[Federal Register Volume 70, Number 79 (Tuesday, April 26, 2005)]
[Notices]
[Pages 21449-21470]
From the Federal Register Online via the Government Publishing 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, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, [email protected]; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].

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 NPF-85, which require 
Exelon Generation Company, LLC, (Exelon), to report violations of the 
requirements contained in Section 2.C of these licenses. 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: Darrell J. Roberts.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2), 
Beaver County, Pennsylvania

    Date of amendment request: February 11, 2005.
    Description of amendment request: The proposed changes would modify 
the BVPS-1 and 2 Technical Specifications (TSs) to implement the 
relaxed axial offset control (RAOC) and FQ surveillance 
methodologies. These methodologies are used to reduce operator action 
required to maintain conformance with power distribution control TSs, 
and increase the ability to return to power after a plant trip while 
still maintaining margin to safety limits under all operating 
conditions.
    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 will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed changes do not initiate an accident. Evaluations 
and analyses of accidents, which are potentially affected by the 
parameters and assumptions, associated with the RAOC and 
FQ(Z) methodologies have shown that all design standards 
and applicable safety criteria will continue to be met. The 
consideration of these changes does not result in a situation where 
the design, material, or construction standards that were applicable 
prior to the change are altered. Therefore, the proposed changes 
will not result in any additional challenges to plant equipment that 
could increase the probability of any previously evaluated accident.
    The proposed changes associated with the RAOC and 
FQ(Z) methodologies do not affect plant systems such that 
their function in the control of radiological consequences is 
adversely affected. The actual plant configuration, performance of 
systems, or initiating event mechanisms are not being

[[Page 21458]]

changed as a result of the proposed changes. The design standards 
and applicable safety criteria limits will continue to be met, 
therefore, fission barrier integrity is not challenged. The proposed 
changes associated with the RAOC and FQ(Z) methodologies 
have been shown not to adversely affect the plant response to 
postulated accident scenarios. The proposed changes will therefore 
not affect the mitigation of the radiological consequences of any 
accident described in the Updated Final Safety Analysis Report 
(UFSAR).
    Therefore the proposed changes do not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No. The proposed changes will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed change. 
The proposed changes do not challenge the performance or integrity 
of any safety-related system. The possibility for a new or different 
type of accident from any accident previously evaluated is not 
created since the proposed change does not result in a change to the 
design basis of any plant structure, system or component. Evaluation 
of the effects of the proposed changes has shown that all design 
standards and applicable safety criteria continue to be met.
    Equipment important to safety will continue to operate as 
designed and component integrity will not be challenged. The 
proposed changes do not result in any event previously deemed 
incredible being made credible. The proposed changes will not result 
in conditions that are more adverse and will not result in any 
increase in the challenges to safety systems.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously analyzed.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No. The proposed changes will not involve a 
significant reduction in a margin of safety. The proposed changes 
will assure continued compliance within the acceptance limits 
previously reviewed and approved by the NRC for RAOC and 
FQ(Z) methodologies. All of the appropriate acceptance 
criteria for the various analyses and evaluations will continue to 
be met.
    The impact associated with the implementation of RAOC on peak 
cladding temperature (PCT) has been evaluated for the planned 
extended power uprate. This evaluation has determined that 
implementation of RAOC at the extended power uprate power level will 
not result in a significant reduction in a margin of safety for 
either unit.
    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: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Section Chief: Richard J. Laufer.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2), 
Beaver County, Pennsylvania

    Date of amendment request: February 17, 2005.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3.7.7.1, ``Control Room Emergency 
Habitability Systems'' (BVPS-1), and TS 3.7.7, ``Control Room Emergency 
Air Cleanup and Pressurization System'' (BVPS-2), by dividing each 
specification into two specifications, addressing control room 
emergency ventilation and control room air cooling functions 
separately. Other minor changes are proposed to improve consistency 
with the Standard TSs and consistency between BVPS-1 and 2.
    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 changes do not adversely affect accident initiators 
or precursors or alter the design assumptions, conditions or 
configuration of the facility. The proposed changes do not alter or 
prevent the ability of structures, systems, or components to perform 
their intended function to mitigate the consequences of an 
initiating event within the assumed acceptance limits. The proposed 
change revises the TSs for the control room ventilation systems 
which are mitigating systems designed to minimize inleakage, to 
filter the control room atmosphere and to provide heat removal for 
the control room envelope. These functions maintain the control room 
temperature within design limits and protect the control room 
personnel following accidents previously analyzed. The proposed 
changes do not alter or reduce the capability of the affected 
systems to maintain the control room temperature and protect the 
control room personnel consistent with the assumptions of the 
applicable safety analyses. Therefore, the probability of any 
accident previously evaluated is not significantly increased. The 
proposed change continues to assure [that] adequate system and 
component testing is performed to verify the operability of the 
control room habitability systems to ensure mitigation features are 
capable of performing the assumed functions. Therefore, the 
consequences of any accident previously evaluated are not 
significantly increased.
    Therefore, it is concluded that the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No.
    The proposed changes will not adversely impact the accident 
analysis. The changes will not alter the requirements of the control 
room ventilation systems or their functions during accident 
conditions. No new or different accidents result from the 
application of the revised TS requirements. The changes do not 
involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a significant 
change in the methods governing normal plant operation. The changes 
do not alter assumptions made in the safety analyses. The proposed 
changes are consistent with the safety analyses assumptions and 
current plant operating practices.
    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?
    No.
    The proposed changes do 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 affected by these changes. The proposed changes will not 
result in plant operation in a configuration outside the design 
basis for an unacceptable period of time without compensatory 
measures. The proposed changes do not adversely affect systems that 
respond to safely shut down the plant and to maintain the plant in a 
safe shutdown condition.
    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: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Section Chief: Richard J. Laufer.

[[Page 21459]]

Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of amendment request: February 28, 2005.
    Description of amendment request: The proposed amendments would 
allow the use of the Small Break Loss of Coolant Accident (SBLOCA) 
methodology described in Westinghouse WCAP 10054-P-A Addendum 2 
Revision 1, ``Addendum to the Westinghouse Small Break emergency core 
cooling system (ECCS) Evaluation Model Using the NOTRUMP Code: Safety 
Injection into the Broken Loop and COSI Condensation Model'' dated July 
1997. This revised methodology determines the core response following a 
SBLOCA event and will be used to assure compliance with the post Loss 
of Coolant Accident (LOCA) acceptance criteria specified in 10 CFR 
50.46.
    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.
    The proposed amendment will change the Prairie Island Nuclear 
Generating Plant licensing basis by allowing the use of the approved 
NOTRUMP SBLOCA Evaluation Model described in Westinghouse WCAP 
10054-P-A Addendum 2 Revision 1, ``Addendum to the Westinghouse 
Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety 
Injection into the Broken Loop and COSI Condensation Model''.
    The methodology used to perform small break loss of coolant 
accident (SBLOCA) analyses is not an accident initiator, thus 
changing the methodology does not increase the probability of an 
accident.
    The fuel heat-up results generated by the proposed methodology 
will be utilized to demonstrate that the loss of coolant accident 
(LOCA) criteria for design basis for fission product barriers as 
described in 10 CFR Part 50.46 are not exceeded. The proposed 
methodology does not alter the nuclear reactor core, reactor coolant 
system, or equipment used directly in mitigation of a Small Break 
LOCA, thus radioactive releases due to a SBLOCA accident are not 
affected by the proposed change in analysis methodology. Therefore, 
this change does not increase the consequences of an accident 
previously evaluated.
    Therefore, the proposed 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.
    The proposed amendment will change the Prairie Island Nuclear 
Generating Plant licensing basis by allowing the use of the approved 
NOTRUMP SBLOCA Evaluation Model described in Westinghouse WCAP 
10054-P-A Addendum 2 Revision 1, ``Addendum to the Westinghouse 
Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety 
Injection into the Broken Loop and COSI Condensation Model''.
    The analysis of a SBLOCA accident using the proposed methodology 
does not alter the nuclear reactor core, reactor coolant system, or 
equipment used directly in mitigation of a Small Break LOCA.
    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.
    The proposed amendment will change the Prairie Island Nuclear 
Generating Plant licensing basis by allowing the use of the approved 
NOTRUMP SBLOCA Evaluation Model described in Westinghouse WCAP 
10054-P-A Addendum 2 Revision 1, ``Addendum to the Westinghouse 
Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety 
Injection into the Broken Loop and COSI Condensation Model''.
    The methodology in the proposed licensing basis change has 
previously been reviewed and approved by the Nuclear Regulatory 
Commission as a conservative methodology. The Prairie Island 
configuration is representative of the modeling used in the 
methodology. Therefore, the proposed licensing basis change will 
result in a conservative calculation of fuel conditions following a 
SBLOCA event. This will ensure that there is no reduction in the 
margin of safety for Prairie Island SBLOCA analyses that utilize 
this methodology.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Section Chief: L. Raghavan.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: March 31, 2005.
    Description of amendment request: The proposed amendment will 
increase the licensed power level to 1522 megawatts thermal (MWt) or 
1.50 percent greater than the current power level of 1500 MWt. The 
requested increase in licensed rated power is the result of a 
measurement uncertainty recapture (MUR) power uprate. The information 
provided in support of this request is based on the NRC's Regulatory 
Issue Summary 2002-03, ``Guidance on the Content of Measurement 
Uncertainty Recapture Power Uprate Applications,'' dated January 31, 
2002.
    On July 18, 2003, the licensee submitted, and the NRC subsequently 
approved, an MUR power uprate amendment to increase the licensed power 
level to 1524 MWt or 1.6 percent greater than the current level of 1500 
MWt. Problems during implementation resulted in the submission of an 
exigent license amendment request (LAR), which returned the licensed 
power to its original level (1500 MWt). The current LAR references the 
analysis from the July 18, 2003 submittal.
    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.
    Response: No.
    There are no changes as a result of the MUR power uprate to the 
design or operation of the plant that could affect system, 
component, or accident functions. All systems and components 
function as designed and the performance requirements have been 
evaluated and found to be acceptable.
    The reduction in power measurement uncertainty allows for safety 
analyses to continue to be used without modification. This is 
because those safety analyses were performed or evaluated at 102% of 
1500 MWt (1530 MWt) or higher. Analyses at these power levels 
support a core power level of 1522 MWt with a measurement 
uncertainty of 0.5%. Radiological consequences of USAR [Updated 
Safety Analysis Report] Chapter 14 accidents were assessed 
previously using the alternate source term methodology (Reference 
10.2 [Agencywide Documents Access Management System accession number 
ML013410095]). These analyses were performed at 102% of 1500 MWt 
(1530 MWt) and continue to be bounding. Updated Safety

[[Page 21460]]

Analysis Report (USAR) Chapter 14 analyses and accident analyses 
continue to demonstrate compliance with the relevant accident 
analyses' acceptance criteria. Therefore, there is no significant 
increase in the consequences of any accident previously evaluated.
    The primary loop components (reactor vessel, reactor internals, 
control element drive mechanisms, loop piping and supports, reactor 
coolant pumps, steam generators, and pressurizer) were evaluated at 
an uprated core power level of 1524 MWt and continue to comply with 
their applicable structural limits. These analyses also demonstrate 
the components will continue to perform their intended design 
functions. Changing the heatup and cooldown curves is based on 
uprated fluence values. This does not have a significant effect on 
the reactor vessel integrity. Thus, there is no significant increase 
in the probability of a structural failure of the primary loop 
components. The LBB [leak before break] analysis conclusions remain 
valid and the breaks previously exempted from structural 
consideration remain unchanged.
    All of the NSSS [nuclear steam system supplier] systems will 
continue to perform their intended design functions during normal 
and accident conditions. The auxiliary systems and components 
continue to comply with the applicable structural limits and will 
continue to perform their intended functions. The NSSS/BOP [nuclear 
steam system supplier/balance of plant] interface systems were 
evaluated at 1522 MWt and will continue to perform their intended 
design functions. Plant electrical equipment was also evaluated and 
will continue to perform their intended functions. Therefore, the 
proposed change does not involve a significant increase in 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.
    Response: No.
    No new accident scenarios, failure mechanisms, or single 
failures are introduced as a result of the proposed change. All 
systems, structures, and components previously required for the 
mitigation of an event remain capable of fulfilling their intended 
design function at the uprated power level. The proposed change has 
no adverse effects on any safety related systems or component and 
does not challenge the performance or integrity of any safety 
related system. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Response: No.
    Operation at 1522 MWt core power does not involve a significant 
reduction in the margin of safety. The current accident analyses 
have been previously performed with a 2% power measurement 
uncertainty or at uprated core powers that exceed the MUR uprated 
core power. System and component analyses have been completed at the 
MUR uprated core power conditions. Analyses of the primary fission 
product barriers at uprated core powers have concluded that all 
relevant design basis criteria remain satisfied in regard to 
integrity and compliance with the regulatory acceptance criteria. As 
appropriate, all evaluations have been both reviewed and approved by 
the NRC, or are currently under review (the proposed Pressure-
Temperature Limits Report). Therefore, the proposed change does not 
involve a significant reduction in 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: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California

    Date of amendment requests: December 28, 2004.
    Description of amendment requests: The proposed amendments would 
relocate reactor coolant system related cycle-specific parameters from 
the Technical Specifications to the Core Operating Limits Report.
    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 proposed changes are programmatic and administrative in 
nature, which do not physically alter safety related systems, nor 
affect the way in which safety related systems perform their 
functions. More specific requirements regarding the safety limits 
(i.e., departure from nucleate boiling ratio limit and peak fuel 
centerline temperature limit) are being imposed in Technical 
Specification (TS) 2.1.1, ``Reactor Core SLs [Safety Limits],'' 
which replace the reactor core safety limits figure and are 
consistent with the values stated in the Final Safety Analysis 
Report Update (FSARU). The proposed changes remove cycle-specific 
parameters from TS 3.4.1 and relocate them to the Core Operating 
Limits Report (COLR), which do not change the plant design or affect 
system operating parameters. In addition, the minimum limit for 
reactor coolant system (RCS) total flow rate is being retained in TS 
3.4.1 to assure that a lower flow rate than reviewed by the NRC will 
not be used. The proposed changes do not, by themselves, alter any 
of the parameters. The removal of the cycle-specific parameters from 
the TS does not eliminate existing requirements to comply with the 
parameters.
    The proposed changes to TS 5.6.5b to reference only the topical 
report number and title for three of the topical reports do not 
alter the use of the analytical methods used to determine core 
operating limits that have been reviewed and approved by the NRC. 
This method of referencing topical reports would allow the use of 
current topical reports to support limits in the COLR without having 
to submit a request for an amendment to the operating license. 
Implementation of revisions to these topical reports would still be 
reviewed in accordance with 10 CFR 50.59 and, where required, 
receive NRC review and approval.
    Although the relocation of the cycle-specific parameters to the 
COLR would allow revision of the affected parameters without prior 
NRC approval, there is no significant effect on the probability or 
consequences of an accident previously evaluated. Future changes to 
the COLR parameters could result in event consequences which are 
either slightly less or slightly more severe than the consequences 
for the same event using the present parameters. The differences 
would not be significant and would be bounded by the existing 
requirement of TS 5.6.5c to meet the applicable limits of the safety 
analyses.
    The cycle-specific parameters being transferred from the TS to 
the COLR will continue to be controlled under existing programs and 
procedures. The FSARU accident analyses will continue to be examined 
with respect to changes in the cycle-dependent parameters obtained 
using NRC reviewed and approved reload design methodologies, 
ensuring that the transient evaluation of new reload designs are 
bounded by previously accepted analyses. This examination will 
continue to be performed pursuant to 10 CFR 50.59 requirements, 
ensuring that future reload designs will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. Additionally, the proposed changes do not 
allow for an increase in plant power levels, do not increase the 
production, nor alter the flow path or method of disposal of 
radioactive waste or byproducts. Therefore, the proposed changes do 
not change the type or increase the amount of any effluents released 
offsite.
    Therefore, the proposed changes do not involve a significant 
increase in 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.
    The proposed changes that retain the minimum limit for RCS total 
flow rate in the TS, and that relocate certain cycle-specific 
parameters from the TS to the COLR, thus removing the requirement 
for prior NRC approval of revisions to those parameters, do not 
involve a physical change to the plant. No new equipment is being 
introduced, and

[[Page 21461]]

installed equipment is not being operated in a new or different 
manner. There are no changes being made to the parameters within 
which the plant is operated, other than their relocation to the 
COLR. There are no set points affected by the proposed changes at 
which protective or mitigative actions are initiated. The proposed 
changes will not alter the manner in which equipment operation is 
initiated, nor will the function demands on credited equipment be 
changed. No alteration in the procedures which ensure the plant 
remains within analyzed limits is being proposed, and no change is 
being made to the procedures relied upon to respond to an off-normal 
event. As such, no new failure modes are being introduced.
    The proposed changes to reference only the topical report number 
and title do not alter the use of the analytical methods used to 
determine core operating limits that have been reviewed and approved 
by the NRC. This method of referencing topical reports would allow 
the use of current topical reports to support limits in the COLR 
without having to submit a request for an amendment to the operating 
license. Implementation of revisions to topical reports would still 
be reviewed in accordance with 10 CFR 50.59 and, where required, 
receive NRC review and approval.
    Relocation of cycle-specific parameters has no influence or 
impact on, nor does it contribute in any way to the possibility of a 
new or different kind of accident. The relocated cycle-specific 
parameters will continue to be calculated using the NRC reviewed and 
approved methodology. The proposed changes do not alter assumptions 
made in the safety analysis, and operation within the core operating 
limits will continue.
    Therefore, the proposed changes do not create a new or different 
kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The margin of safety is established through equipment design, 
operating parameters, and the set points at which automatic actions 
are initiated. The proposed changes do not physically alter safety-
related systems, nor do they affect the way in which safety-related 
systems perform their functions. The set points at which protective 
actions are initiated are not altered by the proposed changes. 
Therefore, sufficient equipment remains available to actuate upon 
demand for the purpose of mitigating an analyzed event. As the 
proposed changes to relocate cycle-specific parameters to the COLR 
will not affect plant design or system operating parameters, there 
is no detrimental impact on any equipment design parameter, and the 
plant will continue to operate within prescribed limits.
    The development of cycle-specific parameters for future reload 
designs will continue to conform to NRC reviewed and approved 
methodologies, and will be performed pursuant to 10 CFR 50.59 to 
assure that the plant operates within cycle-specific parameters.
    The proposed changes to reference only the topical report number 
and title do not alter the use of the analytical methods used to 
determine core operating limits that have been reviewed and approved 
by the NRC. This method of referencing topical reports would allow 
the use of current NRC-approved topical reports to support limits in 
the COLR without having to submit a request for an amendment to the 
operating license. Implementation of revisions to topical reports 
would still be reviewed in accordance with 10 CFR 50.59 and, where 
required, receive NRC review and approval.
    Therefore, the proposed changes do 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 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 Section Chief: Robert A. Gramm.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California

    Date of amendment requests: December 31, 2004.
    Description of amendment requests: The proposed amendments would 
revise Technical Specification 3.4.10, ``Pressurizer Safety Valves'' to 
add a separate Action and associated Completion Times for one or more 
inoperable pressurizer safety valves for the condition where the valves 
are inoperable solely due to loop seal temperatures being outside of 
design limits.
    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.
    This proposed change revises Technical Specification (TS) 
3.4.10, ``Pressurizer Safety Valves,'' to add a separate Action and 
associated Completion Times (CTs) for one or more inoperable 
pressurizer safety valves (PSV) for the condition where the valves 
are inoperable solely due to loop seal temperatures being outside of 
design limits. Currently, when a PSV is in such a condition, it is 
conservatively declared inoperable and TS 3.4.10 Condition A is 
entered which has a CT of 15 minutes. A CT of 15 minutes normally 
provides insufficient time for restoring a PSV loop seal temperature 
to within limits. The new Action will provide CTs of 12 hours for 
exceeding the high temperature limit and 24 hours (MODES 1 and 2) or 
72 hours (MODES 3 and 4) for exceeding the low temperature limit. In 
addition, two new PSV loop seal temperature surveillance 
requirements are proposed to assist in assuring PSV operability.
    Loop seals are provided in the PSV inlet piping to maintain PSV 
body temperature within vendor recommended limits. This prevents PSV 
seat leakage that can result from spring relaxation with increased 
temperature. However, the water in the loop seals must be maintained 
at or above a minimum temperature to allow it to flash to steam when 
a PSV lifts. Because of the low density and low mass flow rate, PSV 
steam relief imposes minimal loading on the discharge piping 
ensuring acceptable pipe stresses. However, if cooler water is 
maintained in the loop seals, it may not flash completely, and a 
water and steam mixture could be discharged when a PSV lifts. 
Because of the higher density and higher mass flow rate, PSV relief 
of water and steam could impose increased loading and could result 
in unacceptably high pipe stresses on the discharge piping which 
could render the PSVs inoperable and/or damage the discharge piping.
    The concern with the PSV opening during liquid relief conditions 
or with the loop seal temperature outside design limits, is the 
ability to ensure the valve reseats properly and no leakage occurs 
after the valve closes. However, even under liquid relief 
conditions, PSVs are still capable of providing their required 
relief capacity.
    Failure of the PSV to reseat following discharge would result in 
an unisolable reactor coolant system leak. The consequences of such 
a leak are bounded by existing Final Safety Analysis Report Update 
(FSARU) accident analyses. Probabilistic risk assessment methods and 
a deterministic analysis have been utilized to determine there is no 
significant increase in core damage frequency or large early release 
frequency.
    Therefore, the proposed change does not involve a significant 
increase in 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.
    Failure of one or more PSVs to reseat following discharge would 
result in an unisolable reactor coolant system leak. The 
consequences of such a leak are bounded by existing FSARU accident 
analyses and no new failure modes are introduced.
    Therefore, the proposed change does not create the possibility 
of a new or different accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change is based upon both a deterministic 
evaluation and a risk-informed assessment.
    The deterministic evaluation concluded that even with the loop 
seal temperature outside of design limits, causing one or more PSVs 
to be declared inoperable, the PSVs

[[Page 21462]]

would still lift on demand to perform their safety function. Failure 
of one or more PSVs to reseat following discharge, resulting in an 
unisolable reactor coolant system leak, is an event bounded by 
existing FSARU accident analyses.
    The risk assessment performed to support this license amendment 
request concluded that the increase in plant risk is small and 
consistent with the NRC's Safety Goal Policy Statement, ``Use of 
Probabilistic Risk Assessment Methods in Nuclear Activities: Final 
Policy Statement,'' Federal Register, Volume 60, p. 42622, August 
16, 1995 and guidance contained in of Regulatory Guides (RG) 1.174, 
``An Approach for Using Probabilistic Risk Assessment in Risk-
Informed Decisions on Plant-Specific Changes to the Licensing 
Basis,'' dated July 1998 and RG 1.177, ``An Approach for Plant-
Specific, Risk-Informed Decisionmaking: Technical Specifications,'' 
dated August 1998.
    Together, the deterministic evaluation and the risk-informed 
assessment provide high assurance that the PSVs will meet their 
design requirements.
    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 Section Chief: Robert A. Gramm.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California

    Date of amendment requests: March 11, 2005.
    Description of amendment requests: The proposed amendment would 
modify Technical Specification (TS) 5.5.9, ``Steam Generator (SG) Tube 
Surveillance Program,'' and 5.6.10, ``Steam Generator (SG) Tube 
Inspection Report,'' to allow the use of the SG tube W star (W*) 
alternate repair criteria (ARC) on a permanent basis. The W* ARC allows 
axial primary water stress corrosion cracking indications in the 
Westinghouse explosive tube expansion (WEXTEX) region to remain in 
service if the indication is located below the bottom of the WEXTEX 
transition. In addition, TS 5.6.10.d for NRC notification requirements 
of the voltage-based ARC would be revised.
    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.
    Of the various accidents previously evaluated, the permanent use 
of the steam generator (SG) tube W star (W*) alternate repair 
criteria (ARC) only affects the steam generator tube rupture (SGTR) 
accident evaluation and the postulated main steam line break (MSLB) 
accident evaluation. Loss-of-coolant accident (LOCA) conditions 
cause a compressive axial load to act on the tube. Therefore, since 
the LOCA tends to force the tube into the tubesheet rather than pull 
it out, it is not a factor in this evaluation.
    For the SGTR accident, the required structural margins of the SG 
tubes will be maintained by the presence of the tubesheet. Tube 
rupture is precluded for cracks in the Westinghouse explosive tube 
expansion (WEXTEX) region due to the constraint provided by the 
tubesheet. Therefore, Regulatory Guide (RG) 1.121, ``Bases for 
Plugging Degraded PWR Steam Generator Tubes,'' margins against burst 
are maintained for both normal and postulated accident conditions.
    WCAP-14797-P, Revision 2, defines a length, W*, of degradation-
free expanded tubing that provides the necessary resistance to tube 
pullout due to the pressure-induced forces (with applicable safety 
factors applied). The W* length supplies the necessary resistive 
force to preclude pullout loads under both normal operating and 
accident conditions. The contact pressure results from the WEXTEX 
expansion process, thermal expansion mismatch between the tube and 
tubesheet and from the differential pressure between the primary and 
secondary side as offset at higher tubesheet elevations by bow of 
the tubesheet. The proposed changes do not affect other systems, 
structures, components, or operational features. Therefore, the 
proposed change results in no significant increase in the 
probability of the occurrence of an SGTR or MSLB accident.
    The consequences of an SGTR accident are affected by the 
primary-to-secondary leakage flow during the accident. Primary-to-
secondary leakage flow through a postulated broken tube is not 
affected by the proposed changes since the tubesheet enhances the 
tube integrity in the region of the WEXTEX expansion by precluding 
tube deformation beyond its initial expanded outside diameter. The 
resistance to both tube rupture and collapse is strengthened by the 
tubesheet in that region. At normal operating pressures, leakage 
from primary water stress corrosion cracking in the W* length is 
limited by both the tube-to-tubesheet crevice and the limited crack 
opening permitted by the tubesheet constraint. No leakage has been 
observed in any in situ test of W* indications to date. 
Consequently, negligible normal operating leakage is expected from 
cracks within the tubesheet region.
    MSLB leakage is limited by leakage flow restrictions resulting 
from the crack and tubesheet that provide a restricted leakage path 
and also limit the degree of crack face opening compared to free 
span indications. The total leakage, that is, the combined leakage 
for all such tubes, plus the combined leakage developed by any other 
ARC and non-ARC degradation, is limited to less than the maximum 
allowable MSLB accident dose analysis leak rate limit, such that 
offsite dose is maintained less than the guideline value in Title 10 
to the Code of Federal Regulations (10 CFR) Part 100 and control 
room dose is maintained less than the value in General Design 
Criterion (GDC) 19 of Appendix A to 10 CFR Part 50. In addition, the 
editorial changes made to Technical Specifications 5.5.9 and 5.6.10 
have no impact on the MSLB leakage [and the SGTR].
    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 changes do not introduce any changes or mechanisms 
that create the possibility of a new or different kind of accident. 
Tube bundle integrity is expected to be maintained for all plant 
conditions upon continued implementation of the W* ARC.
    Axial indications left in service shall have the upper crack tip 
below the top of the tubesheet (TTS) by at least the value of the 
nondestructive examination (NDE) uncertainty and crack growth 
allowance, such that at the end of the subsequent operating cycle 
the entire crack remains below the tubesheet secondary face, thereby 
minimizing the potential for free span cracking and demonstrating 
that an acceptable level of risk is maintained for tubes returned to 
service under W* ARC. This repair criterion is in addition to 
ensuring that the upper crack tip is located below the bottom of the 
WEXTEX transition by at least the NDE measurement uncertainty. 
Condition monitoring will verify that all tube cracks returned to 
service under W* ARC remain below the TTS, including an allowance 
for NDE uncertainty.
    These changes do not introduce any new equipment or any change 
to existing equipment. No new effects on existing equipment are 
created nor are any new malfunctions introduced.
    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.
    The proposed changes maintain the required structural margins of 
the SG tubes for both normal and accident conditions. RG 1.121 is 
used as the basis in the development of the W* ARC for determining 
that SG tube integrity considerations are maintained within 
acceptable limits. RG 1.121 describes a method acceptable to the NRC 
staff for meeting General Design Criteria 14, 15, 31,

[[Page 21463]]

and 32 by reducing the probability and consequences of an SGTR. RG 
1.121 concludes that by determining the limiting safe conditions of 
tube wall degradation beyond which tubes with unacceptable cracking, 
as established by inservice inspection, should be removed from 
service or repaired, the probability and consequences of a SGTR are 
reduced. This RG uses safety factors on loads for tube-burst that 
are consistent with the requirements of Section III of the ASME 
Code.
    For primarily axially oriented cracking located within the 
tubesheet, tubeburst is precluded due to the presence of the 
tubesheet. WCAP-14797-P, Revision 2, defines a length, W*, of 
degradation free expanded tubing that provides the necessary 
resistance to tube pullout due to the pressure induced forces (with 
applicable safety factors applied). Application of the W* ARC will 
preclude unacceptable primary-to-secondary leakage during all plant 
conditions. The methodology for determining MSLB leakage due to 
indications within the tubesheet region provides for large margins 
between calculated and actual leakage values. In addition, the total 
leakage, including leakage due to use of other ARC, is maintained 
below the maximum allowable MSLB accident dose analysis leak rate 
limit, such that offsite dose is maintained less than the guideline 
value in 10 CFR Part 100 and control room dose is maintained less 
than the value in GDC 19. In addition, the editorial changes made to 
Technical Specifications 5.5.9 and 5.6.10 have no impact on the 
determination of MSLB leakage [and the SGTR].
    Plugging of the SG tubes reduces the reactor coolant flow margin 
for core cooling. Continued implementation of W* ARC will result in 
maintaining the margin of flow that may have otherwise been reduced 
by tube plugging.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above evaluation, PG&E [Pacific Gas and Electric 
Company] concludes that the proposed change presents no significant 
hazards consideration under the standards set forth in 10 CFR 
50.92(c), and accordingly, a finding of ``no significant hazards 
consideration'' is justified.

    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 Section Chief: Robert Gramm.

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 9, 2004.
    Description of amendment request: The proposed amendments would 
change the SSES 1 and 2 Technical Specifications (TSs) 3.8.4, ``DC 
Sources-Operating,'' 3.8.5, ``DC Sources-Shutdown,'' 3.8.6, ``Battery 
Cell Parameters,'' and add a new TS Section, 5.5.13, ``Battery 
Monitoring and Maintenance Program.'' These changes are consistent with 
Technical Specifications Change Traveler (TSTF) 360, Revision 1 to 
request new actions with increased completion times for an inoperable 
battery chargers and alternate battery charger testing criteria for 
limiting condition for operation (LCO) 3.8.4 and LCO 3.8.5.
    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 changes restructure the Technical 
Specifications (TSs) for the DC Electrical Power Systems. The 
proposed changes add actions to specifically address battery charger 
inoperability. This change will rely upon the capability of 
providing the battery charger function by an alternate means (e.g., 
a 125 volts direct current (VDC) portable battery charger or a 250 
VDC portable battery charger) to justify the proposed Completion 
Times. The DC electrical power systems, including associated battery 
chargers, are not initiators to any accident sequence analyzed in 
the Final Safety Analysis Report (FSAR). Operation in accordance 
with the proposed TS ensures that the DC electrical power systems 
are capable of performing functions as described in the FSAR. 
Therefore the mitigative functions supported by the DC Power Systems 
will continue to provide the protection assumed by the analysis.
    The relocation of preventive maintenance surveillance, and 
certain operating limits and actions to a newly-created, licensee-
controlled TS 5.5.13, ``Battery Monitoring and Maintenance 
Program,'' will not challenge the ability of the DC electrical power 
systems to perform their design functions. The maintenance and 
monitoring required by current TS, which are based on industry 
standards, will continue to be performed. In addition, the DC Power 
Systems are within the scope of 10 CFR 50.65, ``Requirements for 
Monitoring the Effectiveness of Maintenance at Nuclear Power 
Plants,'' which will ensure the control of maintenance activities 
associated with the DC electrical power systems. The integrity of 
fission product barriers, plant configuration, and operating 
procedures as described in the FSAR will not be affected by the 
proposed changes.
    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?
    No. The proposed changes involve restructuring the TS for the DC 
electrical power systems. These changes will rely upon the 
capability of providing the battery charger function by an alternate 
means to justify the proposed completion times when a normal battery 
charger is inoperable. The DC electrical power systems, which 
include the associated battery chargers, are not initiators to any 
accident sequence analyzed in the FSAR. Rather, the DC electrical 
power systems are used to supply equipment used to mitigate an 
accident. These mitigative functions, supported by the DC electrical 
power systems are not affected by these changes and they will 
continue to provide the protection assumed by the safety analysis 
described in the FSAR. There are no new types of failures or new or 
different kinds of accidents or transients that could be created by 
these changes.
    Therefore, the proposed changes 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?
    No. The margin of safety is established through equipment 
design, operating parameters, and the set points at which automatic 
actions are initiated. The proposed changes will not adversely 
affect operation of plant equipment. These changes will not result 
in a change to the set points at which protective actions are 
initiated. Sufficient DC electrical system capacity is ensured to 
support operation of mitigation equipment. The changes associated 
with the new Battery Maintenance and Monitoring Program will ensure 
that the station batteries are maintained in a highly reliable 
state. The use of spare battery chargers will increase the 
reliability of the DC electrical systems during periods of normal 
battery charger inoperability. The equipment fed by the DC 
electrical sources will continue to provide adequate power to safety 
related loads in accordance with analysis assumptions. 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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Section Chief: Richard J. Laufer.

[[Page 21464]]

Southern Nuclear Operating Company, Inc., Docket No. 50-364, Joseph M. 
Farley Nuclear Plant, Unit 2, Houston County, Alabama

    Date of amendment request: January 19, 2005.
    Description of amendment request: The proposed amendments would 
revise the Updated Final Safety Analysis Report to allow the use of 
fire rated electrical cable for fire areas 2-013 and 2-042 in lieu of a 
one hour rated electrical cable raceway fire barrier enclosure as 
described by Title 10 of the Code of Federal Regulations (10 CFR) Part 
50, Appendix R, Section III.G.2 for protection of safe shutdown 
circuits.
    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 does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. 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. This is a revision to the FSAR to use [mineral 
insulated] MI cable in fire areas 2-013 and 2-042. The MI cable has 
been tested to applicable requirements and the implementation design 
reflects the test results. Therefore, the probability of any 
accident previously evaluated is not increased. Equipment required 
to mitigate an accident remain capable of performing the assumed 
function. Therefore, the consequences of any accident previously 
evaluated are not increased.
    Therefore, it is concluded that this change does not 
significantly 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 accident previously evaluated?
    The proposed change will not alter the requirements or function 
for systems required during accident conditions. No new or different 
accidents result from implementing MI cable for fire areas 2-013 and 
2-042. The MI cable has been tested to applicable requirements, and 
the implementation design reflects the test results. The use of MI 
cable is not a significant change in the methods governing normal 
plant operation. The proposed change is consistent with the safety 
analysis assumptions and current plant operating practice.
    Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated is not created.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    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 affected by this change. The proposed change will not result 
in plant operation in a configuration outside the design basis for 
an unacceptable period of time without mitigating actions. The 
proposed change does not affect systems that respond to safely 
shutdown the plant and to maintain the plant in a safe shutdown 
condition.
    Therefore, it is concluded that this 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: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Section Chief: John A. Nakoski.

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station, Units 1 and 2, Somervell County, Texas

    Date of amendment request: March 24, 2005.
    Brief description of amendments: These proposed changes would 
revise Technical Specification (TS) 3.3.1 entitled ``Reactor Trip 
System Instrumentation'' (RTS) and TS 3.3.2 entitled ``Engineered 
Safety Feature Actuation System Instrumentation'' (ESFAS) Required 
Action Notes to reflect the wording in Standard Technical 
Specifications (STS) for plants with bypass capability per TS Task 
Force Traveler 418, Revision 2.
    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.
    [Westinghouse Topical Report] WCAP-14333 provided the technical 
justification for relaxing various RTS and ESFAS Instrumentation 
bypass test times, Completion Times, and Surveillance Frequencies 
located in TS 3.3.1 and 3.3.2. As such, the proposed changes do not 
represent a significant hazards consideration or present a reduction 
in the margin of safety.
    The protection system performance will remain within the bounds 
of the previously performed accident analyses since no hardware 
changes are proposed. The same Reactor Trip System (RTS) 
Instrumentation and Engineered Safety Feature Actuation (ESFAS) 
Instrumentation will continue to be used and remain unchanged. The 
protection systems will continue to function in a manner consistent 
with the plant design basis. These changes to the TS do not result 
in a condition where the design, material, and construction 
standards, which were applicable prior to these changes, are 
altered.
    The proposed changes will not modify any system interface. The 
proposed changes will not affect the probability of any event 
initiators. There will be no degradation in the performance of or an 
increase in the number of challenges imposed on safety-related 
equipment assumed to function during an accident situation. There 
will be no change to normal plant operating parameters or accident 
mitigation performance. The proposed changes will not alter any 
assumptions or change any mitigation actions in the radiological 
consequence evaluations in the FSAR [final safety analysis report].
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, or 
configurations of the facility or change the manner in which the 
plant is operated and maintained. The proposed changes do 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 changes will not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of an accident previously evaluated. 
The proposed changes are consistent with safety analysis assumptions 
and resultant consequences.
    Therefore, the proposed 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.
    There are no hardware changes nor is there any change in the 
method by which any safety-related plant system performs its safety 
function. The proposed changes will not affect the normal method of 
plant operation. No performance requirements will be affected or 
eliminated. The proposed changes will not result in physical 
alteration to any plant system nor will there be any change in the 
method by which any safety-related plant system performs its safety 
function.
    There will be no setpoint changes or changes to accident 
analysis assumptions. No new accident scenarios, transient 
precursors, failure mechanisms, or limiting single failures are 
introduced as a result of these changes. There will be no adverse 
effect or

[[Page 21465]]

challenges imposed on any safety-related system as a result of these 
changes.
    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.
    The proposed changes do not affect the acceptance criteria for 
any analyzed event nor is there a change to any Safety Analysis 
Limit (SAL). There will be no effect on the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined nor will there be any effect on those plant 
systems necessary to assure the accomplishment of protection 
functions. The radiological dose consequence acceptance criteria 
listed in the Standard Review Plan will continue to be met.
    Redundant RTS and ESFAS trains are maintained and diversity, 
with regard to the signals that provide reactor trip and engineered 
safety features actuation, is also maintained. All signals are 
credited as primary or secondary and all operator actions credited 
in the accident analyses will remain the same. The proposed changes 
will not result in plant operation in a configuration outside the 
design basis.
    Therefore, the proposed changes do not involve a 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: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Allen G. Howe.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: March 1, 2005.
    Description of amendment request: The proposed changes to the 
Technical Specifications (TS) would revise the frequency for the Trip 
Actuating Device Operational Test of the P-4 Interlock Function and add 
Mode 4 to the Applicability for TS 3.3.2.
    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 changes involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    The proposed changes do not significantly increase the 
probability or consequences of an accident previously evaluated in 
the UFSAR [Updated Final Safety Analysis Report]. These interlocks 
and the associated testing do not directly initiate an accident. The 
consequences of accidents previously evaluated in the UFSAR are not 
adversely affected by these proposed changes because the changes are 
made to accurately reflect the design of the ESFAS [Engineered 
Safety Features Actuation System] system. Therefore, the proposed 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Do changes create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident already evaluated in 
the UFSAR. No new accident scenarios, failure mechanisms, or 
limiting single failures are introduced as a result of the proposed 
changes. The proposed changes do not challenge the performance or 
integrity of any safety-related systems. Therefore, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Do changes involve a significant reduction in the margin of 
safety?
    The proposed changes do not involve a significant reduction in a 
margin of safety. The proposed changes are made to accurately 
reflect the design of the ESFAS system. The nominal actuation set 
points specified by the Technical Specifications and the safety 
analysis limits assumed in the transient and accident analysis are 
unchanged. Therefore, the proposed changes will not significantly 
reduce the margin of safety as defined in the Technical 
Specifications.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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. Lillian M. Cuoco, Esq., Senior Counsel, 
Dominion Resources Services, Inc., Millstone Power Station, Building 
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
    NRC Section Chief: John A. Nakoski.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: March 8, 2005.
    Description of amendment request: The proposed changes would revise 
the auxiliary feedwater (AFW) operability requirements and add an AFW 
allowed outage time and required actions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed revision to the AFW pump and flowpath requirements, 
as well as the revision of AFW surveillances, does not increase the 
probability of accidents previously evaluated since the AFW System 
is not required to operate until after the occurrence of the 
previously evaluated accidents. The change does not impact any of 
the initiators of the accidents. The proposed change does not 
involve a significant increase in the consequences of an accident 
previously evaluated because the AFW System will continue to perform 
its intended safety function for these accidents. The operation of 
the AFW System with the revised required action statements and added 
surveillances continues to meet the applicable design criteria.
    2. Create the possibility of a new or different type of accident 
from any accident previously identified.
    The safety function of the AFW System continues to be the same 
and is met using the same equipment. The change does not involve any 
plant modifications and does not revise the design of the plant or 
the AFW System. Operation of the AFW System with the revised 
required action statements and revised surveillances continues to 
meet the applicable design criteria and is consistent with the Surry 
accident analyses. Therefore, the proposed change does not introduce 
any new failures that could create the possibility of a new or 
different kind of accident from any accident previously identified.
    3. Involve a significant reduction in a margin of safety.
    The revised requirements for the AFW pumps and flowpaths, as 
well as the revision of AFW surveillances, continue to assure that 
the margins of safety assumed in the accidents and transients that 
rely upon operation of the AFW System are maintained. The proposed 
required action statements appropriately place the plant in a safe 
condition for the circumstances being addressed. Therefore, this 
proposed revision does not affect 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. Lillian M. Cuoco, Esq., Senior Counsel, 
Dominion Resources Services, Inc., Millstone Power Station, Building 
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
    NRC Section Chief: John A. Nakoski.

[[Page 21466]]

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: March 17, 2005.
    Description of amendment request: The proposed change would 
incorporate a license condition that would permit irradiation of the 
fuel assemblies to a lead rod average burnup of 62,000 MWD/MTU.
    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 probability of occurrence or the consequences of an 
accident previously evaluated is not significantly increased.
    For most of the accidents analyzed in the UFSAR [Updated Final 
Safety Analysis Report] (e.g., LOCA [loss-of-coolant accident], 
Steam Line Break, etc.) the fuel design has no impact on the 
likelihood of initiation of an accident. Fuel performance is 
evaluated as a consequence of the accident. The only accident where 
the fuel design may have an impact on the likelihood of a Chapter 14 
accident is the Fuel Handling Accident discussed in Chapter 14.4.1 
of the Surry UFSAR. The activity being evaluated is a slight 
increase in the lead rod average burnup limit for the fuel 
assemblies. No change in fuel design or fuel enrichment will be 
required to increase the lead rod average burnup. The fuel rods at 
the extended lead rod average burnup will continue to meet the 
design limits with respect to fuel rod growth, clad fatigue, rod 
internal pressure and corrosion. Thus, there will be no impact on 
the capability to engage the fuel assemblies with the handling 
tools. Therefore, it is concluded that the change will not result in 
more than a minimal increase in the frequency of occurrence of any 
accident previously evaluated in the UFSAR. The impact of extending 
the lead rod average burnup to 62,000 MWD/MTU from 60,000 MWD/MTU on 
the Core Kinetics Parameter, Core Thermal-Hydraulics/DNBR [Departure 
from Nucleate Boiling Ratio], Specific Accident Considerations, and 
Radiological Consequences was considered. Based on the evaluation of 
these considerations, it is concluded that increasing the lead rod 
average burnup limit to 62,000 MWD/MTU will not result in a 
significant increase in the consequences of the accidents previously 
evaluated in the Surry UFSAR.
    2. The possibility for a new or different type of accident from 
any accident previously evaluated is not created.
    The fuel is the only component affected by the change in the 
burnup limit. The change does not affect the thermal hydraulic 
response to any transient or accident. The fuel rod design criteria 
[will] continue to be met at the higher burnup limit. Thus, the 
change does not create the possibility of an accident of a different 
type.
    3. The margin of safety as defined in the Bases to the Surry 
Technical Specifications is not significantly reduced.
    The operation of the Surry cores with a limited number of fuel 
assemblies with some fuel rods irradiated to a lead rod average 
burnup of 62,000 MWD/MTU will not change the performance 
requirements of any system or component such that any design 
criteria will be exceeded. The normal limits on core operation 
defined in the Surry Technical Specifications will remain applicable 
for the irradiation of the fuel to a lead rod average burnup of 
62,000 MWD/MTU. Therefore, the margin of safety as defined in Bases 
to the Surry Technical Specifications is not significantly reduced.

    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. Lillian M. Cuoco, Esq., Senior Counsel, 
Dominion Resources Services, Inc., Millstone Power Station, Building 
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
    NRC Section Chief: John A. Nakoski.

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].

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power 
Station, Unit No. 2, New London County, Connecticut

    Date of application for amendment: July 6, 2004, as supplemented by 
letters dated September 21, and December 23, 2004.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) to allow a one-time change in the Appendix J, Type 
A, Containment Integrated Leak Rate Test from the required 10 years to 
15 years.
    Date of issuance: April 6, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 285.
    Facility Operating License No. DPR-65: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: February 1, 2005 (70 FR 
5237). The September 21 and December 23, 2004, letters provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination or expand the 
application beyond the scope of the original Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 6, 2005.
    No significant hazards consideration comments received: No.

[[Page 21467]]

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 application for amendments: April 6, 2004, as supplemented 
by letter dated August 5, 2004.
    Brief description of amendments: The amendments revised the 
Technical Specifications to allow a diesel generator battery to remain 
operable with no more than one cell less than 1.36 Volts DC on float 
charge.
    Date of issuance: March 29, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 221 and 216.
    Renewed Facility Operating License Nos. NPF-35 and NPF-52: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 14, 2004 (69 
FR 55469). The supplement dated August 5, 2004 provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 29, 2005.
    No significant hazards consideration comments received: No.

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 application for amendments: September 28, 2004.
    Brief description of amendments: The amendments eliminate the 
technical specification requirements to submit monthly operating 
reports and annual occupational radiation exposure reports.
    Date of issuance: March 31, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 222 and 217.
    Renewed Facility Operating License Nos. NPF-35 and NPF-52: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: November 23, 2004 (69 
FR 68182).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 31, 2005.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of application for amendments: June 3, 2003, as supplemented 
by letters dated July 29 and December 7, 2004, and January 18, 2005.
    Brief description of amendments: The amendments revise TS 3.6.14 to 
allow a pressurizer enclosure hatch between the upper and lower 
containment volumes to be open for up to 6 hours to facilitate 
inspections of components such as the power operated relief valve block 
valves.
    Date of issuance: April 5, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 228/210.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: July 22, 2003 (68 FR 
43383). The supplemental letters dated July 29 and December 7, 2004, 
and January 18, 2005, provided clarifying information that did not 
change the initial proposed no significant hazards consideration 
determinations.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 5, 2005.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of application for amendments: September 20, 2004.
    Brief description of amendments: The amendments deleted the 
Technical Specifications associated with hydrogen recombiners and 
hydrogen monitors.
    Date of issuance: April 4, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 227 and 209.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 1, 2005 (70 FR 
5239)
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 4, 2005.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of application for amendments: September 28, 2004.
    Brief description of amendments: The amendments eliminate the 
technical specification requirements to submit monthly operating 
reports and annual occupational radiation exposure reports.
    Date of issuance: March 31, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 226 and 208.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: November 23, 2004 (69 
FR 68182).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 31, 2005.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of application of amendments: October 16, 2003, as 
supplemented by letters dated May 11, 2004, and January 10, 2005.
    Brief description of amendments: The amendments revised the 
Technical Specification (TS) 3.4.9 and the associated Bases to change 
the minimum pressurizer heater capacity from 126 kW to 400 kW to 
correct a non-conservative TS associated with a pressurizer design-
basis deficiency.
    Date of Issuance: March 28, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 343, 345, & 344.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 20, 2004 (69 FR 
2740).
    The supplements dated May 11, 2004, and January 10, 2005, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register on January 20, 2004 
(69 FR 2740).

[[Page 21468]]

    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 28, 2005.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of application of amendments: September 20, 2004.
    Brief description of amendments: The amendments delete the 
Technical Specifications associated with hydrogen monitors.
    Date of Issuance: April 4, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days after completion of the Spring 2005 refueling outage for 
Unit 1.
    Amendment Nos.: 344, 346 & 345.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 1, 2005 (70 FR 
5239).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 4, 2005.
    No significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of application for amendment: August 5, 2004.
    Brief description of amendment: This amendment revises Technical 
Specification Section 5.5.12, ``Primary Containment Integrity,'' to 
allow a one-time extension of its Appendix J, Type A, Containment 
Integrated Leak Rate Test interval from the current 10-year interval to 
a proposed 15-year interval.
    Date of issuance: April 12, 2005.
    Effective date: April 12, 2005, and shall be implemented within 30 
days.
    Amendment No.: 191.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 31, 2004 (69 FR 
53102).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 12, 2005.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point 
Nuclear Generating Unit No. 3, Westchester County, New York

    Date of application for amendment: June 24, 2004.
    Brief description of amendment: The amendment modifies Technical 
Specification (TS) requirements to adopt the provisions of Industry/TS 
Task Force (TSTF) change TSTF-359, ``Increased Flexibility in Mode 
Restraints.''
    Date of issuance: April 6, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 226.
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 26, 2004 (69 FR 
62474).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 6, 2005.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of application for amendment: December 30, 2004.
    Brief description of amendment: The amendment changes the frequency 
for Technical Specification surveillance requirement (SR) 3.1.4.2, 
which verifies each tested control rod scram time is within limits with 
reactor steam dome pressure >= 800 psig. Specifically, the SR frequency 
increases from 120 days to 200 days of cumulative operation in MODE 1 
(power operation).
    Date of issuance: April 5, 2005.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 283.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 1, 2005 (70 FR 
5241).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 5, 2005.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of application for amendment: September 2, 2004.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 4.5.B.2.2 to change the surveillance requirement 
frequency for air testing the drywell and suppression pool spray 
headers and nozzles from ``once per 5 years'' to ``following 
maintenance that could result in nozzle blockage.''
    Date of issuance: April 12, 2005.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 214.
    Facility Operating License No. DPR-35: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: December 21, 2004 (69 
FR 76490).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 12, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: April 13, 2004.
    Brief description of amendments: The amendments eliminate the 
requirements in Technical Specifications (TSs) associated with hydrogen 
recombiners, and hydrogen and oxygen monitors.
    Date of issuance: April 13, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 173 and 135.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the TSs.
    Date of initial notice in Federal Register: June 8, 2004 (69 FR 
32073).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 13, 2005.
    No significant hazards consideration comments received: No.

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 30, 2004.
    Brief description of amendments: The amendments modify technical 
specification (TS) requirements to adopt the provisions of Industry/TS 
Task Force (TSTF) change TSTF-359, ``Increased Flexibility in Mode 
Restraints.''
    Date of issuance: April 11, 2005.
    Effective date: As of the date of issuance, to be implemented 
within 180 days.
    Amendment Nos.: 252 and 255.
    Renewed Facility Operating License Nos. DPR-44 and DPR-56: The 
amendments revised the Technical Specifications.

[[Page 21469]]

    Date of initial notice in Federal Register: October 12, 2004 (69 FR 
60681).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 11, 2005.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit 1, Lake County, Ohio

    Date of application for amendment: August 31, 2004.
    Brief description of amendment: The amendment revised Technical 
Specification 3.4.1, ``Recirculation Loops Operating,'' associated with 
single recirculation loop operation by incorporating limits for the 
linear heat generation rate fuel thermal limit into the limiting 
condition for operation.
    Date of issuance: March 31, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 134.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 4, 2005 (70 FR 
401).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 31, 2005.
    No significant hazards consideration comments received: No.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

    Date of application for amendment: September 21, 2004.
    Brief description of amendment: The amendment deletes the Technical 
Specifications associated with hydrogen monitors.
    Date of issuance: April 5, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 216.
    Facility Operating License No. DPR-72: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 1, 2005 (70 FR 
5245).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 5, 2005.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear 
Power Plant, Kewaunee County, Wisconsin

    Date of application for amendment: July 6, 2004, as supplemented 
January 27, 2005.
    Brief description of amendment: The amendment relocates the 
calibration requirement of Table TS 4.1-1, Item 22, ``Accumulator Level 
and Pressure,'' and the surveillance requirements of Table TS 4.1-1, 
Item 25, ``Portable Radiation Survey Instruments,'' from the Technical 
Specifications to licensee-controlled documents.
    Date of issuance: April 6, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 182.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 31, 2004 (69 FR 
53112).
    The supplement dated January 27, 2005, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the Nuclear 
Regulatory Commission staff's original proposed no significant hazards 
consideration. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 6, 2005.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

    Date of application for amendments: July 23, 2004, as supplemented 
January 6, 2005.
    Brief description of amendments: The amendments modified the 
Technical Specification (TS) definition OPERABLE with respect to 
requirements for availability of normal and emergency power. 
Additionally, required actions for shutdown power TSs were modified.
    Date of issuance: April 1, 2005.
    Effective date: As of date of issuance, to be implemented within 60 
days.
    Amendment Nos.: 264 and 246.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the TSs.
    Date of initial notice in Federal Register: March 1, 2005 (70 FR 
9983).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 1, 2005.
    No significant hazards consideration comments received: Comments 
received were addressed in the Safety Evaluation dated April 1, 2005.

R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of application for amendment: July 26, 2004, as supplemented 
on March 7, 2005.
    Brief description of amendment: The amendment revised the Technical 
Specifications by eliminating the requirements to provide the NRC 
monthly operating reports and annual occupational radiation exposure 
reports.
    Date of issuance: April 13, 2005.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 89.
    Renewed Facility Operating License No. DPR-18: Amendment revised 
the Technical Specifications and/or License.
    Date of initial notice in Federal Register: October 12, 2004 (69 FR 
60685). The supplemental letter provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the staff's original proposed 
no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 13, 2005.
    No significant hazards consideration comments received: No.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego 
County, California

    Date of application for amendments: December 10, 2004.
    Brief description of amendments: These amendments delete the 
Technical Specifications associated with hydrogen monitors.
    Date of issuance: March 29, 2005.
    Effective date: March 29, 2005, to be implemented within 60 days of 
issuance.
    Amendment Nos.: Unit 2--194; Unit 3--185.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 18, 2005 (70 FR 
2896).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 29, 2005.
    No significant hazards consideration comments received: No.

[[Page 21470]]

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke 
County, Georgia

    Dates of application for amendments: February 26 and April 28, 
2008, as supplemented by letters dated July 8 and October 20, 2004.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) Section 5.6.6, Reactor Coolant System (RCS) Pressure 
Temperature Limits Report (PTLR), to facilitate future licensee-
controlled changes to the PTLR. The changes include a revised PTLR that 
provides new heatup and cooldown limits and Cold Overpressure 
Protection System (COPS) set points, and to recalculate the minimum 
size of the pressurizer power operated relief valve orifice of the RCS 
vent. In addition, the changes relocate the COPS arming temperature to 
the PTLR, and lower the COPS arming temperature from 350 [deg]F to 220 
[deg]F. The licensee also included TS bases changes to support the 
changes to the TSs.
    Date of issuance: March 28, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 136 (Unit 1) and 115 (Unit 2).
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 2004 (69 FR 
19575) and April 22, 2004 (69 FR 34707).
    The supplements dated July 8 and October 20, 2004, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 28, 2005.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: October 14, 2004.
    Brief description of amendments: The amendments eliminate the 
technical specification requirements to submit monthly operating 
reports and annual occupational radiation exposure reports.
    Date of issuance: April 5, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 300 and 289.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revised the technical specifications.
    Date of initial notice in Federal Register: February 1, 2005 (70 FR 
5250).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 5, 2005.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of application for amendment: November 8, 2004.
    Brief description of amendment: The amendment eliminates the 
requirements in Technical Specifications to submit monthly operating 
reports and annual occupational radiation exposure reports.
    Date of issuance: March 21, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days of issuance.
    Amendment No.: 57.
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: January 18, 2005 (70 FR 
2902).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 21, 2005.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units 1 and 2, Louisa County, Virginia

    Date of application for amendment: September 8, 2004.
    Brief description of amendment: These amendments delete the 
Technical Specifications associated with hydrogen recombiners and 
hydrogen monitors.
    Date of issuance: March 22, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 238 and 219.
    Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: January 18, 2005 (70 FR 
2902).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 22, 2005.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia

    Date of application for amendments: June 23, 2004.
    Brief Description of amendments: These amendments revise the 
Technical Specifications Section 3.16, ``Emergency Power System,'' 
requirements for verifying the operability of the remaining emergency 
diesel generator (EDG) when either unit's dedicated EDG or the shared 
backup EDG is inoperable.
    Date of issuance: April 5, 2005.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment Nos.: 241 and 240.
    Renewed Facility Operating License Nos. DPR-32 and DPR-37: 
Amendments change the Technical Specifications.
    Date of initial notice in Federal Register: August 19, 2004 (69 FR 
51490).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 5, 2005.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia

    Date of application for amendments: December 21, 2004.
    Brief Description of amendments: These amendments revise the 
Technical Specifications by eliminating the requirements to submit 
monthly operating reports and occupational radiation exposure reports.
    Date of issuance: March 22, 2005.
    Effective date: March 22, 2005.
    Amendment Nos.: 240 and 239.
    Renewed Facility Operating License Nos. DPR-32 and DPR-37: 
Amendments change the Technical Specifications.
    Date of initial notice in Federal Register: January 18, 2005 (70 FR 
2903).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 22, 2005.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 18th day of April 2005.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 05-8166 Filed 4-25-05; 8:45 am]
BILLING CODE 7590-01-P