at no load
conditions to cool the unit to shutdown cooling entry conditions,
(5) the proposed change does not change or introduce any new
setpoints at which mitigating functions are initiated, (6) no
changes to the design parameters of the AFW are being proposed, and
(7) no changes in system operation that would impact an established
safety margin are being proposed by this change
Therefore, based on the above, the proposed amendment does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT
06385.
NRC Branch Chief: Darrell J. Roberts.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: March 1, 2006.
Description of amendment request: The proposed amendments would
revise the Technical Specifications to reconcile the 10 CFR Part 50 and
10 CFR Part 72 criticality requirements for the loading and unloading
of dry spent fuel storage canisters in the spent fuel pool (SFP).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The applicable accidents are the dropped fuel assembly and drop
of the 100 ton spent fuel cask into the SFP. This amendment request
does not change the fuel assemblies or any of the Part 50
structures, systems, or components involved in fuel assembly or cask
handling or any of the operations involved. Therefore, this
amendment request does not affect the probability of an accident
previously evaluated.
The proposed change does not increase the consequences of an
accident previously evaluated for the following reasons: there is no
increase in radiological source terms for the fuel; there is no
change to the SFP water level; subcriticality is maintained for
normal and accident conditions for the spent fuel storage racks and
for cask loading and unloading; and the same boron concentrations
that were previously credited for the spent fuel storage racks are
assumed in the criticality analysis for cask loading and unloading.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated.
Handling of fuel assemblies and the NUHOMS[reg] spent fuel cask
have been previously evaluated for Oconee. These activities lead to
evaluation of the fuel handling accident (dropped fuel assembly) and
drop of the 100 ton spent fuel cask onto spent fuel stored in the
Oconee SFP. These elements of the license amendment request (LAR)
are not new, and thus do not create the potential for new or
different kinds of accidents.
The new element of this LAR is the application of additional
criticality controls (i.e., minimum burnup requirements for the fuel
assemblies) beyond the 10 CFR 72 controls already in place for the
NUHOMS[reg] spent fuel cask. However, application of such
criticality controls is not a new activity for Oconee, since similar
criticality controls are currently applied to the spent fuel storage
racks. Fuel assembly misloading is not a new accident; as discussed
in Enclosure 3, Section 6.5, fuel assembly misloading has been
considered previously for the NUHOMS[reg] spent fuel cask and for
the Oconee spent fuel pool racks. Furthermore, the criticality
analysis for cask loading and unloading evaluates the same boron
concentrations, moderator temperatures, and misloading scenario as
previously evaluated for the spent fuel storage racks. The analysis
demonstrates that a criticality accident does not occur under these
conditions. It is concluded that the possibility of a criticality
accident is not created since application of criticality controls is
not new and the analysis demonstrates that criticality does not
occur. More generally, this supports the conclusion that the
potential for new or different kinds of accidents is not created.
(3) Involve a significant reduction in a margin of safety.
This LAR involves the application of additional criticality
controls (minimum burnup requirements) to the 10 CFR 72 controls
already in place for the NUHOMS[reg] spent fuel cask. The
criticality analysis demonstrates subcriticality margins are
maintained for normal and accident conditions consistent with 10 CFR
50.68(b) and other NRC guidance. Margins previously established for
Oconee's spent fuel storage racks are not altered. Therefore, this
LAR does not result in a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Branch Chief: Evangelos C. Marinos.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
2, Pope County, Arkansas
Date of amendment request: February 14, 2006.
Description of amendment request: The proposed change will modify
the Arkansas Nuclear One, Unit 2 (ANO-2) Technical Specification (TS)
Surveillance Requirement 4.6.1.1.a. Specifically, the proposed change
will eliminate the requirement to verify containment isolation valves
that are maintained locked, sealed, or otherwise secured closed from
the monthly
[[Page 18374]]
position verification. The proposed change will result in reducing
radiological exposure to Operations, Health Physics, and Security
personnel.
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 accident mitigation features of the plant for previously
evaluated accidents are not affected by the proposed change. No
changes are proposed to the physical components or to the
containment isolation function.
Repositioning of manual containment isolation valves is
procedurally controlled and governed by the note that is contained
in TS 3.6.3.1, Containment Isolation Valves, which allows opening
locked or sealed closed valves on an intermittent basis. The valve
position is tracked until it is restored to its original position
(locked or deactivated position, as appropriate). While the valve
remains open, an individual, in constant communication with the
control room staff, is stationed at the valve. If an accident were
to occur, the control room staff would direct the individual
stationed at the valve to close the valve thereby precluding the
release of radioactivity outside containment.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not change the design, method of
operation, or configuration of the plant. The procedural controls
that establish the ANO-2 containment valve program controls and
include the administrative controls that are associated with the
note in TS 3.6.3.1, ensure containment integrity is appropriately
established such that no new or different types of accidents are
created.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not change the design basis for any
equipment in the plant. The proposed change will exclude
verification of the normally locked, sealed, or otherwise secured
closed valves, blind flanges, and the deactivated automatic valves;
however, the administrative controls applied to these components
ensure that containment integrity is maintained.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1700 K Street, NW., Washington, DC 20006-3817.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment requests: March 7, 2006.
Description of amendment requests: The proposed amendments would
modify the Technical Specifications (TS) of the units by expanding
Section 5.5.2, Leakage Monitoring Program, to include the Liquid Waste
Disposal System, the Waste Gas System, and the Post-Accident
Containment Hydrogen Monitoring System. These systems are currently in
the licensee's own leakage monitoring program but are not listed in TS
Section 5.5.2. The licensee also proposed to make an editorial change
to the section.
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. The NRC staff reviewed the licensee's analysis, and
performed its own as follows:
(1) Does the proposed change involve a significant increase in
the probability of occurrence or consequences of any accident
previously evaluated?
No. The proposed change would only add the three subject systems
to the listing in Section 5.5.2. The licensee is currently
performing leakage monitoring of these systems under its own
program. Leakage monitoring of these three systems, whether listed
in the TS or not, does not have any impact on the initiation of any
accident previously analyzed, or on the scenarios and radiological
consequences of these accidents. Consequently, the proposed change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
(2) Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The proposed change is purely administrative, and does not
involve any change to the design or operation of a system,
structure, or component. Consequently, the proposed change leads to
no possibility to create 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?
No. The proposed change would not change any assumption,
analysis method, calculation model, or acceptance criterion.
Accordingly, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the NRC staff's analysis, it appears that the three
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment requests involve no
significant hazards consideration.
Attorney for licensee: James M. Petro, Jr., Esquire, One Cook
Place, Bridgman, MI 49106.
NRC Branch Chief: L. Raghavan.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: February 16, 2006.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) requirements related to steam
generator (SG) tube integrity. The change is consistent with NRC-
approved Revision 4 to Technical Specification Task Force (TSTF)
Standard Technical Specification Change Traveler, TSTF-449, ``Steam
Generator Tube Integrity.'' The availability of this TS improvement was
announced in the Federal Register on May 6, 2005 (70 FR 24126) as part
of the consolidated line item improvement process (CLIIP).
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on March 2,
2005 (70 FR 10298) as part of the CLIIP. The licensee affirmed the
applicability of the model NSHC determination in its application dated
February 16, 2006.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change requires a SG Program that includes
performance criteria that will provide reasonable assurance that the
SG tubing will retain integrity over the full range of operating
conditions (including startup, operation in the power range, hot
[[Page 18375]]
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 SGTR [steam generator 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 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 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 to the TS identify
the standards against which tube integrity is to be measured.
Meeting the performance criteria provides reasonable assurance that
the SG tubing will remain capable of fulfilling its specific safety
function of maintaining reactor coolant pressure boundary integrity
throughout each operating cycle and in the unlikely event of a
design basis accident. The performance criteria are only a part of
the SG Program required by the proposed change to the TS. The
program, defined by NEI [Nuclear Energy Institute] 97-06, Steam
Generator Program Guidelines, includes a framework that incorporates
a balance of prevention, inspection, evaluation, repair, and leakage
monitoring. The proposed changes do not, therefore, significantly
increase the probability of an accident previously evaluated.
The consequences of design basis accidents are, in part,
functions of the dose equivalent I-131 in the primary coolant and
the primary to secondary leakage rates resulting from an accident.
Therefore, limits are included in the plant technical specifications
for operational leakage and for dose equivalent I-131 in primary
coolant to ensure the plant is operated within its analyzed
condition. The typical analysis of the limiting design basis
accident assumes that primary to secondary leak rate after the
accident is 1 gallon per minute 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 I-131 are at the TS
values before the accident.
The proposed change does not affect the design of the SGs, their
method of operation, or primary coolant chemistry controls. The
proposed approach updates the current TSs and enhances the
requirements for SG inspections. The proposed change does not
adversely impact any other previously evaluated design basis
accident and is an improvement over the current TSs.
Therefore, the proposed change does not affect the consequences
of a SGTR accident and the probability of such an accident is
reduced. In addition, the proposed changes do not affect the
consequences of an MSLB, rod ejection, or a reactor coolant pump
locked rotor event, or other previously evaluated accident.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed performance based requirements are an improvement
over the requirements imposed by the current technical
specifications. Implementation of the proposed SG Program will not
introduce any adverse changes to the plant design basis or
postulated accidents resulting from potential tube degradation. The
result of the implementation of the SG Program will be an
enhancement of SG tube performance. Primary to secondary leakage
that may be experienced during all plant conditions will be
monitored to ensure it remains within current accident analysis
assumptions.
The proposed change does not affect the design of the SGs, their
method of operation, or primary or secondary coolant chemistry
controls. In addition, the proposed change does not impact any other
plant system or component. The change enhances SG inspection
requirements.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
isolate the radioactive fission products in the primary coolant from
the secondary system. In summary, the safety function of an SG is
maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change does not affect tube design or operating environment. The
proposed change is expected to result in an improvement in the tube
integrity by implementing the SG Program to manage SG tube
inspection, assessment, repair, and plugging. The requirements
established by the SG Program are consistent with those in the
applicable design codes and standards and are an improvement over
the requirements in the current TSs.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed change to the
TS.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Branch Chief: L. Raghavan.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of amendment request: February 13, 2006.
Description of amendment request: The proposed amendments would
make miscellaneous administrative changes by revising Technical
Specifications (TS) 3.0 ``Surveillance Requirement (SR)
Applicability''; and TS Chapter 5.0, ``Administrative Controls''. The
proposed changes will improve TS usability, conformance with the
industry standard, NUREG-1431, ``Standard Technical Specifications,
Westinghouse Plants'', Revision 3.0 (NUREG-1431) and accuracy.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
This license amendment request proposes administrative changes
to the Prairie Island Nuclear Generating Plant Technical
Specifications as follows: Technical Specification 3.0,
``Surveillance Requirement (SR) Applicability'', revise page headers
and correct capitalization; and Technical Specification Chapter 5.0,
``Administrative Controls'', correct Topical Report numbers and make
format corrections.
The proposed changes are administrative and do not affect plant
operation maintenance or testing. These changes do not affect any
plant systems which are accident initiators and thus these changes
do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This license amendment request proposes administrative changes
to the Prairie Island Nuclear Generating Plant Technical
Specifications as follows: Technical Specification 3.0,
``Surveillance Requirement (SR) Applicability'', revise page headers
and
[[Page 18376]]
correct capitalization; and Technical Specification Chapter 5.0,
``Administrative Controls'', correct Topical Report numbers and make
format corrections.
The proposed changes are administrative and thus do not create
new failure modes or mechanisms and do not generate new accident
precursors. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
This license amendment request proposes administrative changes
to the Prairie Island Nuclear Generating Plant Technical
Specifications as follows: Technical Specification 3.0,
``Surveillance Requirement (SR) Applicability'', revise page headers
and correct capitalization; and Technical Specification Chapter 5.0,
``Administrative Controls'', correct Topical Report numbers and make
format corrections.
The proposed Technical Specification changes are administrative
and do not affect plant operation, maintenance or testing.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Branch Chief: L. Raghavan.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of amendment request: February 16, 2006.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) requirements related to steam
generator tube integrity. The change is consistent with NRC-approved
Revision 4 to Technical Specification Task Force (TSTF) Standard
Technical Specification Change Traveler, TSTF-449, ``Steam Generator
Tube Integrity.'' The availability of this TS improvement was announced
in the Federal Register on May 6, 2005 (70 FR 24126) as part of the
consolidated line item improvement process (CLIIP).
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on March 2,
2005 (70 FR 10298) as part of the CLIIP. The licensee affirmed the
applicability of the model NSHC determination in its application dated
February 16, 2006.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1--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 SGTR event is one of the design basis accidents that are
analyzed as part of a plant's licensing basis. In the analysis of a
SGTR event, a bounding primary to secondary leakage rate equal to
the operational leakage rate limits in the licensing basis plus the
leakage rate associated with a double-ended rupture of a single tube
is assumed.
For other design basis accidents such as MSLB, rod ejection, and
reactor coolant pump locked rotor the tubes are assumed to retain
their structural integrity (i.e., they are assumed not to rupture).
These analyses typically assume that primary to secondary leakage
for all SGs is 1 gallon per minute or increases to 1 gallon per
minute as a result of accident induced stresses. The accident
induced leakage criterion introduced by the proposed changes
accounts for tubes that may leak during design basis accidents. The
accident induced leakage criterion limits this leakage to no more
than the value assumed in the accident analysis.
The SG performance criteria proposed change to the TS identify
the standards against which tube integrity is to be measured.
Meeting the performance criteria provides reasonable assurance that
the SG tubing will remain capable of fulfilling its specific safety
function of maintaining reactor coolant pressure boundary integrity
throughout each operating cycle and in the unlikely event of a
design basis accident. The performance criteria are only a part of
the SG Program required by the proposed change to the TS. The
program, defined by NEI 97-06, Steam Generator Program Guidelines,
includes a framework that incorporates a balance of prevention,
inspection, evaluation, repair, and leakage monitoring. The proposed
changes do not, therefore, significantly increase the probability of
an accident previously evaluated.
The consequences of design basis accidents are, in part,
functions of the dose equivalent 1-131 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 1-131 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 1-131 are at the TS
values before the accident.
The proposed change does not affect the design of the SGs, their
method of operation, or primary coolant chemistry controls. The
proposed approach updates the current TSs and enhances the
requirements for SG inspections. The proposed change does not
adversely impact any other previously evaluated design basis
accident and is an improvement over the current TSs.
Therefore, the proposed change does not affect the consequences
of a SGTR accident and the probability of such an accident is
reduced. In addition, the proposed changes do not affect the
consequences of an MSLB, rod ejection, or a reactor coolant pump
locked rotor event, or other previously evaluated accident.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed performance based requirements are an improvement
over the requirements imposed by the current technical
specifications. Implementation of the proposed SG Program will not
introduce any adverse changes to the plant design basis or
postulated accidents resulting from potential tube degradation. The
result of the implementation of the SG Program will be an
enhancement of SG tube performance. Primary to secondary leakage
that may be experienced during all plant conditions will be
monitored to ensure it remains within current accident analysis
assumptions.
The proposed change does not affect the design of the SGs, their
method of operation, or primary or secondary coolant chemistry
controls. In addition, the proposed change does not impact any other
plant system or component. The change enhances SG inspection
requirements.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
isolate the
[[Page 18377]]
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. Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Branch Chief: L. Raghavan.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: February 21, 2006.
Description of amendment request: The amendment would revise
Technical Specification 5.5.9, ``Steam Generator (SG) Tube Surveillance
Program,'' to exclude portions of the SG tube below the top of the
tubesheet in the SGs from periodic tube inspections based on the
application of structural analysis and leak rate evaluation results to
re-define the primary-to-secondary pressure boundary.
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 previously analyzed accidents are initiated by the failure
of plant structures, systems, or components. The proposed change
that alters the steam generator inspection criteria does not have a
detrimental impact on the integrity of any plant structure, system,
or component that initiates an analyzed event. The proposed change
will not alter the operation of, or otherwise increase the failure
probability of any plant equipment that initiates an analyzed
accident.
Of the applicable accidents previously evaluated, the limiting
transients[,] with respect to the proposed [change] to the steam
generator tube inspection criteria, are the steam generator tube
rupture (SGTR) event and the steam line break (SLB) accident.
During the SGTR event, the required structural integrity margins
of the steam generator tubes will be maintained by the presence of
the steam generator tubesheet. Steam generator tubes are
hydraulically expanded in the tubesheet area. Tube rupture in tubes
with cracks in the tubesheet is precluded by the constraint provided
by the tubesheet. This constraint results from the hydraulic
expansion process, thermal expansion mismatch between the tube and
the tubesheet[,] and from the differential pressure between the
primary and secondary side [of the steam generator]. Based on this
design, the structural margins against burst, discussed in
Regulatory Guide (RG) 1.121, ``Bases for Plugging Degraded PWR
[Pressurized-Water Reactor] Steam Generator Tubes,'' are maintained
for both normal and postulated accident conditions.
The proposed change does not affect other systems, structures,
components or operational features. Therefore, the proposed [change
results] in no significant increase in the probability [or] the
occurrence of a[n] SGTR accident.
At normal operating pressures, leakage from primary water stress
corrosion cracking (PWSCC) [of a tube] below the proposed inspection
depth is limited by both the tube-to-tubesheet crevice and the
limited crack opening permitted by the tubesheet constraint.
Consequently, negligible normal operating leakage is expected from
cracks within the tubesheet region. The consequences of an SGTR
event are affected by the primary-to-secondary leakage flow during
the event. Primary-to-secondary leakage flow through a postulated
ruptured tube is not affected by the proposed change since the
tubesheet enhances the tube integrity in the region of the hydraulic
expansion by precluding tube deformation beyond its initial
hydraulically expanded outside diameter.
The probability of an SLB is unaffected by the potential failure
of a steam generator tube as this failure is not an initiator for an
SLB.
The consequences of an SLB are also not significantly affected
by the proposed change. During an SLB accident, the reduction in
pressure above the tubesheet on the secondary side of the steam
generator creates a uniformly distributed axial (out of plane) load
on the tubesheet due to the reactor coolant system pressure on the
primary [side] of the tubesheet. The resulting bending action causes
contraction of the tube holes below the tubesheet neutral axis,
adding to the constraint of the tubes in the tubesheet, thereby
further restricting primary-to-secondary leakage.
Primary-to-secondary leakage from tube degradation in the
tubesheet area during the limiting accident (i.e., an SLB) is
limited by flow restrictions resulting from the crack and tube-to-
tubesheet contact pressures that provide a restricted leakage path
above the indications and also limit the degree of potential crack
face opening as compared to free span indications. The primary-to-
secondary leak rate from tube degradation in the tubesheet region
during postulated SLB accident conditions will be no more than twice
that allowed during normal operating conditions when the pressure
boundary is relocated [by the amendment] to the lesser of the H* or
B* [tubesheet inspection] depths. Since normal operating leakage
would be limited to 300 gpd [gallons per day] (0.2 gpm [gallons per
minute]) through any one steam generator per TS 3.4.13, ``RCS
[Reactor Coolant System] Operational leakage,'' the associated
accident condition leak rate, assuming all leakage to be from lower
tubesheet indications, would be limited to 150 gpd per steam
generator. This value is well within the assumed accident leakage
rate of 1.0 gpm discussed in WCGS [(Wolf Creek Generating Station)]
Updated Safety Analysis Report, Table 15.1-3, ``Parameters Used in
Evaluating the Radiological Consequences of a Main Steam Line
Break.'' Therefore, the consequences of an SLB accident remain
unaffected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
(2) Does the proposed change create the possibility of a new or
different accident from any accident previously evaluated?
Response: No.
The proposed change does not introduce any new equipment, create
new failure modes for existing equipment, or create any new limiting
single failures. Plant operation will not be altered, and all safety
functions will continue to perform as previously assumed in accident
analyses. [Excluding portions of the tube below the proposed
tubesheet inspection depths does not introduce a new or different
kind of accident to the steam generator tube because the required
structural margins of the tubes for both normal and accident
conditions are maintained.] Therefore, the proposed [change does]
not create the possibility of a new or different kind of accident
from any previously evaluated.
(3) Does the proposed change involve a significant reduction in
a margin of safety?
Response: No.
The proposed [change maintains] the required structural margins
of the steam generator tubes for both normal and accident
conditions. Nuclear Energy Institute (NEI) 97-06, ``Steam Generator
Program Guidelines,'' and RG 1.121, ``Bases for Plugging Degraded
PWR Steam Generator Tubes,'' are used as the bases in the
development of the tubesheet inspection depth methodology for
determining that steam generator tube integrity considerations
[[Page 18378]]
are maintained within acceptable limits. RG 1.121 describes a method
acceptable to the NRC for meeting General Design Criteria (GDC) 14,
``Reactor coolant pressure boundary,'' GDC 15, ``Reactor coolant
system design,'' GDC 31, ``Fracture prevention of reactor coolant
pressure boundary,'' and GDC 32, ``Inspection of reactor coolant
pressure boundary,'' by reducing the probability and consequences of
a[n] SGTR. RG 1.121 concludes that by determining the limiting safe
conditions for tube wall degradation[,] the probability and
consequence of a[n] SGTR are reduced. This RG uses safety factors on
loads for tube burst that are consistent with the requirements of
Section III of the American Society of Mechanical Engineers (ASME)
Code.
For axially oriented cracking located within the tubesheet, tube
burst is precluded due to the presence of the tubesheet. For
circumferentially oriented cracking, Westinghouse letter LTR-CDME-
05-209-P, ``Steam Generator Tube Alternate Repair Criteria for the
Portion of the Tube Within the Tubesheet at the Wolf Creek
Generating Station,'' [provided in the application,] defines a
length of degradation-free expanded tubing that provides the
necessary resistance to tube pullout due to the pressure induced
forces, with applicable safety factors applied. Application of the
limited tubesheet inspection depth criteria will preclude
unacceptable primary-to-secondary leakage during all plant
conditions. The methodology for determining leakage provides for
large margins between calculated and actual leakage values in the
proposed limited tubesheet inspection depth criteria.
Therefore, the proposed [change does not] involve a significant
reduction in any 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: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, DC 20037.
NRC Branch Chief: David Terao.
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. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of application for amendment: February 25, 2005.
Brief description of amendment: The amendment deleted the reporting
requirement in the Facility Operating License (FOL) related to
reporting violations of other requirements in the operating license.
Date of issuance: February 24, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 172.
Facility Operating License No. NPF-62: The amendment revised the
FOL.
Date of initial notice in Federal Register: April 26, 2005 (70 FR
21450).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 24, 2006.
No significant hazards consideration comments received: No.
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of application for amendment: March 25, 2005.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) to exclude the containment purge valve leakage
rates from the summation of secondary containment bypass leakage rates.
Date of issuance: March 21, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days of the date of issuance.
Amendment No.: 173.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 26, 2005 (70 FR
21451).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 21, 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: July 13, 2005, as supplemented
on November 29, 2005, and January 20 and February 13, 2006.
Brief description of amendments: The amendments revised Technical
Specification (TS) 1.1, ``Definitions,'' TS 3.4.13, ``RCS [reactor
coolant system] Operational Leakage,'' TS 5.5.9, ``Steam Generator Tube
Surveillance Program,'' and TS 5.6.9, ``Steam Generator [SG] Tube
Inspection Report,'' and add a new specification (TS 3.4.18) for SG
Tube Integrity. The changes are consistent with TS Task Force (TSTF)
Change TSTF-449, Revision 4, ``Steam Generator Tube Integrity.''
Date of issuance: March 9, 2006.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment Nos.: 278 and 255.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: December 6, 2005 (70 FR
72669).
The November 29, 2005, and January 20 and February 13, 2006,
supplements provided additional information that clarified the
application, did not expand the scope of the application as originally
[[Page 18379]]
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 these amendments is
contained in a Safety Evaluation dated March 9, 2006.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket Nos. 50-247, Indian Point
Nuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: May 25, 2005, as supplemented by
letter dated January 23, 2006.
Brief description of amendment: The amendment revises the Technical
Specification limit on pressurizer water level in Mode 3 (hot standby).
Date of issuance: March 22, 2006.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 246.
Facility Operating License Nos. DPR-26 and DPR-64: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: June 21, 2005 (70 FR
35736).
The January 23, 2006, supplement 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 as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 22, 2006.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: October 3, 2005.
Brief description of amendment: The amendment revises Technical
Specification Surveillance Requirements to reflect changes to the
Emergency Core Cooling System throttle valves. The amendment adds the
modified throttle valves to the surveillance, removes existing throttle
valves that are now locked closed from the surveillance, and adds
existing valves to the surveillance that are used in a throttle
position when open.
Date of issuance: March 23, 2006.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 230.
Facility Operating License No. DPR-64: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 6, 2005 (70 FR
72670).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 23, 2006.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of application for amendment: February 25, 2005.
Brief description of amendment: The amendments delete the sections
of the Facility Operating Licenses that require reporting of violations
of the requirements in Section 2.C of the Facility Operating Licenses.
Date of issuance: March 13, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 146, 146, 139 and 139.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Facility Operating License.
Date of initial notice in Federal Register: April 26, 2005 (70 FR
21456).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 13, 2006.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station (DNPS), Units 2 and 3, Grundy County, Illinois
Date of application for amendment: April 4, 2005, as supplemented
by letter dated January 13, 2006.
Brief description of amendment: The amendments revised Technical
Specification 3.3.8.1, ``Loss of Power (LOP) Instrumentation,'' and
also revised the Updated Final Safety Analysis Report to implement use
of automatic load tap changers on transformers that provide offsite
power to DNPS, Units 2 and 3.
Date of issuance: March 17, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 219/210.
Facility Operating License Nos. DPR-19 and DPR-25: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 8, 2005 (70 FR
67747).
The January 13, 2006 supplement, contained clarifying information
and did not change the NRC staff's initial proposed finding of no
significant hazards consideration.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 17, 2006.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-352, Limerick Generating
Station, Unit 1, Montgomery County, Pennsylvania
Date of application for amendment: January 10, 2005.
Brief description of amendment: The amendment removed the license
conditions concerning the emergency core cooling system pump suction
strainers from Appendix C of Facility Operating License No. NPF-39.
Date of issuance: March 6, 2006.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 184.
Facility Operating License No. NPF-39.
This amendment revised the License.
Date of initial notice in Federal Register: January 3, 2006 (71 FR
149).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 5, 2006.
No significant hazards consideration comments received: No.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: September 29, 2005.
Description of amendment request: The proposed amendment would
revise the Seabrook Station, Unit No. 1 Technical Specifications (TSs)
to permit a one-time, 6-month addition to the currently approved 5-year
extension to the 10-year test interval for the containment integrated
leak rate test.
Date of issuance: March 24, 2006.
Effective date: As of its date of issuance, and shall be
implemented within 90 days.
Amendment No.: 108.
Facility Operating License No. NPF-86: The amendment revised the
TSs.
Date of initial notice in Federal Register: November 8, 2005 (70 FR
67748).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 24, 2006.
No significant hazards consideration comments received: No.
[[Page 18380]]
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of application for amendment: April 26, 2005.
Brief description of amendment: The amendment revises Technical
Specification 5.6.5.b, ``Core Operating Limits Report,'' to use a
revised fuel assembly growth model for Palisades as described in
Topical Report BAW-2489P, ``Revised Fuel Assembly Growth Correlation
for Palisades,'' Revision 0.
Date of issuance: March 27, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 222.
Facility Operating License No. DPR-20. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: (70 FR 29797).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 27, 2006.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket No. 50-387, Susquehanna Steam Electric
Station, Unit 1 (SSES 1), Luzerne County, Pennsylvania
Date of application for amendment: December 1, 2005, as
supplemented on February 17, 2006.
Brief description of amendment: The amendment changes the SSES 1
Technical Specifications (TSs) by revising the Unit 1 Cycle 15 Minimum
Critical Power Ratio Safety Limit for single-loop operation in TS
2.1.1.2 and the references listed in TS 5.6.5.b.
Date of issuance: March 20, 2006.
Effective date: As of the date of issuance and to be implemented
within 30 days.
Amendment No.: 231.
Facility Operating License No. NPF-14: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 17, 2006 (71 FR
2595).
The supplement dated February 17, 2006, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 20, 2006.
No significant hazards consideration comments received: No.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: October 6, 2004, as supplemented by
letters dated September 16 and November 22, 2005.
Brief description of amendments: The amendments revised the
Technical Specification 3.8.1, ``AC Sources--Operating,'' to remove
mode restrictions on surveillance requirements.
Date of issuance: March 15, 2006.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: 124 and 124.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 15, 2005 (70 FR
12751).
The supplements dated September 16 and November 22, 2005, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 15, 2006.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: October 26, 2005.
Brief description of amendment: The amendment revised Required
Action D.1, in Technical Specification (TS) 3.6.6, ``Containment Spray
and Cooling Systems,'' to require plant shutdown if both containment
cooling trains are out of service, which is more conservative than the
previous requirement that allowed 72 hours to restore one of the
inoperable trains. There are also changes to other required actions in
TS 3.6.6 to reflect the revision to Required Action D.1.
Date of issuance: March 28, 2006.
Effective date: As of its date of issuance, and shall be
implemented within 90 days of the date of issuance.
Amendment No.: 171.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register:January 17, 2006 (71 FR
2597).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 28, 2006.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 3rd day of April 2006.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E6-5086 Filed 4-10-06; 8:45 am]
BILLING CODE 7590-01-P