[Federal Register Volume 72, Number 38 (Tuesday, February 27, 2007)]
[Notices]
[Pages 8800-8808]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-3199]


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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 2, 2007 through February 14, 2007. 
The last biweekly notice was published on February 13, 2007 (72 FR 
6780).

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.

[[Page 8801]]

    Within 60 days after the date of publication of this notice, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, <a href="/cdn-cgi/l/email-protection#afe7caceddc6c1c8ebc0ccc4cadbefc1ddcc81c8c0d9"><span class="__cf_email__" data-cfemail="e3ab8682918a8d84a78c80888697a38d9180cd848c95">[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#e3aca4a0ae828a8fa0868d978691a38d9180cd848c95"><span class="__cf_email__" data-cfemail="68272f2b250901042b0d061c0d1a28061a0b460f071e">[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#60100412200e12034e070f16"><span class="__cf_email__" data-cfemail="b0c0d4c2f0dec2d39ed7dfc6">[email protected]</span></a>.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of amendment request: August 2, 2006.
    Description of amendment request: The proposed amendment will 
modify the statistical summation error term ``Z'' and one of the 
allowable values for certain steam generator water level trip setpoints 
used in the Reactor Trip System and Engineered Safety Feature Actuation 
System instrumentation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


[[Page 8802]]


    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to revise the statistical summation error 
term ``Z'' and one of the allowable values for certain steam 
generator water level (SGWL) reactor protection and engineered 
safety feature actuation functions continues to follow the current 
setpoint methodology previously approved for HNP [Shearon Harris 
Nuclear Power Plant, Unit 1] while addressing newly identified level 
uncertainty considerations. The proposed change does not alter the 
installed plant configuration for the affected instrumentation or 
the associated equipment system interfaces. The proposed change 
continues to maintain the assumptions for the specified instrument 
loops used in the Final Safety Analysis Report (FSAR) for HNP, and 
the channel statistical allowances (CSA) or calculated total loop 
uncertainties remain bounded by the total allowance (TA) values 
presented in the HNP Technical Specifications (TS). The proposed 
change does not alter the accident analyses or the causes for any 
accident described in the FSAR that credit the SGWL setpoint 
actuations. The proposed amendment will not modify, degrade, prevent 
actions or alter any assumptions previously made in evaluating the 
radiological consequences of an accident described in the FSAR.
    Therefore, this amendment 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 to revise the statistical summation error 
term ``Z'' and one of the allowable values for certain SGWL reactor 
protection and engineered safety feature actuation functions 
addresses newly identified level uncertainty considerations. The 
proposed change does not implement any physical changes to the 
systems, structures, or components for the affected instrumentation 
loops or to the associated equipment system interfaces. No new or 
different accident initiators or sequences are created by the 
proposed change. The proposed change continues to maintain the 
safety analysis limits used in the safety analyses that credit the 
specified actuation functions.
    Therefore, this amendment 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 to revise the statistical summation error 
term ``Z'' and one of the allowable values for certain SGWL reactor 
protection and engineered safety feature actuation functions 
addresses newly identified level uncertainty considerations and does 
not involve a reduction in the margin of safety for plant operation. 
Consistent with the requirements of the HNP FSAR, the proposed 
change has been evaluated to ensure that the assumptions for the 
specified instrument loops used in the FSAR continue to be 
maintained and that the CSA or calculated total loop uncertainties 
remain bounded by the TA values presented in the HNP TS. The 
proposed change continues to follow the current setpoint methodology 
previously approved for HNP, and the revised uncertainty analysis 
results in acceptable calculational margin.
    Therefore, this amendment does not involve a significant 
reduction in a margin of safety.

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

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of amendment request: December 20, 2006.
    Description of amendment request: The amendment will revise 
Technical Specification (TS) 6.12 ``High Radiation Area.'' 
Specifically, the proposed amendment would align the requirements with 
the revised 10 CFR 20 as described in Regulatory Guide 8.38, Revision 
1, ``Control of Access to High and Very High Radiation Areas in Nuclear 
Power Plants.''
    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 changes are administrative and affect personnel access 
control requirements for high radiation areas. The changes do not 
affect the operation, physical configuration, or function of plant 
equipment or systems. The changes do not impact the initiators or 
assumptions of analyzed events; nor do they impact the mitigation of 
accidents or transient events. Therefore, these changes do not 
increase the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new of 
[or] different kind of accident from any accident previously 
evaluated?
    Response: No.
    The changes are administrative and affect personnel access 
control requirements for high radiation areas. The changes do not 
alter plant configuration, require installation of new equipment, 
alter assumptions about previously analyzed accidents, or impact the 
operation or function of plant equipment or systems. Therefore, 
these changes will not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The changes are administrative and affect personnel access 
control requirements for high radiation areas. The changes do not 
impact any safety assumptions; nor do the changes have the potential 
to reduce any margin of safety as described in the HNP [Shearon 
Harris Nuclear Power Plant, Unit 1] TS Bases. The proposed changes 
maintain an equivalent level of protection for radiation workers 
and, thereby, provide reasonable assurance that individuals will not 
exceed regulatory dose limits. The proposed changes are consistent 
with: (1) The guidance of Regulatory Guide (RG) 8.38, ``Control of 
Access to High and Very High Radiation Areas in Nuclear Power 
Plants,'' Section C, Regulatory Position 2.4, Alternative Methods 
for Access Control, with the exception that ``should'' has been 
changed to ``shall''; and (2) other nuclear plants' existing TSs 
such as those at Brunswick Steam Electric Plant Units 1 & 2. Based 
on this evaluation, the proposed change does not involve a 
significant reduction in a margin of safety.

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

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

    Date of amendment request: January 4, 2007.
    Description of amendment request: The proposed amendments would 
remove gaseous radioactivity monitoring from the Technical 
Specifications (TSs) as an acceptable option for reactor coolant 
leakage detection.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 8803]]

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:
    Pursuant to 10 CFR 50.91, Duke has made the determination that 
this amendment request does not involve a significant hazards 
consideration by applying the standards established by the NRC 
regulations in 10 CFR 50.92. This ensures 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.
    The removal of the gaseous containment atmosphere radioactivity 
monitor from [the] TS as an acceptable alternative to the 
particulate containment atmosphere radioactivity monitor will not 
reduce the number of operable leak detection channels which the 
Technical Specification LCO [limiting condition for operation] 
currently provides. The gaseous monitor which is being removed from 
[the] Technical Specifications is the least sensitive and has the 
highest response time of the three available leakage monitors 
currently in the Technical Specification. The remaining particulate 
radioactivity monitor will provide greater leak detection capability 
by comparison. Therefore, removal of the gaseous radioactivity 
monitor from the Technical Specification LCO cannot increase the 
probability or consequence of an accident.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    RCS [reactor coolant system] leakage detection instrumentation 
functions to provide control room operators with information which 
is indicative of a degraded RCS pressure boundary. Removal of RIA 49 
from [the] TS will, in effect, remove the ``weakest link'' in the 
leakage detection system requirements of the LCO. It is important to 
note that RIA 49 will remain available. The change only removes it 
from the LCO, not from the plant. So, the result will be an enhanced 
capability for detecting RCS leakage in a timely manner. This 
enhancement, although small, could enable the operator to identify a 
precursor to a LOCA [loss-of-coolant accident] and take actions to 
safely shutdown the plant for repairs prior to actually experiencing 
a significant transient (LOCA). While the leakage detection system 
cannot prevent all LOCAs, these are accidents which have been 
evaluated in the UFSAR [updated final safety analysis report]. In no 
case would this enhancement be capable of creating a new or 
different kind of accident than previously evaluated.
    (3) Involve a significant reduction in a margin of safety.
    The proposed change does not reduce the number of instrument 
channels required by the LCO for the leakage detection system. The 
LCO will still ensure that both a normal sump level instrument and a 
containment atmosphere radioactivity instrument are operable as 
before. It only removes one available option for satisfying the 
requirement for a containment atmosphere radioactivity monitor. The 
remaining containment atmosphere radioactivity monitor has greater 
sensitivity and faster response time than the monitor that is being 
removed from the Technical Specification. No other plant equipment 
is affected by the proposed change. Thus, there is no adverse impact 
on the capability to detect an RCS leak. 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: 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.

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

    Date of amendment request: May 31, 2005, as supplemented by letters 
dated February 8, 2006, and January 5, 2007.
    Description of amendment request: The proposed amendment modifies 
Technical Specification (TS) Sections 3.8.1, ``AC [Alternating Current] 
Sources--Operating,'' 3.8.4, ``DC [Direct Current] Sources--
Operating,'' 3.8.5, ``DC Sources--Shutdown,'' 3.8.6, ``Battery Cell 
Parameters,'' and 5.5, ``Programs and Manuals.'' The proposed change 
incorporates clarifying requirements in surveillance testing of diesel 
generators and new actions for an inoperable battery charger. The 
proposed change includes a revision to the Administrative Program to be 
consistent with Institute of Electrical and Electronics Engineers 
Standard 450-2002, and changes consistent with TS Task Force (TSTF) 
Traveler TSTF-360, Revision 1, ``DC Electrical Rewrite,'' and TSTF-283, 
Revision 3, ``Modify Section 3.8 Mode Restriction Notes.''
    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 emergency diesel generators (DGs) and their associated 
emergency loads are accident-mitigating features. As such, testing 
of the DGs themselves is not associated with any potential accident 
initiating mechanism. Each DG is dedicated to a specific vital bus 
and these buses and DGs are independent of each other. There is no 
common mode failure provided by the testing changes proposed in this 
license amendment request (LAR) that would cause multiple bus 
failures. Therefore, there will be no significant impact on any 
accident probabilities by the approval of the requested amendment.
    SR [surveillance requirement] changes that are consistent with 
Industry/Technical Specification Task Force (TSTF) Standard 
Technical Specification (STS) change TSTF-283, Revision 3 have been 
approved by the NRC and the online tests allowed by the TSTF are 
only to be performed for the purpose of establishing operability. 
Performance of these SRs during normally restricted modes will 
require an assessment to assure plant safety is maintained or 
enhanced.
    The proposed changes restructure the TS for the direct current 
(DC) electrical power system, consistent with TSTF-360, Revision 1. 
The proposed changes add actions to specifically address battery and 
battery charger inoperability. The DC electrical power system, 
including associated battery chargers, is not an initiator of any 
accident sequence analyzed in the Final Safety Analysis Report 
(FSAR). Operation in accordance with the proposed TS ensures that 
the DC electrical power system is capable of performing its function 
as described in the FSAR. Therefore, the mitigating functions 
supported by the DC electrical power system will continue to provide 
the protection assumed by the analysis.
    The relocation of preventive maintenance surveillances, and 
certain operating limits and actions, to a newly-created licensee-
controlled Battery Monitoring and Maintenance Program will not 
challenge the ability of the DC electrical power system to perform 
its design function. Appropriate monitoring and maintenance, 
consistent with industry standards, will continue to be performed. 
In addition, the DC electrical power system is within the scope of 
10 CFR 50.65, ``Requirements for monitoring the effectiveness of 
maintenance at nuclear power plants,'' which will ensure the control 
of maintenance activities associated with the DC electrical power 
system.
    The integrity of fission product barriers, plant configuration, 
and operating procedures as described in the FSAR will not be 
affected by the proposed changes. Therefore, the consequences of 
previously analyzed accidents will not increase by implementing 
these changes.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.

[[Page 8804]]

    The proposed changes involve restructuring the TS for the DC 
electrical power system. The DC electrical power system, including 
associated battery chargers, is not an initiator to any accident 
sequence analyzed in the FSAR. Rather, the DC electrical power 
system is used to supply equipment used to mitigate an accident.
    The proposed change would create no new accidents since no 
changes are being made to the plant that would introduce any new 
accident causal mechanisms. Diesel Generators will be operated in 
the same configuration currently allowed by other DG SRs that allow 
testing in plant Modes 1 and 2 and 3. This license amendment request 
does not impact any plant systems that are accident initiators or 
adversely impact any accident mitigating systems.
    Therefore, the proposed changes do 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 proposed change does not involve a significant reduction in 
the margin of safety. The margin of safety is related to the ability 
of the fission product barriers to perform their design functions 
during and following an accident situation. These barriers include 
the fuel cladding, the reactor coolant system, and the containment 
system. The proposed changes to the testing requirements for the 
plant DGs do not affect the operability requirements for the DGs, as 
verification of such operability will continue to be performed as 
required. Continued verification of operability supports the 
capability of the DGs to perform their required function of 
providing emergency power to plant equipment that supports or 
constitutes the fission product barriers. Consequently, the 
performance of these fission product barriers will not be impacted 
by implementation of this proposed amendment.
    In addition, the margin of safety is established through 
equipment design, operating parameters, and the setpoints at which 
automatic actions are initiated. The proposed changes will not 
adversely affect operation of plant equipment. These changes will 
not result in a change to the setpoints at which protective actions 
are initiated. Sufficient AC and DC capacity to support operation of 
mitigation equipment is ensured. The changes associated with the new 
battery maintenance and monitoring program will ensure that the 
station batteries are maintained in a highly reliable manner. The 
equipment fed by the DC electrical sources will continue to provide 
adequate power to safety related loads in accordance with analysis 
assumptions.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant, 
Van Buren County, Michigan

    Date of amendment request: September 25, 2006.
    Description of amendment request: The proposed amendment would 
revise the Palisades Nuclear Plant (PNP) licensing bases to adopt the 
alternative source term (AST) as described in Title 10 of the Code of 
Federal Regulations (CFR) Section 50.67 following the guidance provided 
in Regulatory Guide (RG) 1.183. This application includes an amendment 
to the Technical Specifications, Definition 1.1, Dose Equivalent I-131.
    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 amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Response: No.
    Alternative source term calculations have been performed for PNP 
that demonstrate the dose consequences remain below limits specified 
in NRC Regulatory Guide 1.183 and 10 CFR 50.67. The proposed change 
does not modify the design or operation of the plant. The use of an 
AST changes only the regulatory assumptions regarding the analytical 
treatment of the design basis accidents and has no direct effect on 
the probability of any accident.
    The AST has been utilized in the analysis of the limiting design 
basis accidents listed above [Loss-of-Coolant Accident, Main Steam 
Line Break, Steam Generator Tube Rupture, Small Line Break Outside 
Containment, Control Rod Ejection, Fuel Handling Accident, and Spent 
Fuel Cask Drop]. The results of the analyses, which include the 
proposed changes to the Technical Specifications, demonstrate that 
the dose consequences of these limiting events are all within the 
regulatory limits.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Response: No.
    The proposed change does not affect any plant structures, 
systems, or components. The proposed operation of plant systems and 
equipment affected by this change does not create the possibility of 
a new or different kind of accident previously evaluated. The 
proposed modifications and post-modification testing are intended to 
enhance the capability of the plant to comply with the revised post 
accident dose results presented in this submittal. Since the 
alternative source term is a revised methodology used to estimate 
resulting accident doses, it is not an accident initiator.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    Response: No.
    The proposed implementation of the alternative source term 
methodology is consistent with NRC Regulatory Guide 1.183. 
Conservative methodologies, per the guidance of RG 1.183, have been 
used in performing the accident analyses. The radiological 
consequences of these accidents are all within the regulatory 
acceptance criteria associated with use of the alternative source 
term methodology.
    The proposed changes continue to ensure that the doses at the 
exclusion area and low population zone boundaries and in the control 
room are within the corresponding regulatory limits of RG 1.183 and 
10 CFR 50.67. The margin of safety for the radiological consequences 
of these accidents is considered to be that provided by meeting the 
applicable regulatory limits, which are set at or below the 10 CFR 
50.67 limits. An acceptable margin of safety is inherent in these 
limits.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

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

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: December 14, 2006.
    Description of amendment request: The proposed amendments would 
revise the reference to ``trash racks and screens'' in Technical 
Specification (TS) 3.5.2, ``ECCS [Emergency Core Cooling System]--
Operating'', Surveillance Requirement (SR) 3.5.2.8 and revise the

[[Page 8805]]

required Refueling Water Storage Tank (RWST) level in TS 3.5.4, 
``Refueling Water Storage Tank (RWST).'' This License Amendment Request 
(LAR) fulfills the commitment made in the supplement to Nuclear 
Management Company Response to Generic Letter 2004-02, ``Potential 
Impact of Debris Blockage on Emergency Recirculation During Design 
Basis Accidents at Pressurized-Water Reactors,'' to submit an LAR to 
revise SR 3.5.2.8 by December 31, 2006.
    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 revise the Technical 
Specifications by changing the containment sump inlet debris 
interceptor description in Surveillance Requirement 3.5.2.8 and 
increasing the Refueling Water Storage Tank level in Surveillance 
Requirement 3.5.4.1 to 265,000 gallons which corresponds to 
approximately 90% indicated instrumentation level. These changes 
support resolution of containment sump blockage issues raised in 
Nuclear Regulatory Commission Bulletin 2003-01, ``Potential Impact 
Of Debris Blockage On Emergency Sump Recirculation At Pressurized-
Water Reactors'' and Generic Letter 2004-02, ``Potential Impact Of 
Debris Blockage On Emergency Recirculation During Design Basis 
Accidents At Pressurized-Water Reactors.''
    The containment sump inlet debris interceptor is a plant design 
feature which mitigates accidents and does not initiate accidents. 
Therefore, the proposed change does not involve a significant 
increase in the probability of an accident. The new sump strainers 
for use as debris interceptors have been evaluated to withstand the 
applicable post accident loads without trash racks and thus the 
description change in Surveillance Requirement 3.5.2.8 does not 
involve a significant increase in the consequences of an accident 
previously evaluated.
    The Refueling Water Storage Tank is required for accident 
mitigation and is not an accident initiator, thus requiring 
additional water volume in the tank does not involve a significant 
increase in the probability of an accident previously evaluated. 
Since the proposed change increases the water volume in the 
Refueling Water Storage Tank available for accident mitigation, this 
change may decrease the consequences of an accident.
    Thus, the changes proposed in this license amendment request 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 revise the Technical 
Specifications by changing the containment sump inlet debris 
interceptor description in Surveillance Requirement 3.5.2.8 and 
increasing the Refueling Water Storage Tank level in Surveillance 
Requirement 3.5.4.1 to 265,000 gallons which corresponds to 
approximately 90% indicated instrumentation level. These changes 
support resolution of containment sump blockage issues raised in 
Nuclear Regulatory Commission Bulletin 2003-01, ``Potential Impact 
Of Debris Blockage On Emergency Sump Recirculation At Pressurized-
Water Reactors'' and Generic Letter 2004-02, ``Potential Impact Of 
Debris Blockage On Emergency Recirculation During Design Basis 
Accidents At Pressurized-Water Reactors.''
    The proposed Technical Specification containment sump suction 
inlet debris interceptor description revision does not create the 
possibility of a new or different kind of accident. There are no new 
failure modes or mechanisms created by the new strainers and there 
are no new accident precursors generated due to this change. The new 
strainers do not change the way in which the plant is operated.
    The proposed Technical Specification Refueling Water Storage 
Tank level increase does not involve a change in system operation or 
the use of the Refueling Water Storage Tank. It does increase the 
quantity of water in the Refueling Water Storage Tank available for 
accident mitigation. There are no new failure modes or mechanisms 
created by the availability or use of an additional water volume in 
the Refueling Water Storage Tank as proposed by this Technical 
Specification change. There are no new accident precursors generated 
with the storage of additional water in the Refueling Water Storage 
Tank.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    This license amendment request proposes to revise the Technical 
Specifications by changing the containment sump inlet debris 
interceptor description in Surveillance Requirement 3.5.2.8 and 
increasing the Refueling Water Storage Tank level in Surveillance 
Requirement 3.5.4.1 to 265,000 gallons which corresponds to 
approximately 90% indicated instrumentation level. These changes 
support resolution of containment sump blockage issues raised in 
Nuclear Regulatory Commission Bulletin 2003-01, ``Potential Impact 
Of Debris Blockage On Emergency Sump Recirculation At Pressurized-
Water Reactors'' and Generic Letter 2004-02, ``Potential Impact Of 
Debris Blockage On Emergency Recirculation During Design Basis 
Accidents At Pressurized-Water Reactors.''
    The proposed Technical Specification containment sump debris 
interceptor description revision does not involve a significant 
reduction in a margin of safety. The new sump strainers for use as 
debris interceptors have been evaluated to withstand the applicable 
post accident loads without trash racks and thus do not involve a 
significant reduction in a margin of safety. The new strainers 
provide additional debris interceptor flow area to the sump and thus 
may improve plant margins of safety.
    The proposed change will increase the required water volume to 
be stored in the Refueling Water Storage Tank which means additional 
water will be available to mitigate accidents. This change does not 
involve a decrease in the margin of safety, but may involve an 
increase in the margin of safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: February 21, 2006.
    Brief description of amendments: The amendments revise the 
Technical Specification (TS) 1.1, ``Definitions,'' and TS 3.4.16, ``RCS 
[Reactor Coolant System] Specific Activity,'' by removing the current 
TS 3.4.16 limits on RCS gross-specific activity with a new dose 
equivalent XE-133 definition that would replace the current E-bar 
average disintegration energy definition. In addition, the current dose 
equivalent I-131 definition would be revised to allow the use of 
alternate, NRC-approved thyroid dose conversion factors.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to add new thyroid dose conversion factor 
reference[s] to the definition of DOSE EQUIVALENT I-131, eliminate 
the definition of E--AVERAGE DISINTEGRATION ENERGY, add a new

[[Page 8806]]

definition of DOSE EQUIVALENT XE-133, replace the Technical 
Specification (TS) 3.4.16 limit on reactor coolant system (RCS) 
gross specific activity with a limit on noble gas specific activity 
in the form of a Limiting Condition for Operation (LCO) on DOSE 
EQUIVALENT XE-133, replace TS Figure 3.4.16-1 with a maximum limit 
on DOSE EQUIVALENT I-131, extend the Applicability of LCO 3.4.16, 
and make corresponding changes to TS 3.4.16 to reflect all of the 
above are not accident initiators and have no impact on the 
probability of occurrence for any design basis accidents.
    The proposed changes will have no impact on the consequences of 
a design basis accident because they will limit the RCS noble gas 
specific activity to be consistent with the values assumed in the 
radiological consequence analyses. The changes will also limit the 
potential RCS iodine concentration excursion to the value currently 
associated with full power operation, which is more restrictive on 
plant operation than the existing allowable RCS iodine specific 
activity at lower power levels.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not alter any physical part of the plant 
nor do they affect any plant operating parameters besides the 
allowable specific activity in the RCS. The changes which impact the 
allowable specific activity in the RCS are consistent with the 
assumptions assumed in the current radiological consequence 
analyses.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The acceptance criteria related to the proposed changes involve 
the allowable Control Room and offsite radiological consequences 
following a design basis accident. The proposed changes will have no 
impact on the radiological consequences of a design basis accident 
because they will limit the RCS noble gas specific activity to be 
consistent with the values assumed in the radiological consequence 
analyses. The changes will also limit the potential RCS iodine 
specific activity excursion to the value currently associated with 
full power operation, which is more restrictive on plant operation 
than the existing allowable RCS iodine specific activity at lower 
power levels.
    Therefore 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: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Branch Chief: David Terao.

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

    Date of amendment request: January 31, 2007.
    Description of amendment request: The proposed change revises the 
Technical Specification (TS) surveillance requirements (SR) for 
addressing a missed surveillance, and is consistent with the Nuclear 
Regulatory Commission (NRC) approved Revision 6 of Technical 
Specification Task Force (TSTF) Standard Technical Specifications (STS) 
Change Traveler TSTF-358, ``Missed Surveillance Requirements.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed change to incorporate the requirements of improved 
STS SR 3.0.1 and SR 3.0.3 into corresponding Surry TS SR 4.0.1 and 
SR 4.0.3, respectively, does not affect the design or operation of 
the plant. The proposed change involves revising the existing Surry 
custom TS to be consistent with NUREG-1431, Revision 3, to 
facilitate the incorporation of TSTF-358 into the TS. The proposed 
change involves no technical changes to the existing TS as it merely 
clarifies how SRs are met. As such, these changes are administrative 
in nature and do not affect initiators of analyzed events or assumed 
mitigation of accident or transient events. Therefore, the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change to incorporate the requirements of improved 
STS SR 3.0.1 and SR 3.0.3 into corresponding Surry TS SR 4.0.1 and 
SR 4.0.3, respectively, does not involve a physical alteration to 
the plant (no new or different type of equipment will be installed) 
or changes in methods governing normal plant operation. The proposed 
change revises the existing Surry TS to be consistent with NUREG-
1431, Revision 3, to clarify how SRs are met and facilitates the 
incorporation of TSTF-358 for addressing missed surveillances. As 
such, the proposed change will not impose any new or different 
requirements or eliminate any existing requirements. 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 to incorporate the requirements of improved 
STS SR 3.0.1 and SR 3.0.3 into corresponding Surry TS SR 4.0.1 and 
SR 4.0.3, respectively, does not affect plant operation or safety 
analysis assumptions in any way. The change provides additional 
clarification on how a surveillance is met and facilitates the 
incorporation of TSTF-358 for addressing missed surveillances. The 
change is administrative in nature and does not affect the operation 
of safety-related systems, structures, or components. 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: Lillian M. Cuoco, Esq., Senior Counsel, 
Dominion Resources Services, Inc., Millstone Power Station, Building 
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
    NRC Branch Chief: Evangelos C. Marinos.

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

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

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

    Date of amendment request: January 19, 2007.
    Brief description of amendment request: The proposed amendment 
would modify Technical Specification

[[Page 8807]]

(TS) 5.5.9 to add steam generator (SG) alternate repair criteria and TS 
5.6.8 to add additional SG reporting requirements.
    Date of publication of individual notice in Federal Register: 
January 30, 2007 (72 FR 4300).
    Expiration date of individual notice: April 2, 2007.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) The 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to <a href="/cdn-cgi/l/email-protection#d3a3b7a193bda1b0fdb4bca5"><span class="__cf_email__" data-cfemail="03736771436d71602d646c75">[email protected]</span></a>.

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

    Date of application for amendment: February 7, 2006, as 
supplemented by letters dated August 14 and November 16, 2006.
    Brief description of amendment: The amendment revised the Millstone 
Power Station, Unit No. 2 Technical Specifications to permit an 
increase in the allowed outage time from 72 hours to 7 days for the 
inoperability of the steam supply to the turbine-driven auxiliary 
feedwater (AFW) pump or the inoperability of the turbine-driven AFW 
pump under certain operating mode restrictions.
    Date of issuance: January 31, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 297.
    Facility Operating License No. DPR-65: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 11, 2006 (71 FR 
18372). The supplements dated August 14, and 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.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 31, 2007.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 17, 2006.
    Brief description of amendment: This amendment changed the method 
for calculating fuel pool decay heat load from the original licensing 
basis methodology of ORIGEN to ORIGEN-ARP.
    Date of issuance: February 8, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 200.
    Facility Operating License No. NPF-21: The amendment revised the 
Final Safety Analysis Report.
    Date of initial notice in Federal Register: May 23, 2006 (71 FR 
29674). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 8, 2007.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 18, 2006.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) Surveillance Requirement (SR) 3.6.1.1.2 by changing 
the test frequency of the drywell-to-suppression chamber bypass leakage 
test from 24 months to 120 months. The amendment also added new TS SRs 
3.6.1.1.3 and 3.6.1.1.4, to test the suppression chamber-to-drywell 
vacuum breakers on a 24-month frequency.
    Date of issuance: February 9, 2007.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 201.
    Facility Operating License No. NPF-21: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: May 23, 2006 (71 FR 
29674). 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.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: June 14, 2006, as supplemented by letter 
dated November 7, 2006.
    Brief description of amendment: The amendment approved the removal 
of Surveillance Requirement 4.8.1.1.2.f from the Waterford Steam 
Electric Station, Unit 3, Technical Specifications. Entergy Operations, 
Inc. has committed to relocate this surveillance requirement, which is 
associated with vendor recommended inspections of the emergency diesel 
generators, to the Technical Requirements Manual.
    Date of issuance: February 6, 2007.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 211.
    Facility Operating License No. NPF-38: The amendment revised the 
Operating License and the Technical Specifications.
    Date of initial notice in Federal Register: August 15, 2006 (71 FR

[[Page 8808]]

46931). The November 7, 2006, supplemental letter provided additional 
clarifying information, did not expand the scope of the application as 
originally noticed, and did not change the NRC staff's original 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 6, 2007.
    No significant hazards consideration comments received: No.

FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of application for amendment: March 1, 2006, as supplemented 
by letter dated August 17, 2006.
    Brief description of amendment: The amendment modifies Special 
Operations Limiting Condition for Operation (LCO) 3.10.1, ``System 
Leakage and Hydrostatic Testing Operation,'' to allow more efficient 
testing during a refueling outage. Specifically, the LCO 3.10.1 
allowance for operation with the average reactor coolant temperature 
greater than 212 [deg]F (while considering operational conditions to be 
in Mode 4), is extended to include operations where temperature exceeds 
212 [deg]F: (1) As a consequence of maintaining adequate reactor 
pressure for a system leakage or hydrostatic test; or (2) as a 
consequence of maintaining adequate reactor pressure for control rod 
scram time testing initiated in conjunction with a system leakage or 
hydrostatic test. This change is based on the NRC-approved Technical 
Specification Task Force (TSTF) standard TS change TSTF-484, Revision 
0.
    Date of issuance: February 5, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 264
    Facility Operating License No. DPR-49: The amendment revises the 
TSs.
    Date of initial notice in Federal Register: (71 FR 70560) December 
5, 2006. The 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 on December 5, 2006 (71 FR 70560).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 5, 2007.
    No significant hazards consideration comments received: No.

GPU Nuclear, Inc., Docket No. 50-320, Three Mile Island Nuclear 
Station, Unit 2, Dauphin County, Pennsylvania

    Date of amendment request: October 10, 2006.
    Brief description of amendment: The amendment revises Technical 
Specification Surveillance Requirement 4.1.1.3 to extend the 
containment airlock surveillance frequency from once per year to once 
every five years.
    Date of issuance: February 7, 2007.
    Effective date: February 7, 2007.
    Amendment No.: 61.
    Possession Only License No. DPR-73: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 5, 2006 (71 FR 
70560). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation Report, dated February 7, 2007.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 14, 2006, as 
supplemented on December 28, 2006.
    Brief description of amendment: The amendment revised Table 
3.3.5.1-1, ``Emergency Core Cooling System Instrumentation,'' of the 
MNGP Technical Specifications, to permit a one-time extension of the 
quarterly surveillance interval (i.e., from 92 days to 140 days), for 
three low pressure coolant injection loop select logic functions.
    Date of issuance: January 18, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No: 149.
    Renewed Facility Operating License No. DPR-22: Amendment revised 
the Renewed Facility Operating License and Technical Specifications.
    The supplemental letter contained clarifying information and did 
not change the initial no significant hazards consideration 
determination, and did not expand the scope of the original Federal 
Register notice.
    Date of initial notice in Federal Register: December 19, 2006 (71 
FR 75995). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 18, 2007.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-259 Browns Ferry Nuclear 
Plant, Unit 1, Limestone County, Alabama

    Date of application for amendment: November 9, 2006 (TS-458).
    Brief description of amendment: The amendment deleted the Technical 
Specification (TS) Surveillance Requirement to verify the position of a 
low pressure coolant injection crosstie valve.
    Date of issuance: February 6, 2007.
    Effective date: Effective as of the date of issuance, to be 
implemented within 30 days.
    Amendment No.: 268.
    Renewed Facility Operating License No. DPR-33: Amendment revised 
the TSs.
    Date of initial notice in Federal Register: November 20, 2006 (71 
FR 671600). 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.

    Dated at Rockville, Maryland, this 15th day of February 2007.

    For the Nuclear Regulatory Commission.
John W. Lubinski,
Acting Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
 [FR Doc. E7-3199 Filed 2-26-07; 8:45 am]
BILLING CODE 7590-01-P


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