[Federal Register Volume 73, Number 58 (Tuesday, March 25, 2008)]
[Notices]
[Pages 15780-15796]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E8-5734]



[[Page 15780]]

-----------------------------------------------------------------------

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 28, 2008 to March 12, 2008. The 
last biweekly notice was published on March 11, 2008 (73 FR 13021).

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.
    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, 
person(s) 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 
via electronic submission through the NRC E-Filing system 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 person(s) 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

[[Page 15781]]

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 hearing or a petition for leave to intervene must be 
filed in accordance with the NRC E-Filing rule, which the NRC 
promulgated in August 28, 2007 (72 FR 49139). The E-Filing process 
requires participants to submit and serve documents over the internet 
or in some cases to mail copies on electronic storage media. 
Participants may not submit paper copies of their filings unless they 
seek a waiver in accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
five (5) days prior to the filing deadline, the petitioner/requestor 
must contact the Office of the Secretary by e-mail at: 
[email protected], or by calling (301) 415-1677, to request (1) a 
digital ID certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and/or (2) 
creation of an electronic docket for the proceeding (even in instances 
in which the petitioner/requestor (or its counsel or representative) 
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms ViewerTM 
to access the Electronic Information Exchange (EIE), a component of the 
E-Filing system. The Workplace Forms ViewerTM is free and is 
available at: http://www.nrc.gov/site-help/e-submittals/install-viewer.html. Information about applying for a digital ID certificate is 
available on NRC's public Web site at :http://www.nrc.gov/site-help/e-submittals/apply-certificates.html.
    Once a petitioner/requestor has obtained a digital ID certificate, 
had a docket created, and downloaded the EIE viewer, it can then submit 
a request for hearing or petition for leave to intervene. Submissions 
should be in Portable Document Format (PDF) in accordance with NRC 
guidance available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the 
time the filer submits its documents through EIE. To be timely, an 
electronic filing must be submitted to the EIE system no later than 
11:59 p.m. Eastern Time on the due date. Upon receipt of a 
transmission, the E-Filing system time-stamps the document and sends 
the submitter an e-mail notice confirming receipt of the document. The 
EIE system also distributes an e-mail notice that provides access to 
the document to the NRC Office of the General Counsel and any others 
who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically may seek assistance through the 
``Contact Us'' link located on the NRC Web site at: http://www.nrc.gov/site-help/e-submittals.html or by calling the NRC technical help line, 
which is available between 8:30 a.m. and 4:15 p.m., Eastern Time, 
Monday through Friday. The help line number is (800) 397-4209 or 
locally, (301) 415-4737.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file a motion, in accordance 
with 10 CFR 2.302(g), with their initial paper filing requesting 
authorization to continue to submit documents in paper format. Such 
filings must be submitted 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; or (2) courier, express mail, or expedited 
delivery service to the Office of the Secretary, Sixteenth Floor, One 
White Flint North, 11555 Rockville, Pike, Rockville, Maryland, 20852, 
Attention: Rulemaking and Adjudications Staff. Participants filing a 
document in this manner are responsible for serving the document on all 
other participants. Filing is considered complete by first-class mail 
as of the time of deposit in the mail, or by courier, express mail, or 
expedited delivery service upon depositing the document with the 
provider of the service.
    Non-timely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission, the presiding 
officer, or the Atomic Safety and Licensing Board that the petition 
and/or request should be granted and/or the contentions should be 
admitted, based on a balancing of the factors specified in 10 CFR 
2.309(c)(1)(i)-(viii). To be timely, filings must be submitted no later 
than 11:59 p.m. Eastern Time on the due date.
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at: 
http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant 
to an order of the Commission, an Atomic Safety and Licensing Board, or 
a Presiding Officer. Participants are requested not to include personal 
privacy information, such as social security numbers, home addresses, 
or home phone numbers in their filings. With respect to copyrighted 
works, except for limited excerpts that serve the purpose of the 
adjudicatory filings and would constitute a Fair Use application, 
participants are requested not to include copyrighted materials in 
their submission.
    For further details with respect to this amendment action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to: [email protected].

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

    Date of amendment request: December 11, 2007.
    Description of amendment request: The amendments would revise the

[[Page 15782]]

Technical Specifications (TSs) permitting relaxation of the allowed 
bypass test times and completion times for various systems in 
accordance with Technical Specification Task Force Traveler (TSTF) 418, 
Revision 2, ``RPS and ESFAS Test Times and Completion Times (WCAP-
14333).
    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:

First Standard

    Does operation of the facility in accordance with the proposed 
amendment involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to the Completion Times, bypass test time, 
and Surveillance Frequencies reduces the potential for inadvertent 
reactor trips and spurious actuations, and therefore do not increase 
the probability of any accident previously evaluated. The proposed 
changes to the Completion Times and bypass test time do not change 
the response of the plant to any accidents and have an insignificant 
impact on the reliability of the reactor trip system and engineered 
safety feature actuation system (RTS and ESFAS) signals. The RTS and 
ESFAS will remain highly reliable and the proposed changes will not 
result in a significant increase in the risk of plant operation. 
This is demonstrated by showing that the impact on plant safety as 
measured by core damage frequency (CDF) is less than 1.0E-06 per 
year and the impact on large early release frequency (LERF) is less 
than 1.0E-07 per year. In addition, for the Completion Time change, 
the incremental conditional core damage probabilities (ICCDP) and 
incremental conditional large early release probabilities (ICLERP) 
are less than 5.0E-07 and 5.0E-08, respectively. These changes meet 
the acceptance criteria in Regulatory Guides 1.174 and 1.177. 
Therefore, since the RTS and ESFAS will continue to perform their 
functions with high reliability as originally assumed, and the 
increase in risk as measured by CDF, LERF, ICCDP, and ICLERP is 
within the acceptance criteria of existing regulatory guidance, 
there will not be a significant increase in the consequences of any 
accidents.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, or 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not alter or 
prevent the ability of structures, systems, and components (SSCs) 
from performing their intended function to mitigate the consequences 
of an initiating event within the assumed acceptance limits. The 
proposed changes do not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of an accident previously evaluated. 
Further, the proposed changes do not increase the types or amounts 
of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures. The proposed changes are consistent with safety 
analysis assumptions and resultant consequences.
    The determination on risk impacts that the results of the 
proposed changes are acceptable was established in the NRC Safety 
Evaluations prepared for WCAP-14333-P-A (issued by letter dated July 
15, 1998) and for WCAP-15376-P-A (issued by letter dated December 
20, 2002). Implementation of the proposed changes will result in an 
insignificant risk impact. Applicability of these conclusions has 
been verified through plant-specific reviews and implementation of 
the generic analysis results in accordance with the respective NRC 
Safety Evaluation conditions.
    The proposed changes based on TSTF-246 do not involve any 
physical alteration of plant SSCs. The remaining intermediate range 
and power range nuclear instruments remain operable and have 
required actions that ensure compliance with applicable safety 
analyses.
    Therefore, it is concluded that this change does not increase 
the probability of occurrence of a malfunction of equipment 
important to safety.

Second Standard

    Does operation of the facility in accordance with the proposed 
amendment create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not result in a change in the manner in 
which the RTS or ESFAS provide plant protection. The RTS and ESFAS 
will continue to have the same setpoints after the proposed changes 
are implemented. There are no design changes associated with the 
license amendment. The changes to Completion Times, bypass test 
times, and Surveillance Frequencies do not change any existing 
accident scenarios, nor create any new or different accident 
scenarios. 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 alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumptions and current plant operating practice.
    The proposed changes do not introduce new failure mechanisms for 
systems, structures, or components not already considered in the 
UFSAR. Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated is not created 
because no new failure mechanisms or initiating events have been 
introduced.

Third Standard

    Does operation of the facility in accordance with the proposed 
amendment involve a significant reduction in the margin of safety?
    Response: No.
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not impacted by these changes.
    Redundant RTS and ESFAS trains are maintained, and diversity 
with regard to the signals that provide reactor trip and ESFAS is 
also maintained. Signals credited as primary or secondary and 
operator actions credited in the accident analyses will remain the 
same. The proposed changes will not result in plant operation in a 
configuration outside design basis. The calculated impact on risk is 
insignificant and meets the acceptance criteria contained in 
Regulatory Guides 1.174 and 1.177. Although there was no attempt to 
quantify any positive human factors benefit due to increased 
Completion Times and bypass test time, it is expected that there 
would be a net benefit due to a reduced potential for spurious 
reactor trips and actuations associated with testing.
    Implementation of the proposed changes is expected to result in 
an overall improvement in safety, as follows:
    a. Reduced testing will result in fewer inadvertent reactor 
trips, less frequent actuation of ESFAS components, less frequent 
distraction of operations personnel without significantly affecting 
RTS and ESFAS reliability.
    b. Improvements in the effectiveness of the operating staff in 
monitoring and controlling plant operation will be realized. This is 
due to less frequent distraction of the operators and shift 
supervisor to attend to instrumentation Required Actions with short 
Completion Times.
    c. Longer repair times associated with increased Completion 
Times will lead to higher quality repairs and improved reliability.
    d. The Completion Time extensions for the reactor trip breakers 
will provide the utilities additional time to complete test and 
maintenance activities while at power, potentially reducing the 
number of forced outages related to compliance with reactor trip 
breaker Completion Times, and provide consistency with the 
Completion Times for the logic trains.
    Therefore, it is concluded that this change does not involve a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: 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: Melanie Wong.

[[Page 15783]]

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: December 11, 2007.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications permitting relaxation of the 
allowed bypass test times and completion times for various systems in 
accordance with Technical Specification Task Force Traveler (TSTF) 418, 
Revision 2, ``RPS and ESFAS Test Times and Completion Times (WCAP-
14333).
    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:

First Standard

    Does operation of the facility in accordance with the proposed 
amendment involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to the Completion Times, bypass test time, 
and Surveillance Frequencies reduces the potential for inadvertent 
reactor trips and spurious actuations, and therefore do not increase 
the probability of any accident previously evaluated. The proposed 
changes to the Completion Times and bypass test time do not change 
the response of the plant to any accidents and have an insignificant 
impact on the reliability of the reactor trip system and engineered 
safety feature actuation system (RTS and ESFAS) signals. The RTS and 
ESFAS will remain highly reliable and the proposed changes will not 
result in a significant increase in the risk of plant operation. 
This is demonstrated by showing that the impact on plant safety as 
measured by core damage frequency (CDF) is less than 1.0E-06 per 
year and the impact on large early release frequency (LERF) is less 
than 1.0E-07 per year. In addition, for the Completion Time change, 
the incremental conditional core damage probabilities (ICCDP) and 
incremental conditional large early release probabilities (ICLERP) 
are less than 5.0E-07 and 5.0E-08, respectively. These changes meet 
the acceptance criteria in Regulatory Guides 1.174 and 1.177. 
Therefore, since the RTS and ESFAS will continue to perform their 
functions with high reliability as originally assumed, and the 
increase in risk as measured by CDF, LERF, ICCDP, and ICLERP is 
within the acceptance criteria of existing regulatory guidance, 
there will not be a significant increase in the consequences of any 
accidents.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, or 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not alter or 
prevent the ability of structures, systems, and components (SSCs) 
from performing their intended function to mitigate the consequences 
of an initiating event within the assumed acceptance limits. The 
proposed changes do not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of an accident previously evaluated. 
Further, the proposed changes do not increase the types or amounts 
of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures. The proposed changes are consistent with safety 
analysis assumptions and resultant consequences.
    The determination that the results of the proposed changes are 
acceptable was established in the NRC Safety Evaluations prepared 
for WCAP-14333-P-A (issued by letter dated July 15, 1998) and for 
WCAP-15376-P-A (issued by letter dated December 20, 2002). 
Implementation of the proposed changes will result in an 
insignificant risk impact. Applicability of these conclusions has 
been verified through plant-specific reviews and implementation of 
the generic analysis results in accordance with the respective NRC 
Safety Evaluation conditions.
    The proposed changes based on TSTF-246 do not involve any 
physical alteration of plant systems, structures, or components. The 
remaining intermediate range and power range nuclear instruments 
remain operable and have required actions that ensure compliance 
with applicable safety analyses.
    Therefore, it is concluded that this change does not increase 
the probability of occurrence of a malfunction of equipment 
important to safety.

Second Standard

    Does operation of the facility in accordance with the proposed 
amendment create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not result in a change in the manner in 
which the RTS or ESFAS provide plant protection. The RTS and ESFAS 
will continue to have the same setpoints after the proposed changes 
are implemented. There are no design changes associated with the 
license amendment. The changes to Completion Times, bypass test 
times, and Surveillance Frequencies do not change any existing 
accident scenarios, nor create any new or different accident 
scenarios. 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 alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumptions and current plant operating practice.
    The proposed changes do not introduce new failure mechanisms for 
systems, structures, or components not already considered in the 
UFSAR. Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated is not created 
because no new failure mechanisms or initiating events have been 
introduced.

Third Standard

    Does operation of the facility in accordance with the proposed 
amendment involve a significant reduction in the margin of safety?
    Response: No.
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not impacted by these changes.
    Redundant RTS and ESFAS trains are maintained, and diversity 
with regard to the signals that provide reactor trip and ESFAS is 
also maintained. Signals credited as primary or secondary and 
operator actions credited in the accident analyses will remain the 
same. The proposed changes will not result in plant operation in a 
configuration outside design basis. The calculated impact on risk is 
insignificant and meets the acceptance criteria contained in 
Regulatory Guides 1.174 and 1.177. Although there was no attempt to 
quantify any positive human factors benefit due to increased 
Completion Times and bypass test time, it is expected that there 
would be a net benefit due to a reduced potential for spurious 
reactor trips and actuations associated with testing.
    Implementation of the proposed changes is expected to result in 
an overall improvement in safety, as follows:
    e. Reduced testing will result in fewer inadvertent reactor 
trips, less frequent actuation of ESFAS components, less frequent 
distraction of operations personnel without significantly affecting 
RTS and ESFAS reliability.
    f. Improvements in the effectiveness of the operating staff in 
monitoring and controlling plant operation will be realized. This is 
due to less frequent distraction of the operators and shift 
supervisor to attend to instrumentation Required Actions with short 
Completion Times.
    g. Longer repair times associated with increased Completion 
Times will lead to higher quality repairs and improved reliability.
    h. The Completion Time extensions for the reactor trip breakers 
will provide the utilities additional time to complete test and 
maintenance activities while at power, potentially reducing the 
number of forced outages related to compliance with reactor trip 
breaker Completion Times, and provide consistency with the 
Completion Times for the logic trains.
    Therefore, it is concluded that this change does not involve a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration. 
Attorney for licensee: Ms. Lisa F.

[[Page 15784]]

Vaughn, Associate General Counsel and Managing Attorney, Duke Energy 
Carolinas, LLC, 526 South Church Street, EC07H, Charlotte, NC 28202. 
NRC Branch Chief: Melanie Wong.

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: January 22, 2008.
    Description of amendment request: The proposed amendments would 
modify Technical Specification (TS) requirements related to control 
room envelope habitability in accordance with Technical Specification 
Task Force (TSTF)-448, Revision 3, ``Control Room Habitability.'' For 
McGuire Nuclear Station, Units 1 and 2, this TSTF revises TS 3.7.9, 
Control Room Area Ventilation System (CRAVS), and adds a new 
administrative controls program, TS 5.5.16, Control Room Envelope 
Habitability Program.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on October 17, 2006 (71 FR 61075) on possible license 
amendments adopting TSTF-448 using the NRC's consolidated line item 
improvement process (CLIIP) for amending the licensee's TSs, which 
included a model safety evaluation (SE) and model no significant 
hazards consideration (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 January 17, 
2007 (72 FR 2022), which included the resolution of public comments on 
the model SE. The licensee has affirmed the applicability of the 
following NSHC determination in its application.
    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

    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change revises the TS for the control room 
envelope (CRE) emergency ventilation system, which is a mitigation 
system designed to minimize unfiltered air leakage into the CRE and 
to filter the CRE atmosphere to protect the CRE occupants in the 
event of accidents previously analyzed. An important part of the CRE 
emergency ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated. Therefore, the probability of any accident 
previously evaluated is not increased. Performing tests to verify 
the operability of the CRE boundary and implementing a program to 
assess and maintain CRE habitability ensure that the CRE emergency 
ventilation system is capable of adequately mitigating radiological 
consequences to CRE occupants during accident conditions, and that 
the CRE emergency ventilation system will perform as assumed in the 
consequence analyses of design basis accidents. Thus, the 
consequences of any accident previously evaluated are not increased. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

Criterion 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 impact the accident analysis. The 
proposed change does not alter the required mitigation capability of 
the CRE emergency ventilation system, or its functioning during 
accident conditions as assumed in the licensing basis analyses of 
design basis accident radiological consequences to CRE occupants. No 
new or different accidents result from performing the new 
surveillance or following the new program. The proposed change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a significant 
change in the methods governing normal plant operation. The proposed 
change does not alter any safety analysis assumptions and is 
consistent with current plant operating practice. Therefore, this 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.

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

    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff 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: Melanie C. Wong.

Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point 
Nuclear Generating Unit No. 2 (IP2), Westchester County, New York

    Date of amendment request: December 13, 2007.
    Description of amendment request: The proposed amendment would add 
some Emergency Core Cooling System (ECCS) valves and remove other ECCS 
valves from Surveillance Requirement (SR) 3.5.2.1. The purpose of the 
SR is to verify that ECCS valves whose single failure could cause loss 
of the ECCS function are in the required position with power removed so 
that the single failure could not occur. The valves being added are 
currently controlled administratively. The valves being removed have 
been evaluated to demonstrate that a single failure would not cause 
loss of the ECCS function.
    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 adds three ECCS valves and removes four ECCS 
valves from IP2 SR 3.5.2.1. The purpose of the surveillance is to 
assure that the valves are in their required position with power 
removed so that misalignment or single failure cannot prevent 
completion of the ECCS function. The performance of the SR does not 
involve any actions related to the initiation of an accident and 
therefore the proposed changes cannot increase the probability of an 
accident. Misalignment or single failure of one of the three valves 
being added to TS could cause a loss of the ECCS function so the 
change will not increase the consequences of an accident but rather 
provide assurance that no such increase can occur. Removal of the 
four valves has been evaluated and the evaluation demonstrates that 
the misalignment or single failure of one of the valves will not 
affect the ECCS function and therefore will not increase the 
consequences of an accident. 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?

[[Page 15785]]

    Response: No.
    The proposed change adds three ECCS valves and removes four ECCS 
valves from IP2 SR 3.5.2.1. The purpose of the surveillance is to 
assure that the valves are in their required position with power 
removed so that misalignment or single failure cannot prevent 
completion of the ECCS function. The removal of valves from the 
surveillance allows power to be maintained to the valves during 
normal operation but does not otherwise affect the function of the 
valves or the design and operation of plant systems. The addition of 
power does mean that the valves could fail open but this does not 
create the possibility of a new or different type of accident since 
such a failure mode is currently evaluated. The performance of the 
SR for added valves does not affect the function of the valves or 
the manner in which the valves or their systems are operated or any 
procedures used for valve or system operation. The change assures 
that the valves will be in their correct position and does not 
introduce any new failure modes or the possibility of a different 
accident. 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 
the margin of safety?
    Response: No.
    The proposed change adds three ECCS valves and removes four ECCS 
valves from IP2 SR 3.5.2.1. The purpose of the surveillance is to 
assure that the valves are in their required position with power 
removed so that misalignment or single failure cannot prevent 
completion of the ECCS function. The addition of the three valves to 
the TS provides additional assurance that operation will be with 
power removed and the valves in the correct position. This increases 
safety margin. Removal of valves from the surveillance is based on 
analysis of the effects of misalignment or single failure on the 
ECCS function. Analysis demonstrates that the misalignment or single 
failure would not adversely affect the ECCS function and therefore 
there is no significant reduction in the margin of safety. The 
margin of safety remains adequate to assure the ECCS function is 
performed.

    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: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Mark G. Kowal.

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 amendment request: December 18, 2007.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) requirements related to control 
room envelope habitability by adding a Control Room Envelope 
Habitability Program and then referencing this program in place of 
existing surveillances. It also standardizes terminology and modifies 
other TS related to the control room envelope.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-448, Revision 3. The NRC staff 
issued a notice of opportunity for comment in the Federal Register on 
October 17, 2006 (71 FR 61075), on possible amendments concerning TSTF-
448, including a model safety evaluation and model no significant 
hazards (NSHC) determination, using the consolidated line item 
improvement process (CLIIP). The NRC staff subsequently issued a notice 
of availability of the models for referencing in license amendment 
applications in the Federal Register on January 17, 2007 (72 FR 2022). 
The licensee affirmed the applicability of the following NSHC 
determination in its application dated December 18, 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

    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits.
    The proposed change revises the TS for the CRE emergency 
ventilation system, which is a mitigation system designed to 
minimize unfiltered air leakage into the CRE and to filter the CRE 
atmosphere to protect the CRE occupants in the event of accidents 
previously analyzed. An important part of the CRE emergency 
ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated. Therefore, the probability of any accident 
previously evaluated is not increased.
    Performing tests to verify the operability of the CRE boundary 
and implementing a program to assess and maintain CRE habitability 
ensure that the CRE emergency ventilation system is capable of 
adequately mitigating radiological consequences to CRE occupants 
during accident conditions, and that the CRE emergency ventilation 
system will perform as assumed in the consequence analyses of design 
basis accidents. Thus, the consequences of any accident previously 
evaluated are not increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

Criterion 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 impact the accident analysis. The 
proposed change does not alter the required mitigation capability of 
the CRE emergency ventilation system, or its functioning during 
accident conditions as assumed in the licensing basis analyses of 
design basis accident radiological consequences to CRE occupants. No 
new or different accidents result from performing the new 
surveillance or following the new program. The proposed change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a significant 
change in the methods governing normal plant operation. The proposed 
change does not alter any safety analysis assumptions and is 
consistent with current plant operating practice.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.

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

    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed this 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: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White 
Plains, NY 10601.

[[Page 15786]]

    NRC Branch Chief: Mark G. Kowal.

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 amendment request: December 20, 2007.
    Description of amendment request: The proposed amendment would 
modify Technical Specifications (TS), to replace the current limits on 
primary coolant gross specific activity with limits on primary coolant 
noble gas activity. The noble gas activity would be based on DOSE 
EQUIVALENT XE-133 and would take into account only the noble gas 
activity in the primary coolant.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-490. The NRC staff issued a 
notice of opportunity for comment in the Federal Register on November 
20, 2006 (71 FR 67170), on possible amendments concerning TSTF-490, 
including a model safety evaluation and model no significant hazards 
(NSHC) determination, using the consolidated line item improvement 
process (CLIIP). The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on March 15, 2007 (72 FR 12217). 
The licensee affirmed the applicability of the following NSHC 
determination in its application dated December 20, 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

    Reactor coolant specific activity is not an initiator for any 
accident previously evaluated. The Completion Time when primary 
coolant gross activity is not within limit is not an initiator for 
any accident previously evaluated. The current variable limit on 
primary coolant iodine concentration is not an initiator to any 
accident previously evaluated. As a result, the proposed change does 
not significantly increase the probability of an accident. The 
proposed change will limit primary coolant noble gases to 
concentrations consistent with the accident analyses. The proposed 
change to the Completion Time has no impact on the consequences of 
any design basis accident since the consequences of an accident 
during the extended Completion Time are the same as the consequences 
of an accident during the Completion Time. As a result, the 
consequences of any accident previously evaluated are not 
significantly increased.

Criterion 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 in specific activity limits does not alter 
any physical part of the plant nor does it affect any plant 
operating parameter. The change does not create the potential for a 
new or different kind of accident from any previously evaluated.

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

    The proposed change revises the limits on noble gas 
radioactivity in the primary coolant. The proposed change is 
consistent with the assumptions in the safety analyses and will 
ensure the monitored values protect the initial assumptions in the 
safety analyses.

    The NRC staff has reviewed this 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: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Mark G. Kowal.

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

    Date of amendment request: January 31, 2008.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to change the description of 
fuel assemblies specified in TS 4.2.1, and add the Framatome Advanced 
Nuclear Power, Inc. (ANP) report, BAW-10240(P)-A, ``Incorporation of M5 
Properties in Framatome ANP Approved Methods,'' to the analytical 
methods referenced in TS 5.6.5.b to permit the use of M5 alloy for fuel 
rod cladding and fuel assembly structural components in future 
operating cycles.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed license amendment adds a Nuclear Regulatory 
Commission approved analytical method, BAW-10240(P)-A, 
``Incorporation of M5 Properties in Framatome ANP Approved 
Methods,'' used to determine the core operating limits, to Technical 
Specification (TS) 5.6.5.b and changes the description of fuel 
assemblies specified in TS 4.2.1 to allow use of the M5 alloy. The 
proposed amendment does not affect the acceptance criteria for any 
Final Safety Analysis Report (FSAR) safety analysis analyzed 
accidents and anticipated operational occurrences. As such, the 
proposed amendment does not increase the probability or consequences 
of an accident. The proposed amendment does not involve operation of 
the required structures, systems or components (SSCs) in a manner or 
configuration different from those previously recognized or 
evaluated.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    Use of M5 clad fuel will not result in changes in the operation 
or configuration of the facility. Topical report BAW-10240(P)-A 
describes, by reference, that the material properties of the M5 
alloy are similar or better than those of zircaloy-4. Therefore, M5 
fuel rod cladding and fuel assembly structural components will 
perform similarly to those fabricated from zircaloy-4, thus 
precluding the possibility of the fuel becoming an accident 
initiator and causing a new or different type of accident.
    Since the material properties of M5 alloy are similar or better 
than those of zircaloy-4, there will be no significant changes in 
the types of any effluents that may be released off-site. There will 
not be a significant increase in occupational or public radiation 
exposure.
    The proposed amendment does not involve operation of any 
required SSCs in a manner or configuration different from those 
previously recognized or evaluated. No new failure mechanisms will 
be introduced by the changes being requested.
    Therefore, 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?
    Response: No.
    The proposed change will not involve a significant reduction in 
the margin of safety because it has been demonstrated that the 
material properties of the M5 alloy are not significantly different 
from those of zircaloy-4. M5 alloy is expected to perform similarly 
or better than zircaloy-4 for all normal operating and accident 
scenarios, including both loss-of-coolant accident (LOCA) and non-
LOCA scenarios. The proposed changes do not affect the acceptance 
criteria for any FSAR safety analysis analyzed accidents or 
anticipated operational occurrences. All required safety limits 
would continue to be analyzed using methodologies approved by the 
Nuclear Regulatory Commission.
    Therefore, the proposed amendment would not involve a 
significant reduction in a margin of safety.


[[Page 15787]]


    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: Mr. William Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White 
Plains, NY 10601.
    NRC Acting Branch Chief: Patrick D. Milano.

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

    Date of amendment request: February 1, 2008.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 5.5.16.a, ``Containment Leakage 
Rate Testing Program,'' to add an exception to Regulatory Guide 1.163 
to allow the use of Standard ANSI/ANS 56.8-2002, and to revise TS 
5.5.16.b to specify both a lower peak calculated containment internal 
pressure following a large-break loss-of-coolant accident (LOCA) and 
containment design pressure.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. [Do] the proposed change[s] involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to TS 5.5.16.a adds an exception to 
Regulatory Guide 1.163 to specify use of Standard ANSI/ANS-56.8-
2002, rather than ANSI/ANS-56.8-1994.
    The proposed change to TS 5.5.16.b specifies both the peak 
calculated containment internal pressure with margin following a 
large-break LOCA and the containment design pressure.
    These changes only affect the applicable version of the standard 
(2002 in place of 1994) and the test pressures for containment leak-
rate tests, and do not involve the modification of any plant 
equipment or have any effect on plant operation. The changes are 
made based on the safety analysis and containment design, and do not 
have any adverse effect on accidents previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. [Do] the proposed change[s] create the possibility of a new 
or different accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not involve a physical alteration to the 
plant or a change in the methods governing normal plant operation. 
The changes are made based on the safety analysis and containment 
design, and do not affect any previously evaluated accidents.
    Therefore, the proposed change[s] [do] not create the 
possibility of a new or different accident from any accident 
previously evaluated.
    3. [Do] the proposed change[s] involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not affected by these changes, and the changes will not result 
in plant operation in a configuration outside the design basis.
    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: Jennifer Post, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Branch Chief: Thomas G. Hiltz.

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

    Date of amendment request: February 29, 2008.
    Description of amendment request: The proposed amendments would 
modify Technical Specification (TS) requirements related to control 
room envelope (CRE) habitability in accordance with the Nuclear 
Regulatory Commission (NRC)-approved Revision 3 of Technical 
Specification Task Force (TSTF) Standard Technical Specifications (STS) 
Change Traveler TSTF-448, ``Control Room Habitability.''
    The NRC staff published a notice of opportunity for comment in the 
Federal Register on October 17, 2006 (71 FR 61075), on possible license 
amendments adopting TSTF-448 using the NRC's consolidated line-item 
improvement process (CLIIP) for amending licensees' TSs, which included 
a model safety evaluation (SE) and model no significant hazards 
consideration (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 January 17, 2007 (72 
FR 2022), which included the resolution of public comments on the model 
SE and model NSHC determination. The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated February 29, 2008.
    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

    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change revises the TS for the CRE emergency 
ventilation system, which is a mitigation system designed to 
minimize unfiltered air leakage into the CRE and to filter the CRE 
atmosphere to protect the CRE occupants in the event of accidents 
previously analyzed. An important part of the CRE emergency 
ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated. Therefore, the probability of any accident 
previously evaluated is not increased. Performing tests to verify 
the operability of the CRE boundary and implementing a program to 
assess and maintain CRE habitability ensure that the CRE emergency 
ventilation system is capable of adequately mitigating radiological 
consequences to CRE occupants during accident conditions, and that 
the CRE emergency ventilation system will perform as assumed in the 
consequence analyses of design basis accidents. Thus, the 
consequences of any accident previously evaluated are not increased. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

Criterion 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 impact the accident analysis. The 
proposed change does not alter the required mitigation capability of 
the CRE emergency ventilation system, or its functioning during 
accident conditions as assumed in the licensing basis analyses of 
design basis accident radiological consequences to CRE occupants. No 
new or different accidents result from performing the new 
surveillance or following the new program. The proposed change does 
not

[[Page 15788]]

involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a significant 
change in the methods governing normal plant operation. The proposed 
change does not alter any safety analysis assumptions and is 
consistent with current plant operating practice. Therefore, this 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.

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

    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation as determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition. Therefore, the proposed change does not involve 
a significant reduction in a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Branch Chief: Melanie C. Wong.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: January 28, 2008.
    Description of amendment request: The amendments would revise the 
Technical Specifications (TS) to establish an Action in TS 3.3.1, 
``Reactor Trip Instrumentation,'' for two inoperable channels of 
extended range neutron flux instrumentation. The licensee also proposes 
a minor correction to revise ACTION c of TS 3.4.1.4.2, ``Reactor 
Coolant System, Cold Shutdown--Loops Not Filled,'' to change the 
requirement for verification of boron concentration to verification of 
shutdown margin.
    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 change[s] involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The extended range neutron flux monitoring instrumentation that 
is the subject of the proposed change performs a monitoring function 
and of itself has no potential as an accident initiator. The 
proposed requirement for the condition where both channels of the 
function are inoperable establishes actions that preserve the design 
basis where no actions previously existed. This is a more 
restrictive change and thus does not increase the probability or 
consequences of an accident previously evaluated.
    The proposed change[s] to TS 3.4.1.4.2 ACTION c. clarification 
regarding the verification of shutdown margin [do] not result in any 
technical change in the way the TS ACTION is applied. Therefore this 
proposed change does not increase the probability or consequences of 
an accident previously evaluated.
    The proposed change[s] [include] formatting changes that are 
administrative and consequently have no effect on accident analyses.
    Therefore, the proposed change[s] [do] not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. [Do] the proposed change[s] create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not involve any physical alteration of 
plant equipment and [do] not change the method by which any safety 
related structure, system, or component performs its function or is 
tested. As such, no new or different types of equipment will be 
installed, and the basic operation of installed equipment is 
unchanged. The methods governing plant operation and testing remain 
consistent with current safety analysis assumptions.
    The proposed change[s] [include] formatting changes that are 
administrative and consequently have no effect on accident analyses.
    Therefore, the proposed change[s] will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. [Do] the proposed change[s] involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes do not negate any existing requirement, and 
d[o] not adversely affect existing plant safety margins or the 
reliability of the equipment assumed to operate in the safety 
analysis. The purpose of the proposed changes is to provide greater 
assurance that the design basis is maintained. There are no changes 
being made to safety analysis assumptions, safety limits or safety 
system settings that would adversely affect plant safety as a result 
of the proposed change[s].
    The proposed change[s] [include] formatting changes that are 
administrative and consequently have no effect on accident analyses.
    Therefore, the proposed change[s] [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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Branch Chief: Thomas G. Hiltz.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: January 23, 2008.
    Description of amendment request: The amendments would revise the 
Technical Specification (TS) 3.6.1.3 Actions to (1) allow entry and 
exit through the containment air lock doors, even if the applicable 
action requires the containment air lock door to be closed, and (2) 
expand the current guidance provided to address inoperable air lock 
components.
    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 change[s] involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed Technical Specification changes to revise the 
action requirements associated with the containment air lock will 
not cause an accident to occur and will not result in any change in 
the operation of the associated accident mitigation equipment. The 
containment air lock is not an accident initiator. The proposed 
changes will not revise the operability requirements (e.g., leakage 
limits) for the containment air lock. Proper operation of the 
containment air lock will still be verified. As a result, the design 
basis accidents will remain the same postulated events described in 
the South Texas Project Unit 1 and Unit 2 Updated Final Safety 
Analysis Report, and the consequences of the design basis accidents 
will remain the same.
    Therefore, the proposed changes will not increase the 
probability or consequences of an accident previously evaluated.
    2. [Do] the proposed change[s] create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes to the Technical Specifications do not 
impact any system or component that could cause an accident. The 
proposed changes will not alter the plant configuration (no new or 
different type of

[[Page 15789]]

equipment will be installed) or require any unusual operator 
actions. The proposed changes will not alter the way any structure, 
system, or component functions, and will not significantly alter the 
manner in which the plant is operated. The response of the plant and 
the operators following an accident will not be different. In 
addition, the proposed changes do not introduce any new failure 
modes.
    Therefore, the proposed changes will not create the possibility 
of a new or different kind of accident from any accident previously 
analyzed.
    3. [Do] the proposed change[s] involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed Technical Specification changes to revise the 
action requirements associated with the containment air lock will 
not cause an accident to occur and will not result in any change in 
the operation of the associated accident mitigation equipment. The 
operability requirements for the containment air lock have not been 
changed. The containment air lock will continue to function as 
assumed in the safety analysis. In addition, the proposed changes 
will not adversely affect equipment design or operation, and there 
are no changes being made to the Technical Specification required 
safety limits or safety system settings that would adversely affect 
plant safety.
    Therefore, the proposed changes will not result in 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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Branch Chief: Thomas G. Hiltz

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

    Date of amendment request: December 28, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specification Administrative Controls Section 5.5.8, 
``Inservice Testing Program,'' to indicate that the Inservice Testing 
Program shall include testing frequencies applicable to the American 
Society of Mechanical Engineers (ASME) Code for Operation and 
Maintenance, and to indicate that there may be some non-standard 
frequencies specified as 2 years or less in the Inservice Testing 
Program to which the provisions of Surveillance Requirement 3.0.2 is 
applicable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises TS 5.5.8, ``Inservice Testing 
Program,'' for consistency with the requirements of 10 CFR 
50.55a(f)(4) regarding the inservice testing of pumps and valves. 
The proposed change incorporates revisions to the ASME Code that 
result in a net improvement in the measures for testing pumps and 
valves. The proposed change does not impact any accident initiators 
or analyzed events or assumed mitigation of accident or transient 
events, nor does it involve the addition or removal of any 
equipment, or any design changes to the facility. Therefore, the 
proposed change does not represent a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No
    The proposed change revises TS 5.5.8, ``lnservice Testing 
Program,'' for consistency with the requirements of 10 CFR 
50.55a(f)(4) regarding the inservice testing of pumps and valves. 
The proposed change incorporates revisions to the ASME Code that 
result in a net improvement in the measures for testing pumps and 
valves.
    The proposed change does not involve a modification to the 
physical configuration of the plant (i.e., no new equipment will be 
installed) or change in the methods governing normal plant 
operation. The proposed change will not impose any new or different 
requirements or introduce a new accident initiator, accident 
precursor, or malfunction mechanism. Additionally, there is no 
change in the types or increases in the amounts of any effluent that 
may be released off-site, and there is no increase in individual or 
cumulative occupational exposure. Therefore, this proposed change 
does not create the possibility of an accident of a different kind 
than previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No
    The proposed change revises TS 5.5.8, ``Inservice Testing 
Program, '' for consistency with the requirements of 10 CFR 
50.55a(f)(4) regarding the inservice testing of pumps and valves. 
The proposed change incorporates revisions to the ASME Code that 
result in a net improvement in the measures for testing pumps and 
valves. The safety functions of the affected pumps and valves will 
be maintained. Therefore, this 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: John O'Neill, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: Thomas G. Hiltz.

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

    Date of amendment request: December 28, 2007.
    Description of amendment request: The amendment would revise 
Technical Specifications (TS) 3.7.2, to add the Main steam isolation 
valve (MSIV) bypass valves to the scope of the TS. The proposed changes 
include a revision to the APPLICABILITY for the TS and a revision to 
footnote (i) in Table 3.3.2-1 of TS 3.3.2, ``ESFAS Instrumentation,'' 
to make it consistent with the revised Applicability of LCO 3.7.2. The 
amendment would also add new TS 3.7.19, ``Secondary System Isolation 
Valves (SSIVs),'' to include Limiting Conditions for Operation and 
Surveillance Requirements for the secondary system isolation valves: 
Main steam low point drain isolation valves, steam generator chemical 
injection isolation valves, steam generator blowdown isolation valves, 
and steam generator sample line isolation valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change adds requirements to the TS to ensure that 
systems and components are maintained consistent with the safety 
analysis and licensing basis.
    Requirements are incorporated into the TS for secondary system 
isolation valves. These changes do not involve any design or 
physical changes to the facility, including the SSIVs themselves. 
The design and functional performance requirements, operational 
characteristics, and reliability of the SSIVs are unchanged. There 
is no impact on the design safety function of MSIVs, MFIVs, MFRVs or 
MFRVBVs [main steam isolation valves, main feedwater isolation 
valves, main feedwater regulating valves, main isolation feedwater 
regulating valve bypass valves] to close (either as an accident 
mitigator or as a potential transient initiator). Since no failure 
mode or initiating condition that could cause an accident (including 
any plant transient) evaluated per the FSAR [final safety analysis 
report]-described safety analyses is created or

[[Page 15790]]

affected, the change cannot involve a significant increase in the 
probability of an accident previously evaluated.
    With regard to the consequences of an accident and the equipment 
required for mitigation of the accident, the proposed changes 
involve no design or physical changes to components in the main 
steam supply system or feedwater system. There is no impact on the 
design safety function of MSIVs, MFIVs, MFRVs, or MFRVBVs or any 
other equipment required for accident mitigation. Adequate equipment 
availability would continue to be required by the TS. The 
consequences of applicable, analyzed accidents (such as a main steam 
line break of feedline break) are not impacted by the proposed 
changes.
    The change in APPLICABILITY for the MSIVs is consistent with the 
Westinghouse Standard Technical Specification 3.7.2. The change to 
footnote (i) in TS Table 3.3.2-1 makes the provisions of that note 
for the affected instrumentation consistent with the revised 
APPLICABILITY of TS 3.7.2. These changes involve no physical changes 
to the facility and do not adversely affect the availability of the 
safety functions assumed for the MSIVs and SSIVs. Therefore, they do 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated. Based on the above 
considerations, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes add requirements to the TS that support or 
ensure the availability of the safety functions assumed or required 
for the MSIVs and SSIVs. The changes do not involve a physical 
alteration of the plant (no new or different type of equipment will 
be installed) or changes in controlling parameters. Additional 
requirements are being imposed, but they are consistent with the 
assumptions made in the safety analysis and licensing basis. The 
addition of Conditions, Required Actions and Completion Times to TS 
for the SSIVs does not involve a change in the design, 
configuration, or operational characteristics of the plant. Further, 
the proposed changes do not involve any changes in plant procedures 
for ensuring that the plant is operated within analyzed limits. As 
such, no new failure modes or mechanisms that could cause a new or 
different kind of accident from any previously evaluated are 
introduced.
    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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed addition of Conditions, Required Actions and 
Completion Times for SSIVs, as well as the proposed change to the 
APPLICABILITY for the MSIV TS (and the corresponding change to the 
footnote for the ESFAS Instrumentation in TS 3.3.2) does not alter 
the manner in which safety limits or limiting safety system settings 
are determined. No changes to instrument/system actuation setpoints 
are involved. The safety analysis acceptance criteria are not 
impacted and the proposed change will not permit plant operation in 
a configuration outside the design basis. The changes are consistent 
with the safety analysis and licensing basis for the facility.
    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: John O'Neill, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: Thomas G. Hiltz.

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

    Date of amendment request: December 28, 2007.
    Description of amendment request: The amendment would incorporate 
changes in the Technical Specifications (TS). Specifically, a footnote 
associated with Table 3.3.2-1 of Technical Specification 3.3.2, 
``Engineered Safety Feature Actuation System (ESFAS) Instrumentation,'' 
would be revised to make the exception allowed by the footnote 
consistent with the scope and Applicability of TS 3.7.3, ``Main 
Feedwater Isolation Valves (MFIVs) and Main Feedwater Regulating Valves 
(MFRVs) and Main Feedwater Regulating Valve Bypass Valves (MFRVBVs)'' 
and a Note connected with each of two Surveillance Requirements (SRs), 
i.e., SR 3.7.2.1 and SR 3.7.2.2 under TS 3.7.2, ``Main Steam Isolation 
Valves (MSIVs),'' would be deleted as it is no longer needed or 
appropriate for the affected SRs.
    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
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since there are 
no design changes. All design, material, and construction standards 
that were applicable prior to this amendment request will be 
maintained. There will be no changes to any design or operating 
limits.
    The proposed changes will not change accident initiators or 
precursors assumed or postulated in the final safety analysis report 
(FSAR)-described accident analyses, nor will they alter the design 
assumptions, conditions, and configuration of the facility or the 
manner in which the plant is operated and maintained. The proposed 
changes will not alter or prevent the ability of structures, 
systems, and components (SSCs) from performing their intended 
functions to mitigate the consequences of an initiating event within 
the assumed acceptance limits.
    The proposed changes do not physically alter safety-related 
systems, nor do they affect the way in which safety-related systems 
perform their functions.
    All accident analysis acceptance criteria will continue to be 
met with the proposed changes. The proposed changes will not affect 
the source term, containment isolation, or radiological release 
assumptions used in evaluating the radiological consequences of an 
accident previously evaluated. The proposed changes will not alter 
any assumptions or change any mitigation actions in the radiological 
consequence evaluations in the FSAR. The applicable radiological 
dose acceptance criteria will continue to be met.
    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
    There are no proposed design changes, nor are there any changes 
in the method by which any safety-related plant structure, system, 
or component (SSC) performs its specified safety function. The 
proposed changes will not affect the normal method of plant 
operation or change any operating parameters. No equipment 
performance requirements will be affected. The proposed changes will 
not alter any assumptions made in the safety analyses.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures will be introduced as a 
result of this amendment. There will be no adverse effect or 
challenges imposed on any safety-related system as a result of this 
amendment.
    The proposed amendment will not alter the design or performance 
of the 7300 Process Protection System, Nuclear Instrumentation 
System, or Solid State Protection System used in the plant 
protection systems.
    Therefore, the proposed changes do 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
    There will be no effect on those plant systems necessary to 
assure the accomplishment of protection functions.

[[Page 15791]]

There will be no impact on the overpower limit, departure from 
nucleate boiling ratio (DNBR) limits, heat flux hot channel factor 
(FQ), nuclear enthalpy rise hot channel factor (FAH), loss of 
coolant accident peak cladding temperature (LOCA PCT), peak local 
power density, or any other margin of safety. The applicable 
radiological dose consequence acceptance criteria for design-basis 
transients and accidents will continue to be met.
    The proposed changes do not eliminate any surveillance or alter 
the frequency of surveillances required by the Technical 
Specifications. None of the acceptance criteria for any accident 
analysis will be changed.
    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: John O'Neill, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: Thomas G. Hiltz.

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

    Date of amendment request: November 29, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.4.10, ``Pressurizer Safety 
Valves,'' and TS 3.4.11, ``Pressurizer Power Operated Relief Valves 
(PORVs),'' to modify the completion times for default conditions in 
both TSs and to allow separate condition entry for PORV block valves in 
TS 3.4.11. The amendment request is adopting the following two Nuclear 
Regulatory Commission (NRC)-approved TS Task Force (TSTF) travelers to 
the standard TSs: TSTF-247-A and TSTF-352-A.
    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 change[s] involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since there are 
no design changes. All design, material, and construction standards 
that were applicable prior to this amendment request will be 
maintained. There will be no changes to the design and operating 
temperature and pressure limits placed on the reactor coolant 
system.
    The proposed changes will not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, and configuration of the facility or the manner in which 
the plant is operated and maintained. The proposed changes will not 
alter or prevent the ability of structures, systems, and components 
(SSCs) from performing their intended functions to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits.
    The proposed changes do not physically alter safety-related 
systems nor affect the way in which safety-related systems perform 
their functions.
    All accident analysis acceptance criteria will continue to be 
met with the proposed changes. The proposed changes will not affect 
the source term, containment isolation, or radiological release 
assumptions used in evaluating the radiological consequences of an 
accident previously evaluated. The proposed changes will not alter 
any assumptions or change any mitigation actions in the radiological 
consequence evaluations in the FSAR [Final Safety Analysis Report 
for the plant]. The applicable radiological dose acceptance criteria 
will continue to be met.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. [Do] the proposed change[s] create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    There are no proposed design changes nor are there any changes 
in the method by which any safety-related plant SSC performs its 
safety function. The proposed changes will not affect the normal 
method of plant operation or change any operating parameters. No 
equipment performance requirements will be affected. The proposed 
changes will not alter any assumptions made in the safety analyses.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures will be introduced as a 
result of this amendment. There will be no adverse effect or 
challenges imposed on any safety-related system as a result of this 
amendment.
    The proposed amendment will not alter the design or performance 
of the 7300 Process Protection System, Nuclear Instrumentation 
System, or Solid State Protection System used in the plant 
protection systems.
    Therefore, the proposed changes do not create the possibility of 
a new or different accident from any accident previously evaluated.
    3. [Do] the proposed change[s] involve a significant reduction 
in a margin of safety?
    Response: No.
    There will be no effect on those plant systems necessary to 
assure the accomplishment of protection functions. There will be no 
impact on the overpower limit, departure from nucleate boiling ratio 
(DNBR) limits, heat flux hot channel factor (FQ), nuclear enthalpy 
rise hot channel factor (F[Delta]H), loss of coolant accident peak 
cladding temperature (LOCA PCT), peak local power density, or any 
other margin of safety. The applicable radiological dose consequence 
acceptance criteria will continue to be met. The proposed changes do 
not eliminate any surveillances or alter the frequency of 
surveillances required by the Technical Specifications. None of the 
acceptance criteria for any accident analysis will be changed.
    The proposed changes will have no impact on the radiological 
consequences of a design basis accident.
    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: John O'Neill, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: Thomas G. Hiltz.

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.

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

    Date of amendment request: February 13, 2008.
    Brief description of amendment request: The amendments propose a 
one time steam generator (SG) tubing eddy current inspection interval 
revision to the Vogtle Electric Generating Plant, Units 1 and 2 (Vogtle 
1 and 2) Technical Specifications (TSs) 5.5.9, ``Steam Generator (SG) 
Program,'' to incorporate an interim alternate repair criterion

[[Page 15792]]

(ARC) in the provisions for SG tube repair criteria during the Vogtle 1 
inspection performed in refueling outage 14 and subsequent operating 
cycle, and during the Vogtle 2 inspection performed in refueling outage 
13 and subsequent 18-month SG tubing eddy current inspection interval 
and subsequent 36-month SG tubing eddy current inspection interval. The 
amendments also revise TS 5.6.10, ``Steam Generator Tube Inspection 
Report,'' where three new reporting requirements are proposed to be 
added to the existing seven requirements.
    Date of publication of individual notice in Federal Register: 
February 26, 2008 (73 FR 10305).
    Expiration date of individual notice: April 28, 2008.

Notice of Issuance of Amendments to Facility Operating Licenses

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

Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc., 
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, 
Louisiana

    Date of amendment request: March 28, 2007, as supplemented by 
letter dated October 24, 2007.
    Brief description of amendment: The amendment revised the required 
wattage specified in the River Bend Station, Unit 1, Technical 
Specification 5.5.7.e, Ventilation Filter Testing Program, for the 
Control Room Fresh Air System (CRFAS) heater for testing. The required 
wattage for testing the CRFAS heater was revised from 23  
2.3 kilowatt (kW) to ``>==15 kW.''
    Date of issuance: February 28, 2008
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 159
    Facility Operating License No. NPF-47: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: May 8, 2007 (72 FR 
26175). The supplement dated October 24, 2007, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register on May 8, 2007 (72 FR 26175).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 28, 2008.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 30, 2007, as supplemented 
by letter dated December 5, 2007.
    Brief description of amendment: The amendment revised Technical 
Specification 3.1.3.4, ``Reactivity Control Systems CEA [Control 
Element Assembly] Drop Time,'' to change the individual rod drop time 
from the fully withdrawn position to 90 percent insertion from less 
than or equal to 3.5 seconds to less than or equal to 3.7 seconds.
    Date of issuance: March 5, 2008.
    Effective date: As of its date of issuance and shall be implemented 
prior to startup following the spring 2008 refueling outage.
    Amendment No.: 275.
    Renewed Facility Operating License No. NPF-6: The amendment revised 
the Technical Specifications and license.
    Date of initial notice in Federal Register: October 9, 2007 (72 FR 
57354). The supplemental letter dated December 5, 2007, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 5, 2008.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 15, 2007.
    Brief description of amendment: The amendment changes Technical 
Specification (TS) Section 1.4 and Section 5. Changes to TS 1.4 
incorporate Nuclear Regulatory Commission (NRC)-approved Technical 
Specification Task Force (TSTF) Standard Technical Specification 
Changes TSTF-284, ``Add `Met vs. Perform' to Specification 1.4, 
Frequency,'' Revision 3, TSTF-485-A, ``Correction Example 1.4-1,'' 
Revision 0, and make administrative changes. Changes to TS Section 5 
incorporate NRC-approved TSTF-258, ``Changes to Section 5.0, 
Administrative Controls,'' Revision 4, NRC-approved TSTF-273, ``[Safety 
Functions Determination Program] SFDP Clarifications,'' Revision 2, as 
amended by Westinghouse Owners Group (WOG) editorial change WOG-ED-23, 
and make administrative changes.
    Date of issuance: March 5, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 231
    Facility Operating License No. DPR-20: Amendment revised the 
Technical Specifications and Renewed License.

[[Page 15793]]

    Date of initial notice in Federal Register: June 19, 2007 (72 FR 
33782).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 5, 2008.
    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: June 18, 2007.
    Brief description of amendments: The amendment revised Technical 
Specification 3.7.5, ``Control Room Area Ventilation Air Conditioning 
(AC) System,'' to add an Action Statement for two inoperable control 
room area ventilation AC subsystems. This operating license improvement 
was made available by the Nuclear Regulatory Commission on March 26, 
2007 (72 FR 14143) as part of the consolidated line item improvement 
process.
    Date of issuance: March 10, 2008
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 188/175
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications and License.
    Date of initial notice in Federal Register: September 1, 2007 (72 
FR 51860). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 10, 2008.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2 and 3, 
York and Lancaster Counties, Pennsylvania

    Date of amendment request: November 17, 2006, as supplemented by 
letters dated September 21, 2007, December 21, 2007, February 1, 2008, 
and February 14, 2008.
    Brief description of amendment: The amendment revises Technical 
Specification Surveillance Requirement 3.3.1.1.8 to increase the 
frequency interval between Local Power Range Monitor (LPRM) 
calibrations from 1000 megawatt days per ton (MWD/T) average core 
exposure to 2000 MWD/T average core exposure. The LPRM system provides 
signals to associated nuclear instrumentation systems that serve to 
detect conditions in the core that have the potential to threaten the 
overall integrity of the fuel barrier. The LPRM system also 
incorporates features designed to diagnose and display various system 
trip and inoperative conditions.
    Date of issuance: February 29, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 266 and 270
    Facility Operating License Nos. DPR-44 and DPR-56: Amendment 
revised the License and Technical Specifications.
    Date of initial notice in Federal Register: August 28, 2007 (72 FR 
49577). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 29, 2008.
    No significant hazards consideration comments received: No.

FPL Energy, Point Beach, 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 12, 2007, as 
supplemented by letters dated December 12, and December 21, 2007.
    Brief description of amendments: The amendments revises Technical 
Specification 5.5.15 ``Containment Leakage Rate Testing Program,'' for 
Units 1 and 2. The proposed change allows a one-time interval extension 
of no more than 5 years for the Type A, Integrated Leakage Rate Test.
    Date of issuance: February 26, 2008
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 232, 237
    Renewed Facility Operating License Nos. DPR-24 and DPR-27: 
Amendments revised the Technical Specifications/License.
    Date of initial notice in Federal Register: December 4, 2007 (72 FR 
68217). The supplements contained clarifying information and did not 
change the staff's initial proposed finding of no significant hazards 
consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 26, 2008.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 15, 2006
    Brief description of amendments: The amendments revised Action Q of 
Technical Specifications Section 3.3.1, ``Reactor Trip System (RTS) 
Instrumentation,'' to reflect deletion of the power range neutron flux 
high negative rate trip function previously approved by Amendment Nos. 
293 (for Unit 1) and 275 (for Unit 2).
    Date of issuance: March 5, 2008
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 302 (for DCCNP-1) and 285 (for DCCNP-2)
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Renewed Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: November 21, 2006 (71 
FR 67396).
    The Commission's related evaluation of the amendment is contained 
in a safety evaluation dated March 5, 2008.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 22, 2007, as supplemented by letter 
dated December 5, 2007.
    Brief description of amendments: The amendments revised the 
Technical Requirements Surveillance 13.3.33.2, Cycling Frequency for 
the Turbine Stop and Control Valves. The change will increase the valve 
cycle frequency interval from 12 to 26 weeks.
    Date of issuance: February 29, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: Unit 1-143; Unit 2-143
    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: August 14, 2007 (72 FR 
45462). The supplement dated December 5, 2007, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register on August 14, 2007 (72 FR 45462).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 29, 2008.

[[Page 15794]]

    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 30, 2007, as supplemented 
by letters dated October 16, 2007, and November 2, 2007.
    Brief description of amendment: The amendment changes the NMP2 
Technical Specifications to reflect an expanded operating domain 
resulting from implementation of Average Power Range Monitor/Rod Block 
Monitor/Technical Specifications/Maximum Extended Load Line Analysis 
(ARTS/MELLLA). The Average Power Range Monitor (APRM) flow-biased 
simulated thermal power allowable value (AV) would be revised to permit 
operation in the MELLLA region. The current flow-biased Rod Block 
Monitor (RBM) would be replaced by a power dependent RBM, which also 
would require new AVs. The flow-biased APRM simulated thermal power 
setdown requirement would be replaced by more direct power and flow 
dependent thermal limits administration. The Surveillance Requirement 
for the standby liquid control (SLC) system would be revised to require 
each SLC pump to deliver required flow at a discharge pressure >=1325 
psig in lieu of >=1320 psig; the SLC relief valve setpoint would be 
increased from 1394 psig to 1400 psig. Finally, the proposed amendment 
employs a new model for performing the anticipated transients without 
scram analysis for ARTS/MELLLA conditions.
    Date of issuance: February 27, 2008
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 123
    Renewed Facility Operating License No. NPF-69: Amendment revises 
the License and Technical Specifications.
    Date of initial notice in Federal Register: May 22, 2007 (72 FR 
28721). The supplements dated October 16, 2007, and November 2, 2007, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the Nuclear Regulatory Commission staff's initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 27, 2008.
    No significant hazards consideration comments received: No

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

    Date of application for amendment: January 30, 2007, as 
supplemented by letter dated December 28, 2007.
    Brief description of amendment: The amendment revised Technical 
Specifications (TSs) Surveillance Requirement (SR) 3.5.1.3.b to 
correctly state that the required pressure at which the Alternate 
Nitrogen System is determined to be operable should be greater than or 
equal to 410 psig, not the former stated pressure of greater than or 
equal to 220 psig. The safety-related Alternate Nitrogen System 
provides an alternate pressure source to equipment required during or 
following an accident. The licensee determined that the former 
acceptance value specified by SR 3.5.1.3.b (greater than or equal to 
220 psig ) was non-conservative and needed to be corrected to the 
higher value.
    Date of issuance: February 21, 2008
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 155
    Facility Operating License No. DPR-22: Amendment revised the 
Technical Specifications and the Operating License.
    Date of initial notice in Federal Register: March 27, 2007 (72 FR 
14307). The supplemental letter contained clarifying information, did 
not change the initial no significant hazards consideration 
determination, and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 21, 2008.
    No significant hazards consideration comments received: No

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power 
Plant, Unit 3, Humboldt County, California (TAC. No. J52690)

    Date of application for amendment: May 17, 2006, supplemented 
January 25, 2008.
    Brief description of amendment: The amendment approves a proposed 
change to the Physical Security Plan related to security post manning 
requirements.
    Date of issuance: February 27, 2008
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 42
    Facility Operating License No. DPR-7: This amendment revises the 
License.
    Date of initial notice in Federal Register: February 13, 2007 (72 
FR 6788).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 27, 2008.
    No significant hazards consideration comments received: No

PSEG Nuclear LLC, Docket No. 50-311, Salem Nuclear Generating Station, 
Unit No. 2, Salem County, New Jersey

    Date of application for amendment: October 17, 2007, as 
supplemented on January 11, 2008.
    Brief description of amendment: The amendment allows a one-time 
revision to the requirements for fuel decay time prior to commencing 
movement of irradiated fuel in the reactor. Specifically, the proposed 
amendment revises Technical Specification (TS) 3/4.9.3 to allow fuel 
movement to commence at 86 hours after the reactor is subcritical. The 
proposed change is only applicable to Salem Unit 2 refueling outage 
2R16 which is scheduled to commence on March 11, 2008.
    Date of issuance: March 5, 2008
    Effective date: As of the date of issuance, to be implemented 
within 7 days.
    Amendment No.: 271
    Facility Operating License No. DPR-75: The amendment revises the 
TSs and the license.
    Date of initial notice in Federal Register: December 4, 2007 (72 FR 
68218). The letter dated January 11, 2008, provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination or expand the application beyond 
the scope of the original Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 5, 2008.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 28, 2007, as 
supplemented on October 9, 2007, December 21, 2007, January 18, 2008, 
and January 30, 2008.
    Brief description of amendments: The amendments revised the 
``Maximum Power Level'' in paragraph 2.C(1) of the Vogtle Electric 
Generating Plant, Facility Operating Licenses NPF-68 and NPF-81 for 
Unit 1 and Unit 2, respectively. In addition, the amendments revised 
the definition of

[[Page 15795]]

``Rated Thermal Power (RTP)'' in Technical Specification 1.1 for both 
units to reflect the change to the Maximum Power Level. The proposed 
change increased the RTP from 3565 MWt to 3625.6 MWt, resulting in an 
increase of 1.7% from the current reactor output. This increase in 
reactor core power level is referred to as a Measurement Uncertainty 
Recapture (MUR) power uprate.
    Date of issuance: February 27, 2008
    Effective date: As of the date of issuance and shall be implemented 
at the completion of spring 2008 refueling outage for Unit 1 and fall 
2008 refueling outage for Unit 2.
    Amendment Nos.: 149, 129
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: November 20, 2007 (72 
FR 65372). The supplements dated October 9, 2007, December 21, 2007, 
January 18, 2008, and January 30, 2008, 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 February 27, 2008.
    No significant hazards consideration comments received: No

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

    Date of application for amendments: August 28, 2007, as 
supplemented on October 9, 2007, December 21, 2007, January 18, 2008, 
and January 30, 2008.
    Brief description of amendments: The amendments revised the 
``Maximum Power Level'' in paragraph 2.C(1) of the Vogtle Electric 
Generating Plant, Facility Operating Licenses NPF-68 and NPF-81 for 
Unit 1 and Unit 2, respectively. In addition, the amendments revised 
the definition of ``Rated Thermal Power (RTP)'' in Technical 
Specification 1.1 for both units to reflect the change to the Maximum 
Power Level. The proposed change increased the RTP from 3565 MWt to 
3625.6 MWt, resulting in an increase of 1.7% from the current reactor 
output. This increase in reactor core power level is referred to as a 
Measurement Uncertainty Recapture (MUR) power uprate.
    Date of issuance: February 27, 2008
    Effective date: As of the date of issuance and shall be implemented 
at the completion of spring 2008 refueling outage for Unit 1 and fall 
2008 refueling outage for Unit 2.
    Amendment Nos.: 149, 129
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: November 20, 2007 (72 
FR 65372). The supplements dated October 9, 2007, December 21, 2007, 
January 18, 2008, and January 30, 2008, 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 February 27, 2008.
    No significant hazards consideration comments received: No

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: March 22, 2007, as supplemented by 
letters dated April 10, July 18, October 11, November 13, December 13, 
and December 18, 2007.
    Brief description of amendments: The amendments revised the 
licensing basis, pursuant to Title 10 of the Code of Federal 
Regulations, Section 50.67, ``Accident Source Term,'' and approved the 
methodology for evaluating radiological consequences of design-basis 
accidents as described in Regulatory Guide 1.183, ``Alternative 
Radiological Source Terms for Evaluating Design Basis Accidents (DBAs) 
at Nuclear Power Reactors.'' The amendments revised the Technical 
Specifications in support of the revisions to the licensing basis.
    Date of issuance: March 6, 2008
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment Nos.: Unit 1--182; Unit 2--169
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: July 31, 2007 (72 FR 
41788). The supplemental letters dated April 10, July 18, October 11, 
November 13, December 13, and December 18, 2007, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 6, 2008.
    No significant hazards consideration comments received: No.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: March 14, 2007, as supplemented by 
letter dated December 18, 2007.
    Brief description of amendment: The amendment revised TS Table 
3.3.2-1, ``Engineered Safety Features Actuation System 
Instrumentation,'' to separate the automatic actuation logic and 
actuation relays for steam line isolation (Function 4) and main 
feedwater isolation (Function 5) into the solid state protection system 
function and the main steam and feedwater isolation system. There are 
other proposed changes to the TSs and the plant in the application that 
are not being addressed in this amendment. The amendment to revise 
Surveillance Requirements 3.7.2.1 and 3.7.3.1 to replace the valve 
isolation times with the phrase ``within limits'' was issued August 28, 
2007. The remaining TS and plant changes in the application will be 
addressed in future letters to the licensee.
    Date of issuance: March 3, 2008
    Effective date: As of its date of issuance and shall be implemented 
prior to the startup from Refueling Outage 16, scheduled for the spring 
of 2008.
    Amendment No.: 175
    Facility Operating License No. NPF-42: The amendment revised the 
Operating License and Technical Specifications.
    Date of initial notice in Federal Register: The supplemental letter 
dated December 18, 2007, 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 on June 19, 2007 (72 FR 33785).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 3, 2008.
    No significant hazards consideration comments received: No

    Dated at Rockville, Maryland, this 17th day of March 2008.


[[Page 15796]]


    For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. E8-5734 Filed 3-24-08; 8:45 am]
BILLING CODE 7590-01-P