[Federal Register Volume 72, Number 48 (Tuesday, March 13, 2007)]
[Notices]
[Pages 11383-11403]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-4251]


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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 February 15, 2007 through March 1, 2007. 
The last biweekly notice was published on February 27, 2007 (72 FR 
8800).

[[Page 11384]]

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, 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, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of

[[Page 11385]]

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, <a href="/cdn-cgi/l/email-protection#90d8f5f1e2f9fef7d4fff3fbf5e4d0fee2f3bef7ffe6"><span class="__cf_email__" data-cfemail="f8b09d998a91969fbc979b939d8cb8968a9bd69f978e">[email protected]</span></a>; 
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 
<a href="/cdn-cgi/l/email-protection#8bc4ccc8c6eae2e7c8eee5ffeef9cbe5f9e8a5ece4fd"><span class="__cf_email__" data-cfemail="2d626a6e604c44416e484359485f6d435f4e034a425b">[email protected]</span></a>. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to <a href="/cdn-cgi/l/email-protection#51213523113f23327f363e27"><span class="__cf_email__" data-cfemail="a4d4c0d6e4cad6c78ac3cbd2">[email protected]</span></a>.

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 amendment request: February 1, 2007.
    Description of amendment request: The proposed license amendment 
would revise Surveillance Requirement (SR) 3.5.2.8 in Technical 
Specification 3.5.2, ``ECCS [Emergency Core Cooling System]--
Operating,'' to reflect the replacement of the containment 
recirculation sump suction inlet trash racks and screens with 
strainers, in response to Nuclear Regulatory Commission (NRC) Generic 
Letter 2004-02, ``Potential Impact of Debris Blockage on Emergency 
Recirculation during Design Basis Accidents at Pressurized-Water 
Reactors.'' The proposed license amendment would replace ``trash racks 
and screens'' with ``strainers'' in SR 3.5.2.8.
    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 consequences of accidents evaluated in the Updated Final 
Safety Analysis Report [UFSAR] that could be affected by the 
proposed change are those involving the pressurization of 
Containment and associated flooding of the Containment and 
recirculation of this fluid within the Emergency Core Cooling System 
(ECCS) or the Containment Spray System (CSS) (e.g., loss-of-coolant 
accidents [LOCAs]). The proposed change does not impact the 
initiation or probability of occurrence of any accident. Although 
the configurations of the existing containment recirculation sump 
trash racks and screen and the replacement sump strainer cassettes 
are different, they serve the same fundamental purpose of passively 
removing debris from the sump's suction supply of the supported 
system pumps. Removal of trash racks does not impact the adequacy of 
the pump net positive suction head assumed in the safety analysis. 
Likewise, the change does not reduce the reliability of any 
supported systems or introduce any new system interactions. The 
greatly increased surface area of the new strainer is designed to 
reduce head loss and reduce the approach velocity at the strainer 
face significantly, decreasing the risk of impact from large debris 
entrained in the sump flow stream.
    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 containment recirculation sump strainers are a passive 
system used for accident mitigation. As such, they cannot be 
accident initiators. Therefore, there is no possibility that this 
change could create any new or different kind of accident. No new 
accident scenarios, transient precursors, or limiting single 
failures are introduced as a result of the proposed change. There 
will be no adverse effect or challenges imposed on any safety-
related system as a result of the change. Therefore, the possibility 
of a new or different [kind] of accident is not created.
    There are no changes which would cause the malfunction of 
safety-related equipment, assumed to be OPERABLE in the accident 
analyses, as a result of the proposed Technical Specification 
change. No new equipment performance burdens are imposed. The 
possibility of a malfunction of safety-related equipment with a 
different result is not created.
    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?
    The proposed change does not affect the acceptance criteria for 
any analyzed event nor is there a change to any safety analysis 
limit. There will be no effect on the manner in which safety limits, 
limiting safety system settings, or limiting conditions for 
operation are determined nor will there be any effect on those plant 
systems necessary to assure the accomplishment of protection 
functions. The proposed change does not adversely affect the fuel, 
fuel cladding, Reactor Coolant System, or containment integrity. The 
radiological dose consequence acceptance criteria listed in the 
Updated Final Safety Analysis Report will continue to be met.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Carey Fleming, Esquire, Senior Counsel--
Nuclear Generation, Constellation Generation Group, LLC, 750 East Pratt 
Street, 17th floor, Baltimore, MD 21202.
    NRC Acting Branch Chief: John P. Boska.

Carolina Power & Light Company, Docket Nos. 50-325 and 5-324 Brunswick 
Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina

    Date of amendments request: December 21, 2006.
    Description of amendment request: The proposed amendment would

[[Page 11386]]

modify technical specification (TS) requirements of TS 3.4.1, 
``Recirculation Loops Operating,'' to require the recirculation loops 
be operated with matched flows versus recirculation pump speeds as 
currently required. This change affects the Limiting Condition for 
Operation (LCO) requirements and Surveillance Requirements (SRs) of TS 
3.4.1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment implements more conservative requirements 
associated with recirculation loop operation. Specifically, the LCO 
requirements of TS 3.4.1 and SR 3.4.1.1 are being revised to 
directly monitor recirculation loop jet pump flows versus 
recirculation pump speed, eliminating potential non-conservatism 
associated with relating recirculation loop jet pump flow to 
recirculation pump speed. These requirements assure that the 
mismatch between recirculation loop jet pump flows are bounded by 
the existing design bases analyses. As a result, the proposed change 
ensures that the consequences of a design bases LOCA [loss-of-
coolant accident] remain within the existing evaluation.
    The proposed change does not involve a physical change to the 
Reactor Recirculation system, nor does it alter the assumptions of 
the accident analyses. Therefore the probability of an accident 
previously evaluated is not affected.
    Based on the above, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical change to the 
Reactor Recirculation system, nor does it alter the assumptions of 
the accident analyses.
    The implementation of more conservative requirements associated 
with recirculation loop operation does not introduce any new failure 
modes. 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 amendment implements more conservative requirements 
associated with recirculation loop operation. These requirements 
ensure that the Reactor Recirculation system is operated consistent 
with the initial conditions of the existing design bases analyses. 
Since the design bases analyses assumptions are unchanged, the 
proposed change does not involve a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Branch Chief: L. Raghavan.

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

    Date of amendment request: December 15, 2006.
    Description of amendment request: The amendment would incorporate 
changes to the Technical Specifications (TS) associated with previously 
approved industry initiatives. The first change would relocate the 
Safety Limit Violation specifications from the administrative controls 
TS section to the safety limit TS sections as approved by TSTF-05-A, 
``Deletion of Safety Limit Violation Requirements.'' The second change 
would incorporate generic position titles, as approved by TSTF-65-A, 
``Use of Generic Titles for Utility Positions,'' and incorporates 
changes approved by NRC Administrative Letter (AL) 95-06, ``Relocation 
of Technical Specification Administrative Controls Related to Quality 
Assurance.''
    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?
    No. The proposed amendment consists of changes to and relocation 
of administrative TS requirements that were previously generically 
approved by the NRC. The proposed amendment would not change any of 
the previously evaluated accidents in the updated safety analysis 
report (USAR). The administrative controls that are affected by the 
proposed amendment do not have any function related to preventing or 
mitigating any of these previously evaluated accidents. The proposed 
amendment does not affect any systems, structures, or components 
(SSCs) that have the function of preventing or mitigating any of 
these previously evaluated accidents. The proposed amendment does 
not increase the likelihood of the malfunction of an SSC, thus the 
potential impact on analyzed accidents need not be considered.
    Because the proposed amendment is a relocation of administrative 
requirements that are not associated with preventing or mitigating 
the consequences of any previously evaluated accidents, there is no 
affect on the probability or consequences of an accident previously 
evaluated.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. The proposed amendment consists of changes to and relocation 
of administrative TS requirements previously generically approved by 
the NRC. This amendment will not change the design function of any 
SSC or the manner that any SSC is operated. Because this amendment 
does not change the design function or operation of any SSC, the 
amendment would not create the possibility of a new or different 
kind of accident due to credible new failure mechanisms, 
malfunctions, or accident initiators not considered in the design 
and licensing bases.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The proposed amendment consists of changes to and relocation 
of administrative TS requirements previously generically approved by 
the NRC. The amendment does not alter any design basis safety limit 
and no safety margins are affected.
    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: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Acting Branch Chief: P. Milano.

Duke Power Company LLC, et al., Docket No. 50-413, Catawba Nuclear 
Station, Unit 1 (Catawba), York County, South Carolina

    Date of amendment request: November 22, 2006.
    Description of amendment request: The amendment would revise the 
Catawba Unit 1 Facility Operating License (FOL) to provide for an 
extension of the time limit to complete the required modification to 
the Emergency Core Cooling System (ECCS) sump.

[[Page 11387]]

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed license amendment delineates a new Unit 1 FOL 
condition to implement a completion date associated with the ECCS 
sump strainer modification. The proposed license amendment is 
administrative in nature and is being submitted to fulfill a 
commitment made in previous Duke licensing correspondence. 
Therefore, the proposed license amendment has no effect upon either 
the probability or consequences of an accident previously evaluated.
    2. The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    As stated above, the proposed license amendment is 
administrative in nature and does not change the manner in which 
Unit 1 is designed or operated. Therefore, the proposed license 
amendment cannot create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. The proposed license amendment does not involve a significant 
reduction in a margin of safety.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their intended functions. 
These barriers include the fuel cladding, the reactor coolant 
system, and the containment. The performance of these barriers will 
not be affected by the addition of the proposed FOL condition. Being 
administrative in nature, the proposed license amendment therefore 
does not involve a significant reduction in any safety margin.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Associate General 
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South 
Church Street, EC07H, Charlotte, NC 28202.
    NRC Branch Chief: Evangelos C. Marinos.

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

    Date of amendment request: April 11, 2006.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TSs) related to the organizational 
description in TS 5.2.1
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided it's analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below.
Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated
    The proposed change revises an organizational description in TS 
5.2.1 to reflect the change of the title of the Vice President Nuclear 
Generation. The change is solely administrative in nature and has no 
impact on any accident probabilities or consequences. The change does 
not affect structures or components in the plant. The change has no 
affect on any accident previously evaluated. Therefore the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From any Accident Previously 
Evaluated
    There are no new accident causal mechanisms created as a result of 
this proposed change. No changes are being made to the plant that will 
introduce any new accident causal mechanisms. The change is solely 
administrative in nature and does not impact any plant systems that are 
accident initiators. Therefore, no new accidents or a different 
accident than previously evaluated is being created.
Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in a Margin of Safety.
    Margin of safety is related to confidence in the ability of the 
fission product barriers to perform their design functions during and 
following an accident situation. The proposed change is solely 
administrative in nature and does not affect the performance of the 
barriers. Consequently, no safety margins will be impacted. Therefore, 
the proposed change does not involve a significant reduction in a 
margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied, therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Duke Power Company LLC, 
422 South Church Street, Charlotte, North Carolina 28201-1006.
    NRC Branch Chief: Evangelos C. Marinos.

Exelon Generation Company, LLC (EGC), Docket Nos. STN 50-454 and STN 
50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

    Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 
1 and 2, Will County, Illinois.
    Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 
2 and 3, Grundy County, Illinois.
    Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 
2, LaSalle County, Illinois.
    Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, 
Units 2 and 3, York and Lancaster Counties, Pennsylvania.
    Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, 
Units 1 and 2, Rock Island County, Illinois.
    Date of amendment request: December 15, 2006.
    Description of amendment request: The proposed amendment would 
modify the technical specifications (TSs) by replacing the term 
``plant-specific titles'' with ``generic titles'' in TS Section 
5.2.1.a, ensuring the TS description is consistent with the EGC Quality 
Assurance Topical Report (QATR). The proposed amendment will also 
revise the Peach Bottom TS Section 5.2.1.a, to replace the reference to 
the Updated Final Safety Analysis Report with reference to the EGC 
QATR. This will align the Peach Bottom TS wording with the rest of the 
EGC fleet.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change is a word replacement in TS 5.2.1, ``Onsite 
and Offsite Organizations.'' The proposed change involves no changes 
to plant systems or accident analyses. The proposed change is 
administrative in nature and, as such, does not affect initiators of 
analyzed events or assumed mitigation of accidents or transients.
    Therefore, the proposed change does not involve any increase in 
the probability or

[[Page 11388]]

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.
    Creation of the possibility of a new or different kind of 
accident would require creating one or more new accident precursors. 
New accident precursors may be created by modifications of plant 
configuration, including changes in allowable modes of operation. 
The proposed change does not involve a physical alteration of the 
plant, add any new equipment, or allow any existing equipment to be 
operated in a manner different from the present method of 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?
    Response: No.
    The proposed change is administrative in nature and has no 
impact on equipment design or method of operation. There are no 
changes being made to safety limits or safety system allowable 
values 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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Michael L. Marshall, Jr.

Exelon Generation Company, LLC (EGC) Docket Nos. 50-254 and 50-265, 
Quad Cities Nuclear Power Station, Unit 1, Rock Island County, Illinois

    Date of amendment request: January 16, 2007.
    Description of amendment request: The proposed amendment revises 
the values of the safety limit minimum critical power ratio (SLMCPR) in 
the Quad Cities Nuclear Power Station (QCNPS), Unit 1, Technical 
Specification (TS) Section 2.1.1, ``Reactor Core SLs [Safety Limits].'' 
Specifically, the proposed change would require that for QCNPS, Unit 1, 
the minimum critical power ratio shall be greater than 1.11 for two 
recirculation loop operation, or greater than 1.13 for single 
recirculation loop operation. This change is needed to support the next 
cycle of operation for QCNPS, Unit 1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. The 
consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences. Limits have been established consistent with NRC-
approved methods to ensure that fuel performance during normal, 
transient, and accident conditions is acceptable. The proposed 
change conservatively establishes the SLMCPR for QCNPS, Unit 1, 
Cycle 20 such that the fuel is protected during normal operation and 
during plant transients or anticipated operational occurrences 
(AOOs).
    Changing the SLMCPR does not increase the probability of an 
evaluated accident. The change does not require any physical plant 
modifications, physically affect any plant components, or entail 
changes in plant operation. Therefore, no individual precursors of 
an accident are affected.
    The proposed change revises the SLMCPR to protect the fuel 
during normal operation as well as during plant transients or AOOs. 
Operational limits will be established based on the proposed SLMCPR 
to ensure that the SLMCPR is not violated. This will ensure that the 
fuel design safety criterion (i.e., that at least 99.9% of the fuel 
rods do not experience transition boiling during normal operation 
and AOOs) is met. Since the proposed change does not affect 
operability of plant systems designed to mitigate any consequences 
of accidents, the consequences of an accident previously evaluated 
are not expected to increase.
    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.
    Creation of the possibility of a new or different kind of 
accident requires creating one or more new accident precursors. New 
accident precursors may be created by modifications of plant 
configuration, including changes in allowable modes of operation. 
The proposed change does not involve any plant configuration 
modifications or changes to allowable modes of operation. The 
proposed change to the SLMCPR assures that safety criteria are 
maintained for QCNPS, Unit 1, Cycle 20.
    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 SLMCPR provides a margin of safety by ensuring that at least 
99.9% of the fuel rods do not experience transition boiling during 
normal operation and AOOs if the MCPR limit is not violated. The 
proposed change will ensure the current level of fuel protection is 
maintained by continuing to ensure that at least 99.9% of the fuel 
rods do not experience transition boiling during normal operation 
and AOOs if the MCPR limit is not violated. The proposed SLMCPR 
values were developed using NRC-approved methods. Additionally, 
operational limits will be established based on the proposed SLMCPR 
to ensure that the SLMCPR is not violated. This will ensure that the 
fuel design safety criterion (i.e., that no more than 0.1% of the 
rods are expected to be in boiling transition if the MCPR limit is 
not violated) is met.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based upon the above, EGC concludes that the proposed amendment 
presents no significant hazards consideration under the standards 
set forth in 10 CFR 50.92(c), and, accordingly, a finding of no 
significant hazards consideration is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Michael L. Marshall, Jr.

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

    Date of amendment request: December 29, 2006.
    Description of amendment request: The proposed amendment revises 
Technical Specification (TS) 3.1.8, ``Scram Discharge Volume (SDV) Vent 
and Drain Valves,'' to allow a vent or drain line with one inoperable 
valve to be isolated instead of requiring the valve to be restored to 
operable status within 7 days.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on February 24, 2003 (68 FR 8637), on possible 
amendments concerning the consolidated line item implement process 
(CLIIP), including a model safety evaluation and a model no significant 
hazards consideration

[[Page 11389]]

(NSHC) determination. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on April 15, 2003 (68 FR 18294), 
as part of the CLIIP. In its application dated December 29, 2006, the 
licensee affirmed the applicability of the following determination.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    A change is proposed to allow the affected SDV vent and drain 
line to be isolated when there are one or more SDV vent or drain 
lines with one valve inoperable instead or requiring the valve to be 
restored to operable status within 7 days. With one SDV vent or 
drain valve inoperable in one or more lines, the isolation function 
would be maintained since the redundant valve in the affected line 
would perform its safety function of isolating the SDV. Following 
the completion of the required action, the isolation function is 
fulfilled since the associated line is isolated. The ability to vent 
and drain the SDVs is maintained and controlled through 
administrative controls. This requirement assures the reactor 
protection system is not adversely affected by the inoperable 
valves. With the safety functions of the valves being maintained, 
the probability or consequences of an accident previously evaluated 
are not significantly increased.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The proposed change ensures that the safety functions of the SDV 
vent and drain valves are fulfilled. The isolation function is 
maintained by redundant valves and by the required action to isolate 
the affected line. The ability to vent and drain the SDVs is 
maintained through administrative controls. In addition, the reactor 
protection system will prevent filling of an SDV to the point that 
it has insufficient volume to accept a full scram. Maintaining the 
safety functions related to isolation of the SDV and insertion of 
control rods ensures that the proposed change does not involve a 
significant reduction in the margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Michael L. Marshall, Jr.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry 
Nuclear Power Plant, Unit 1, Lake County, Ohio, and Docket Nos. 50-334 
and 50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver 
County, Pennsylvania

    Date of amendment request: January 11, 2007.
    Description of amendment request: The proposed license amendments 
would modify technical specification (TS) requirements for inoperable 
snubbers by adding Limiting Condition for Operation (LCO) 3.0.8. The 
proposed license amendments also modify LCO 3.0.1 to incorporate the 
addition of LCO 3.0.8. This change is based on the TS Task Force (TSTF) 
Traveler, TSTF-372, Revision 4. A notice of availability for this TS 
improvement using the consolidated line item improvement process was 
published in the Federal Register on May 4, 2005.
    The Nuclear Regulatory Commission (NRC) staff issued a notice of 
availability of a model no significant hazards consideration (NSHC) 
determination for referencing license amendment applications in the 
Federal Register on November 24, 2004 (69 FR 68412), and May 4, 2005 
(70 FR 23252). The licensee affirmed the applicability of the model 
NSHC determination in its application dated January 11, 2007.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The Proposed Change Does Not Involve a Significant Increase 
in the Probability or Consequences of an Accident Previously 
Evaluated.
    The proposed change allows a delay time for entering a supported 
system technical specification (TS) when the inoperability is due 
solely to an inoperable snubber if risk is assessed and managed. The 
postulated seismic event requiring snubbers is a low-probability 
occurrence and the overall TS system safety function would still be 
available for the vast majority of anticipated challenges. 
Therefore, the probability of an accident previously evaluated is 
not significantly increased, if at all. The consequences of an 
accident while relying on allowance provided by proposed LCO 3.0.8 
are no different than the consequences of an accident while relying 
on the TS required actions in effect without the allowance provided 
by proposed LCO 3.0.8. Therefore the consequences of an accident 
previously evaluated are not significantly affected by this change. 
The addition of a requirement to assess and manage the risk 
introduced by this change will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. The Proposed Change Does Not Create the Possibility of a New 
or Different Kind of Accident From Any Previously Evaluated.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Allowing delay times for entering supported system TS when 
inoperability is due solely to inoperable snubbers, if risk is 
assessed and managed, will not introduce new failure modes or 
effects and will not, in the absence of other unrelated failures, 
lead to an accident whose consequences exceed the consequences of 
accidents previously evaluated. The addition of a requirement to 
assess and manage the risk introduced by this change will further 
minimize possible concerns. Thus, this change does not create the 
possibility of a new or different kind of accident from an accident 
previously evaluated.
    3. The Proposed Change Does Not Involve a Significant Reduction 
in the Margin of Safety.
    The proposed change allows a delay time for entering a supported 
system TS when the inoperability is due solely to an inoperable 
snubber, if risk is assessed and managed. The postulated seismic 
event requiring snubbers is a low-probability occurrence and the 
overall TS system safety function would still be available for the 
vast majority of anticipated challenges. The risk impact of the 
proposed TS changes was assessed following the three-tiered approach 
recommended in RG 1.177. A bounding risk assessment was performed to 
justify the proposed TS changes. This application of LCO 3.0.8 is 
predicated upon the licensee's performance of a risk assessment and 
the management of plant risk. The net change to the margin of safety 
is insignificant. Therefore, this change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Michael L. Marshall, Jr.

[[Page 11390]]

Nine Mile Point Nuclear Station (NMPNS), LLC, Docket No. 50-220, Nine 
Mile Point Nuclear Station Unit No. 1 (NMP1), Oswego County, New York

    Date of amendment request: December 14, 2006.
    Description of amendment request: The proposed license amendment 
would revise the accident source term used in the NMP1 design basis 
radiological consequence analyses in accordance with 10 CFR 50.67. The 
revised accident source term replaces the current methodology that is 
based on TID-14844, ``Calculation of Distance Factors for Power and 
Test Reactor Sites,'' with the alternative source term (AST) 
methodology described in Regulatory Guide (RG) 1.183, ``Alternative 
Source Terms for Evaluating Design Basis Accidents at Nuclear Power 
Reactors.'' The amendment request is for full implementation of the AST 
as described in RG 1.183, with the exception that TID-14844 will 
continue to be used as the radiation dose basis for equipment 
qualification and vital area access.
    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.
    Adoption of the AST and those plant systems affected by 
implementing AST do not initiate DBAs [design-basis accidents]. The 
AST does not affect the design or manner in which the facility is 
operated; rather, for postulated accidents, the AST is an input to 
calculations that evaluate the radiological consequences. The AST 
does not by itself affect the post-accident plant response or the 
actual pathway of the radiation released from the fuel. It does, 
however, better represent the physical characteristics of the 
release, so that appropriate mitigation techniques may be applied. 
Implementation of the AST has been incorporated in the analyses for 
the limiting DBAs at NMP1.
    The structures, systems and components affected by the proposed 
change mitigate the consequences of accidents after the accident has 
been initiated. Application of the AST does result in changes to 
NMP1 Updated Final Safety Analysis Report (UFSAR) functions (e.g., 
Liquid Poison system). As a condition of the application of AST, 
NMPNS is proposing to use the Liquid Poison system to control the 
suppression pool pH following a LOCA [loss-of-coolant accident]. The 
proposed changes also revise operability requirements for the 
secondary containment and certain post-accident filtration systems 
while handling irradiated fuel that has decayed for greater than 24 
hours and during core alterations. These changes have been included 
within the AST evaluations. These changes do not require any 
physical changes to the plant. As a result, the proposed changes do 
not involve a revision to the parameters or conditions that could 
contribute to the initiation of a DBA discussed in Chapter XV of the 
NMP1 UFSAR. Since design basis accident initiators are not being 
altered by adoption of the AST, the probability of an accident 
previously evaluated is not affected.
    Plant-specific AST radiological analyses have been performed 
and, based on the results of these analyses, it has been 
demonstrated that the dose consequences of the limiting events 
considered in the analyses are within the acceptance criteria 
provided by the NRC for use with the AST. These criteria are 
presented in 10 CFR 50.67 and Regulatory Guide 1.183. Even though 
the AST dose limits are not directly comparable to the previously 
specified whole body and thyroid dose guidelines of General Design 
Criterion 19 and 10 CFR 100.11, the results of the AST analyses have 
demonstrated that the 10 CFR 50.67 limits are satisfied. Therefore, 
it is concluded that adoption of the AST does not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    Based on the above discussion, it is concluded that the proposed 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Implementation of AST and the proposed changes do not alter or 
involve any design basis accident initiators. These changes do not 
involve any physical changes to the plant and do not affect the 
design function or mode of operations of systems, structures, or 
components in the facility prior to a postulated accident. Since 
systems, structures, and components are operated essentially no 
differently after the AST implementation, no new failure modes are 
created by this proposed change.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The changes proposed are associated with a new licensing basis 
for analysis of NMP1 DBAs. Approval of the licensing basis change 
from the original source term to the AST is being requested. The 
results of the accident analyses performed in support of the 
proposed changes are subject to revised acceptance criteria. The 
limiting DBAs have been analyzed using conservative methodologies, 
in accordance with the guidance contained in Regulatory Guide 1.183, 
to ensure that analyzed events are bounding and that safety margin 
has not been reduced. The dose consequences of these limiting events 
are within the acceptance criteria presented in 10 CFR 50.67 and 
Regulatory Guide 1.183. Thus, the proposed changes continue to 
ensure that the doses at the exclusion area boundary and low 
population zone boundary, as well as in the control room, are within 
corresponding regulatory criteria.
    Therefore, by meeting the applicable regulatory criteria for 
AST, it is concluded that the proposed changes do not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006.
    NRC Acting Branch Chief: John P. Boska.

Nine Mile Point Nuclear Station (NMPNS), LLC, Docket No. 50-410, Nine 
Mile Point Nuclear Station Unit No. 2 (NMP2), Oswego County, New York

    Date of amendment request: January 4, 2007.
    Description of amendment request: The proposed license amendment 
would revise Technical Specification (TS) 3.7.1, ``Service Water (SW) 
System and Ultimate Heat Sink (UHS),'' as follows: Revise the existing 
Limiting Condition for Operation (LCO) statement to require four 
operable SW pumps to be in operation when SW subsystem supply header 
water temperature is <=82 [deg]F; add a requirement that five operable 
SW pumps be in operation when SW subsystem supply header water 
temperature is >82 [deg]F and <=84 [deg]F; delete Condition G and the 
associated Required Actions and Completion Times; revise Surveillance 
Requirement 3.7.1.3 to increase the maximum allowed SW subsystem supply 
header water temperature from 82 [deg]F to 84 [deg]F; and modify the 
requirements for increasing the surveillance frequency as the 
temperature approaches the limit.
    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.

[[Page 11391]]

    The proposed change eliminates the requirement to perform 
temperature averaging when the UHS temperature is >82 [deg]F, 
establishes 84 [deg]F as the design limit for UHS water temperature 
for operation on a continuous basis, and revises the frequency for 
verifying that the UHS temperature is within the prescribed limit. 
The TS currently allow operation with the UHS water temperature 
temporarily exceeding 82 [deg]F, up to a maximum of 84 [deg]F. The 
UHS temperature itself is not an initiator of accidents analyzed in 
the Updated Safety Analysis Report (USAR). Raising the maximum 
temperature limit and revising the associated surveillance 
requirement frequency do not involve any plant hardware changes or 
new operator actions that could serve to initiate an accident. 
Continuous operation with the elevated UHS temperature may result in 
a few balance-of-plant equipment high temperature alarms. Operator 
response to these alarms would be in accordance with established 
alarm response procedures. In all cases, trip setpoints leading to a 
reactor scram or a power runback will not be reached, and the 
likelihood of component failures that could initiate an accident 
will not be significantly increased.
    The potential impact of the proposed change on the ability of 
the plant to mitigate postulated accidents has been evaluated. These 
evaluations demonstrate that safety-related systems and components 
that rely on the UHS as the cooling medium or as a pump suction 
source are capable of performing their intended safety functions at 
the higher UHS temperature, and that containment integrity and 
equipment qualification are maintained. The calculated post-accident 
dose consequences reflected in the USAR do not directly utilize UHS 
temperature as an input and thus are not impacted by the proposed 
change.
    Based on the above, the proposed change will have no adverse 
effect on plant operation or the availability or operation of any 
accident mitigation equipment. 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 will not alter the current plant 
configuration (no new or different type of equipment will be 
installed) or require any new or unusual operator actions. The 
proposed change will not alter the way any structure, system, or 
component functions and will not cause an adverse effect on plant 
operation or accident mitigation equipment. The response of the 
plant and the operators following a design-basis accident is 
unaffected by the change. The proposed change does not introduce any 
credible new failure mechanisms, malfunctions, or accident 
initiators not considered in the design and licensing bases. 
Analyses have shown that the design basis heat removal capability of 
the affected safety-related components is maintained at the 
increased UHS water temperature limit.
    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 margin of safety is determined by the design and 
qualification of the plant equipment, the operation of the plant 
within analyzed limits, and the point at which protective or 
mitigative actions are initiated. The proposed change does not 
impact these factors. An evaluation of the safety systems has been 
performed to ensure their safety functions can be met for operation 
with a UHS water temperature of 84 [deg]F on a continuous basis. 
Operation with the UHS water temperature temporarily exceeding 82 
[deg]F, up to a maximum of 84 [deg]F, is currently allowed. 
Operating on a continuous basis at the higher UHS temperature 
represents a slight reduction in design margins in terms of the 
ability of affected systems to remove accident heat loads. However, 
the evaluation has demonstrated that the proposed change does not 
have a significant impact on the capability of the affected systems 
to perform their safety-related post-accident functions and to 
mitigate accident consequences. The design limits for the 
containment and fuel cladding will not be exceeded, and equipment 
qualification will be maintained. No protection setpoints are 
affected by the proposed change. The revised frequency for 
performing the TS surveillance to verify that the UHS temperature is 
within the prescribed limit will continue to assure that plant 
operators are aware of and are monitoring increasing UHS temperature 
trends prior to reaching a value of 82 [deg]F, when a fifth SW pump 
must be placed in operation. This action is no different than that 
required by the current TS.
    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: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006.
    NRC Acting Branch Chief: John P. Boska.

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of amendment request: January 29, 2007.
    Description of amendment request: The proposed amendment would 
revise Table 3.3.5.1-1, ``Emergency Core cooling System 
Instrumentation,'' of the Technical Specifications (TSs) to extend the 
quarterly surveillance interval from quarterly to a nominal 24-month 
interval for three low pressure coolant injection loop select logic 
functions. Consistent with the extended test interval, the licensee 
also proposed to change the allowable values associated with each of 
the three logic functions (i.e., response time in seconds). The 
licensee stated that the quarterly surveillance requirement was 
inappropriately introduced when the TSs was converted from its previous 
custom format to the current Improved Technical Specification format by 
Amendment No. 146. Before the conversion, there was no such quarterly 
surveillance requirement. Furthermore, the plant was not designed to 
have these three logic functions tested while on-line.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) Part 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration (NSHC). 
The NRC staff reviewed the licensee's analysis, and has performed its 
own analysis as follows:
    (1) Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed amendment would extend the performance interval 
from quarterly to a 24-month interval, and change the associated 
allowable values for the three logic functions. The performance of 
these surveillances, or the failure to perform, as well as the 
surveillance finding (i.e., response time in seconds) are not 
precursors to, and do not affect the probability of, an accident. There 
is no design or operation change associated with the proposed 
amendment. Therefore, the proposed amendment does not increase the 
probability of an accident previously evaluated.
    A delay in performing these surveillances would not result in a 
system being unable to perform its required function. The extended 
surveillance and associated changed allowable values will not affect 
the three logic functions to operate as designed. Therefore, the plant 
systems required to mitigate accidents will remain capable of 
performing their design function. As a result, the proposed amendment 
will not lead to any significant change in the consequences of any 
accident.
    (2) Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?

[[Page 11392]]

    No. The proposed amendment does not involve a physical alteration 
of any system, structure, or component (SSC) or a change in the way any 
SSC is operated. The proposed amendment does not involve operation of 
any SSCs in a manner or configuration different from those previously 
recognized or evaluated. No new failure mechanisms will be introduced 
by the extended surveillance interval and associated allowable values. 
Thus, the proposed amendment does 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 amendment only changes the surveillance interval 
and associated allowable values for the three logic functions. There 
will be no modification of any TSs limiting condition for operation, no 
change to any limit on previously analyzed accidents, no change to how 
previously analyzed accidents or transients would be mitigated, no 
change in any methodology used to evaluate consequences of accidents, 
and no change in any operating procedure or process. The 
instrumentation and components involved in this proposed amendment have 
exhibited reliable operation based on the results of their performance 
during past periodic emergency core cooling system functional testing. 
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 
the NRC staff's own analysis above, it appears that the three standards 
of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the proposed amendment involves no significant hazards 
consideration.
    Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Branch Chief: L. Raghavan.

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

    Date of amendment request: January 29, 2007.
    Description of amendment request: The proposed amendments would 
revise technical specification (TS) 3.5.3, ``ECCS (Emergency Core 
Cooling Systems)--Shutdown'' operability requirements for the Safety 
Injection (SI) subsystem. These revisions will allow the required SI 
pump to be rendered incapable of injecting into the Reactor Coolant 
System (RCS) during low temperature (MODE 4) operations due to a single 
action or automatic signal. The capability of the plant operators to 
initiate SI flow on a timely basis will be maintained.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    This license amendment request proposes to add a new Note to 
Technical Specification 3.5.3, ``Emergency Core Cooling System--
Shutdown''. This Note will allow the Safety Injection system to be 
considered operable within the Limiting Condition for Operation 
requirements while the system is not capable of automatic injection 
provided it is capable of being manually aligned for injection.
    This Emergency Core Cooling System is not an accident initiator, 
thus the proposed changes do not increase the probability of an 
accident. The current licensing basis, Technical Specifications and 
Bases do not require automatic initiation instrumentation for the 
Emergency Core Cooling System in Mode 4, but rather assume operator 
action to mitigate an accident. With the proposed Technical 
Specification and Bases changes, the Emergency Core Cooling System 
will continue to be operable for manual initiation. Since the 
changes proposed in this license amendment request do not impact the 
performance of the system, these changes do not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    The changes proposed in this license amendment do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This license amendment request proposes to add a new Note to 
Technical Specification 3.5.3, ``Emergency Core Cooling System--
Shutdown''. This Note will allow the Safety Injection system to be 
considered operable within the Limiting Condition for Operation 
requirements while the system is not capable of automatic injection 
provided it is capable of being manually aligned for injection.
    The changes proposed for the Emergency Core Cooling System 
Technical Specifications do not change any system operations, 
maintenance activities or testing requirements. The Limiting 
Condition for Operation will continue to be met, no new failure 
modes or mechanisms are created and no new accident precursors are 
generated by this change. The Technical Specification changes 
proposed in this license amendment do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    This license amendment request proposes to add a new Note to 
Technical Specification 3.5.3, ``Emergency Core Cooling System--
Shutdown''. This Note will allow the Safety Injection system to be 
considered operable within the Limiting Condition for Operation 
requirements while the system is not capable of automatic injection 
provided it is capable of being manually aligned for injection.
    The current licensing basis, Technical Specifications and Bases 
rely upon operator actions to initiate safety injection to mitigate 
an accident in Mode 4 and do not require operability of any process 
instrumentation capable of automatically initiating the Emergency 
Core Cooling System. With the changes proposed in this license 
amendment request, the safety injection system will continue to be 
operable and the plant will continue to rely on operator actions for 
safety injection initiation. Thus, the Technical Specification 
changes proposed in this license amendment request do not involve a 
significant reduction in a margin of safety.

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

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

    Date of amendment request: October 11, 2006, as supplemented on 
October 25, November 21, and December 4, 2006.
    Description of amendment request: The proposed amendments would 
increase the SSES 1 and 2 licensed thermal power to 3952 Mega-watts 
thermal (MWt), which is 20% above the original rated thermal power 
(RTP) of 3293 MWt, and approximately 13% above the current RTP of 3489 
MWt. The proposed amendments would revise the SSES 1 and 2 Operating 
License and Technical Specifications necessary to implement the 
increased power level.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the

[[Page 11393]]

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?
    Extended Power Uprate
    Response: No.
    The probability (frequency of occurrence) of Design Basis 
Accidents occurring is not affected by the increased power level, 
because Susquehanna continues to comply with the regulatory and 
design basis criteria established for plant equipment. A 
probabilistic risk assessment demonstrates that the calculated core 
damage frequencies do not significantly change due to Constant 
Pressure Power Uprate (CPPU). Scram setpoints (equipment settings 
that initiate automatic plant shutdowns) are established such that 
there is no significant increase in scram frequency due to CPPU. No 
new challenges to safety-related equipment result from CPPU.
    The changes in consequences of postulated accidents, which would 
occur from 102% of the CPPU (rated thermal power) RTP compared to 
those previously evaluated, are acceptable. The results of CPPU 
accident evaluations do not exceed the NRC-approved acceptance 
limits. The spectrum of postulated accidents and transients has been 
investigated, and are shown to meet the plant's currently licensed 
regulatory criteria. In the area of fuel and core design, for 
example, the Safety Limit Minimum Critical Power Ratio (SLMCPR) and 
other applicable Specified Acceptable Fuel Design Limits (SAFDLS) 
are still met. Continued compliance with the SLMCPR and other SAFDLs 
will be confirmed on a cycle specific basis consistent with the 
criteria accepted by the NRC.
    Challenges to the Reactor Coolant Pressure Boundary were 
evaluated at CPPU conditions (pressure, temperature, flow, and 
radiation) were found to meet their acceptance criteria for 
allowable stresses and overpressure margin.
    Challenges to the containment have been evaluated, and the 
containment and its associated cooling systems continue to meet 10 
CFR [Part] 50, Appendix A, Criterion 16, Containment Design; 
Criterion 38, Containment Heat Removal; and Criterion 50, 
Containment Design Basis. The increase in the calculated post LOCA 
[loss-of-coolant accident] suppression pool temperature above the 
currently assumed peak temperature was evaluated and determined to 
be acceptable.
    Radiological release events (accidents) have been evaluated, and 
shown to meet the guidelines of 10 CFR 50.67.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

LPRM [Local Power Range Monitor] Calibration Interval Technical 
Specification SR [Surveillance Requirement] Frequency Change

    Response: No.
    The revised surveillance interval continues to ensure that the 
LPRM signal is adequately calibrated. This change will not alter the 
basic operation of process variables, structures, systems, or 
components as described in the SSES FSAR [final safety analysis 
report], and no new equipment is introduced by the change in LPRM 
surveillance interval. The performance of the APRM [average power 
range monitor] and RBM [rod block monitor] systems is not 
significantly affected by the proposed LPRM surveillance interval 
increase. Therefore, the probability of accidents previously 
evaluated is unchanged.
    The proposed change results in no change in radiological 
consequences of the design basis LOCA as currently analyzed for 
SSES. The consequences of an accident can be affected by the thermal 
limits existing at the time of the postulated accident, but LPRM 
chamber exposure has no significant effect on the calculated thermal 
limits because LPRM accuracy does not significantly deviate with 
exposure. For the extended calibration interval, the assumption in 
the safety limit analysis remains valid, maintaining the accuracy of 
the thermal limit calculation. Therefore, the thermal limit 
calculation is not significantly affected by LPRM calibration 
frequency and the consequences of an accident previously evaluated 
are unchanged.
    The change does not affect the initiation of any event, nor does 
it negatively impact the mitigation of any event. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.

RHR [Residual Heat Removal] Service Water System and Ultimate Heat 
Sink Technical Specification and Methods Change

    Response: No.
    The proposed changes do not involve any new initiators for any 
accidents nor do they increase the likelihood of a malfunction of 
any Structures, Systems or Components (SSCs). Implementation of the 
subject changes reduces the probability of adverse consequences of 
accidents previously evaluated, because inclusion of the manual 
spray array bypass isolation valves and the small spray array 
isolation valves in the Technical Specifications (TS) increases 
their reliability to function for safe shutdown. The use of the ANS/
ANSI-5.1-1979 decay heat model in the UHS [ultimate heat sink] 
performance analysis is not relevant to accident initiation, but 
rather, pertains to the method used to evaluate currently postulated 
accidents. Its use does not, in any way, alter existing fission 
product boundaries, and provides a conservative prediction of decay 
heat. Therefore, the change in decay heat calculational method does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

Containment Analysis Methods Change

    Response: No.
    The use of passive heat sinks, and the ANS/ANSI-5.1-1979 decay 
heat model are not relevant to accident initiation, but rather, 
pertain to the method used to evaluate postulated accidents. The use 
of these elements does not, in any way, alter existing fission 
product boundaries, and provides a conservative prediction of the 
containment response to DBA [design-basis accident]-LOCAs. 
Therefore[,] the Containment Analysis Method Change does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.

Feedwater Pump/Condensate Pump Trip Change

    Response: No.
    Feedwater pump trips and condensate pump trips rarely occur. A 
low water level SCRAM on loss of one feedwater pump or one 
condensate pump is bounded by the loss of all feedwater transient in 
Final Safety Analysis Report (FSAR) Appendix 15E. A trip of one 
feedwater pump or a trip of one condensate pump does not result in 
the loss of all feedwater. The Feedwater Pump / Condensate Pump Trip 
Change is included in the CPPU Probabilistic Risk Assessment (PRA). 
The best estimate for the Susquehanna Steam Electric Station (SSES) 
Core Damage Frequency (CDF) risk increase due to the CPPU is 6E-08 
for Unit 1 and 7E-08 for Unit 2 which are in the lower left corner 
of Region III of Regulatory Guide [sic] (Reference 15) (i.e., very 
small risk changes). The best estimate for the Large Early Release 
Frequency (LERF) increase is 1.0E-09/yr for both units which is also 
in the lower left corner of the Region III range of Regulatory Guide 
1.174. Therefore, the Feedwater Pump/Condensate Pump Trip Change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.

Main Turbine Pressure Regulation System

    Response: No.
    Technical Specification 3.7.8 does not directly or indirectly 
affect any plant system, equipment, component, or change the process 
used to operate the plant. Technical Specification 3.7.8 would 
ensure acceptable performance, since it would establish requirements 
for adhering to the appropriate thermal limits, depending on the 
operability of the main turbine pressure regulation system. Use of 
the appropriate limits assures that the appropriate safety limits 
will not be exceeded during normal or anticipated operational 
occurrences. Thus, Technical Specification 3.7.8 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?

Extended Power Uprate

    Response: No.
    Equipment that could be affected by EPU has been evaluated. No 
new operating mode, safety-related equipment lineup, accident 
scenario, or equipment failure mode was identified. The full 
spectrum of accident

[[Page 11394]]

considerations has been evaluated and no new or different kind of 
accident has been identified. CPPU uses developed technology and 
applies it within capabilities of existing or modified plant safety 
related equipment in accordance with the regulatory criteria 
(including NRC approved codes, standards and methods). No new 
accidents or event precursors have been identified.
    The SSES TS require revision to implement EPU. The revisions 
have been assessed and it was determined that the proposed change 
will not introduce a different accident than that previously 
evaluated. Therefore[,] the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.

LPRM Calibration Interval Technical Specification SR Frequency 
Change

    Response: No.
    The proposed change will not physically alter the plant or its 
mode of operation. The performance of the APRM and RBM systems is 
not significantly affected by the proposed LPRM surveillance 
interval increase. As such, no new or different types of equipment 
will be installed and the basic operation of installed equipment is 
unchanged. The methods of governing plant operation and testing are 
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.

RHR Service Water System and Ultimate Heat Sink Technical 
Specification and Methods Change

    Response: No.
    The subject changes apply Technical Specification controls to 
new UHS manual bypass isolation valves and the existing small spray 
array isolation valves. The design functions of the systems are not 
affected.
    The addition of manually operated valves in the system, 
operational changes and the Technical Specification changes do not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    The use of the ANS/ANSI-5.1-1979 decay heat model is not 
relevant to accident initiation, but rather pertains to the method 
used to evaluate currently postulated accidents. The use of this 
analytical tool does not involve any physical changes to plant 
structures or systems, and does not create a new initiating event 
for the spectrum of events currently postulated in the FSAR. 
Further, it does not result in the need to postulate any new 
accident scenarios. Therefore[,] the decay heat calculational method 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated[.]

Containment Analysis Methods Change

    Response: No.
    The use of passive heat sinks and the ANS/ANSI-5.1-1979 decay 
heat model are not relevant to accident initiation, but pertain to 
the method used to evaluate currently postulated accidents. The use 
of these analytical tools does not involve any physical changes to 
plant structures or systems, and does not create a new initiating 
event for the spectrum of events currently postulated in the FSAR. 
Further, they do not result in the need to postulate any new 
accident scenarios. Therefore, the Containment Analysis Method 
Change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.

Feedwater Pump/Condensate Pump Trip Change

    Response: No.
    The occurrence of a reactor SCRAM is already considered in the 
current licensing basis and is not an accident. A SCRAM resulting 
from the trip of a feedwater pump or a condensate pump is bounded by 
a loss of all feedwater event. The loss of all feedwater transient 
is already considered in the plant licensing basis. The SCRAM due to 
the feedwater or condensate pump trip does not change the results of 
the loss of all feedwater transient in any way. Therefore, the 
Feedwater Pump/Condensate Pump Trip Change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.

Main Turbine Pressure Regulation System

    Response: No.
    Technical Specification 3.7.8 will not directly or indirectly 
affect any plant system, equipment, or component and therefore does 
not affect the failure modes of any of these items. Thus, Technical 
Specification 3.7.8 does not create the possibility of a previously 
unevaluated operator error or a new single failure.
    Therefore, Technical Specification 3.7.8 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?

Extended Power Uprate

    Response: No.
    The CPPU affects only design and operational margins. Challenges 
to the fuel, reactor coolant pressure boundary, and containment were 
evaluated for CPPU conditions. Fuel integrity is maintained by 
meeting existing design and regulatory limits. The calculated loads 
on affected structures, systems and components, including the 
reactor coolant pressure boundary, will remain within their design 
allowables for design basis event categories. No NRC acceptance 
criterion is exceeded. Because the SSES configuration and responses 
to transients and postulated accidents do not result in exceeding 
the presently approved NRC acceptance limits, the proposed changes 
do not involve a significant reduction in a margin of safety.

LPRM Calibration Interval Technical Specification Change

    Response: No.
    The proposed change has no impact on equipment design or 
fundamental operation and there are no changes being made to safety 
limits or safety system allowable values that would adversely affect 
plant safety as a result of the proposed change. The performance of 
the APRM and RBM systems is not significantly affected by the 
proposed LPRM surveillance interval increase. The margin of safety 
can be affected by the thermal limits existing prior to an accident; 
however, uncertainties associated with LPRM chamber exposure have no 
significant effect on the calculated thermal limits. For the 
extended calibration interval, the assumption in the safety limit 
analysis remains valid, maintaining the accuracy of the thermal 
limit calculation.
    Since the proposed change does not affect safety analysis 
assumptions or initial conditions, the margin of safety in the 
safety analyses are maintained. Therefore, the proposed change does 
not involve a significant reduction in a margin of safety.

RHR Service Water System and Ultimate Heat Sink Technical 
Specification and Methods Change

    Response: No.
    Implementation of the subject changes does not significantly 
reduce the margin of safety since these changes add components and 
Technical Specification controls for the components not currently 
addressed in the Technical Specifications. These changes increase 
the reliability of the affected components/systems to function for 
safe shutdown.
    Therefore[,] these changes do not involve a significant 
reduction in margin of safety.
    The ANS/ANSI-5.1-1979 model provides a conservative prediction 
of decay heat. The use of this element is consistent with current 
industry standards, and has been previously accepted by the staff 
for use in containment analysis by other licensees, as described in 
GE Nuclear Energy. ``Constant Pressure Power Uprate,'' Licensing 
Topical Report NEDC-33004P-A, Revision 4, dated July 2003; and the 
letter to Gary L. Sozzi (GE) from Ashok Thandani (NRC) on the Use of 
the SHEX Computer Program and ANSI/ANS 5.1-1979, ``Decay Heat Source 
Term for Containment Long-Term Pressure and Temperature Analysis,'' 
July 13, 1993. Therefore, the decay heat calculational method change 
does not involve a significant reduction in the margin of safety.

Containment Analysis Methods Change

    Response: No.
    The use of passive heat sinks and the ANS/ANSI-5.1-1979 decay 
heat model are realistic phenomena, and provide a conservative 
prediction of the plant response to DBA-LOCAs. The use of these 
elements is consistent with current industry standards, and has been 
previously accepted by the staff for other licensees, as described 
in GE Nuclear Energy: ``Constant Pressure Power Uprate,'' Licensing 
Topical Report NEDC-33004P-A, Revision 4, dated July 2003; the 
letter to Gary L. Sozzi (GE) from Ashok Thandani (NRC) on the Use of 
the SHEX Computer Program; and ANSI/ANS 5.1-1979, ``Decay Heat 
Source Term for Containment Long-Term Pressure and Temperature 
Analysis,'' July 13, 1993. Therefore the Containment Analysis Method 
Change does not involve a significant reduction in [a] margin of 
safety.

[[Page 11395]]

Feedwater Pump/Condensate Pump Trip Change

    Response: No.
    A low water level SCRAM on loss of one feedwater pump or one 
condensate pump is bounded by the loss of all feedwater transient in 
FSAR Appendix 15E. The loss of all feedwater transient is a non-
limiting event that does not contribute to the setting of the fuel 
safety limits. Consequently, a SCRAM resulting from a feedwater pump 
or condensate pump trip does not reduce the margin to fuel safety 
limits. Therefore, the potential for a SCRAM resulting from a 
feedwater pump trip or a condensate pump trip does not involve a 
significant reduction in [a] margin of safety.

Main Turbine Pressure Regulation System

    Since Technical Specification 3.7.8 does not alter any plant 
system, equipment, component, or processes used to operate the 
plant, the proposed change will not jeopardize or degrade the 
function or operation of any plant system or component governed by 
Technical Specifications. Technical Specification 3.7.8 preserves 
the margin of safety by establishing requirements for adhering to 
the appropriate thermal limits.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC (Acting) Branch Chief: Douglas V. Pickett.

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 amendment request: February 2, 2007.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) LCO 3.10.1 to expand its scope to 
include provisions for temperature excursions greater than 212 degrees 
F as a consequence of scram time testing initiated in conjunction with 
an inservice leak or hydrostatic test. During these tests and with 
temperature greater than 212 degrees F, operational conditions are 
considered to be in Mode 4.
    The NRC staff issued a notice of availability of a model safety 
evaluation and model no significant hazards consideration (NSHC) 
determination for referencing in license amendment applications in the 
Federal Register on October 27, 2006 (71 FR 63050). The licensee 
affirmed the applicability of the model NSHC determination in its 
application dated February 2, 2007.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    Technical Specifications currently allow for operation at 
greater than 212 deg F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. Extending the activities that can apply this allowance will not 
adversely impact the probability or consequences of an accident 
previously evaluated. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Accident Previously 
Evaluated

    Technical Specifications currently allow for operation at 
greater than 212 deg F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. No new operational conditions beyond those currently allowed by 
LCO 3.10.1 are introduced. The changes do not involve a physical 
alteration of the plant (i.e., no new or different type of equipment 
will be installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements or eliminate any existing requirements. The 
changes do not alter assumptions made in the safety analysis. The 
proposed changes are consistent with the safety analysis assumptions 
and current plant operating practice. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in a Margin of Safety

    Technical Specifications currently allow for operation at 
greater than 212 deg F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. Extending the activities that can apply this allowance will not 
adversely impact any margin of safety. Allowing completion of 
inspections and testing and supporting completion of scram time 
testing initiated in conjunction with an inservice leak or 
hydrostatic test prior to power operation results in enhanced safe 
operations by eliminating unnecessary maneuvers to control reactor 
temperature and pressure. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Branch Chief: Evangelos C. Marinos.

Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant, 
Unit 2, Hamilton County, Tennessee

    Date of amendment request: January 12, 2007.
    Description of amendment request: The proposed amendment would 
revise the steam generator (SG) program requirements in the Sequoyah 
(SQN) Unit 2 Technical Specifications (TSs) to allow use of an SG 
voltage-based repair criteria probability of detection (POD) method 
using plant-specific SG tube inspection results. The proposed POD 
method is referred to as the probability of prior cycle detection 
(POPCD) method.
    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 use of a revised SG voltage-based repair 
criteria POD method, the POPCD method, to determine the BOC 
[beginning of cycle] indication voltage distribution for the SQN 
Unit 2 operational assessments does not increase the probability of 
an accident. Based on industry and plant-specific bobbin detection 
data for ODSCC [outside diameter stress corrosion cracking] within 
the SG tube support plate (TSP) region, large voltage bobbin 
indications which individually can challenge structural or leakage 
integrity can be detected with near 100 percent certainty. Since 
large voltage outside diameter stress corrosion cracking ODSCC 
bobbin indications within the SG TSP can be detected, they will not 
be left in service, and therefore these indications should not be 
included in the voltage distribution for the purpose of operational 
assessments. The POPCD method improves the estimate of potentially 
undetected indications for operational assessments, but does not 
directly affect the inspection results. Since large voltage 
indications are detected, they will not result in an increase in the 
probability of SG tube rupture accident or an increase in the 
consequences of a tube rupture or main steam line break (MSLB) 
accident.

[[Page 11396]]

    2. Does the proposed change create the possibility of a new or 
different accident from any accident previously evaluated?
    Response: No.
    The use of the POPCD method is associated with numerical 
predictions of probabilities for the steam generator tube rupture 
(SGTR) accident. Since the SGTR accident is considered in SQN's 
Updated Final Safety Analysis Report, there is no possibility to 
create a design basis accident that has not been previously 
evaluated. Therefore, the proposed change does not create the 
possibility of a new or different accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No. The use of the POPCD method to determine the BOC 
voltage distribution for the SQN Unit 2 operational assessments does 
not involve a significant reduction in a margin of safety. The 
applicable margin of safety potentially impacted is the SG tube 
structural and leakage criteria. Based on industry and plant-
specific bobbin detection data for ODSCC within the SG TSP region, 
large voltage bobbin indications that can individually challenge 
structural or leakage integrity can be detected with near 100 
percent certainty and will not be left in service. Therefore, these 
indications should not be included in the voltage distribution for 
the purposes of operational assessments. Since these large voltage 
indications are detected, they will not result in a significant 
increase in the actual EOC [end of cycle] leakage for a MSLB 
accident or the actual EOC probability of burst. The POPCD method 
approach to POD considers the potential for missing indications that 
might challenge structural or leakage integrity by applying the 
POPCD data from successive inspections. If a large indication was 
missed in one inspection, it would continue to grow until detected 
in a later inspection. Accordingly, there is no significant increase 
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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Brenda Mozafari (Acting).

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.

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

    Date of amendment request: February 2, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3.6.1.7, ``Suppression Chamber-to-
Drywell Vacuum Breakers,'' to allow a one-time extension to the current 
closure verification surveillance requirement for one of two redundant 
disks in one of nine vacuum breakers until reliable position indication 
can be restored in the main control room during the next refueling 
outage (R-18), which is scheduled to begin on May 12, 2007.
    Date of publication of individual notice in Federal Register: 
February 12, 2007 (72 FR 6606).
    Expiration date of individual notice: February 26, 2007.

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

    Date of application for amendment: December 27, 2006.
    Brief description of amendment: The proposed amendment would revise 
Limiting Condition for Operation 3.14.A to adopt the Technical 
Specification Task Force 484, Revision 0, ``Use of Technical 
Specification 3.10.1 for Scram Time Testing Activities.''
    Date of publication of individual notice in Federal Register: 
February 20, 2007 (72 FR 7776).
    Expiration date of individual notice: March 22, 2007 (public 
comments) and April 23, 2007 (hearing requests).

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

    Date of application for amendment: January 15, 2007.
    Brief description of amendment: The amendment request supercedes 
the previously submitted license amendment request dated April 12, 
2006, proposing new Pressure-Temperature (PT) curves and to extend the 
applicability of current PT limits expressed in Technical Specification 
Figures 3.6.1, 3.6.2, and 3.6.3 through the end of operating cycle 18.
    Date of publication of individual notice in Federal Register: 
February 12, 2007 (72 FR 6609).
    Expiration date of individual notice: March 14, 2007 (public 
comments) and April 13, 2007 (hearing requests).

PSEG Nuclear LLC, Docket No. 50-272, Salem Nuclear Generating Station, 
Unit No. 1, Salem County, New Jersey

    Date of amendment request: January 18, 2007.
    Brief description of amendment request: The amendment request 
proposes a one-time change to the Technical Specifications (TSs) 
regarding the steam generator (SG) tube inspection and repair required 
for the portion of the SG tubes passing through the tubesheet region. 
Specifically, for Salem Unit No. 1 refueling outage 18 (planned for 
spring 2007) and the subsequent operating cycle, the proposed TS 
changes would limit the required inspection (and repair if degradation 
is found) to the portions of the SG tubes passing through the upper 17 
inches of the approximate 21-inch tubesheet region.
    Date of publication of individual notice in Federal Register: 
January 25, 2007 (72 FR 3427).
    Expiration date of individual notice: February 26, 2007 (public 
comments) and March 26, 2007 (hearing requests).

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

[[Page 11397]]

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, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to <a href="/cdn-cgi/l/email-protection#91e1f5e3d1ffe3f2bff6fee7"><span class="__cf_email__" data-cfemail="7e0e1a0c3e100c1d50191108">[email protected]</span></a>.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos. 
1, 2, and 3, Maricopa County, Arizona

    Date of application for amendments: September 28, 2006.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) requirements for mode change limitations in Limiting 
Condition for Operation (LCO) 3.0.4 and Surveillance Requirement 3.0.4 
to adopt the provisions of Industry/TS Task Force (TSTF) Traveler 
number TSTF-359, ``Increase Flexibility in Mode Restraints.'' The 
amendments also revised TS Example 1.4-1 to reflect the changes made to 
LCO 3.0.4 and to be consistent with TSTF-485, which has been 
incorporated into the Standard Technical Specifications Revision 3.1.
    Date of issuance: February 21, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: Unit 1-165, Unit 2--165, Unit 3--165.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Operating License and Technical Specifications.
    Date of initial notice in Federal Register: November 7, 2006 (71 FR 
65140). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 21, 2007.
    No significant hazards consideration comments received: No.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos. 
1, 2, and 3, Maricopa County, Arizona

    Date of application for amendments: February 14, 2006.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) requirements in the Limiting Condition for Operation 
for TS 3.6.3, ``Containment Isolation Valves,'' and associated Actions 
and Surveillance Requirements to allow for a blind flange to be used 
for containment isolation in each of the two flow paths of the 42-inch 
refueling purge valves in Modes 1 through 4, without remaining in TS 
3.6.3 Condition D.
    Date of issuance: February 22, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: Unit 1-166, Unit 2-166, Unit 3-166.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Operating License and Technical Specifications.
    Date of initial notice in Federal Register: March 14, 2006 (71 FR 
13171).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 22, 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: April 26, 2006.
    Brief Description of amendments: Revised the Technical 
Specification (TS) requirements for inoperable snubbers by adding 
Limiting Condition for Operation 3.0.8.
    Date of issuance: February 15, 2007.
    Effective date: February 15, 2007, implement within 90 days.
    Amendment Nos.: 241 and 269.
    Renewed Facility Operating License Nos. DPR-71 and DPR-62: 
Amendments change the TSs.
    Date of initial notice in Federal Register: June 6, 2006 (71 FR 
32603).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 15, 2007.

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

    Date of application for amendments: April 11, 2006, as supplemented 
November 29, 2006.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) related to steam generator tube 
integrity. The changes are consistent with the consolidated line-item 
improvement process, Nuclear Regulatory Commission's approved Technical 
Specification Task Force (TSTF) Standard Specification Change Traveler, 
TSTF-449, Revision 4, ``Steam Generator Tube Integrity.''
    Date of issuance: March 1, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: 237, 218.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: December 5, 2006 (71 FR 
70557) The supplement dated November 29, 2006, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated March 1, 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 20, 2006.
    Brief description of amendment: The amendment removed ANO-2 reactor 
coolant structural integrity requirements contained in TS 3.4.10.1. The 
TS change is consistent with NUREG-1432, ``Standard Technical 
Specifications Combustion Engineering Plants,'' Revision 3.1. The Bases 
for TS 3.4.10.1 will be deleted and performed under the ANO-2 TS Bases 
Control Program, and is not included with the submittal. The amendment 
also renumbers TS pages 3/4 4-22a, 23, 23a, and 23b as TS pages 3/4 4-
23, 24, 25, and 26, respectively.
    Date of issuance: March 1, 2007.

[[Page 11398]]

    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 270.
    Renewed Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications/license.
    Date of initial notice in Federal Register: May 9, 2006 (71 FR 
26999). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 1, 2007.
    No significant hazards consideration comments received: No.

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 application for amendment: November 1, 2006.
    Brief description of amendment: The amendment modified technical 
specification requirements for inoperable snubbers by adding Limiting 
Condition of Operation 3.0.8 using the Consolidated Line Item 
Improvement Process.
    Date of issuance: February 20, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No: 171.
    Facility Operating License No. NPF-29: The amendment revises the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: December 5, 2006 (71 FR 
70558). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 20, 2007.
    No significant hazards consideration comments received: No.

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 application for amendment: November 13, 2006.
    Brief description of amendment: The amendment revised Grand Gulf 
Nuclear Station, Unit 1, Technical Specification (TS) Limiting 
Condition of Operation 3.10.1, and the associated TS Bases, to expand 
its scope to include provisions for temperature excursions greater than 
200 [deg]F as a consequence of inservice leak and hydrostatic testing, 
and as a consequence of scram time testing initiated in conjunction 
with an inservice leak or hydrostatic test, while considering 
operational conditions to be in MODE 4.
    Date of issuance: February 21, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No: 172.
    Facility Operating License No. NPF-29: The amendment revises the 
Operating License and Technical Specifications.
    Date of initial notice in Federal Register: December 19, 2006 (71 
FR 75993). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 21, 2007.
    No significant hazards consideration comments received: No.

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 application for amendment: May 8, 2006, as supplemented by 
letter dated November 16, 2006.
    Brief description of amendment: The change added an NRC-approved 
topical report to the analytical methods referenced in Technical 
Specification Section 5.6.5, ``Core Operating Limits Report (COLR).''
    Date of issuance: February 22, 2007.
    Effective date: As of the date of issuance and shall be implemented 
prior to Cycle 16 operation.
    Amendment No: 173.
    Facility Operating License No. NPF-29: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: June 20, 2006 (71 FR 
35458). The supplement dated November 16, 2006, 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 February 22, 2007.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian 
Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New 
York

    Date of application for amendment: May 31, 2006, as supplemented by 
letter dated August 30, 2006.
    Brief description of amendment: The amendments revise the Technical 
Specifications (TSs) associated with steam generator tube integrity 
consistent with Revision 4 to the TS Task Force (TSTF) Standard 
Technical Specification Change Document TSTF-449, ``Steam Generator 
Tube Integrity.''
    Date of issuance: February 20, 2007.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 251 and 233.
    Facility Operating License Nos. DPR-26 and DPR-64: The amendment 
revised the License and the TSs.
    Date of initial notice in Federal Register: August 1, 2006 (71 FR 
43531). The August 30, 2006, supplement 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 amendment is contained 
in a Safety Evaluation dated February 20, 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: April 4, 2006.
    Brief description of amendments: The amendments add one NRC-
approved topical report reference to the list of analytical methods in 
Technical Specification (TS) Section 5.6.5, ``Core Operating Limits 
Report (COLR),'' that can be used to determine core operating limits 
and delete seven obsolete references from the same TS section.
    Date of issuance: February 15, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 181/168.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications and License.
    Date of initial notice in Federal Register: August 15, 2006 (71 FR 
46933). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 15, 2007.
    No significant hazards consideration comments received: No.

[[Page 11399]]

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

    Date of application for amendments: February 25, 2005, as 
supplemented by letters dated November 11, 2005, April 19, July 10, 
2006, September 1, October 24, December 7, 2006, and February 1, 2007.
    Brief description of amendments: The amendment converts the current 
Technical Specifications to the Improved Technical Specifications 
(ITSs) format and relocates certain requirements to other licensee-
controlled documents. The ITSs are based on NUREG-1431, ``Standard 
Technical Specifications--Westinghouse Plants,'' Revision 2, with the 
Technical Specification Task Force changes to make the Beaver Valley 
Power Station Unit Nos. 1 and 2 (BVPS-1 and 2) ITS more consistent with 
Revision 3; the Commission's Final Policy Statement, ``NRC Final Policy 
Statement on Technical Specification Improvements for Nuclear Power 
Reactors,'' dated July 22, 1993 (58 FR 39132); and 10 CFR 50.36, 
``Technical specifications.'' The purpose of the conversion is to 
provide clearer and more readily understandable requirements in the TSs 
for BVPS-1 and 2 to ensure safe operation. In addition, the amendment 
includes a number of issues that were considered beyond the scope of 
NUREG-1431.
    Date of issuance: February 1, 2007.
    Effective date: As of the date of issuance, and shall be 
implemented within 150 days.
    Amendment Nos.: 278 and 161.
    Facility Operating License Nos. DPR-66 and NPF-73: The amendment 
revised the License and the Technical Specifications.
    Date of initial notice in Federal Register: March 22, 2006 (71 FR 
14554). The letters dated November 11, 2005, April 19, July 10, 2006, 
September 1, October 24, December 7, 2006, and February 1, 2007, 
supplement 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 amendment is contained 
in a Safety Evaluation dated February 1, 2007.
    No significant hazards consideration comments received: No.

FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire

    Date of amendment request: April 28, 2006.
    Description of amendment request: The amendment revised the 
Seabrook Technical Specifications (TSs) Limiting Condition for 
Operation 3.0.4 and Surveillance Requirement (SR) 4.0.4 to adopt the 
provisions of Industry/TS Task Force (TSTF) change TSTF-359, Revision 
9, ``Increased Flexibility in Mode Restraints.'' TSTF-359 is part of 
the consolidated line item improvement process. Specifically, the 
proposed change allows, for systems and components, mode changes into a 
TS condition that has a specific required action and completion time.
    Date of issuance: February 9, 2007.
    Effective date: As of its date of issuance, and shall be 
implemented within 90 days.
    Amendment No.: 114.
    Facility Operating License No. NPF-86: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: July 5, 2006 (71 FR 
38182). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 9, 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: October 23, 2006.
    Brief description of amendments: The amendments to the Technical 
Specifications (TSs) eliminate the use of the defined term CORE 
ALTERATIONS in the TSs.
    Date of issuance: February 15, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 224 & 230.
    Renewed Facility Operating License Nos. DPR-24 and DPR-27: 
Amendments revised the Technical Specifications/License.
    Date of initial notice in Federal Register: December 5, 2006 (71 FR 
70562). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 15, 2007.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: February 13, 2006.
    Brief description of amendments: The amendments revise Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Technical 
Specifications (TS) to change the wording in TS 3.0, ``Surveillance 
Requirement (SR) Applicability'' and change format and titles in TS 
5.0, ``Administrative Controls.'' The proposed changes improve the TS 
usability, conformance with the industry standard, NUREG-1431, 
``Standard Technical Specifications, Westinghouse Plants,'' Revision 
3.0 and accuracy.
    Date of issuance: February 13, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 176 and 166.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 11, 2006 (71 FR 
18375). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 13, 2007.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: November 13, 2006.
    Brief description of amendment: The amendment relocated the 
requirements of Technical Specification (TS) 2.22, ``Toxic Gas 
Monitors,'' and TS Table 3-3, Item 29, to the Fort Calhoun Station, 
Unit No. 1, Updated Safety Analysis Report.
    Date of issuance: February 28, 2007.
    Effective date: As of its date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment No.: 248.
    Renewed Facility Operating License No. DPR-40: The amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 19, 2006 (71 
FR 75996). The Commission's related evaluation of the amendment is 
contained in a safety evaluation dated February 28, 2007.
    No significant hazards consideration comments received: No.

[[Page 11400]]

PPL Susquehanna, LLC, Docket No. 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: April 28, 2006.
    Brief description of amendments: The amendments revise the SSES 1 
and 2 Technical Specifications 3.1.7, ``Standby Liquid Control (SLC) 
System,'' to modify the SLC system for single loop pump operation and 
the use of enriched sodium pentaborate solution.
    Date of issuance: February 28, 2007.
    Effective date: As of the date of issuance and to be implemented 
prior to the startup following the SSES 1 Spring 2008 15th refueling 
outage and SSES 2 Spring 2007 13th refueling outage for Units 1 and 2, 
respectively.
    Amendment Nos.: 240 and 217.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the TSs and license.
    Date of initial notice in Federal Register: August 15, 2006 (71 FR 
46936). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 28, 2007.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-259 Browns Ferry Nuclear 
Plant, Unit 1, Limestone County, Alabama

    Date of application for amendment: May 1, 2006 (TS-455), as 
supplemented by letters dated September 1, and November 6, 2006.
    Brief description of amendment: The amendment revises the numeric 
values of the safety limit critical power ratio (SLMCPR) in the 
Technical Specification (TS) Section 2.1.1.2 for one and two reactor 
recirculation loop operation to incorporate the results of the Cycle 7 
SLMCPR analysis.
    Date of issuance: February 6, 2007.
    Effective date: Date of issuance, to be implemented within 60 days.
    Amendment No.: 267.
    Facility Operating License Nos. DPR-33: Amendment revised the TSs.
    Date of initial notice in Federal Register: August 15, 2006 (71 FR 
46937). The supplements dated September 1, and November 6, 2006, 
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 February 6, 2007.
    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: August 22, 2005, as supplemented by 
letters dated September 18 and October 23, 2006.
    Brief description of amendments: The amendments revised the Final 
Safety Evaluation Report Sections 1, 6, and 15. The changes reflect the 
licensee's adoption of Nuclear Regulatory Commission's Regulatory Guide 
1.195, ``Methods and Assumptions for Evaluating Radiological 
Consequences of Design Basis Accidents at Light-Water Reactors,'' for 
calculating radiological consequences and replacement of steam 
generators for Comanche Peak Steam Electric Station, Unit 1, in the 
spring of 2007.
    Date of issuance: February 20, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: 130/130.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Final Safety Analysis Report and Facility Operating 
Licenses.
    Date of initial notice in Federal Register: November 8, 2005 (70 FR 
67754). The supplements dated September 18 and October 23, 2006, 
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 February 20, 2007.
    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: February 21, 2006, as supplemented by 
letters dated September 12 and December 14, 2006.
    Brief description of amendments: The amendments increased the 
allowable values (AVs) for steam generator (SG) water level trip 
setpoints and the required minimum SG secondary side water inventory in 
shutdown modes for the replacement SGs in Comanche Peak Steam Electric 
Station (CPSES), Unit 1. For CPSES Unit 2, the corresponding AVs and 
the SG secondary water inventory in the current TSs remain unchanged 
since the existing SGs in Unit 2 will continue to be used.
    Date of issuance: February 20, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: NPF-87--131; NPF-89--131.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: June 6, 2006 (71 FR 
32609). The supplements dated September 12 and December 14, 2006, 
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 February 20, 2007.
    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: December 16, 2005, as supplemented by 
letters dated August 31 and September 29, 2006.
    Brief description of amendments: The amendments revised Technical 
Specifications (TSs) 1.1 and 5.6.6 consistent with the Nuclear 
Regulatory Commission (NRC)-approved Technical Specification Task Force 
(TSTF) Standard Technical Specification Change Traveler, TSTF-419, 
``Revise PTLR [Pressure Temperature Limits Report] Definition and 
References in ISTS [Improved Standard Technical Specification] 5.6.6.
    Date of issuance: February 22, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: NPF-87-132 and NPF-89-132.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: March 14, 2006 (71 FR

[[Page 11401]]

13182). The supplements dated August 31 and September 29, 2006, 
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 February 22, 2007.
    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: December 12, 2005.
    Brief description of amendments: The amendments revise the 
Technical Specification (TS) Surveillance Requirements (SRs) 3.3.1.2 
and 3.3.1.3, ``Reactor Trip System (RTS) Instrumentation.'' The license 
amendment request is based on Technical Specification Task Force (TSTF) 
Traveler, TSTF-371-A, Revision 1, ``NIS [Nuclear Instrumentation 
System] Power Range Channel Daily SR TS Change to Address Low Power 
Decalibration.'' TSTF-371-A, Revision 1, revised the requirements for 
performing a daily surveillance adjustment of the power range 
channel(s) to address industry concern that compliance with SR 3.3.1.2 
and SR 3.3.1.3 may result in a non-conservative channel calibration 
during reduced-power operations. The changes resolved the issue of non-
conservatism.
    Date of issuance: February 26, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: NPF-87-133, NPF-89-133.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: March 28, 2006 (71 FR 
15490).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 26, 2007.
    No significant hazards consideration comments received: No.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application for amendment: May 30, 2006, as supplemented by 
letters dated November 22 and December 19, 2006.
    Brief description of amendment: The amendment revised Surveillance 
Requirements (SRs) 3.5.2.8 and 3.6.7.1 due to (1) the future 
replacement of the existing containment recirculation sump suction 
inlet trash racks and screens with strainers, (2) the resulting 
relocation of the recirculation fluid pH control (RFPC) system from the 
sump, and (3) the removal of details from SR 3.6.7.1, including the 
relocation of the name of the RFPC chemical to a license condition in 
Appendix C to the license. The modifications will be done in the 
refueling outage scheduled for the spring of 2007. The amendment also 
deleted the footnote to the frequency for SR 3.5.2.5 because it is no 
longer applicable.
    Date of issuance: February 21, 2007.
    Effective date: As of its date of issuance, and shall be 
implemented prior to entry into Mode 4 during the plant startup from 
the refueling outage scheduled for the spring of 2007.
    Amendment No.: 180.
    Facility Operating License No. NPF-30: The amendment revised the 
Operating License and the Technical Specifications.
    Date of initial notice in Federal Register: August 15, 2006 (71 FR 
46940). The supplemental letters dated November 22 and December 19, 
2006, 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 published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 21, 2007.
    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.

[[Page 11402]]

    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 51.12(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 
application for amendment, (2) the amendment to Facility Operating 
License, 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 System's (ADAMS) Public Electronic Reading Room 
on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems 
in accessing the documents located in ADAMS, contact the PDR Reference 
staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to <a href="/cdn-cgi/l/email-protection#166672645678647538717960"><span class="__cf_email__" data-cfemail="5b2b3f291b352938753c342d">[email protected]</span></a>.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. 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, and electronically on the Internet at the NRC Web site, 
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are 
problems in accessing the document, contact the PDR Reference staff at 
1 (800) 397-4209, (301) 415-4737, or by e-mail to <a href="/cdn-cgi/l/email-protection#eb9b8f99ab859988c58c849d"><span class="__cf_email__" data-cfemail="e4948096a48a9687ca838b92">[email protected]</span></a>. If a 
request for a hearing or petition for leave to intervene is filed by 
the above date, 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 identify 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 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to those specific sources and 
documents of which the petitioner is aware and on which the petitioner 
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.\1\ 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 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.
---------------------------------------------------------------------------

    \1\ To the extent that the applications contain attachments and 
supporting documents that are not publicly available because they 
are asserted to contain safeguards or proprietary information, 
petitioners desiring access to this information should contact the 
applicant or applicant's counsel and discuss the need for a 
protective order.
---------------------------------------------------------------------------

    Each contention shall be given a separate numeric or alpha 
designation within one of the following groups:
    1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the 
applications.
    2. Environmental--primarily concerns/issues relating to matters 
discussed or referenced in the environmental analysis for the 
applications.
    3. Miscellaneous--does not fall into one of the categories outlined 
above.
    As specified in 10 CFR 2.309, if two or more petitioners/requestors 
seek to co-sponsor a contention, the petitioners/requestors shall 
jointly designate a representative who shall have the authority to act 
for the petitioners/requestors with respect to that contention. If a 
petitioner/requestor seeks to adopt the contention of another 
sponsoring petitioner/requestor, the petitioner/requestor who seeks to 
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that 
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the 
petitioners/requestors with respect to that contention.
    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. Since the Commission has made a final determination that the 
amendment involves no significant hazards consideration, if a hearing 
is requested, it will not stay the effectiveness of the amendment. Any 
hearing held would take place while the amendment is in effect.
    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

[[Page 11403]]

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, 
<a href="/cdn-cgi/l/email-protection#feb69b9f8c979099ba919d959b8abe908c9dd0999188"><span class="__cf_email__" data-cfemail="a4ecc1c5d6cdcac3e0cbc7cfc1d0e4cad6c78ac3cbd2">[email protected]</span></a>; 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 <a href="/cdn-cgi/l/email-protection#c9868e8a84a8a0a58aaca7bdacbb89a7bbaae7aea6bf"><span class="__cf_email__" data-cfemail="c08f87838da1a9ac83a5aeb4a5b280aeb2a3eea7afb6">[email protected]</span></a>. 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 or 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).

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

    Date of amendment request: February 2, 2007.
    Description of amendment request: The amendment revised Technical 
Specification 3.6.1.7, ``Suppression Chamber-to-Drywell Vacuum 
Breakers,'' to allow a one-time extension to the current closure 
verification surveillance requirement for one of two redundant disks in 
one of nine vacuum breakers until reliable position indication can be 
restored in the main control room during the next refueling outage (R-
18), which is scheduled to begin on May 12, 2007.
    Date of issuance: February 27, 2007.
    Effective date: As of its date of issuance and shall be implemented 
within 14 days from the date of issuance.
    Amendment No.: 202.
    Facility Operating License No.: NPF-21: Amendment revises the 
technical specifications and license.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes. 72 FR 6606, published February 12, 2007. The 
notice provided an opportunity to submit comments on the Commission's 
proposed NSHC determination. No comments have been received. The notice 
also provided an opportunity to request a hearing within 60 days after 
the date of publication of the notice, but indicated that if the 
Commission makes a final NSHC determination, any such hearing would 
take place after issuance of the amendment.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, state consultation, and final NSHC determination 
are contained in a safety evaluation dated February 27, 2007.
    Attorney for licensee: William A. Horin, Esq., Winston & Strawn, 
1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: David Terao.

    Dated at Rockville, Maryland, this 2nd day of March 2007.

    For the Nuclear Regulatory Commission.
Michael C. Cheok,
Acting Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. E7-4251 Filed 3-12-07; 8:45 am]
BILLING CODE 7590-01-P


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