Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 21449-21470 [05-8166]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 70, Number 79 (Tuesday, April 26, 2005)]
[Notices]
[Pages 21449-21470]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 05-8166]


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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 April 1, 2005, through April 14, 2005. The 
last biweekly notice was published on April 12, 2005 (70 FR 19110).

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

[[Page 21450]]

Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    Within 60 days after the date of publication of this notice, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, https://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, HearingDocket@nrc.gov; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois

    Date of amendment request: February 25, 2005.
    Description of amendment request: The proposed change would delete 
Section 2.G of the Clinton's Facility Operating License (FOL), NPF-62, 
which requires AmerGen Energy Company, LLC, to report violations of the 
requirements contained in Section 2.C of this license. The proposed 
change will reduce unnecessary regulatory burden and will allow AmerGen 
to take full advantage of the revisions to Title 10, Code of Federal 
Regulations (10

[[Page 21451]]

CFR), Section 50.72, ``Immediate notification requirements for 
operating nuclear power reactors,'' and 10 CFR 50.73, ``Licensee event 
report system.''
    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?
    The proposed change involves an administrative change only. The 
proposed change does not involve the modification of any plant 
equipment or affect plant operation. The proposed change will have 
no impact on any safety related structures, systems or components. 
The reporting requirement section of the FOL is not required because 
the requirements are either adequately addressed by 10 CFR 50.72 and 
10 CFR 50.73, or other regulatory requirements, or are not required 
based on the nature of the Condition.
    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?
    The proposed change has no impact on the design, function or 
operation of any plant structure, system or component. The proposed 
change is administrative in nature and does not affect plant 
equipment or accident analyses. The reporting requirement section of 
the FOL is not required because the requirements are either 
adequately addressed by 10 CFR 50.72 and 10 CFR 50.73, or other 
regulatory requirements, or are not required based on the nature of 
the Condition.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The proposed change is administrative in nature, does not negate 
any existing requirement, and does not adversely affect existing 
plant safety margins or the reliability of the equipment assumed to 
operate in the safety analysis. As such, there is no change being 
made to safety analysis assumptions, safety limits or safety system 
settings that would adversely affect plant safety as a result of the 
proposed change. Margins of safety are unaffected by deletion of the 
reporting requirement that is adequately addressed elsewhere.
    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 
requested amendment involves no significant hazards consideration.
    Attorney for licensee: Mr. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Gene Y. Suh.

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois

    Date of amendment request: March 25, 2005.
    Description of amendment request: The proposed change would revise 
Technical Specification Surveillance Requirement (SR) 3.6.1.3.8 to add 
a note excluding leakage through primary containment penetrations 1MC-
101 and 1MC-102 from the secondary containment bypass leakage total 
specified in the SR.
    Implementation of this proposed change will provide operational 
flexibility by allowing Clinton Power Station (CPS) to utilize the 
additional margin in the regulatory dose limit analysis that supports 
the implementation of the alternative source term.
    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 amendment adds a note excluding the leakage through 
the primary containment purge lines from the secondary containment 
bypass leakage based on separate analysis of these paths using the 
assumptions in the alternative source term (AST) revision to the 
loss of coolant accident (LOCA) analysis.
    The proposed change does not require modification to the 
facility. The proposed change in secondary containment bypass 
leakage does not affect the operation of any facility equipment, the 
interface between facility systems, or the reliability of any 
equipment. In addition, secondary containment bypass leakage does 
not constitute an initiator of any previously evaluated accidents. 
Therefore, the proposed amendment does not involve a significant 
increase in the probability of an accident previously evaluated.
    The radiological consequences of the LOCA analysis using the 
primary containment purge line leakage as separate from the 
secondary containment bypass leakage, has been evaluated as part of 
the application of AST assumptions. The results conclude that the 
radiological consequences remain within applicable regulatory 
limits.
    Therefore, the proposed change does 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.
    The proposed change does not affect the design, functional 
performance or operation of the facility. No new equipment is being 
introduced and installed equipment is not being operated in a new or 
different manner. Similarly, the proposed change does not affect the 
design or operation of any structures, systems or components 
involved in the mitigation of any accidents, nor does it affect the 
design or operation of any component in the facility such that new 
equipment failure modes are created. There are no set points at 
which protective or mitigative actions are initiated that are 
affected by this proposed action. No change is being made to 
procedures relied upon to respond to an off-normal event.
    As such 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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Margins of safety are established in the design of components, 
the configuration of components to meet certain performance 
parameters, and in the establishment of set points to initiate 
alarms or actions. The proposed change adds a note excluding the 
leakage through the primary containment purge lines from the 
secondary containment bypass leakage based on separate analysis of 
these paths using the assumptions in the AST revision to the LOCA 
analysis. There is no change in the design of the affected systems, 
no alteration of the set points at which alarms or actions are 
initiated, and no change in plant configuration from original 
design.
    The margin of safety is considered to be that provided by 
meeting the applicable regulatory limits. The AST analysis indicates 
that the doses following a LOCA remain within the regulatory limits, 
and therefore, there is not a significant reduction in a margin of 
safety. The AST analysis confirms the change continues to ensure 
that the doses at the exclusion area and low population zone 
boundaries, as well as the control room, are within the 
corresponding regulatory limits.
    Therefore, operation of CPS in accordance with the proposed 
change will 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 
requested amendment involves no significant hazards consideration.
    Attorney for licensee: Mr. Thomas S. O'Neill, Associate General 
Counsel,

[[Page 21452]]

Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Section Chief: Gene Y. Suh.

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois

    Date of amendment request: April 1, 2005.
    Description of amendment request: The proposed changes would 
incorporate into the Technical Specifications (TSs) the Oscillation 
Power Range Monitor (OPRM) instrumentation that will be declared 
operable within 30 days after completion of the February 2006 refueling 
outage. The proposed changes would add TS Section 3.3.1.3, 
``Oscillation Power Range Monitor (OPRM) Instrumentation,'' and would 
revise TS Sections 3.4.1, ``Recirculation Loops Operating,'' and 5.6.5, 
``Core Operating Limits Report (COLR).'' In addition, the changes would 
insert a new TS section for the OPRM instrumentation, delete the 
current thermal-hydraulic instability administrative requirements, and 
add the appropriate references for the OPRM trip set points and 
methodology. Clinton Power Station (CPS) will activate the automatic 
reactor protection system (i.e., scram) outputs of the OPRM 
instrumentation upon implementation of these proposed TS changes.
    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 specify limiting conditions for operation, 
required actions and surveillance requirements for the OPRM system, 
and allows operation in regions of the power to flow map currently 
restricted by the requirements of the Interim Corrective Actions 
(ICAs) and certain limiting conditions of operation of TS Section 
3.4.1, ``Recirculation Loops Operating.'' The restrictions of the 
ICAs and TS Section 3.4.1 were imposed to ensure adequate capability 
to detect and suppress conditions consistent with the onset of 
thermal-hydraulic oscillations that may develop into a thermal-
hydraulic instability event. A thermal-hydraulic instability event 
has the potential to challenge the Minimum Critical Power Ratio 
(MCPR) safety limit. The OPRM system can automatically detect and 
suppress conditions necessary for thermal-hydraulic instability. 
With the activation of the OPRM system, the restrictions of the ICAs 
and TS Section 3.4.1 will no longer be required.
    This proposed change has no impact on any of the existing 
neutron monitoring functions. When the OPRM is operable with 
operating limits as specified in the Core Operating Limits Report 
(COLR), the OPRM can automatically detect the imminent onset of 
local power oscillations and generate a trip signal. Actuation of a 
Reactor Protection System (RPS) trip (i.e., scram) will suppress 
conditions necessary for thermal-hydraulic instability and decrease 
the probability of a thermal-hydraulic instability event. In the 
event the trip capability of the OPRM is not maintained, the 
proposed changes limit the period of time before an alternate method 
to detect and suppress thermal-hydraulic oscillations is required. 
CPS intends to utilize the ICAs as the alternative method for 
ensuring thermal-hydraulic oscillations do not occur. Since the 
duration of this period of time is limited, the increase in the 
probability of a thermal-hydraulic instability event is not 
significant.
    Activation of the OPRM scram function will replace the current 
methods that require operators to insert an immediate manual reactor 
scram in certain reactor operating regions where thermal hydraulic 
instabilities could potentially occur. While these regions will 
continue to be avoided during normal operation, certain transients, 
such as a reduction in reactor recirculation flow, could place the 
reactor in these regions. During these transient conditions, with 
the OPRM instrumentation scram function activated; an immediate 
manual scram will no longer be required. This may potentially cause 
a marginal increase in the probability of occurrence of an 
instability event. This potential increase in probability is 
acceptable because the OPRM function will automatically detect the 
instability condition and initiate a reactor scram before the 
Minimum Critical Power Ratio (MCPR) Safety Limit is reached. 
Consequences of the potential instability event are reduced because 
of the more reliable automatic detection and suppression of an 
instability event, and the elimination of dependence on the manual 
operator actions. Operators monitor for indications of thermal 
hydraulic instability when the reactor is operating in regions of 
potential instability as a backup to the OPRM instrumentation.
    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 replace procedural actions that were 
established to avoid operating conditions where reactor 
instabilities might occur with an NRC approved automatic detect and 
suppress function (i.e., OPRM).
    Potential failures in the OPRM trip function could result in 
either failure to take the required mitigating action or an 
unintended reactor scram. These are the same potential effects of 
failure of the operator to take the correct appropriate action under 
the current procedural actions. The effects of failure of the OPRM 
equipment are limited to reduced or failed mitigation, but such 
failure cannot cause an instability event or other type of accident.
    The OPRM system uses input signals shared with the Average Power 
Range Monitor (APRM) system and rod block functions to monitor core 
conditions and generate a Reactor Protection System (RPS) trip when 
required. Quality requirements for software design, testing, 
implementation and module self-testing of the OPRM system provide 
assurance that no new equipment malfunctions due to software errors 
are created. The design of the OPRM system also ensures that neither 
operation nor malfunction of the OPRM system will adversely impact 
the operation of the other systems and no accident or equipment 
malfunction of these other systems could cause the OPRM system to 
malfunction or cause a different kind of accident. No new failure 
modes of either the new OPRM equipment or of the existing APRM 
equipment have been introduced.
    Operation in regions currently restricted by the ICAs and TS 
Section 3.4.1 is within the nominal operating domain and ranges of 
plant systems and components for which postulated equipment and 
accidents have been evaluated. Therefore, operation within these 
regions does not create the possibility of a new or different kind 
of accident from any previously evaluated.
    These proposed changes which specify limiting conditions for 
operations, required actions and surveillance requirements of the 
OPRM system and allow operation in certain regions of the power-to-
flow map do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The OPRM system monitors small groups of Local Power Range 
Monitor (LPRM) signals for indication of local variations of core 
power consistent with thermal-hydraulic oscillations and generates 
an RPS trip when conditions consistent with the onset of 
oscillations are detected. An unmitigated thermal-hydraulic 
instability event has the potential to result in a challenge to the 
MCPR safety limit. The OPRM system provides the capability to 
automatically detect and suppress conditions that might result in a 
thermal-hydraulic instability event and thereby maintains the margin 
of safety by providing automatic protection for the MCPR safety 
limit while reducing the burden on the control room operators 
significantly. The OPRM trip provides a trip output of the same type 
as currently used for the APRM. Its failure modes and types are 
similar to those for the present APRM output. Since the MCPR Safety 
Limit will not be exceeded as a result of an instability event 
following implementation of the OPRM trip function, it is concluded 
that the proposed change does not reduce the margin of safety.
    Operation in regions currently restricted by the requirements of 
the ICAs and TS Section 3.4.1 is within the nominal operating domain 
assumed for identifying the range of initial

[[Page 21453]]

conditions considered in the analysis of anticipated operational 
occurrences and postulated accidents. Therefore, operation in these 
regions does not involve a significant reduction in the margin of 
safety.
    The proposed changes, which specify limiting conditions for 
operations, required actions and surveillance requirements of the 
OPRIVI system and allow operation in certain regions of the power to 
flow map, 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 
requested amendment involves no significant hazards consideration.
    Attorney for licensee: Mr. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Gene Y. Suh.

AmerGen Energy Company, LLC, Docket No. 50-219, Oyster Creek Nuclear 
Generating Station, Ocean County, New Jersey

    Date of amendment request: February 25, 2005.
    Description of amendment request: The proposed change would delete 
Section 2.E of the Oyster Creek's Facility Operating License (FOL), 
DPR-16, which requires AmerGen Energy Company, LLC, to report 
violations of the requirements contained in Section 2.C of this 
license. The proposed change will reduce unnecessary regulatory burden 
and will allow AmerGen to take full advantage of the revisions to Title 
10, Code of Federal Regulations (10 CFR), Section 50.72, ``Immediate 
notification requirements for operating nuclear power reactors,'' and 
10 CFR 50.73, ``Licensee event report system.''
    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 involves an administrative change only. The 
proposed change does not involve the modification of any plant 
equipment or affect plant operation. The proposed change will have 
no impact on any safety related structures, systems or components. 
The reporting requirement section of the FOL is not required because 
the requirements are either adequately addressed by 10 CFR 50.72 and 
10 CFR 50.73, or other regulatory requirements, or are not required 
based on the nature of the Condition.
    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 has no impact on the design, function or 
operation of any plant structure, system or component. The proposed 
change is administrative in nature and does not affect plant 
equipment or accident analyses. The reporting requirement section of 
the FOL is not required because the requirements are either 
adequately addressed by 10 CFR 50.72 and 10 CFR 50.73, or other 
regulatory requirements, or are not required based on the nature of 
the Condition.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change is administrative in nature, does not negate 
any existing requirement, and does not adversely affect existing 
plant safety margins or the reliability of the equipment assumed to 
operate in the safety analysis. As such, there is no change being 
made to safety analysis assumptions, safety limits or safety system 
settings that would adversely affect plant safety as a result of the 
proposed change. Margins of safety are unaffected by deletion of the 
reporting requirement that is adequately addressed elsewhere.
    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 
requested amendment involves no significant hazards consideration.
    Attorney for licensee: Mr. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Richard J. Laufer.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: March 17, 2005.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3.4.10, ``Reactor Coolant System (RCS) 
Pressure and Temperature (P/T) Limits,'' to replace the combination 
figure with separate P/T limit figures for each one of the three 
categories of operation: hydrostatic pressure test [Curve A], non-
nuclear heatup and cooldown [Curve B], and nuclear (core critical) 
operation [Curve C]. The new curves also provide composite limits for 
all reactor pressure vessel (RPV) regions including core beltline 
region. RPV bottom head individual limit curves are superimposed on 
Curves A and B. In addition, two sets of curves are calculated; one for 
32 effective full power years (EFPY) which represents the end of the 
current 40-year plant license and the other one is for 24 EFPY which 
has been selected as an intermediate point between the current EFPY and 
32 EFPY.
    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. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The revised P/T curves are based on the 1998 Edition of the 
American Society of Mechanical Engineers (ASME) Boiler and Pressure 
Vessel (B&PV) Code, Section XI, including the 2000 Addenda. This 
edition of the Code has been approved for use in both 10 CFR 50.55a 
and Regulatory Guide (RG) 1.147. The revised curves are also based 
on updated fluence calculations performed utilizing NRC-approved 
methodology consistent with RG 1.190 for calculating Reactor 
Pressure Vessel (RPV) neutron fluence. Revised fluence calculations 
are applicable for 24 and for 32 Effective Full Power Years (EFPY). 
The 32 EFPY represents a conservative exposure level at the end of 
the current 40-year plant operating license. The proposed change 
incorporates adjustment of the reference temperature for all 
beltline material to account for irradiation effects and provide a 
comparable level of protection as previously evaluated and approved. 
The adjusted reference temperature calculations were performed in 
accordance with the requirements of 10 CFR 50 Appendix G using the 
guidance contained in RG 1.99, Revision 2, to provide operating 
limits for up to 32 EFPY.
    There are no changes being made to the RCS pressure boundary or 
to RCS material, design or construction standards. The proposed P/T 
curves define limits that continue to ensure the prevention of 
nonductile failure of the RCS pressure boundary. The revision of the 
P/T curves does not alter any assumptions previously made in the 
radiological consequence evaluations since the integrity of the RCS 
pressure boundary is unaffected. Therefore, the proposed changes 
will not significantly increase the probability or consequences of 
an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.

[[Page 21454]]

    The revised P/T curves are based on a later edition and addenda 
of the ASME Code that incorporates current industry standards for 
the curves. The revised curves are also based on an RPV fluence that 
has been recalculated in accordance with the methodology of RG 
1.190. The proposed change does not involve a modification to plant 
structures, systems or components. There is no effect on the 
function of any plant system, and no newly introduced system 
interactions. The proposed change does not create new failure modes 
or cause any systems, structures or components to be operated beyond 
their design bases. Therefore, the proposed change does not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The proposed P/T curves define the limits of operation to 
prevent nonductile failure of the RPV upper vessel, bottom head and 
beltline region. The new curves conform to the guidance contained in 
RG 1. 190, ``Calculational and Dosimetry Methods for Determining 
Pressure Vessel Neutron Fluence,'' and RG 1.99, Revision 2, 
``Radiation Embrittlement of Reactor Vessel Materials,'' and 
maintain the safety margins specified in 10 CFR 50 Appendix G. 
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: David G. Pettinari, Legal Department, 688 
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279
    NRC Section Chief: L. Raghavan.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: March 17, 2005. This amendment request 
supercedes, in its entirety, a previous application dated March 19, 
2004, published in the Federal Register on June 22, 2004 (69 FR 34698).
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.3.6.1, ``Primary Containment 
Isolation Instrumentation,'' to correct a formatting error introduced 
during conversion to Improved Technical Specifications (ITS) by 
replacing ``1 per room'' with ``2'' for the required channels per trip 
system for the reactor water cleanup (RWCU) area ventilation 
differential temperature--high primary containment isolation 
instrumentation.
    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 proposed change restores the number of Required Channels Per 
Trip System of the RWCU Area Ventilation Differential Temperature--
High isolation, Function 5.c of Table 3.3.6.1-1 of TS 3.3.6.1, 
Primary Containment Isolation Instrumentation, to its pre-ITS value 
and adds a note to Table 3.3.6.1-1 of TS 3.3.6.1, Primary 
Containment Isolation Instrumentation, that ensures, during 
surveillance testing and normal operation, there will always be at 
least one instrument monitoring for a small leak in all RWCU 
locations. No changes in operating practices or physical plant 
equipment are created as a result of 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 type of accident from any accident previously evaluated?
    Response: No.
    The proposed change restores the number of Required Channels Per 
Trip System of the RWCU Area Ventilation Differential Temperature--
High isolation, Function 5.c of Table 3.3.6.1-1 of TS 3.3.6.1, 
Primary Containment Isolation Instrumentation, to its pre-ITS value 
and adds a note to Table 3.3.6.1-1 of TS 3.3.6.1, Primary 
Containment Isolation Instrumentation, that ensures, during 
surveillance testing and normal operation, there will always be at 
least one instrument monitoring for a small leak in all RWCU 
locations. No physical change in plant equipment will result from 
this proposed change. Therefore, the proposed change does not create 
the possibility of a new or different type 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 restores the number of Required Channels Per 
Trip System of the RWCU Area Ventilation Differential Temperature--
High isolation, Function 5.c of Table 3.3.6.1-1 of TS 3.3.6.1, 
Primary Containment Isolation Instrumentation, to its pre-ITS value 
and adds a note to Table 3.3.6.1-1 of TS 3.3.6.1, Primary 
Containment Isolation Instrumentation, that ensures, during 
surveillance testing and normal operation, there will always be at 
least one instrument monitoring for a small leak in all RWCU 
locations. 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: David G. Pettinari, Legal Department, 688 
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279
    NRC Section Chief: L. Raghavan.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of amendment request: November 16, 2004.
    Description of amendment request: The amendments would revise 
Technical Specifications (TS) 3.5.2, ``Emergency Core Cooling System,'' 
TS 3.6.6, ``Containment Spray System,'' TS 3.6.17, ``Containment Valve 
Injection Water System,'' TS 3.7.5, ``Auxiliary Feedwater System,'' TS 
3.7.7, ``Component Cooling Water System,'' TS 3.7.8, ``Nuclear Service 
Water System (NSWS),'' TS 3.7.10, ``Control Room Area Ventilation 
System'' TS 3.7.12, ``Auxiliary Building Filtered Ventilation Exhaust 
System,'' and TS 3.8.1, ``AC Sources-Operating'' for Catawba, Units 1 
and 2. The revisions would allow for the ``A'' and ``B'' NSWS headers 
to be take out of service for up to 14 days each for system upgrades.
    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 operation of the facility in accordance with the 
proposed amendment involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    Response: No.
    The pipe repair project for the [nuclear service water system] 
NSWS and proposed [technical specifications] TS changes have been 
evaluated to assess their impact on normal operation of the systems 
affected and to ensure that the design basis safety functions are 
preserved. During the pipe repair the other NSWS train will be 
operable and no major maintenance or testing will be done on the 
operable train. The operable train will be protected to help ensure 
it would be available if called upon.
    This pipe repair project will enhance the long term structural 
integrity in the NSWS system. This will ensure that the NSWS headers 
maintain their integrity to ensure its ability to comply with design 
basis requirements and increase the overall reliability for many 
years.
    The increased NSWS train unavailability as a result of the 
implementation of this

[[Page 21455]]

amendment does involve a one time increase in the probability or 
consequences of an accident previously evaluated during the time 
frame the NSWS headers are out of service for pipe repair. 
Considering this small time frame for the NSWS train outages with 
the increased reliability and the decrease in unavailability of the 
NSWS system in the future because of this project, the overall 
probability or consequences of an accident previously evaluated will 
decrease.
    Therefore, because this is a temporary and not a permanent 
change, the time averaged risk increase is acceptable. The increase 
in the overall reliability of the NSWS along with the decreased 
unavailability in the future because of the pipe repair project will 
result in an overall increase in the safety of both Catawba units. 
Therefore, the consequences of an accident previously evaluated 
remains unaffected and there will be minimal impact on any accident 
consequences.
    2. Does operation of the facility in accordance with the 
proposed amendment create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    Response: No.
    Implementation of this amendment would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. The proposed temporary TS changes do not 
affect the basic operation of the [emergency core cooling system] 
ECCS, [containment spray system] CSS, [containment valve injection 
water system] CVIWS, NSWS, [auxiliary feedwater] AFW, [component 
cooling water] CCW, [control room area ventilation system] [sic] 
CRAVS, [auxiliary building filtered ventilation exhaust system] 
ABFVES, or [emergency diesel generator] EDG systems. The only change 
is increasing the required action time frame from 72 hours (ECCS, 
CSS, NSWS, AFW, CCW, and EDG) or 168 hours (CVIWS, CRAVS and ABFVES) 
to 336 hours. The train not undergoing maintenance will be operable 
and capable of meeting its design requirements. Therefore, only the 
redundancy of the above systems is affected by the extension of the 
required action to 336 hours. During the project, contingency 
measures will be in place to provide additional assurance that the 
affected systems will be able to complete their design functions.
    No new accident causal mechanisms are created as a result of NRC 
approval of this amendment request. No changes are being made to the 
plant, which will introduce any new accident causal mechanisms.
    3. Does operation of the facility in accordance with the 
proposed amendment involve a significant reduction in the margin of 
safety?
    Response: No.
    Implementation of this amendment would not involve a significant 
reduction in a margin of safety. Margin of safety is related to the 
confidence in the ability of the fission product barriers to perform 
their design functions during and following an accident situation. 
These barriers include the fuel cladding, the reactor coolant 
system, and the containment system. The performance of these fission 
product barriers will not be impacted by implementation of this 
proposed temporary TS amendment. During the NSWS train outages, the 
affected systems will still be capable of performing their required 
functions and contingency measures will be in place to provide 
additional assurance that the affected systems will be maintained in 
a condition to be able to complete their design functions. No safety 
margins will be impacted.
    The probabilistic risk analysis conducted for this proposed 
amendment demonstrated that the [core damage probability] CDP 
associated with the outage extension is judged to be acceptable for 
a one-time or rare evolution. Therefore, there is not 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: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: March 8, 2005.
    Description of amendment request: The proposed amendment would 
enable the licensee to make changes to the Updated Safety Analysis 
Report (USAR) to reflect the use of the non-single-failure-proof Fuel 
Building Cask Handling Crane (FBCHC) for dry spent fuel cask component 
lifting and handling operations.
    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. Will operation of the facility in accordance with this 
proposed amendment involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment introduces no new mode of plant 
operations and does not affect Structures, Systems, and Components 
(SSCs) associated with power production, accident mitigation, or 
safe plant shutdown. The SSCs affected by this proposed amendment 
are the Fuel Building Cask Handling Crane (FBCHC), the spent fuel 
storage canister, the spent fuel transfer cask, and the spent fuel 
inside the storage canister. A hypothetical 30 ft. drop of a loaded 
spent fuel shipping cask from the FBCHC is part of the River Bend 
Station (RBS) current licensing basis. With the proposed spent fuel 
transfer cask design and procedural changes implemented, the FCHC 
will be used to lift and handle a fuel-loaded spent fuel transfer 
cask of the same maximum weight and approximately the same 
dimensions as previously evaluated in the RBS USAR. The proposed 
amendment involves the use of redundant crane rigging during most 
lateral moves with a loaded spent fuel transfer cask, which provides 
temporary single-failure proof design features to provide protection 
against an uncontrolled lowering of the load or load drop. In those 
cases where the spent fuel transfer cask is not supported with 
redundant rigging, certain hypothetical, non-mechanistic load drops 
have been postulated and evaluated, with due consideration of the 
use of impact limiters in some locations.
    With this amendment, the probability of a loaded spent fuel 
transfer cask drop is actually less likely than previously evaluated 
because the capacity of the spent fuel multi-purpose canister [MPC] 
(68 fuel assemblies) is larger than the capacity of the shipping 
cask described in the current licensing basis (18 fuel assemblies), 
which means that fewer casks will be required to be loaded, lifted, 
and handled for a given population of spent fuel assemblies. The 
consequences of the hypothetical spent fuel transfer cask load drops 
on plant SSCs are bounded by those previously evaluated for a 
shipping cask. That is, there is no significant damage to the Fuel 
Building structure or any SSCs used for safe plant shutdown. New 
analyses of hypothetical drops of a loaded transfer cask or canister 
confirm that there is no release of radioactive material from the 
storage canister and no unacceptable damage to the fuel, MPC, or 
transfer cask.
    The hypothetical drop of a spent fuel canister lid into an open, 
fuel-filled canister in the spent fuel pool during fuel loading has 
also been evaluated. Again, this hypothetical accident is no more 
likely to occur than previously considered due to the higher 
capacity of the spent fuel transfer cask over the spent fuel 
shipping cask (i.e., fewer casks will need to be loaded for a given 
number of fuel assemblies). The radiological consequences of this 
event due to the potential damage of spent fuel assemblies in the 
canister onto which the lid could be dropped have been evaluated. 
While more total fuel assemblies could potentially be damaged from a 
spent fuel canister lid drop compared to that assumed for the fuel 
handling accident described in the RBS current licensing basis, the 
significantly longer decay time of the spent fuel assemblies in the 
canister results in a much smaller source term, such that the 
existing fuel handling accident described in USAR Section 15.7.4 
provides a bounding evaluation for the radiological consequences MPC 
lid drop. There is no rearrangement of the fuel or deformation of 
the fuel basket in the canister such that a critical geometry is 
created as a result of an MPC lid drop.
    The likelihood of a spent fuel canister lid drop due to the 
failure of a crane component due to overload is very unlikely 
because the rated load of the crane (250,000 lbs) is

[[Page 21456]]

approximately 16 times the weight of components lifted to install 
the canister lid.
    2. Will operation of the facility in accordance with this 
proposed amendment create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment introduces no new mode of plant 
operations and does not affect SSCs associated with power 
production, accident mitigation, or safe plant shutdown. The SSCs 
affected by this proposed amendment are the non-single-failure-proof 
FBCHC, the spent fuel canister, the spent fuel transfer cask, and 
the spent fuel inside the canister. The design function of the FBCHC 
is not changed. The proposed amendment does not create the 
possibility of a new or different kind of accident due to credible 
new failure mechanisms, malfunctions, or accident initiators. The 
proposed amendment creates a new initiator of two accidents 
previously evaluated and caused by the non-mechanistic single 
failure of a component in the FBCHC load path.
    The current licensing basis accidents for which new initiators 
are created by this amendment are the spent fuel shipping cask drop 
and the fuel handling accident. The RBS current licensing basis 
includes evaluations of the consequences of a spent fuel shipping 
cask drop and the consequences of the drop of a spent fuel assembly 
into the reactor core shortly after shutdown and reactor head 
removal. The new initiators include the drop of a spent fuel 
transfer cask of the same maximum weight and approximately the same 
dimensions as the shipping cask, and the drop of a spent fuel 
canister lid into an open, fuel filled canister in the spent fuel 
pool. Both of these new initiators create hypothetical accidents 
that are comparable in consequences to those previously evaluated. 
For the drop of a spent fuel transfer cask, the consequences are 
bounded by the current licensing basis analysis of the spent fuel 
shipping cask drop. That is, there is no significant damage to the 
Fuel Building structure or any SSCs used for safe plant shutdown, 
and there is no release of radioactive material. New analyses of the 
drop of a loaded transfer cask confirm that there is no release of 
radioactive material from the storage canister and no unacceptable 
damage to the fuel, MPC, or transfer cask.
    For the drop of the spent fuel canister lid, the significantly 
longer decay time of the spent fuel assemblies in the canister 
compared to a spent fuel assembly in a recently shutdown reactor 
results in doses to the public that are less than the previously 
analyzed fuel handling accident. There is no rearrangement of the 
fuel in the canister such that a critical geometry is created as a 
result of an MPC lid drop.
    3. Will operation of the facility in accordance with this 
proposed amendment involve a significant reduction in a margin of 
safety?
    Response: No.
    The proposed amendment introduces no new mode of plant 
operations and does not affect SSCs associated with power 
production, accident mitigation, or safe plant shutdown. The SSCs 
affected by this proposed amendment are the non-single-failure-proof 
FBCHC, the spent fuel storage canister, the spent fuel transfer 
cask, and the spent fuel inside the canister. Therefore, this 
amendment does not affect the reactor or fuel during power 
operations, the reactor coolant pressure boundary, or primary or 
secondary containment. All activities associated with this amendment 
occur in the Fuel Building or in the adjacent outdoor truck bay 
area. The design function of the FBCHC is not changed. The proposed 
changes to plant operating procedures needed to implement dry spent 
fuel storage at RBS do not exceed or alter a design basis or safety 
limit associated with plant operation, accident mitigation, or safe 
shutdown. The FBCHC is used to lift and handle the spent fuel 
canister lid over spent fuel in the canister while in the spent fuel 
pool, and to lift and handle the spent fuel transfer cask, both when 
it is empty and after it is loaded with spent fuel in the spent fuel 
pool.
    This proposed amendment results in a net safety benefit because 
a larger capacity cask is being used to move spent fuel out of the 
spent fuel pool that was previously evaluated (68 fuel assemblies 
versus 18 fuel assemblies), while maintaining the same maximum 
analyzed cask weight described in the USAR. This yields fewer casks 
to be loaded, fewer heavy load lifts, and, as a result, fewer 
opportunities for events such as load drops. Because the maximum 
weight of the loaded spent fuel transfer cask is the same as that 
assumed for the shipping cask and for which the FBCHC was designed, 
all design safety margins for use of the FBCHC remain unchanged. The 
rated capacity of the FBCHC is approximately 16 times that of 
components lifted to place the spent fuel canister lid, yielding 
significant safety margins for that particular lift.
    Based on the above review, it is concluded that: (1) the 
proposed amendment does not constitute a significant hazards 
consideration as defined by 10 CFR 50.92; and (2) there is 
reasonable assurance that the health and safety of the public will 
not be endangered by the proposed amendment; and (3) this action 
will not result in a condition which significantly alters the impact 
of the station on the environment as described in the NRC Final 
Environmental Impact Statement.

    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: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Allen G. Howe.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

    Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 
and 2, Ogle County, Illinois
    Docket No. 50-237, Dresden Nuclear Power Station, Unit 2, Grundy 
County, Illinois
    Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 
2, LaSalle County, Illinois
    Date of amendment request: February 25, 2005.
    Description of amendment request: The proposed change would delete 
the applicable sections of the Facility Operating Licenses (FOLs); NPF-
72, NPF-77, NPF-37, NPF-66, DPR-19, NPF-11, and NPF-18, respectively; 
which require Exelon Generation Company, LLC, to report violations of 
the requirements contained in Section 2.C of the Braidwood Station, 
Units 1 and 2, and Byron Station, Units 1 and 2 FOLs; Section 2.C of 
the Dresden Nuclear Power Station, Unit 2, renewed FOL; and Sections 
2.C and 2.E of the LaSalle County Station, Units 1 and 2, FOLs. The 
proposed change will reduce unnecessary regulatory burden and will 
allow Exelon to take full advantage of the revisions to Title 10, Code 
of Federal Regulations (10 CFR), Section 50.72, ``Immediate 
notification requirements for operating nuclear power reactors,'' and 
10 CFR 50.73, ``Licensee event report system.''
    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 involves an administrative change only. The 
proposed change does not involve the modification of any plant 
equipment or affect plant operation. The proposed change will have 
no impact on any safety related structures, systems or components. 
The reporting requirement section of the FOL is not required because 
the requirements are either adequately addressed by 10 CFR 50.72 and 
10 CFR 50.73, or other regulatory requirements, or are not required 
based on the nature of the Condition.
    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

[[Page 21457]]

accident from any accident previously evaluated?
    Response: No.
    The proposed change has no impact on the design, function or 
operation of any plant structure, system or component. The proposed 
change is administrative in nature and does not affect plant 
equipment or accident analyses. The reporting requirement section of 
the FOL is not required because the requirements are either 
adequately addressed by 10 CFR 50.72 and 10 CFR 50.73, or other 
regulatory requirements, or are not required based on the nature of 
the Condition.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change is administrative in nature, does not negate 
any existing requirement, and does not adversely affect existing 
plant safety margins or the reliability of the equipment assumed to 
operate in the safety analysis. As such, there is no change being 
made to safety analysis assumptions, safety limits or safety system 
settings that would adversely affect plant safety as a result of the 
proposed change. Margins of safety are unaffected by deletion of the 
reporting requirement that is adequately addressed elsewhere.
    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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Gene Y. Suh.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Unit Nos. 1 and 2, Montgomery County, Pennsylvania

    Date of amendment request: February 25, 2005.
    Description of amendment request: The proposed change would delete 
the applicable sections of the Limerick Generating Station, Units 1 and 
2, Facility Operating Licenses (FOLs), NPF-39 and NP