Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 60032-60041 [E7-20679]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 72, Number 204 (Tuesday, October 23, 2007)]
[Notices]
[Pages 60032-60041]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E7-20679]


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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 September 27, 2007, to October 10, 2007. The 
last biweekly notice was published on October 9, 2007 (72 FR 57352).

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,

[[Page 60033]]

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

[[Page 60034]]

the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or by e-
mail to pdr@nrc.gov.

Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power 
Station, Kewaunee County, Wisconsin

    Date of amendment request: September 24, 2007.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to add a reference to 
Dominion Topical Report DOM-NAF-5, ``Application of Dominion Nuclear 
Core Design and Safety Analysis Methods to the Kewaunee Power Station 
(KPS),'' to the list of approved analytical methods. The proposed 
changes would permit the application of the Dominion nuclear core 
design and safety analysis methods, including the methodology to 
perform core thermal-hydraulic analysis to predict critical heat flux 
and departure from nucleate boiling ratio for the Westinghouse 422 V+ 
fuel design. The proposed amendment would also: (1) Accommodate the use 
of the methodologies proposed in DOM-NAF-5, (2) delete one approved 
analytical method that will no longer be used, and (3) delete date and 
revision numbers from the current TS list of approved analytical 
methods, consistent with TS Task Force (TSTF) Change Traveler TSTF-363-
A, Revision 0, ``Revise Topical Report References in ITS [improved TSs] 
5.6.5, COLR [Core Operating Limits Report],'' dated August 4, 2003.
    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 analysis methods of DOM-NAF-5 do not make any contribution 
to the potential accident initiators and thus do not increase the 
probability of any accident previously evaluated. The use of the 
approved Dominion analysis methodologies will not increase the 
probability of an accident because plant systems, structures, and 
components (SSC) will not be affected or operated in a different 
manner, and system interfaces will not change.
    Since the applicable safety analysis and nuclear core design 
acceptance criteria will be satisfied when the Dominion analysis 
methods are applied to KPS, the use of the approved Dominion 
analysis methods does not increase the potential consequences of any 
accident previously evaluated. The use of the approved Dominion 
methods will not result in a significant impact on normal operating 
plant releases, and will not increase the predicted radiological 
consequences of postulated accidents described in the USAR [updated 
safety analysis report].
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or the consequences of any accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different type of accident from any accident previously 
evaluated?
    Response: No.
    The use of Dominion analysis methods and the Dominion 
statistical design limit (SDL) for fuel departure from nucleate 
boiling ratio (DNBR) and fuel critical heat flux (CHF) does not 
impact any of the applicable core design criteria. All pertinent 
licensing basis limits and acceptance criteria will continue to be 
met. Demonstrated adherence to these limits and acceptance criteria 
precludes new challenges to SSCs that might introduce a new type of 
accident. All design and performance criteria will continue to be 
met and no new single failure mechanisms will be created. The use of 
the Dominion methods does not involve any alteration to plant 
equipment or procedures that might introduce any new or unique 
operational modes or accident precursors.
    Therefore, the proposed amendment does not create 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.
    Nuclear core design and safety analysis acceptance criteria will 
continue to be satisfied with the application of Dominion methods. 
Meeting the analysis acceptance criteria and limits ensures that the 
margin of safety is not significantly reduced. Nuclear core design 
and safety analysis acceptance criteria will continue to be 
satisfied with the application of Dominion methods. In particular, 
use of VIPRE-D with the proposed SDL provides at least a 95% 
probability at a 95% confidence level that DNBR will not occur (the 
95/95 DNBR criterion). The required DNBR margin of safety for KPS, 
which is the margin between the 95/95 DNBR criterion and clad 
failure, is therefore not reduced.
    Therefore, the proposed amendment 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: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar Street, Richmond, VA 23219.
    NRC Acting Branch Chief: Travis L. Tate.

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

    Date of amendment request: August 28, 2007.
    Brief description of amendments: Revision to the Operating License 
and Technical Specification (TS) 1.0, ``Use and Application, and TS 
3.7.17'', ``Spent Fuel Assembly Storage,'' to Revise Rated Thermal 
Power from 3458 megawatts thermal (MWt) to 3612 MWt.
    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 impacts of the proposed Stretch Power Uprate (SPU) on plant 
systems, structures, and components (SSCs) were reviewed with 
respect to SSC design capability, and it was determined that 
following completion of plant changes to support the SPU, no system, 
structure, or component would exceed its design conditions or 
limits. Evaluations supporting those conclusions were performed 
consistent with proposed Technical Specification changes. 
Consequently, equipment reliability and structural integrity will 
not be adversely affected. Control system studies demonstrated that 
plant response to operational transients under SPU conditions will 
not significantly increase reactor trip frequency, so there will be 
no significant increase in the frequency of SSC challenges caused by 
reactor trip.
    New systems are not needed to implement the SPU, and new 
interactions among SSCs are not created. The SPU does not create new 
failure modes for existing SSCs. Modified components do not 
introduce new failure modes relative to those of the components in 
their pre-modified condition. Consequently, new initiators of 
previously analyzed accidents are not created.
    The fission product barriers--fuel cladding, reactor coolant 
pressure boundary, and the containment building--remain unchanged. 
The spectrum of previously analyzed postulated accidents and 
transients was evaluated, and effects on the fuel, the reactor 
coolant pressure boundary, and the containment were determined. 
These analyses were performed consistent with the proposed Technical 
Specification changes. The results demonstrate that existing reactor 
coolant pressure boundary and containment limits are met and that 
effects on the fuel are such that dose consequences meet existing 
criteria at SPU conditions.
    There is no increase in the probability of an accident 
concerning the potential insertion of a fuel assembly in an 
incorrect location in the Spent Fuel Pool Region I/

[[Page 60035]]

Region II racks as a result of the specified storage patterns. 
Luminant Power [Luminant Generation Company LLC] has used 
administrative controls to move fuel assemblies from location to 
location since the initial receipt of fuel on site. Fuel assembly 
placement will continue to be controlled pursuant to approved fuel 
handling procedures and in accordance with the Technical 
Specification for spent fuel rack storage configuration limitations.
    Therefore, the proposed change does 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.
    New systems are not required to implement the SPU, and new 
interactions among SSCs are not created. The SPU does not create new 
failure modes for existing SSCs. Modified components do not 
introduce failures different from those of the components in their 
pre-modified condition. Consequently, no new or different accident 
sequences arise from SSC interactions or failures.
    Training will be provided to address SPU effects, and the 
plant's simulator will be updated consistent with SPU conditions. 
Operating procedure changes are minor and do not result in any 
significant changes in operating philosophy. For these reasons, the 
SPU does not introduce human performance issues that could create 
new accidents or different accident sequences.
    The increase in power level does not create new fission product 
release paths. The fission product barriers (fuel cladding, reactor 
coolant pressure boundary, and the containment building) remain 
unchanged.
    The potential for criticality in the spent fuel pool is not a 
new or different type of accident. The potential criticality 
accidents have been reanalyzed to demonstrate that the pool remains 
subcritical.
    Therefore, the proposed change does 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.
    Structural evaluations performed at SPU conditions demonstrated 
that calculated loads on affected SSCs remain within their design 
for all design basis event categories. American Society of 
Mechanical Engineers (ASME) Code fatigue limits continue to be met.
    Fuel performance evaluations were performed using parameter 
values appropriate for a reload core operating at SPU conditions. 
Those evaluations demonstrate that fuel performance acceptance 
criteria continue to be met. Loss of Coolant Accident (LOCA) and 
non-LOCA safety analyses were performed assuming SPU conditions and 
consistent with the proposed Technical Specification change. 
Emergency core cooling system performance was shown to meet the 
criteria of 10 CFR 50.46. The non-LOCA events identified in the 
Final Safety Analysis Report (FSAR) Chapter 15 were shown to meet 
existing acceptance criteria.
    The containment building response to mass and energy releases 
was evaluated assuming SPU conditions. The evaluations showed that 
temperature and pressure limits were met.
    No plant changes associated with the SPU reduce the degree of 
component or system redundancy. Existing Technical Specification 
operability and surveillance requirements are not reduced by the 
proposed changes.
    The proposed fuel storage requirements in Technical 
Specification 3.7.17 will provide adequate margin to assure that the 
fuel storage array (Region I and Region II) will always remain 
subcritical by the 5% margin recommended by the Nuclear Regulatory 
Commission (NRC).
    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: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Branch Chief: Thomas G. Hiltz.

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

    Date of amendment request: August 16, 2007.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) requirements related to control 
room envelope habitability in TS 3.7.9, ``Control Room Emergency Air 
Treatment System (CREATS),'' and TS section 5.5, ``Programs and 
Manuals.'' The changes are consistent with the Nuclear Regulatory 
Commission approved Industry/Technical Specification Task Force (TSTF)-
448, Revision 3. The availability of this TS improvement was published 
in the Federal Register on January 17, 2007, as part of the 
consolidated line item improvement process.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration adopted by the licensee 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 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) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change revises the TS for the CRE emergency 
ventilation system, which is a mitigation system designed to 
minimize unfiltered air leakage into the CRE and to filter the CRE 
atmosphere to protect the CRE occupants in the event of accidents 
previously analyzed. An important part of the CRE emergency 
ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated. Therefore, the probability of any accident 
previously evaluated is not increased. Performing tests to verify 
the operability of the CRE boundary and implementing a program to 
assess and maintain CRE habitability ensure that the CRE emergency 
ventilation system is capable of adequately mitigating radiological 
consequences to CRE occupants during accident conditions, and that 
the CRE emergency ventilation system will perform as assumed in the 
consequence analyses of design basis accidents. Thus, the 
consequences of any accident previously evaluated are not increased. 
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 impact the accident analysis. The 
proposed change does not alter the required mitigation capability of 
the CRE emergency ventilation system, or its functioning during 
accident conditions as assumed in the licensing basis analyses of 
design basis accident radiological consequences to CRE occupants. No 
new or different accidents result from performing the new 
surveillance or following the new program. The proposed change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a significant 
change in the methods governing normal plant operation. The proposed 
change does not alter any safety analysis assumptions and is 
consistent with current plant operating practice. Therefore, this 
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 change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without

[[Page 60036]]

compensatory measures. The proposed change does 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 analysis adopted by the licensee 
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: Daniel F. Stenger, Ballard Spahr Andrews & 
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC 
20005.
    NRC Branch Chief: Mark G. Kowal.

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

    Date of amendment requests: September 24, 2007.
    Description of amendment requests: Approval of the revision to the 
San Onofre Nuclear Generating Station Unit 3 Technical Specification 
5.5.2.15, ``Containment Leakage Rate Testing Program.'' The request is 
for a one-time extension from the currently approved 15-year interval 
since the last Integrated Leak Rate Test (ILRT) to a 16-year interval 
since the last ILRT.
    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 revision to Technical Specifications adds a one 
time extension to the current interval for Type A testing (10 CFR 
50, Appendix J, Option B, Integrated Leak Rate Testing). The current 
test interval of 15 years, based on past performance, would be 
extended on a one time basis to 16 years from the last Type A test. 
The proposed extension to Type A testing does not involve a 
significant increase in the probability or consequences of an 
accident since research documented in NUREG-1493, ``Performance-
Based Containment System Leakage Testing Requirements,'' September 
1995, has found that, generically, very few potential containment 
leakage paths are not identified by Type B and C tests. The NUREG 
concluded that reducing the Type A testing frequency to once per 
twenty years was found to lead to an imperceptible increase in risk. 
A high degree of assurance is provided through testing and 
inspection that the containment will not degrade in a manner 
detectable only by Type A testing. The most recent Type A test at 
Unit 3 shows leakage to be below acceptance criteria, indicating a 
leak tight containment. Inspections required by the American Society 
of Mechanical Engineers (ASME) Code Section Xl (Subsections IWE and 
IWL) and maintenance rule monitoring (10 CFR 50.65, ``Requirements 
for Monitoring the Effectiveness of Maintenance at Nuclear Power 
Plants) are performed in order to identify indications of 
containment degradation that could affect leak tightness. Type B and 
C testing required by Technical Specifications will identify any 
containment opening such as valves that would otherwise be detected 
by the Type A tests. These factors show that a Type A test extension 
will not represent a significant increase in the consequences of an 
accident.
    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 revision to Technical Specifications adds a one 
time extension to the current interval for Type A testing (10 CFR 
50, Appendix J, Option B, Integrated Leak Rate Testing). The current 
test interval of 16 years, based on past performance, would be 
extended on a one time basis to 16 years from the last Type A test. 
The proposed extension to Type A testing cannot create the 
possibility of a new or different type of accident since there are 
no physical changes being made to the plant and there are no changes 
to the operation of the plant that could introduce a new failure 
mode creating an accident or affecting the mitigation of an 
accident. 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 revision to Technical Specifications adds a one 
time extension to the current interval for Type A testing (10 CFR 
50, Appendix J, Option B, Integrated Leak Rate Testing). The current 
test interval of 15 years, based on past performance, would be 
extended on a one time basis to 16 years from the last Type A test. 
The proposed extension to Type A testing will not significantly 
reduce the margin of safety. The NUREG 1493, ``Performance-Based 
Containment System Leakage Testing Requirements,'' September 1995, 
generic study of the effects of extending containment leakage 
testing found that a 20 year extension in Type A leakage testing 
resulted in an imperceptible increase in risk to the public. NUREG 
1493 found that, generically, the design containment leakage rate 
contributes about 0.1 percent to the individual risk and that the 
decrease in Type A testing frequency would have a minimal [e]ffect 
on this risk since 95% of the potential leakage paths are detected 
by Type C testing. Regular inspections required by the American 
Society of Mechanical Engineers (ASME) Code Section Xl (Subsections 
IWE and IWL) and maintenance rule monitoring (10 CFR 50.65, 
``Requirements for Monitoring the Effectiveness of Maintenance at 
Nuclear Power Plants['']) will further reduce the risk of a 
containment leakage path going undetected.
    Therefore[,] the proposed change does not involve a significant 
reduction in a margin of safety.

    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Branch Chief: Thomas G. Hiltz.

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: September 19, 2007.
    Description of amendment request: The proposed amendments would 
revise various Technical Specification (TS) setting limits and the 
overtemperature [Delta]T/overpower [Delta]T time constants in TS 2.3 
and TS 3.7. The methodology for determining the revised setting limits 
and time constants is in agreement with methods 1 and 2 in ISA-RP67.04, 
Part II.
    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 license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    No. The proposed change revises [Limited Safety System Settings] 
LSSSs and setting limits to ensure that safety limits are not 
exceeded as a result of normal and expected instrument drift between 
calibration intervals. The new allowable values (LSSSs and setting 
limits) were derived to meet the intent of RIS 2006-17, ``NRC Staff 
Position on the Requirements of 10 CFR 50.36, `Technical 
Specifications,' Regarding Limiting Safety System Settings During 
Periodic Testing and Calibration of Instrument Channels,'' dated 
August 24, 2006.
    The proposed TS change does not change any of the previously 
evaluated accidents in the Updated Final Safety Analysis Report 
(UFSAR). Rather, the proposed change ensures that reactor trip 
system and engineered safety function actuation system actuations 
occur as designed and within safety limits. In addition, it 
increases the probability that a malfunctioning instrument channel 
will be identified.
    This change is not considered to represent a significant 
increase in the probability or

[[Page 60037]]

consequences of an accident, since it will decrease the probability 
of the malfunction of a system, structure or component (SSC), 
thereby decreasing the probability or consequences of an accident 
previously evaluated. Specifically, the change is conservative in 
nature since it will increase the likelihood that a malfunctioning 
instrument channel will be identified prior to that channel 
exceeding its safety limit.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    No. The proposed change revises LSSSs and setting limits to 
ensure that safety limits are not exceeded as a result of normal and 
expected instrument drift between calibration intervals.
    The change is conservative and is intended to ensure the safety 
analysis is maintained. Specifically, the proposed change is 
intended to identify a malfunctioning channel prior to its exceeding 
the safety limit sooner than the current instrument setting 
methodology. 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 amendment involve a significant reduction 
in a margin of safety?
    No. The proposed change revises LSSSs and setting limits to 
ensure that safety limits are not exceeded as a result of normal and 
expected instrument drift between calibration intervals. The new 
allowable values (LSSS and setting limits) were derived to meet the 
intent of RIS 2006-17, ``NRC Staff Position on the Requirements of 
10 CFR 50.36, `Technical Specifications,' Regarding Limiting Safety 
System Settings During Periodic Testing and Calibration of 
Instrument Channels,'' dated August 24, 2006.
    Channel statistical allowance (CSA) calculations have been 
performed on channels with an associated safety analysis limit to 
determine the instrument channel uncertainty. Channel operational 
test (COT) errors are associated with those portions of the 
instrument channel tested to verify channel operability. These COT 
errors were extracted from the CSA to derive an allowable value for 
the channel. The allowable value is set at a distance from the 
actual (nominal) trip setpoint equal to the COT errors (with some 
minimal additional margin on some channels). The overall result is a 
reduction in the distance between the allowable value and the 
nominal trip setpoint. Consequently, for a malfunctioning channel, 
the allowable value will be exceeded with less drift and, therefore, 
corrective action will be initiated sooner after implementation of 
the proposed change. This will increase the likelihood that the 
safety analysis limit for the channel is not exceeded.
    The distance between the safety analysis limit and the nominal 
trip setpoint has not been decreased; therefore, the safety margin 
has [not been] reduced. The likelihood that a malfunctioning channel 
is identified prior to exceeding its safety analysis limit has 
increased. Therefore, the proposed amendment 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: 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 Branch Chief: Evangelos C. Marinos.

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

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

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

    Date of amendment request: March 29, 2007.
    Brief description of amendment request: The proposed amendments 
would revise the Catawba Nuclear Station, Units 1 and 2, Technical 
Specification section 3.5.2.8, and the associated Bases and authorize 
changes to the Updated Final Safety Analysis Report concerning 
modifications to the emergency core cooling system sumps.
    Date of publication of individual notice in Federal Register: 
August 13, 2007, (72 FR 45274).
    Expiration date of individual notice: October 15, 2007.

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

    Date of amendment request: April 30, 2007.
    Brief description of amendment request: The proposed amendments 
would revise the Catawba Nuclear Station, Unit 2, Technical 
Specification Section 5.5.9 concerning modifications to the steam 
generator tube repair criteria.
    Date of publication of individual notice in Federal Register: 
August 13, 2007, (72 FR 45272).
    Expiration date of individual notice: October 15, 2007.

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 51.22(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/

[[Page 60038]]

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. 50-289, Three Mile Island 
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

    Date of application for amendment: May 15, 2006, as supplemented by 
letters dated October 6, 2006, December 12, 2006, May 31, 2007, July 
25, 2007, and September 4, 2007.
    Brief description of amendment: The amendment consists of changes 
to various technical specifications (TSs) regarding steam generator 
tube integrity. It is based on Revision 4 to Technical Specification 
Task Force (TSTF) Standard Technical Specification Change Traveler, 
TSTF-449, ``Steam Generator Tube Integrity,'' and is adapted for the 
custom TSs used at TMI-1.
    Date of issuance: September 27, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 261.
    Facility Operating License No. DPR-50: Amendment revised the 
license and the technical specifications.
    Date of initial notice in Federal Register: July 18, 2006 (71 FR 
40744). The supplements dated October 6, 2006, December 12, 2006, May 
31, 2007, July 25, 2007, and September 4, 2007, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the NRC 
staff's original proposed no significant hazards determination. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated September 27, 2007.
    No significant hazards consideration comments received: No.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of application for amendments: May 2, 2007.
    Brief description of amendments: Consistent with the Nuclear 
Regulatory Commission approved Technical Specification Task Force-427, 
Revision 2, the amendments add a new limiting condition for operation 
(LCO) 3.0.9, to the TS. LCO 3.0.9 will allow the licensee to delay 
declaring an LCO not met for equipment supported by barriers unable to 
perform their associated support function for up to 30 days provided 
that risk is assessed and managed.
    Date of issuance: September 27, 2007.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment Nos.: 282 and 259.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the License and Technical Specifications.
    Date of initial notice in Federal Register: June 19, 2007 (72 FR 
33781). The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated September 27, 2007.
    No significant hazards consideration comments received: No.

Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of application for amendments: September 28, 2006, as 
supplemented by letter dated September 20, 2007.
    Brief Description of amendments: The amendments changed Technical 
Specification (TS) 3.8.3, ``Diesel Fuel Oil,'' to allow the main fuel 
oil storage tank to be taken out of service for 14 days for inspection, 
maintenance, and associated repairs on a one-time basis.
    Date of issuance: September 27, 2007.
    Effective date: Date of issuance to be implemented within 60 days.
    Amendment Nos.: 242 and 270.
    Renewed Facility Operating License Nos. DPR-71 and DPR-62: 
Amendments changed the TSs.
    Date of initial notice in Federal Register: January 3, 2007 (72 FR 
148). The supplement dated September 20, 2007, 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 September 
27, 2007.
    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: January 18, 2007.
    Brief description of amendment: The amendment revises the 
expiration time limit of the reactor coolant system Pressure/
Temperature limit graphs in Technical Specifications (TS); revises the 
adjusted reference temperature for the reactor vessel; and revises the 
Low Temperature Overpressure Protection (LTOP) arming temperature value 
specified in TSs. It also makes editorial changes in the use of 
inequality signs in TSs associated with the LTOP arming temperature in 
order to make them consistent.
    Date of issuance: October 4, 2007.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 235.
    Facility Operating License No. DPR-64: The amendment revised the 
License and the Technical Specifications.
    Date of initial notice in Federal Register: April 10, 2007 (72 FR 
17946). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 4, 2007.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear 
Plant, Van Buren County, Michigan

    Date of application for amendment: September 25, 2006, as 
supplemented by letters dated June 15, September 7, September 20, and 
September 21, 2007.
    Brief description of amendment: The amendment provides the 
Technical Specification (TS) changes and evaluations of the 
radiological consequences of design-basis accidents for implementation 
of a full-scope alternative source term methodology.
    Date of issuance: September 28, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 226.
    Renewed Facility Operating License No. DPR-20. Amendment revised 
the TSs and the Operating License.
    Date of initial notice in Federal Register: February 27, 2007 (72 
FR 8804). The supplemental letters contained clarifying information and 
did not change the initial no significant hazards consideration 
determination, and did not expand the scope of the original Federal 
Register notice. The Commission's related evaluation of the amendment 
is contained in a Safety Evaluation dated September 28, 2007.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear 
Plant, Van Buren County, Michigan

    Date of application for amendment: June 29, 2007, as supplemented 
by letter dated August 20, 2007.

[[Page 60039]]

    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.5.5, ``Trisodium Phosphate,'' and the associated 
surveillance requirements by replacing the containment sump buffering 
agent, trisodium phosphate, with sodium tetraborate decahydrate (STB). 
In particular, the amendment revises the TS Limiting Condition for 
Operation (LCO) 3.5.5, with a new weight requirement for STB. The title 
of the TS section is also changed from ``Trisodium Phosphate'' to 
``Containment Sump Buffering Agent and Weight Requirements.''
    Date of issuance: October 2, 2007.
    Effective date: As of the date of issuance and shall be implemented 
during the 2007 refueling outage, prior to Mode 3 entry following 
refueling.
    Amendment No.: 227.
    Renewed Facility Operating License No. DPR-20. Amendment revised 
the TS and License.
    Date of initial notice in Federal Register: July 10, 2007 (72 FR 
37544). 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 amendment 
is contained in a Safety Evaluation dated October 2, 2007.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear 
Plant, Van Buren County, Michigan

    Date of application for amendment: April 18, 2007, as supplemented 
by letters dated July 16 and September 20, 2007.
    Brief description of amendment: The amendment changes Technical 
Specification (TS) Surveillance Requirement (SR) 3.5.2.9, to make the 
surveillance consistent with the plant design following planned 
modifications to the containment sump. Entergy Nuclear Operations' 
(ENO) modification removes the existing emergency core cooling system 
(ECCS) suction inlet screens. In lieu of the ECCS suction inlet 
screens, ENO is installing passive strainer assemblies on the 590 foot 
elevation of containment. The SR change was necessary to reflect the 
change in equipment.
    Date of issuance: October 4, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 228.
    Renewed Facility Operating License No. DPR-20. Amendment revised 
the TS and License.
    Date of initial notice in Federal Register: June 19, 2007 (72 FR 
33782). The supplemental letters contained clarifying information and 
did not change the initial no significant hazards consideration 
determination, and did not expand the scope of the original Federal 
Register notice. The Commission's related evaluation of the amendment 
is contained in a Safety Evaluation dated October 4, 2007.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2 (ANO-2), Pope County, Arkansas

    Date of application for amendment: March 30, 2007, as supplemented 
on June 13, 2007.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.9.12, ``Fuel Storage,'' and its associated tables, 
figures, and surveillance requirements, TS 5.3, ``Fuel Storage,'' and 
adds TS 6.5.17, ``Metamic Coupon Sampling Program.'' The ANO-2 TS 
3.9.12 is changed to: (1) Support higher fuel assembly uranium-235 (U-
235) enrichment; (2) apply the appropriate loading restrictions; and 
(3) delete the dry cask loading restrictions. ANO-2 TS 5.3.1 b is 
changed to reflect a different spent fuel pool boron concentration that 
is needed to assure K-effective remains less than or equal to 0.95. 
ANO-2 TS 5.3.2a is modified to reflect a higher fuel assembly U-235 
enrichment. A new coupon sampling program is added as TS 6.5.17, and TS 
4.9.12.d is added to direct performance of the coupon sampling program.
    Date of issuance: September 28, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 273.
    Renewed Facility Operating License No. NPF-6: Amendment revised the 
Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: May 8, 2007 (72 FR 
26175). The supplement dated June 13, 2007, 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 September 
28, 2007.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 26, 2006, as 
supplemented by letter dated August 8, 2007.
    Brief description of amendment: The amendment revised the technical 
specifications to allow the AREVA NP Inc. Advanced Mark-BW(A) fuel 
assemblies to be loaded into the Braidwood Station, Unit 1 core for 
operating Cycles 15, 16, and 17.
    Date of issuance: October 4, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 145/145.
    Facility Operating License Nos. NPF-72 and NPF-77: The amendment 
revised the Technical Specifications and License.
    Date of initial notice in Federal Register: (72 FR 152; January 3, 
2007). The August 8, 2007, supplement contained clarifying information 
and did not change the NRC staff's initial proposed finding of no 
significant hazards consideration. The Commission's related evaluation 
of the amendments is contained in a Safety Evaluation dated October 4, 
2007.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: October 18, 2006, as 
supplemented by letter dated, March 26, 2007.
    Brief description of amendments: The amendments would modify the 
technical specifications (TS) to risk-inform requirements regarding 
selected required action end states consistent with the Nuclear 
Regulatory Commission (NRC)-approved industry and TS task force (TSTF-
423), Revision 0, ``Technical Specifications End States, NEDC-32988-
A.'' This TSTF was published in the Federal Register on March 23, 2006, 
as part of the consolidated line item improvement.
    Date of issuance: September 27, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment Nos.: 184/171.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications and License.
    Date of initial notice in Federal Register: May 8, 2007 (72 FR 
26177).

[[Page 60040]]

The March 26, 2007, supplement contained clarifying information and did 
not change the NRC staff's initial proposed finding of no significant 
hazards consideration. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated September 27, 
2007.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334, 
Beaver Valley Power Station, Unit No. 1, Beaver County, Pennsylvania

    Date of application for amendment: February 9, 2007, as 
supplemented by letters dated August 8, August 23, and September 13, 
2007.
    Brief description of amendment: The amendment will address Generic 
Safety Issue 191 ``Assessment of Debris Accumulation on PWR Sump 
Performance,'' by implementing Technical Specification (TS) changes 
that reflect the use of a new recirculation spray system pump start 
signal due to a modification to the containment sump screens and 
replace the use of LOCTIC with the Modular Accident Analysis Program-
Design Basis Accident calculation methodology to calculate containment 
pressure, temperature, and condensation rates for input to the SWNAUA 
code, which ultimately changes the aerosol removal coefficients used in 
dose consequence analysis.
    Date of issuance: October 5, 2007.
    Effective date: As of the date of issuance, and shall be 
implemented prior to the first entry into Mode 4 coming out of 1R18, 
which begins September 2007.
    Amendment No: 280.
    Facility Operating License No. DPR-66: The amendment revised the 
License and TS.
    Date of initial notice in Federal Register: April 24, 2007 (72 FR 
20383). The supplements dated August 8, August 23, and September 13, 
2007, 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 October 5, 2007.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of application for amendments: January 26, 2007, as 
supplemented by letter dated July 11, 2007.
    Brief description of amendments: The amendment conforms the license 
to reflect the direct transfer of Wisconsin Electric Power Company's 
ownership interest and the Nuclear Management Company's operating 
authority for the renewed Facility Operating License, Nos. DPR-24 and 
DPR-27 for Point Beach Nuclear Plant, Units 1 and 2 (Point Beach) to 
FPL Energy Point Beach, LLC, as approved by order of the Commission 
order dated July 31, 2007. Transfer of the licenses will also authorize 
FPL Energy Point Beach, LLC, pursuant to the general license 
requirements in 10 CFR 72.210, to store spent fuel in the Independent 
Spent Fuel Storage Installation at Point Beach.
    Date of issuance: September 28, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 228, 233.
    Renewed Facility Operating License Nos. DPR-24 and DPR-27: 
Amendments revised the Technical Specifications/License.
    Date of initial notice in Federal Register: February 28, 2007 (72 
FR 9035). The July 11, 2007, supplement contained clarifying 
information and did not change the staff's initial proposed finding of 
no significant hazards consideration. The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
July 31, 2007.
    No significant hazards consideration comments received: No.

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: October 12, 2006.
    Brief description of amendment: The amendment revises the number of 
fuel assemblies that are allowed to be stored in the spent fuel pool 
(SFP) from 1879 to 1321 in Technical Specification (TS) 4.3.3 and 
removes the reference to Type 4 SFP storage racks in TS limiting 
condition for operation 3.7.13.
    Date of issuance: October 1, 2007.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 103.
    Renewed Facility Operating License No. DPR-18: Amendment revised 
the License and Technical Specifications.
    Date of initial notice in Federal Register: November 7, 2006 (71 FR 
65145). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 1, 2007.
    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: April 28, 2006, and as 
supplemented by letters dated November 13 and December 22, 2006, May 7, 
June 15, July 27, and September 11, 2007.
    Brief description of amendments: The change increased the minimum 
allowed boron concentration of the spent fuel pool and allowed credit 
for soluble boron, guide tube inserts made from borated stainless 
steel, and fuel storage patterns in place of Boraflex.
    Date of issuance: September 27, 2007.
    Effective date: As of the date of issuance, and shall be 
implemented within 180 days of issuance.
    Amendment Nos.: Unit 2-213; Unit 3-205.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: June 6, 2006 (71 FR 
32606). The supplemental letters dated November 13 and December 22, 
2006, May 7, June 15, July 27, and September 11, 2007, 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 September 27, 2007.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of application for amendments: February 13, 2007.
    Brief description of amendments: The amendments revised the 
Tech