Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 51378-51388 [05-16979]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 70, Number 167 (Tuesday, August 30, 2005)]
[Notices]
[Pages 51378-51388]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 05-16979]


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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 August 5, 2005, to August 18, 2005. The last 
biweekly notice was published on August 16, 2005 (70 FR 48201).

Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination. Within 60 days after the date of publication of this 
notice, the licensee may file a request for a hearing with respect to 
issuance of the amendment to the subject facility operating license and 
any person whose interest may be affected by this proceeding and who 
wishes to participate as a party in the proceeding must file a written 
request for a hearing and a petition for leave to intervene.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    Within 60 days after the date of publication of this notice, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, https://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in

[[Page 51379]]

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 basis for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, HearingDocket@nrc.gov; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

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

    Date of amendment request: July 14, 2005.
    Description of amendment request: The proposed change would modify 
the Millstone Power Station, Unit No. 2 reactor coolant system (RCS) 
heatup and cooldown limits Technical Specification (TS) 3.4.9.1, 
``Reactor Coolant System''. The associated TS bases will be updated to 
address the proposed 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 amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes are a result of the new analysis of the RCS 
P-T [pressure-temperature] limits and associated heatup/cooldown 
rates. These changes will support plant operation to 54 EFPY 
[effective full-power years] and provide flexibility during plant 
heatup and cooldown, especially during equipment manipulations such 
as securing RCPs [reactor coolant pumps], swapping SDC [shutdown 
cooling] heat exchangers, and initiating SDC.
    The hydrostatic and leak test limit will now be administratively 
controlled by the heatup limit. Administratively limiting 
hydrostatic and leak tests to the heatup limit provides additional 
margin to the Appendix G requirements. Table 3.4-2 has been modified 
to remove the Inservice Hydrostatic and Leak Testing item and to add 
a note indicating heatup limitations also apply to hydrostatic and 
leak test conditions. The requirement to remain isothermal (rate `` 
5 [deg]F/hour) for 1 hour prior [to] and during hydrostatic and leak 
test [s] above the heatup curve is no longer needed as operation 
above the heatup curve is no longer allowed.
    The proposed changes to the RCS P-T limits and rates of 
temperature change are based on the new analysis. This analysis uses 
standard approved methods that ensure the margins of safety required 
by 10 CFR 50, Appendix G are maintained. The other changes discussed 
are more restrictive enhancements to technical specification 
requirements. Therefore, the proposed changes will not result in 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 changes will not alter the plant configuration (no 
new or different type of equipment will be installed) or require any 
new or unusual operator actions. They do not alter the way any 
structure, system, or component functions. The increased heatup and 
cooldown rates are bounded by the existing accident analysis. The 
proposed changes do not introduce any new failure modes. Therefore, 
the proposed changes 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?

[[Page 51380]]

    Response: No.
    The proposed changes will modify the RCS P-T limits, and the RCS 
heatup and cooldown rate limits. The proposed changes are being made 
as a result of the new P-T and LTOP [low-temperature overpressure 
protection] analyses performed. The new P-T curves and heatup and 
cooldown rates are developed in accordance with the requirements and 
methods described in 10 CFR 50 Appendix G and are consistent with 
the criteria contained in the Standard Review Plan Section 5.3.2. 
This will ensure the integrity of the reactor vessel is maintained 
during all aspects of plant operation. Therefore, there is no 
significant effect on the probability or consequences of any 
accident previously evaluated and no significant impact on offsite 
doses associated with previously evaluated accidents. This license 
amendment request does not result in a reduction of the margin of 
safety as defined in the bases for the technical specifications 
addressed by the proposed changes.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Section Chief: Darrell J. Roberts.

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

    Date of amendment request: May 24, 2005.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification allowances to bypass the rod worth 
minimizer consistent with previously-approved standards.
    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 special operation allowances do not 
involve the modification of any plant equipment or affect basic 
plant operation. The relevant design basis accident is the control 
rod drop accident (CRDA), which involves multiple failures to 
initiate the event. Control rod decoupling and remaining stuck full-
in while its drive mechanism is withdrawn are required initiators. 
The proposed special operations have no adverse impact on control 
rod coupling or control rod performance. As such, there is no 
significant increase in the probability of an accident previously 
evaluated.
    The CRDA analysis consequences and related initial conditions 
remain unchanged when invoking the proposed special operation 
allowance. The control rod withdrawal sequence is assumed to limit 
individual control rod worths as another initial condition for the 
CRDA. However, consistent with existing requirements for control rod 
withdrawal operations, all control rod withdrawal sequences are 
analyzed to meet this criterion and are implemented under the 
control of the rod worth minimizer or by independent verification by 
a second licensed operator or other qualified member of the 
technical staff. The consequences of analyzed events are therefore 
not affected. Therefore, the proposed change does not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No. The proposed change does not involve any physical 
alteration of plant equipment and does not change the method by 
which any safety-related system performs its function. As such, no 
new or different types of equipment will be installed, and the basic 
operation of installed equipment is unchanged. The methods governing 
plant operation and testing remain consistent with current safety 
analysis assumptions. Therefore, the proposed change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No. The proposed special operation allowances do not 
involve the modification of any plant equipment or affect basic 
plant operation. The relevant design basis accident is the control 
rod drop accident (CRDA), which involves multiple failures to 
initiate the event. Additionally, CRDA analysis consequences and 
related initial conditions remain unchanged when invoking the 
proposed special operation allowance. These changes do not negate 
any existing requirement, and do not adversely affect existing plant 
safety margins or the reliability of the equipment assumed to 
operate in the safety analysis. As such, there are no changes 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. 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: J.M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts, 02360-5599.
    NRC Section Chief: Darrell Roberts.

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

    Date of amendment request: May 24, 2005.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) applicability requirements related 
to primary containment oxygen concentration and drywell-to-suppression 
chamber differential pressure 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. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No. The proposed applicability and associated default 
actions being revised do not involve the modification of any plant 
equipment or affect basic plant operation. Additionally, the 
associated limitations are not assumed to be an initiator of any 
analyzed event. Therefore, the proposed change does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    The limits imposed by the associated specifications remain 
unchanged. The consequences of analyzed events are therefore not 
affected. Brief periods where the requirements for maintaining these 
limits are relaxed are currently considered in the TS and associated 
licensing basis. The proposed change clarifies and modifies the 
definition of these periods, however, any changes are not considered 
significant and are supported by remaining [definitions] consistent 
with the recommended allowances of NUREG-1433, Rev. 3, ``Standard 
Technical Specifications, General Electric Plants, BWR [boiling-
water reactor]/4.'' Therefore, the proposed change does not involve 
a significant increase in the consequences of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No. The proposed change does not involve any physical 
alteration of plant equipment and does not change the method by 
which any safety-related system performs its function. As such, no 
new or different types of equipment will be installed, and the basic 
operation of installed equipment is unchanged. The methods governing 
plant operation and testing remain consistent with

[[Page 51381]]

current safety analysis assumptions. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No. The proposed applicability and associated default 
actions being revised do not involve the modification of any plant 
equipment or affect basic plant operation. Additionally, the 
associated limitations remain unchanged. These changes do not negate 
any existing requirement, and do not adversely affect existing plant 
safety margins or the reliability of the equipment assumed to 
operate in the safety analysis. As such, there are no changes 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. The revised plant conditions reflecting the 
applicability and the duration allowed to restore limits are not 
credited in any design basis event. These changes do not reflect any 
significant adverse impact to the overall risk of operating during 
brief periods without the required primary containment oxygen 
concentration since the total time for any occurrence is only 
marginally extended and reflects times recommended by NUREG-1433. 
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: J.M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts, 02360-5599.
    NRC Section Chief: Darrell Roberts.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

    Date of amendment request: June 27, 2005.
    Description of amendment request: The proposed amendment revises 
the Facilities Operating License to change technical specification (TS) 
3.6.1.3, Required Actions A.1 and B.1, to add closed relief valves as 
acceptable isolation devices provided that the relief setpoint is 
greater than 1.5 times containment design pressure.
    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.
    Primary Containment Isolation Valves (PCIVs) are accident 
mitigating features designed to limit releases from the containment 
following an accident. The Technical Specifications (TS) specify 
actions to be taken to preserve the containment isolation function 
if a PCIV is inoperable. These actions include isolating the 
penetration flow path by specific methods. The proposed TS change 
adds closed relief valves with acceptable relief setpoints as 
another method to isolate the penetration flowpath. The use of 
relief valves with relief setpoints greater than 1.5 times the 
containment design pressure meets the Standard Review Plan options 
for acceptable isolation devices. This relief setpoint provides [a] 
sufficient margin to minimize the potential for premature opening 
due to containment post-accident pressures. The proposed change does 
not affect any initiators to accidents previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not introduce any new modes of plant 
operation or adversely affect the design function or operation of 
safety features. The proposed TS change allows use of existing plant 
equipment as compensatory measures to maintain the containment 
isolation design intent when the normal isolation features are 
inoperable. Since relief valves used for this purpose will not be 
disabled by blind flanges, the system piping overpressure protection 
design feature will also be preserved.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The safety margin associated with this change is that associated 
with preserving the containment integrity. NUREG-0800, the Standard 
Review Plan, recognizes that relief valves with relief setpoints 
greater than 1.5 times containment design pressure are acceptable as 
containment isolation devices. Closed relief valves with relief 
setpoints of this margin provide an isolation alternative that is 
less susceptible to a single failure (i.e., inadvertent opening) yet 
still preserves the overpressure protection that the component was 
intended to provide.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1700 K Street, NW., Washington, DC 20006-3817.
    NRC Section Chief: David Terao.

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of amendment request: June 15, 2005.
    Description of amendment request: The proposed change revises 
Technical Specification (TS) 3.3.2.2 ``Feedwater System and Main 
Turbine High Water Level Trip Instrumentation,'' to reflect a design 
change to the instrumentation logic.
    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 revises TS 3.3.2.2 to reflect a design 
change to the instrumentation logic that trips the three feedwater 
pumps and main turbine. The design change will add a redundant high 
reactor water level trip channel to both trip systems. The Feedwater 
System and main turbine high water level trip is credited in the 
QCNPS [Quad Cities Nuclear Power Station] accident analysis to 
function during an increase in feedwater flow transient. 
Specifically, the instrumentation and associated trip limits the 
reactor water level increase resulting from a feedwater controller 
failure during maximum flow demand, thus preventing a nuclear fuel 
minimum critical power ratio violation associated with increased 
subcooling and resultant pressure transient. Additionally, this trip 
function prevents excessive water inventory from entering the main 
steam system and damaging steam-handling equipment.
    TS requirements that govern operability or routine testing of 
plant instruments are not assumed to be initiators of any analyzed 
event because these instruments are intended to detect, prevent, or 
mitigate accidents. The Feedwater System and main turbine trip 
instrumentation serves to mitigate transients that result in 
increased reactor water level. The trip instrumentation associated 
with the proposed changes and design change are independent from the 
instrumentation and logic used in the Feedwater Control System and 
Turbine Control System. Therefore, the proposed change does not 
involve a

[[Page 51382]]

significant increase in the probability of an accident previously 
evaluated.
    The proposed design change to add a redundant high reactor water 
level trip channel to both trip systems, and the associated TS 
changes, do not adversely impact the instrumentation's ability to 
perform the functions described above. The design change will 
utilize installed spare trip units and relay contacts of the same 
design as those presently credited to meet TS 3.3.2.2 requirements. 
The method in which the reactor water level is sensed and the 
reactor water level setpoints at which a trip is initiated are not 
impacted. The instrumentation response times and the instrumentation 
output to the equipment being tripped remains the same. Therefore, 
the proposed change does not involve a significant increase in the 
consequences of an accident previously evaluated. Furthermore, there 
will be no change in the types or significant increase in the 
amounts of any effluents released offsite.
    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 does not alter the parameters within which 
the plant is operated. There are no setpoints at which protective or 
mitigative actions are initiated that are affected by the proposed 
change. This proposed change will not alter the manner in which 
equipment operation is initiated nor will the demands on mitigating 
equipment be changed. The proposed change to TS 3.3.2.2 adds 
redundant instrumentation to improve system reliability, and 
increase maintenance and testing flexibility. The instrumentation 
being added to the trip logic utilizes the same transmitters, and 
the same type of trip units and trip relays, as presently used to 
monitor reactor water level and initiate Emergency Core Cooling 
System operation.
    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?
    Margins of safety are established in the design of components, 
the configuration of components to meet certain performance 
parameters, and in the establishment of setpoints to initiate alarms 
or actions. The proposed amendment supports a change to the logic 
that trips the three feedwater pumps and the main turbine from a 
two-out-of-two initiation logic to a one-out-of-two twice initiation 
logic. The proposed amendment does not alter the setpoints at which 
the trip function occurs, the response time of the trip initiation 
logic, or the plant response following a valid trip signal. The 
proposed changes to the TS 3.3.2.2 Required Actions and Completion 
Times are consistent with other instrumentation TS that incorporate 
a one-out-of-two twice initiation logic.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
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.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of amendment request: August 4, 2005.
    Description of amendment request: The proposed amendment would 
revise Hope Creek Generating Station Technical Specification 3.7.1.3, 
``Ultimate Heat Sink,'' to allow a 24-hour average temperature to be 
used if ultimate heat sink temperature exceeds 89.5 [deg]F provided the 
ultimate heat sink temperature or safety auxiliary cooling system 
temperature does not exceed 95 [deg]F.
    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 ultimate heat sink (UHS) is not an accident indicator. An 
increase in UHS temperature will not increase the probability of 
occurrence of an accident. The proposed change will allow plant 
operation to continue if temperature of the UHS exceeds 89.5 [deg]F 
provided that UHS temperature averaged over the previous 24-hour 
period is less than 89.5 [deg]F and the UHS temperature and safety 
auxiliary cooling system (SACS) temperatures do not exceed 95 
[deg]F. Maintaining these temperatures less than or equal to 95 
[deg]F ensures that accident mitigation equipment will continue to 
perform its required function.
    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 of 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change will not install any new or different 
equipment or modify equipment in the plant. The proposed change will 
not alter the operation or function of structures, systems or 
components. The response of the plant and the operators following a 
design basis accident is unaffected by this change. The proposed 
change does not introduce any new failure modes and the design basis 
heat removal capability of the safety related components is 
maintained at the increased UHS temperature limit.
    Therefore, the proposed chage will not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    The increase to the UHS temperature will not adversely affect 
design basis accident mitigation equipment. Ensuring that SACS 
temperature remains below 95 [deg]F when UHS is above 89.5 [deg]F 
ensures that heat removal capability is within the current analyzed 
limits. Accident mitigation equipment will continue to function as 
assumed in the accident analysis. Therefore, the proposed change 
does not involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for Licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: Darrell J. Roberts.

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

[[Page 51383]]

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, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

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

    Date of application for amendment: December 17, 2004.
    Brief description of amendment: The amendment revised Appendix B, 
Environmental Protection Plan (non-radiological) of the Clinton 
Facility Operating License.
    Date of issuance: August 9, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 166.
    Facility Operating License No. NPF-62: The amendment revised the 
Environmental Protection Plan.
    Date of initial notice in Federal Register: April 12, 2005 (70 FR 
19112).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 9, 2005.
    No significant hazards consideration comments received: No.

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

    Date of application for amendment: October 20, 2004.
    Brief description of amendment: The amendment revised Table 4.1-1, 
``Instrument Surveillance Requirements,'' of the Technical 
Specifications and associated Bases to extend the functional testing 
surveillance interval from monthly to a semi-annual interval for 
reactor trip system instrumentation channels, and from the current 
monthly to quarterly for the reactor trip devices.
    Date of issuance: August 11, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 255.
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 23, 2004 (69 
FR 68181)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 11, 2005.
    No significant hazards consideration comments received: No.

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

    Date of application for amendment: June 24, 2004, as supplemented 
February 24, 2005.
    Brief description of amendment: The amendment revised TMI-1 
Technical Specification (TS) 4.0.2 to adopt the provisions of Technical 
Specification Task Force (TSTF) Traveller TSTF-358, Revision 6, 
revising the required actions and time constraints regarding missed 
surveillances. The amendment also added a new Section 6.18 to the TSs 
incorporating a Technical Specifications Bases Control Program.
    Date of issuance: August 12, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 256.
    Facility Operating License No. DPR-50. Amendment revised the TSs.
    Date of initial notice in Federal Register: March 1, 2005 (70 FR 
9987).
    The supplement dated February 24, 2005, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the NRC 
staff's original proposed no significant hazards consideration 
determination . The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 12, 2005.
    No significant hazards consideration comments received: No.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of application for amendment: June 21, 2004.
    Brief description of amendment: The amendment changes Technical 
Specification Section 5.5.14, ``Technical Specifications (TS) Bases 
Control Program,'' to incorporate changes in Section 50.59 of Title 10 
of the Code of Federal Regulations terminology. The amendment also 
revises Section 5.7.1, ``High Radiation Area,'' by adding wording that 
was inadvertently deleted with the issuance of the Improved Standard 
Technical Specifications in Amendment No. 176.
    Date of issuance: August 2, 2005.
    Effective date: August 2, 2005.
    Amendment No.: 205.
    Renewed Facility Operating License No. DPR-23. Amendment revises 
the Technical Specifications.
    Date of initial notice in Federal Register: August 31, 2004 (69 FR 
53101).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 2, 2005.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power 
Station, Unit No. 3, New London County, Connecticut

    Date of application for amendment: December 23, 2004.
    Brief description of amendment: The amendment relocated certain 
Technical Specifications (TSs) to the Millstone Power Station, Unit No. 
3 Technical Requirements Manual.
    Date of issuance: August 11, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 225.
    Facility Operating License No. NPF-49: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: May 24, 2005 (70 FR 
29788).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 11, 2005.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: September 30, 2004, as supplemented by 
letters dated April 26 and June 8, 2005.

[[Page 51384]]

    Brief description of amendment: The amendment changes the existing 
steam generator (SG) tube surveillance program to be consistent with 
that being proposed by the Technical Specification (TS) Task Force 
(TSTF) in TSTF-449. These changes revise definitions in TS 1.1, reactor 
coolant system operational leakage in TS 3.4.13, SG program in TS 
5.5.9, and SG tube inspection reports in TS 5.6.7, and add a new TS 
3.4.16 on SG tube integrity. Also, as a result of the licensee 
replacing the SGs with SGs having a new Alloy 690 thermally treated 
tubing design, the TSs are revised to reflect this replacement.
    Date of issuance: August 10, 2005.
    Effective date: As of the date of issuance and shall be implemented 
prior to resumption of operation from the 1R19 refueling outage 
scheduled for the fall of 2005.
    Amendment No.: 224.
    Renewed Facility Operating License No. DPR-51: Amendment revised 
the Technical Specifications.
    Date of notices in Federal Register: November 9, 2004 (69 FR 64987) 
and May 24, 2005 (70 FR 29790). The supplement dated June 8, 2005, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 10, 2005.
    No significant hazards consideration comments received: No.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont.

    Date of application for amendment: December 6, 2004, as 
supplemented on June 14, 2005.
    Brief description of amendment: The amendment makes administrative 
and other miscellaneous changes to the Facility Operating License (FOL) 
and Technical Specifications (TSs) including correction of references 
and deleting obsolete or redundant TS requirements and surveillances.
    Date of Issuance: August 15, 2005.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 226.
    Facility Operating License No. DPR-28: The amendment revised the 
FOL and TSs.
    Date of initial notice in Federal Register: January 18, 2005 (70 FR 
2888). The supplement contained clarifying information only, and did 
not change the initial no significant hazards consideration 
determination or expand the scope of the initial Federal Register 
notice.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated August 15, 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: September 15, 2004.
    Brief description of amendments: The amendments deleted the 
Technical Specification requirements to maintain hydrogen recombiners 
and hydrogen/oxygen monitors and related Surveillance Requirements.
    Date of issuance: August 11, 2005.
    Effective date: As of the date of issuance, to be implemented 
within 120 days.
    Amendments Nos.: 256 and 259.
    Renewed Facility Operating License Nos. DPR-44 and DPR-56: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 1, 2005, (70 
FR 5244). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 11, 2005.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of application for amendment: April 22, 2005.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) related to fuel handling and storage. 
Specifically, the changes revised TS 3/4.9.11, ``Storage Pool Water 
Level,'' TS 3/4.9.12, ``Storage Pool Ventilation,'' TS 3/4.9.13, 
``Spent Fuel Assembly Storage,'' and TS 5.6, ``Fuel Storage,'' to 
reflect that spent fuel storage racks are no longer installed in the 
cask pit or transfer pit. Fuel storage racks were permitted to be 
temporarily installed in the cask pit and transfer pit during a project 
to increase spent fuel pool (SFP) storage capacity. All temporarily 
installed fuel storage racks have now been moved into the SFP. 
Additionally, the changes relocated the requirements of TS 3/4.9.7, 
``Crane Travel--Fuel Handling Building,'' to the Davis-Besse Nuclear 
Power Station Technical Requirements Manual. The changes to TS 3/4.9.13 
and TS 5.6 also reflected that there are no longer low density fuel 
storage racks in the SFP. The changes made TS requirements consistent 
with the current fuel storage design.
    Date of issuance: August 16, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 266.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 24, 2005 (70 FR 
29795).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 16, 2005.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of application for amendment: February 22, 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: August 16, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 267.
    Facility Operating License No. NPF-3: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 10, 2005 (70 FR 
24651).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 16, 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: February 22, 2005.
    Brief description of amendment: The amendment revised the Technical 
Specifications by eliminating the requirements to submit the NRC 
monthly operating reports and annual occupational radiation exposure 
reports.

[[Page 51385]]

    Date of issuance: August 16, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 136.
    Facility Operating License No. NPF-58: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 10, 2005 (70 FR 
24651).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 16, 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: October 14, 2004.
    Brief description of amendment: The amendment revises a technical 
specification surveillance requirement to change the required frequency 
of the reactor building spray nozzle surveillance from once every 10 
years to ``following maintenance that could result in nozzle 
blockage.''
    Date of issuance: August 4, 2005.
    Effective date: August 4, 2005.
    Amendment No.: 219.
    Facility Operating License No. DPR-72: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: January 18, 2005 (70 FR 
2891).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 4, 2005.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of application for amendments: February 25, 2005, as 
supplemented June 2, 2005.
    Brief description of amendments: The amendments modify the 
Technical Specifications by revising the near-end-of-life moderator 
temperature coefficient (MTC) surveillance requirement by placing a set 
of conditions on core performance, which, if met, would allow 
conditional exemption from the required MTC measurement.
    Date of issuance: August 8, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 288, 270.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 29, 2005 (70 FR 
15943).
    The supplement dated June 2, 2005, provided additional information 
that clarified the application, did not expand the scope of the 
application as originally noticed, and did not change the NRC staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register. The Commission's related evaluation 
of the amendments is contained in a Safety Evaluation dated August 8, 
2005.
    No significant hazards consideration comments received: No.

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 application for amendments: September 1, 2004, as 
supplemented by letter dated May 17, 2005.
    Brief description of amendments: The amendments approve the use of 
Generation of Thermal-Hydraulic Information Containment Version 7.1 
patch 1 (GOTHIC), for licensing analyses for the Prairie Island Nuclear 
Generating Plants to (1) evaluate the short-term peak pressure and 
temperature response of the containment atmosphere to large pipe breaks 
in high energy piping systems--the design-basis loss-of-coolant 
accident (LOCA) and the design-basis main steam line break, and (2) to 
evaluate the long-term containment response following a design-basis 
LOCA.
    Date of issuance: August 12, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 171,161.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Updated Safety Analysis Report.
    Date of initial notice in Federal Register: September 28, 2004 (69 
FR 57990).
    The supplemental letter contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 12, 2005.
    No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

    Date of application for amendments: March 5, 2004.
    Brief description of amendments: The amendments revised Technical 
Specifications Surveillance Requirement (SR) 3.6.4.1.3 to require that 
only one secondary containment access door in each access opening be 
verified closed. In addition, SR 3.6.4.1.3 allows entry and exit access 
between required secondary containment zones that have a single door.
    Date of issuance: August 16, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 224 and 201.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 27, 2004 (69 FR 
22882).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 16, 2005.
    No significant hazards consideration comments received: No.

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 application for amendments: September 8, 2004.
    Brief description of amendments: The amendments revised the SSES 1 
and 2 Technical Specification 3.8.7, ``Distribution Systems-
Operating,'' to add an action note to address the potential for 
deenergized Class 1E battery chargers, and correct three unrelated 
editorial changes.
    Date of issuance: August 17, 2005.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment Nos.: 225 and 202.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 24, 2005 (70 FR 
29798).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 17, 2005.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of application for amendment: October 1, 2004.
    Brief description of amendment: This amendment deleted the 
Technical Specifications (TSs) associated with hydrogen recombiners, 
and hydrogen and oxygen monitors.
    Date of issuance: August 9, 2005.

[[Page 51386]]

    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 160.
    Facility Operating License No. NPF-57: This amendment revised the 
TSs.
    Date of initial notice in Federal Register: March 15, 2005 (70 FR 
12749).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 9, 2005.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: March 15, 2005.
    Brief description of amendments: The amendments revise the 
Technical Specification 3.7.10, ``Control Room Emergency Filtration/
Pressurization System (CREFS).'' The revision allows a one-time 
extension from 24 hours to 14 days of the allowable duration of 
operation with control room boundary inoperable.
    Date of issuance: August 11, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: 120, 120.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 24, 2005 (70 FR 
29801).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 11, 2005.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    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 for the 
amendment 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.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective 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.