[Federal Register Volume 71, Number 30 (Tuesday, February 14, 2006)]
[Notices]
[Pages 7804-7817]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 06-1162]
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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 January 20, 2006, to February 2, 2006. The
last biweekly notice was published on January 31, 2006 (71 FR 5078).
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
[[Page 7805]]
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 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 email to [email protected].
Dairyland Power Cooperative, Docket No. 50-409, La Crosse Boiling Water
Reactor, Genoa, Wisconsin
Date of amendment request: December 13, 2005.
Description of amendment requests: The La Crosse Boiling Water
Reactor (LACBWR) is currently undergoing limited decommissioning and
dismantlement. The proposed license amendment would revise Technical
Specifications (TS) to allow waste processing components or fixtures to
be handled over the Fuel Element Storage Well (FESW), limiting the
weight of such items to 50 tons (the weight of the heavy load drop
found acceptable in the cask drop analyses performed for the LACBWR
FESW). The proposed wording changes to the TS would allow processing
and shipment of Class B and Class C radioactive waste currently stored
in the FESW, which will require a cask similar to the spent fuel
shipping cask reflected in the current TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR Part 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No.
The shipping cask, whether it is a spent fuel shipping cask or a
waste shipping cask, will be handled with the same equipment, under
essentially the same LACBWR crane operating procedures and
precautions, and will be conservatively enveloped by previous
accident evaluations that assumed a heavy load drop weighing 50
tons. Allowing the placement of typical waste processing equipment
in the FESW and the handling of a waste shipping cask limited to
weighing less than 50 tons over the FESW may increase the number of
cask movements over the FESW slightly but will not increase the
probability nor consequences of an accident previously evaluated
during a given cask handling.
(2) Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No.
Simply changing the name of the heavy load handled over the FESW
from ``spent fuel shipping cask'' to the generic term ``shipping
cask,'' as long as the heavy loads are limited to the analyzed drop
weight of 50 tons and their methods of handling are essentially
equivalent, does not create the possibility of a new or different
kind of accident from any accident previously evaluated. Other waste
processing equipment will likewise be limited to the analyzed drop
weight.
[[Page 7806]]
(3) Does the proposed change involve a significant reduction in
a margin of safety? No.
Any shipping cask or other waste processing equipment to be
handled over the LACBWR FESW will be conservatively enveloped by the
load and conditions in the heavy load drop analysis, which assumed a
drop weight of 50 tons, performed for the LACBWR FESW and,
therefore, the TS change will not involve a significant reduction in
a margin of safety.
The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the
licensee's analysis and, based on this review, it appears that the
three standards of 10 CFR Part 50.92(c) are satisfied. Therefore, NRC
staff proposes to determine that the amendment request involves no
significant hazards consideration.
NRC Section Chief: Claudia Craig.
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power
Station, Kewaunee County, Wisconsin
Date of amendment request: January 12, 2006.
Description of amendment request: The proposed changes to the
Technical Specifications (TSs) are necessary in order to implement the
guidance for the industry initiative on NEI 97-06, ``Steam Generator
[SG] Program Guidelines.''
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, via reference to a generic analysis published in the
Federal Register on March 2, 2005 (70 FR 10298). In addition, the
licensee's January 12, 2006, application contains analysis of the issue
of no significant hazards consideration associated with those changes
to the TS needed to adapt the model, generic, TS ( described in NUREG-
1431, Revision 3) addressed in the Federal Register on March 2, 2005,
to the plant-specific TS applicable to Kewaunee Power Station. The
analysis 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 [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 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 [Iodine 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.
The proposed change involves rewording of certain Technical
Specification sections to be consistent with NUREG-1431, Revision 3.
These modifications involve no technical changes to the existing
Technical Specifications. As such, these changes are administrative
in nature and do not affect initiators of analyzed events or assumed
mitigation of accident or transient events.
Therefore, these changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
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.
The proposed change involves rewording of certain Technical
Specification sections to be consistent with NUREG-1431, Revision 3.
The change does not involve a physical alteration of the plant (no
new or different type of equipment will be installed) or changes in
methods governing normal plant operation. The changes will not
impose any new or different requirements or eliminate any existing
requirements from those already approved in the CLIIP.
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
[[Page 7807]]
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.
The proposed change involves rewording of certain Technical
Specification sections to be consistent with NUREG-1431, Revision 3.
The changes are administrative in nature and will not involve any
technical changes. The changes will not reduce a margin of safety
because they have no impact on any safety analysis assumptions. In
addition, since these changes are administrative in nature, no
question of safety is involved.
Therefore, the changes do not involve a significant reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
Acting NRC Branch Chief: T. Kobetz.
Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian
Point Nuclear Generating Unit Nos. 2 and 3 (IP2 & IP3), Westchester
County, New York
Date of amendment request: December 27, 2005
Description of amendment request: The proposed amendment changes
consist of:
Adoption of Technical Specification Task Force (TSTF)-258,
Revision 4; regarding changes to Section 5.0, Administrative Controls .
Adoption of TSTF-308, Revision 1; regarding the
determination of cumulative and projected dose contributions in the
Radioactive Effluents Control Program (RECP).
Revision of IP2 definition for dose equivalent 1-131 based
on NUREG-1431, Revision 3.
Revision of IP2 RECP requirements based on NUREG-1431,
Revision 3.
Revision of IP3 Explosive Gas and Storage Tank
Radioactivity Monitoring Program requirements based on NUREG-1431.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes are administrative in nature and have no
affect on accident scenarios previously evaluated. Affected sections
include Unit Staff requirements, the Radioactive Effluent Controls
Program (RECP), and High Radiation Areas. In addition, a definition
is being revised for IP2. The proposed changes will result in
consistent wording for the affected sections in the Indian Point 2
and Indian Point 3 Technical Specifications, based on wording used
in the latest version of the Standard Technical Specifications. This
will facilitate the implementation of common programs and
administrative procedures for the Indian Point site. The proposed
changes do not affect initiating events for accidents previously
evaluated and do not affect modified plant systems or procedures
used to mitigate the progression or outcome of those accident
scenarios.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not involve the installation of new
plant equipment or modification of existing plant equipment. No
system or component setpoints are being changed and there are no
changes being proposed for the way that the plant is operated. There
are no new accident initiators or equipment failure modes resulting
from the proposed changes. The proposed changes are administrative
in nature and support the implementation of common programs and
administrative procedures for the two nuclear units located at the
same site.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes revise a definition and the description of
certain administrative control programs. There are no changes
proposed to equipment operability requirements, setpoints, or
limiting parameters specified in the plant Technical Specifications.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Branch Chief: Richard J. Laufer.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
and Lancaster Counties, Pennsylvania
Date of application for amendments: April 28, 2005.
Description of amendment request: The proposed changes will modify
Technical Specifications (TSs) 3.3.4.2, ``End of Cycle Recirculation
Pump Trip (EOC-RPT) Instrumentation''; 3.4.1,''Recirculation Loops
Operating''; and 3.7.6, ``Main Turbine Bypass System'' to add a
requirement for the linear heat generation rate (LHGR) limits specified
in the Core Operating Limits Report (COLR).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident. The
consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. The LHGR is a measure of the heat generation rate of a
fuel rod in a fuel assembly at any axial location.
Limits on the LHGR are specified to ensure that fuel design
limits are not exceeded anywhere in the core during normal
operation, including anticipated operational occurrences, and to
ensure that the peak cladding temperature (PCT) during a postulated
design basis Loss-of-Coolant Accident (LOCA) does not exceed the
limits specified in 10 CFR 50.46.
LHGR limits have been established consistent with the NRC-
approved GESTAR methodology to ensure that fuel performance during
normal, transient, and accident conditions is acceptable. The
proposed changes establish a requirement for LHGR limits to be
modified, as specified in the COLR, such that the fuel is protected
for the conditions of an inoperable EOC-RPT [end-of-cycle
recirculation pump trip] instrument function, single recirculation
loop operation, or an inoperable Main Turbine Bypass System and
during any plant transients or
[[Page 7808]]
anticipated operational occurrences that may occur while in these
conditions. Modifying the LHGR limits for the above three (3)
condition[s] does not increase the probability of an evaluated
accident. The proposed change[s] [do] not require any physical plant
modifications, physically affect any plant components, or entail
changes in plant operation. Therefore, no individual precursors of
an accident are affected.
Limits on the LHGR are specified to ensure that fuel design
limits are not exceeded anywhere in the core during normal
operation, including anticipated operational occurrences, and to
ensure that the PCT during a postulated design basis LOCA does not
exceed the limits specified in 10 CFR 50.46. This will ensure that
the fuel design safety criteria (i.e., less than 1% plastic strain
of the fuel cladding and no fuel centerline melting) are met and
that the core remains in a coolable geometry following a postulated
design basis LOCA or any anticipated operational occurrence. Since
the operability of plant systems designed to mitigate any
consequences of accidents has not changed and all fuel design limits
continue to be met, the consequences of an accident previously
evaluated are not expected to increase.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
Creation of the possibility of a new or different kind of
accident would require the creation of one or more new precursors of
that accident. New accident precursors may be created by
modifications of the plant configuration, including changes in
allowable modes of operation. The proposed changes do not involve
any modifications of the plant configuration or allowable modes of
operation. Requiring the LHGR limits to be modified for the
conditions of inoperable EOC-RPT instrument function, single
recirculation loop operation, or an inoperable Main Turbine Bypass
System ensures that fuel design limits are not exceeded anywhere in
the core during normal operation, including anticipated operational
occurrences and that the assumptions of the LOCA analyses are met.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. The proposed change[s] will not adversely affect
operation of plant equipment. The change[s] will not result in a
change to the setpoints at which protective actions are initiated.
LHGR limits for the conditions of an inoperable EOC-RPT instrument
function, single recirculation loop operation, or an inoperable Main
Turbine Bypass System are established to ensure that fuel design
limits are not exceeded anywhere in the core during normal
operation, including anticipated operational occurrences and that
the PCT during a postulated design basis LOCA does not exceed the
limits specified in 10 CFR 50.46. This will ensure that the core
remains in a coolable geometry following a postulated design basis
LOCA. The proposed change will ensure the appropriate level of fuel
protection.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for Licensee: Mr. Brad Fewell, Assistant General Counsel,
Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA
19348.
NRC Branch Chief: Darrell J. Roberts.
FPL Energy Seabrook LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: December 19, 2005.
Description of amendment request: The requested change will delete
those parts of Technical Specification (TS) 6.8.1.2, ``Annual
Reports,'' related to occupational radiation exposures and challenges
to pressurizer relief and safety valves, and TS 6.8.1.5, ``Monthly
Operating Reports.'' 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 June 23, 2004 (69 FR 35067). The licensee affirmed the applicability
of the model NSHC determination in its application dated December 19,
2005.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates the Technical Specifications
(TSs) reporting requirements to provide a monthly operating report
of shutdown experience and operating statistics if the equivalent
data is submitted using an industry electronic database. It also
eliminates the TS reporting requirement for an annual occupational
radiation exposure report, which provides information beyond that
specified in NRC regulations. The proposed change involves no
changes to plant systems or accident analyses. As such, the change
is administrative in nature and does not affect initiators of
analyzed events or assumed mitigation of accidents or transients.
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 involve a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This is an administrative change to reporting requirements of
plant operating information and occupational radiation exposure
data, and has no effect on plant equipment, operating practices or
safety analyses assumptions. For these reasons, the proposed change
does not involve a significant reduction in the margin of safety.
Based upon the reasoning presented above, the requested change does
not involve significance hazards consideration.
Attorney for licensee: M.S. Ross, Florida Power & Light Company,
P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: Darrell J. Roberts.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant (MNGP), Wright County, Minnesota
Date of amendment request: September 15, 2005.
Description of amendment request: The licensee proposed to revise
the current licensing basis by incorporating a full-scope application
of the Alternative Source Term (AST) methodology (see Regulatory Guide
1.183, ``Alternative Radiological Source Terms for Evaluating Design
Basis Accidents of Nuclear Power Reactors,'' July 2000) in the analysis
of radiological consequences for design-basis accidents. Approval of
this amendment by the Nuclear Regulatory Commission (NRC) staff would
result in updating various portions of the MNGP Technical
Specifications to reflect the assumptions and parameters used in the
AST methodology. Also, upon approval of the proposed amendment, the
licensee will make conforming changes to the MNGP Updated Final Safely
Analysis Report.
[[Page 7809]]
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's own analysis is presented below:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
No. The licensee's proposed application of AST methodology to
the licensing basis is analytical in nature (i.e., in Chapter 14 of
the MNGP Updated Final Safety Analysis Report), and does not lead to
nor is it a result of modifications to plant equipment or method of
operation. Since there is no change to plant equipment or method of
operation, there can thus be no change in the probability of
occurrence of an accident, and no change to the accident scenarios
documented in the MNGP licensing basis and previously evaluated by
the NRC staff. Consequently, the actual accident radiological
consequences would not be any different whether or not AST
methodology is used in predicting radiological consequences.
(2) Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
No. The proposed amendment does not introduce new equipment
operating modes, nor does it alter existing system and component
design. Accordingly, the proposed amendment to apply AST methodology
does not introduce new failure modes, nor does it alter the
equipment required for accident mitigation. The postulated accident
scenarios previously evaluated are not changed in any way.
Therefore, the proposed amendment 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
the margin of safety?
No. The proposed amendment would approve the licensee's
application of AST methodology to predict radiological consequences
for various postulated accident scenarios. The AST methodology is an
NRC-approved alternative for this purpose. Other than this change,
which will be reviewed by the NRC staff, the licensee is proposing
no other changes to other analytical models, assumptions,
parameters, or acceptance criteria. Accordingly, the proposed
amendment does not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
its own analysis above, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the proposed amendment involves no significant hazards
consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Acting Branch Chief: T. Kobetz.
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: November 9, 2005.
Description of amendment request: The proposed amendments would
revise Technical Specifications (TS) for the Prairie Island Nuclear
Generating Plant (PINGP) Units 1 and 2, to clarify which TS
Surveillance Requirements (SRs) shall be met for TS systems which
include more components (installed spare components) than are required
to satisfy the TS Limiting Conditions for Operation (LCO).
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 proposes to revise Technical
Specification Surveillance Requirements for event monitoring
instrumentation, containment ventilation isolation instrumentation,
cooling water system, AC sources during plant operations and nuclear
instrumentation during refueling. The affected Surveillance
Requirements may require all possible components in their associated
Technical Specifications to meet the Surveillance Requirements even
though the Technical Specifications Limiting Conditions for
Operation only require some of the possible components to be
operable to satisfy the Limiting Conditions for Operation.
Consistent with industry guidance, the affected Surveillance
Requirements were revised to include some form of ``required'' as a
descriptor of the components which shall meet the Surveillance
Requirements. Minor format and error corrections are also proposed
for some of these Technical Specifications.
The instrumentation and systems which are the subject of the
affected Technical Specifications mitigate accidents or monitor
plant conditions. The instrumentation and systems are not accident
initiators, thus the proposed changes do not involve a significant
increase in the probability of a previously evaluated accident. With
the proposed changes, the Technical Specification Limiting
Conditions for Operation will continue to be met, thus the proposed
changes do not involve a significant increase in the consequences of
a previously evaluated accident. Therefore, 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 proposes to revise Technical
Specification Surveillance Requirements for event monitoring
instrumentation, containment ventilation isolation instrumentation,
cooling water system, AC sources during plant operations and nuclear
instrumentation during refueling. The affected Surveillance
Requirements may require all possible components in their associated
Technical Specifications to meet the Surveillance Requirements even
though the Technical Specifications Limiting Conditions for
Operation only require some of the possible components to be
operable to satisfy the Limiting Conditions for Operation.
Consistent with industry guidance, the affected Surveillance
Requirements were revised to include some form of ``required'' as a
descriptor of the components which shall meet the Surveillance
Requirements. Minor format and error corrections are also proposed
for some of these Technical Specifications.
The proposed Technical Specification changes do not involve a
change in the instrumentation or systems' operation, or the use of
the instrumentation or systems. The Limiting Conditions for
Operation will continue to be met and the instrumentation and
systems will continue to provide their same monitoring or mitigation
function. There are no new failure modes or mechanisms created
through the clarifications of which components must meet the
Surveillance Requirements. There are no new accident precursors
generated by clarifying which components must meet the Surveillance
Requirements. The minor format and error corrections 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 proposes to revise Technical
Specification Surveillance Requirements for event monitoring
instrumentation, containment ventilation isolation instrumentation,
cooling water system, AC sources during plant operations and nuclear
instrumentation during refueling. The affected Surveillance
Requirements may require all possible components in their associated
Technical Specifications to meet the Surveillance Requirements even
though the Technical Specifications Limiting Conditions for
Operation only require some of the possible components to be
operable to satisfy the Limiting Conditions for Operation.
Consistent with industry guidance, the affected Surveillance
Requirements were revised to include some form of ``required'' as a
descriptor of the components which shall meet the Surveillance
Requirements. Minor format and error corrections are also proposed
for some of these Technical Specifications.
The Technical Specification changes proposed in this License
Amendment
[[Page 7810]]
Request are administrative, that is, they do not involve any
substantive changes in plant systems, structures or components and
they do not involve any changes in plant operations. Currently the
affected Technical Specification Limiting Conditions for Operation
do not require all possible components addressed by the Technical
Specifications to be operable. This License Amendment Request
clarifies that the components not required to be operable are not
required to meet the Surveillance Requirements. The Limiting
Conditions for Operation will continue to be met as required by the
Technical Specifications. Minor format and error corrections are
also proposed. Since these changes are administrative, they do not
involve a significant reduction in a margin of safety.
Therefore, based on the considerations given above, 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 Acting Branch Chief: Timothy Kobetz.
Pacific Gas and Electric Company, Docket No. 50-275, Diablo Canyon
Nuclear Power Plant, Unit No. 1, San Luis Obispo County, California
Date of amendment requests: December 16, 2005.
Description of amendment requests: The proposed amendment would
revise Technical Specification 5.6.5, ``Core Operating Limits Report
(COLR),'' by adding WCAP-12945-P-A, Addendum 1-A, Revision 0, ``Method
for Satisfying 10 CFR 50.46 [Section 50.46 of Title 10 of the Code of
Federal Regulations] Reanalysis Requirements for Best Estimate LOCA
[Loss-of-Coolant Accident] Evaluation Models,'' dated December 2004, as
an approved analytical method for determining core operating limits for
Unit 1. Pacific Gas and Electric is performing a plant-specific best-
estimate loss-of-coolant accident analysis for Unit 2 using a
methodology different than the methodology presented in Addendum 1-A to
WCAP-12945-P-A. Therefore, this license amendment applies only to Unit
1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to allow the use of the abbreviated best
estimate loss-of-coolant accident (LOCA) analysis methodology does
not involve a physical alteration of any plant equipment or change
operating practice at Unit 1 of Diablo Canyon Power Plant (DCPP).
Therefore, there will be no increase in the probability of a LOCA.
The consequences of a LOCA are not being increased.
The plant conditions assumed in the analysis are bounded by the
design conditions for all equipment in Unit 1. That is, it is shown
that the emergency core cooling system is designed so that its
calculated cooling performance conforms to the criteria contained in
10 CFR 50.46, paragraph b. No other accident is potentially affected
by this change.
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 would not result in any physical alteration
to any Unit 1 system, and there would not be a change in the method
by which any safety related system performs its function. The
parameters assumed in the analysis are within the design limits of
existing plant equipment.
Therefore, the proposed change does not create the possibility
of a new or different accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
It has been shown that the analytic technique used in the
analysis realistically describes the expected behavior of the DCPP
Unit 1 reactor system during a postulated LOCA. Uncertainties have
been accounted for as required by 10 CFR 50.46. A sufficient number
of LOCAs with different break sizes, different locations, and other
variations in properties have been analyzed to provide assurance
that the most severe postulated LOCAs were analyzed. It has been
shown by the analysis that there is a high level of probability that
all criteria contained in 10 CFR 50.46, paragraph b, are met.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Branch Chief: David Terao.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of amendment request: November 18, 2005.
Description of amendment request: The proposed amendment would
change the SSES 1 and 2 Technical Specifications (TSs) to implement the
Average Power Range Monitor/Rod Block Monitor/Technical Specifications/
Maximum Extended Load Line Limit Analysis (ARTS/MELLLA). Specifically,
the average power range monitor (APRM) flow-biased scram and rod block
trip setpoints would be revised to permit operation in the MELLLA
region. The current flow-biased rod block monitor (RBM) would also be
replaced by a power dependent RBM implemented through the referenced
proposed upgrade to a digital power range neutron monitor system
(PRNMS). The change from the flow-biased RBM to the power-dependent RBM
would also require new trip setpoints. In addition, the flow-biased
APRM scram and rod block trip setdown requirement would be replaced by
more direct power and flow-dependent thermal limits to reduce the need
for APRM gain adjustments, and to allow more direct thermal limits
administration during operation other than rated conditions. Finally,
the proposed amendment would change the methods used to evaluate the
annulus pressurization (AP), mass blowdown, and early release resulting
from the postulated recirculation suction line break (RSLB).
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.
Proposed Change No. 1: The proposed change eliminates the
Average Power Range Monitor (APRM) flow-biased scram and rod block
trip setpoint setdown requirements and substitutes power and flow
dependent adjustments to the Minimum Critical Power Ratio (MCPR) and
Linear Heat Generation Rate (LHGR) thermal limits. Thermal limits
will be determined using NRC approved analytical methods. The
proposed change will have no effect upon any accident initiating
mechanism. The power and flow
[[Page 7811]]
dependent adjustments will ensure that the MCPR safety limit will
not be violated as a result of any Anticipated Operational
Occurrence (AOO), and that the fuel thermal and mechanical design
bases will be maintained. Therefore, the proposed change will not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
Proposed Change No. 2: The proposed change expands the power and
flow operating domain by relaxing the restrictions imposed by the
formulation of the APRM flow-biased scram and rod block trip
setpoints and the replacement of the current flow-biased RBM with a
new power dependent RBM, which will be implemented using a digital
Power Range Neutron Monitoring System (PRNMS). The APRM and RBM are
not involved in the initiation of any accident; and the APRM flow-
biased scram and rod block functions are not credited in any PPL
safety licensing analyses.
The analysis of the instrument line break event resulted in an
insignificant change in the radiological consequences. The change
for the instrument line break was an insignificant increase of 0.1
Rem.
Since the proposed changes will not affect any accident
initiator, or introduce and initial conditions that would result in
NRC approved criteria being exceeded, and since the APRM and RBM
will remain capable of performing their design functions, the
proposed change will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Proposed Change No. 3: The methods used to evaluate Annulus
Pressurization (AP) and mass blowdown and energy releases resulting
from the postulated Recirculation Suction Line Break (RSLB) at the
MELLLA conditions are changed to use more realistic, but still
conservative, methods of analysis to determine an AP mass and energy
release profile for AP loads resulting from the postulated RSLB. The
releases resulting from the RSLB at off-rated conditions have been
demonstrated to be bounded by the current design basis loads. Since
the proposed changes do not affect any accident initiator and since
the RSLB AP releases remain bounded by the current design basis, the
proposed changes do not involve a significant increase in the
probability or radiological consequences of an accident previously
evaluated. Therefore the proposed changes do not involve a
significant increase in the probability or consequences of any
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.
Proposed Change No. 1: The proposed change eliminates the
Average Power Range Monitor (APRM) flow-biased scram and rod block
setpoint setdown requirements and substitutes power and flow
dependent adjustments to the Minimum Critical Power Ratio (MCPR) and
Linear Heat Generation Rate (LHGR) thermal limits. Because the
thermal limits will continue to be met, no analyzed transient event
will escalate into a new or different type of accident due to the
initial starting conditions permitted by the adjusted thermal
limits. Therefore, the proposed change will not create the
possibility of a new or different kind of accident previously
evaluated.
Proposed Change No. 2: The proposed change expands the power and
flow operating domain by relaxing the restrictions imposed by the
formulation of the APRM flow-biased scram and rod block trip
setpoints and the replacement of the current flow-biased RBM with a
new power dependent RBM, which will be implemented using a digital
Power Range Neutron Monitoring System (PRNMS). Changing the
formulation for the APRM flow-biased scram and rod block trip
setpoints and from a flow-biased RBM to a power dependent RBM does
not change their respective functions and manner of operation. The
change does not introduce a sequence of events or introduce a new
failure mode that would create a new or different type of accident.
The APRM flow-biased rod block trip setpoint will continue to block
control rod withdrawal when core power significantly exceeds normal
limits and approaches the scram level. The APRM flow-biased scram
trip setpoint will continue to initiate a scram if the increasing
power/flow condition continue beyond the APRM flow-biased rod block
setpoint. The power dependent RBM will prevent rod withdrawal when
the power dependent RBM rod block setpoint is reached. No new
failure mechanisms, malfunctions, or accident initiators are being
introduced by the proposed changes. In addition, operating within
the expanded power flow map will not require any systems, structures
or components to function differently than previously evaluated and
will not create initial conditions that would result in a new or
different kind of accident from any accident previously evaluated.
Proposed Change No. 3: The methods used to evaluate Annulus
Pressurization (AP) and mass blowdown and energy releases resulting
from the postulated Recirculation Suction Line Break (RSLB) at the
MELLLA conditions are changed to use more realistic, but still
conservative, methods of analysis to determine an AP mass and energy
release profile for AP loads resulting from the postulated RSLB. The
proposed changes to the methods of analysis to determine AP mass and
energy releases resulting from the postulated RSLB do not change the
design function or operation of any plant equipment. No new failure
mechanisms, malfunctions, or accident initiators are being
introduced by the proposed changes. Therefore, the proposed changes
do not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
Proposed Change No. 1: The proposed change eliminates the
Average Power Range Monitor (APRM) flow-biased scram and rod block
setpoint setdown requirements and substitutes power and flow
dependent adjustments to the Minimum Critical Power Ratio (MCPR) and
Linear Heat Generation Rate (LHGR) thermal limits. Replacement of
the APRM setpoint setdown requirement with power and flow dependent
adjustments to the MPR and LHGR thermal limits will ensure that
margins to the fuel cladding Safety Limit are preserved during
operation at other than rated conditions. Thermal limits will be
determined using NRC approved analytical methods. The power and flow
dependent adjustments will ensure that the MPR safety limit will not
be violated as a result of any Anticipated Operational Occurrence
(AOO), and that the fuel thermal and mechanical design bases will be
maintained. The 10 CFR 50.46 acceptance criteria for the performance
of the Emergency Core Cooling System (ECCS) following postulated
Loss-Of-Coolant Accidents (LOCAs) will continue to be met.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety.
Proposed Change No. 2: The proposed change expands the power and
flow operating domain by relaxing the restrictions imposed by the
formulation of the APRM flow-biased scram and rod block trip
setpoints and the replacement of the current flow-biased RBM with a
new power dependent RBM, which will be implemented using a digital
Power Range Neutron Monitoring System (PRNMS). The APRM flow-biased
rod block trip setpoint will continue to block control rod
withdrawal when core power significantly exceeds normal limits and
approaches the scram level. The APRM flow-biased scram trip setpoint
will continue to initiate a scram if the increasing power/flow
condition continues beyond the APRM flow-biased rod block setpoint.
The RBM will continue to prevent rod withdrawal when the power
dependent RBM rod block setpoint is reached. The MPR and LHGR
thermal limits will be developed to ensure that fuel thermal
mechanical design bases shall remain within the licensing limits
during a rod withdrawal error event and to ensure that the MPR
safety limit will not be violated as a result of a rod withdrawal
error event. Operation in the expanded operating domain will not
alter the manner in which safety limits, limiting safety system
settings, or limiting conditions for operation are determined.
Anticipated operational occurrences and postulated accident within
the expanded operating domain will be evaluated using NRC approved
methods. Therefore, the proposed change will not involve a
significant reduction in the margin of safety.
Proposed Change No. 3: The methods used to evaluate Annulus
Pressurization (AP) and mass blowdown and energy releases resulting
from the postulated Recirculation Suction Line Break (RSLB) at the
MELLLA conditions are changed to use more realistic, but still
conservative, methods of analysis to determine an AP mass and energy
release profile for AP loads resulting from the postulated RSLB.
Mass and energy releases for AP loads resulting from the postulated
RSLB remain bounded by the current design basis releases. Therefore,
the proposed change does not involve a significant reduction in a
margin of safety.
[[Page 7812]]
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Branch Chief: Richard J. Laufer.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment requests: November 30, 2005.
Description of amendment requests: The proposed amendment would
revise the Technical Specification (TS) requirements related to steam
generator (SG) tube integrity, based on the NRC-approved Revision 4 to
TS Task Force (TSTF)-449, ``Steam Generator Tube Integrity.''
The NRC staff issued a notice of opportunity for comment in the
Federal Register on March 2, 2005 (70 FR 10298), on possible amendments
adopting TSTF-449, including a model safety evaluation and model no
significant hazards consideration (NSHC) determination, using the
consolidated line item improvement process. The NRC staff subsequently
issued a notice of availability of the models for referencing in
license amendment applications in the Federal Register on May 6, 2005
(70 FR 24126). The licensee affirmed the applicability of the following
NSHC determination in its application dated November 30, 2005.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change requires a[n] 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 MSLB [main steamline 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 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 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[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 any adverse changes to the plant design basis or
postulated accidents resulting from potential tube degradation. The
result of the implementation of the SG Program will be an
enhancement of SG tube performance. Primary to secondary LEAKAGE
that may be experienced during all plant conditions will be
monitored to ensure it remains within current accident analysis
assumptions.
The proposed change does not affect the design of the SGs, their
method of operation, or primary or secondary coolant chemistry
controls. In addition, the proposed change does not impact any other
plant system or component. The change enhances SG inspection
requirements.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
isolate the radioactive fission products in the primary coolant from
the secondary system. In summary, the safety function of an SG is
maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change does not affect tube design or operating environment. The
proposed change is expected to result in an improvement in the tube
integrity by implementing the SG Program to manage SG tube
inspection, assessment, repair, and plugging. The requirements
established by the SG Program are consistent with those in the
applicable design codes and standards and are an improvement over
the requirements in the current TSs.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed change to the
TS.
The NRC staff 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
[[Page 7813]]
amendment requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Branch Chief: David Terao.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of amendment request: December 16, 2005.
Description of amendment request: The proposed amendment would
revise the ACTIONS NOTE for TS 3.7.5, ``Auxiliary Feedwater (AFW)
System,'' based on Industry/Technical Specification Task Force (TSTF)
Standard Technical Specification Change Traveler TSTF-359, Revision 9,
``Increased Flexibility in Mode Restraints.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility or the manner in which
the plant is operated and maintained. The proposed change does not
alter or prevent the ability of structures, systems, and components
(SSCs) from performing their intended function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change does not affect the source term,
containment isolation, or radiological release assumptions used in
evaluating the radiological consequences of an accident previously
evaluated. Further, the proposed change does not increase the types
or amounts of radioactive effluent that may be released offsite, nor
significantly increase individual or cumulative occupational public
radiation exposures. The proposed change is consistent with safety
analysis assumptions and resultant consequences.
Therefore, the proposed change does not increase the probability
or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated?
No. The proposed change does not involve a physical alteration
of the plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. In addition, the change does not impose any new or
different requirements or eliminate any existing requirements. The
change does not alter assumptions made in the safety analysis. The
proposed change is consistent with the safety analysis assumptions
and current plant operating practice.
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?
No. The proposed change does not alter the manner in which
safety limits, limiting safety system settings or limiting
conditions for operation are determined. The safety analysis
acceptance criteria are not impacted by this change. The proposed
change will not result in plant operation in a configuration outside
the design basis.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Branch Chief: Evangelos C. Marinos.
Tennessee Valley Authority (TVA), Docket Nos. 50-327 and 50-328,
Sequoyah Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: December 19, 2005 (TS-05-11).
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) for consistency with the
requirements of 10 CFR 50.55a(f)(4). Title 10 CFR 50.55a(f)(4) provides
reference to the applicable American Society of Mechanical Engineers
(ASME) code for testing pumps and valves that are classified as ASME
Code Class 1, 2, and 3. The proposed change provides consistency with
the 10 CFR 50.55a(f)(4) requirement by replacing the TS reference to
ASME Boiler and Pressure Vessel Code, Section XI, with the ASME Code
for Operation and Maintenance of Nuclear Power Plants (ASME OM Code) as
it applies to the Inservice Test program. This change is based on TSTF-
479, Revision 0, ``Changes to Reflect Revision of 10 CFR 50.55a.''
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.
TVA's proposed change revises TS Surveillance Requirement (SR)
4.0.5 for SQN Units 1 and 2 to conform to the requirements of 10 CFR
50.55a(f) regarding inservice testing of pumps and valves for the
third 10-Year interval. The current TSs reference the ASME Boiler
and Pressure Vessel Code, Section XI, as the requirements for
inservice testing of ASME Code Class 1, 2, and 3 pumps and valves.
The proposed changes would replace current reference to Section XI
of the Boiler and Pressure Vessel Code to the ASME OM Code, which is
consistent with 10 CFR 50.55a(f) and accepted for use by the Nuclear
Regulatory Commission (NRC). The proposed change incorporates
updates to ASME code requirements that result in a net improvement
in the measures for testing pumps and valves.
The proposed change does not involve any hardware changes, nor
does it affect the probability of any event initiators. There will
be no change to normal plant operating parameters, engineered safety
feature actuation setpoints, accident mitigation capabilities, or
accident analysis assumptions or inputs.
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 incorporates ASME code requirements that
result in a net improvement for testing pumps and valves. The
proposed change does not involve a modification to the physical
configuration of the plant (i.e., no new equipment will be
installed) or change in the methods governing normal plant
operation. The proposed change will not impose any new or different
requirements or introduce a new accident initiator, accident
precursor, or malfunction mechanism. Additionally, there is no
change in the types or increases in the amounts of any effluent that
may be released off-site and there is no increase in individual or
cumulative occupational exposure.
Equipment important to safety will continue to operate as
designed. The changes to not result in any event previously deemed
incredible being made credible. The changes do not result in adverse
conditions or result in any increase in the challenges to safety
systems.
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 incorporates revisions to the ASME Code that
result in a net improvement in the measures of testing.
[[Page 7814]]
The safety function of the affected components will be maintained.
There are no new or significant changes to the initial
conditions contributing to accident severity or consequences. The
proposed amendment will not otherwise affect the plant protective
boundaries, will not cause a release of fission products to the
public, nor will it degrade the performance of any other structures,
systems, or components important to safety.
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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: Michael L. Marshall, Jr.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1 Rhea County, Tennessee
Date of amendment request: December 13, 2005 (TS-05-06).
Description of amendment request: The proposed amendment would
change the steam generator (SG) level requirement for Limiting
Condition for Operation (LCO) 3.4.7.b and Surveillance Requirements
(SRs) 3.4.5.2, 3.4.6.3 and 3.4.7.2 from greater than or equal to (>=) 6
percent to >= 32 percent following replacement of the SGs during the
Unit 1 Cycle 7 refueling outage.
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 accidents and transients of interest are those that may
occur in MODE 3, 4 or 5 and that rely upon one or two of the SGs to
be OPERABLE to provide a heat sink for the removal of decay heat
from the reactor vessel. These events include an accidental control
rod withdrawal from subcritical, ejection of a control rod, and
accidental boron dilution. TS [Technical Specification] SRs provide
verification of SG water level which demonstrates that the SG is
OPERABLE and able to act as a heat sink.
The proposed revision to TSs 3.4.5, 3.4.6, and 3.4.7 reflects
the change to the required minimum SG water level necessary to
demonstrate OPERABILITY of the RSGs [Replacement SGs]. Therefore,
since no initiating event mechanisms or OPERABILITY requirements are
being changed, 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.
Operation in MODE 3, 4 or 5 with a SG water level of less than
32% of span is not an initiator of any of the accidents and
transients described in the UFSAR [updated final safety analysis
report]. This situation puts the plant into a LCO [limiting
condition for operation] situation and requires that the plant
initiate actions within a specified timeframe if SG OPERABILITY
cannot be restored within the specified timeframe. The change in the
value of the SG water level reflects the differences between the
OSGs [Old Steam Generators] and the RSGs. The new value will be used
in the same manner as the old one to assess the OPERABILITY of the
SGs.
Therefore, operation in MODE 3, 4 or 5 with a SG water level of
less than 32% of span will not initiate an accident nor create any
new failure mechanisms. The changes to the TSs do not result in any
event previously deemed incredible being made credible. The change
will not result in more adverse conditions and is not expected to
result in any increase in the challenges to safety systems.
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 changes to the affected TSs revise the value of SG
narrow range water level that is needed to demonstrate that
OPERABILITY of the SG to support operation with the RSGs. The change
in the value of the SG water level reflects the differences between
the OSGs and the RSGs. These changes assure that the required
numbers of SGs are OPERABLE with a secondary side narrow range water
level indication high enough to cover the tubes. Therefore, the
acceptance criterion is to provide an indicated level that will
ensure the tubes are covered. Since the same acceptance criteria is
being used for the RSGs as was used for the OSGs, there is no
reduction in the margin of safety.
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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: Michael L. Marshall, Jr.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1 (WBN), Rhea County, Tennessee
Date of amendment request: December 15, 2005 (TS-05-09).
Description of amendment request: The proposed amendment would
revise the Technical Specification Surveillance Requirements to
increase the minimum required average ice basket weight, and thus the
corresponding total weight of the stored ice in the WBN ice condenser.
The changes to the ice basket and total ice weights are due to the
additional energy associated with the Replacement Steam Generators.
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 primary purpose of the ice bed is to provide a large heat
sink to limit peak containment pressure in the event of a release of
energy from a design basis loss-of-coolant [accident] (LOCA) or high
energy line break (HELB) in containment. The LOCA requires the
greatest amount of ice compared to other accident scenarios;
therefore the increase in ice weight is based on the LOCA analysis.
The amount of ice in the bed has no impact on the initiation of an
accident, but rather on the mitigation of the accident.
The containment integrity analysis shows that the proposed
increased ice weight is sufficient to maintain the peak containment
pressure below the containment design pressure, and that the
containment heat removal systems function to rapidly reduce the
containment pressure and temperature in the event of a LOCA.
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 ice condenser serves to limit the peak pressure inside
containment following a LOCA. The revised containment pressure
analysis determined that sufficient ice would be present to maintain
the peak containment pressure below the containment design pressure.
The increased ice weight does not create the possibility of an
accident that is different from any already evaluated in the WBN
Updated Final Safety [Analysis Report]
[[Page 7815]]
(UFSAR). No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of this proposed change.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The containment integrity analysis for increased ice weight
results in a peak containment pressure that is slightly greater than
that in the previous analysis of record, but still less than design
pressure. This increase in peak pressure, along with the ice weight
increase, is due to an increase in RCS [reactor coolant system]
inventory and stored residual heat in the replacement Steam
Generators that will be installed in the Unit 1 Cycle 7 Refueling
Outage.
The revised technical specification ice weight surveillance
limits are based on the ice weight assumed in the containment
integrity analysis, with margins included for sublimation that is
based on actual sublimation data from the first six refueling cycles
at WBN. The analysis further demonstrates that the existing
relationship between ice bed melt-out and containment spray
switchover has been conservatively maintained. With the increased
ice inventory, melt-out of the ice bed following a worst case large
break LOCA has been determined to occur after the switchover of
containment spray to the recirculation mode. Thus, the greater ice
bed mass does not result in a reduction in the margin for operator
action to initiate the switchover.
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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: Michael L. Marshall, Jr.
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: April 1, 2005, as supplemented
September 23, 2005.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) to support the implementation of Oscillation Power
Range Monitor.
Date of issuance: January 26, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days following restart from the February 2006 refueling
outage.
Amendment No.: 171.
Facility Operating License No. NPF-62: The amendment revised the
TSs.
Date of initial notice in Federal Register:April 26, 2005 (70 FR
21452). The supplement dated September 23, 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.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 26, 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 September 16, 2005.
Brief description of amendments: The amendments revised Technical
Specification (TS) 3.1.1, ``Shutdown Margin,'' to modify the
restrictions in Required Action B.1 to allow positive reactivity
additions as long as the shutdown margin requirements in Limiting
Condition for Operations 3.1.1 are maintained. The amendments also
corrected an administrative error regarding an incorrect TS reference
in TS 3.4.17, ``Special Test Exception RCS [reactor coolant system]
Loops--Modes 4 and 5.''
Date of issuance: January 19, 2006.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 277 and 254.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: July 5, 2005 (70 FR
38716).
The September 16, 2005, 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.
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated January 19, 2006.
No significant hazards consideration comments received: No.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: March 17, 2005, as supplemented
by letter dated April 15, 2005.
Brief description of amendment: The amendment revised Technical
Specification
[[Page 7816]]
(TS) 3.4.10, ``RCS [Reactor Coolant System] Pressure and
Temperature (P/T) Limits.'' Specifically, the amendment revised the P/T
curves for the hydrostatic pressure test, non-nuclear heatup and
cooldown, and nuclear (core critical) limits illustrated in TS Figure
3.4.10-1 with six recalculated separate curves for 24 and 32 effective
full power years of reactor operation. In addition, the amendment
revised associated surveillance requirements.
Date of issuance: January 25, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 168.
Facility Operating License No. NPF-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 26, 2005 (70 FR
21453).
The supplement dated April 15, 2005, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the NRC
staff's original proposed no significant hazards determination as
published in the Federal Register on April 26, 2005 (70 FR 21453).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 25, 2006.
No significant hazards consideration comments received: No.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: May 18, 2005, as supplemented by
letter dated August 8, 2005.
Brief description of amendment: The amendment revised the Fermi 2
Technical Specifications to add Actions to limiting condition for
operation [LCO] 3.8.1, ``AC [alternating current] Sources--Operating,''
for one offsite circuit inoperable, for two offsite circuits
inoperable, and for one offsite circuit and one or both emergency
diesel generators in one division inoperable.
Date of issuance: January 31, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 170.
Facility Operating License No. NPF-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 7, 2005 (70 FR
33212).
The supplement dated August 8, 2005, provided additional
information that clarified the application, did not expand the scope of
the application as originally notice, and did not change the NRC
staff's original proposed no significant hazards consideration
determination as published in the Federal Register on June 7, 2005 (70
FR 33212).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 31, 2006.
No significant hazards consideration comments received: No.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: June 29, 2005.
Brief description of amendment: The amendment revised Surveillance
Requirements (SR) 3.6.1.3.11 and 3.6.1.3.12 in TS 3.6.1.3, ``Primary
Containment Isolation Valves (PCIVs).'' Specifically, the proposed
amendment revised the combined secondary containment bypass leakage
rate limit for all bypass leakage paths in SR 3.6.1.3.11 from 0.05 to
0.10 La (the maximum allowable containment leakage rate) and
the combined main steam isolation valve (MSIV) leakage rate limit for
all four main steam lines in SR 3.6.1.3.12 from 150 to 250 standard
cubic feet per hour.
Date of issuance: January 25, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 169.
Facility Operating License No. NPF-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 16, 2005 (70 FR
48203).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 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: January 31, 2005.
Brief description of amendment: The amendment changed Technical
Specifications (TS) 3.8.2.5, ``ELECTRICAL POWER SYSTEMS--Containment
Penetration Conductor Overcurrent Protective Devices.'' The change
relocated the requirements for containment penetration conductor
overcurrent protective devices from the TSs to the licensee's Technical
Requirements Manual (TRM). The Bases for this TS were also relocated to
the TRM.
Date of issuance: January 23, 2006.
Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance.
Amendment No.: 263.
Renewed Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 2, 2005 (70 FR
44401).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 23, 2006.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346,
Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: July 27, 2005.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3/4.10.2, ``Special Test Exceptions--Physics
Tests,'' to increase the allowed time between the flux channel Channel
Functional Tests and the beginning of Mode 2 Physics Tests from 12
hours to 24 hours.
Date of issuance: January 31, 2006.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 271.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications and Surveillance Requirements.
Date of initial notice in Federal Register: September 27, 2005 (70
FR 56502).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 31, 2006.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of application for amendment: August 1, 2005, as supplemented
by letters dated October 11, November 1, November 2, and November 28,
2005.
Brief description of amendment: The amendment conforms the license
to reflect the transfer of Facility Operating License No. DPR-49 to FPL
Energy Duane Arnold, LLC, as approved by order of the Commission dated
December 23, 2005.
Date of issuance: January 27, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 260.
[[Page 7817]]
Facility Operating License No. DPR-49: The amendment revised the
Operating License. Date of initial notice in Federal Register:
September 20, 2005 (70 FR 55175).
The supplemental letters contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 23, 2005.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of application for amendments: August 23, 2004, as
supplemented by letter dated May 20, 2005.
Brief description of amendments: The amendments revised the
Technical Specifications Surveillance Requirements for certain
containment purge valves. The amendments replace requirements for valve
seat replacement every 24 months with a requirement to perform an
Appendix J leakage rate test of the valves at a frequency of at least
once every 30 months.
Date of issuance: January 20, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 248/192.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: January 4, 2005 (70 FR
405).
The supplemental letter contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 20, 2006.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 2nd day of February 2006.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 06-1162 Filed 2-13-06; 8:45 am]
BILLING CODE 7590-01-P