[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