Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 50356-50368 [E8-19369]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 73, Number 166 (Tuesday, August 26, 2008)]
[Notices]
[Pages 50356-50368]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E8-19369]


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 NUCLEAR REGULATORY COMMISSION


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses Involving No Significant Hazards Considerations

I. Background

    Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from July 31, 2008 to August 13, 2008. The last 
biweekly notice was published on August 12, 2008 (73 FR 46926).

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-

[[Page 50357]]

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 6D44, 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, https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request 
for a hearing or petition for leave to intervene is filed within 60 
days, the Commission or a presiding officer designated by the 
Commission or by the Chief Administrative Judge of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the Chief Administrative Judge of the Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for 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 
hearingdocket@nrc.gov, 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 https://www.nrc.gov/site-help/e-submittals/install-
viewer.html.

[[Page 50358]]

Information about applying for a digital ID certificate is available on 
NRC's public Web site at https://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 https://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 https://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 
https://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, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr.resource@nrc.gov.
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power 
Station (KPS), Kewaunee County, Wisconsin
    Date of amendment request: July 7, 2008.
    Description of amendment request: The proposed amendment would 
allow the use of a methodology to determine the seismic loads on the 
recently upgraded Auxiliary Building crane. The upgrade was to make the 
crane single-failure-proof through replacement of the crane trolley, 
and modification of the existing crane bridge. The proposed seismic 
analysis methodology has not been approved for use at KPS, and is thus 
not currently in the KPS Updated Safety Analysis Report. The proposed 
methodology recognizes the inherent propensity for structures not fixed 
to one another (e.g., steel wheels on steel rails) to roll if 
sufficient lateral force is applied. The licensee proposed this seismic 
analysis methodology for use solely on the Auxiliary Building crane 
upgrade. The licensee stated that recognition of wheel rolling between 
the crane trolley and bridge and their respective rails reflects the 
true nature of the installed equipment and its response to horizontal 
forces generated by a seismic event.
    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. The NRC staff reviewed the licensee's analysis and has 
prepared its own as follows:

    (1) Does the proposed amendment involve a significant increase 
in the probability or consequences of an accident previously 
evaluated?
    Response: No.
    This proposed amendment pertains solely to a nonlinear seismic 
analysis method supporting the upgrade of the KPS Auxiliary Building 
crane from a non-single-failure-proof design to a single-failure 
proof design. Specifically, the existing crane trolley has been 
replaced with a state-of-the-art design that is single-failure-
proof. The crane does not interface with operating plant equipment, 
and will continue to be able to withstand a design-basis seismic 
event without an uncontrolled lowering of the load. Thus, the 
probability and consequences of a load drop are not increased by the 
upgrade and proposed change in seismic analysis methodology. 
Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    (2) Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    This proposed amendment pertains to an analysis method 
supporting the upgrade of an existing plant component. This seismic 
analysis methodology is proposed for use solely on the crane upgrade 
and not for any other plant structures, systems, or components. The 
design-rated load of the crane main hoist remains the same (i.e., 
125

[[Page 50359]]

tons). This load bounds the design and supporting analysis. The 
auxiliary hook design rated load has been increased from 10 tons to 
15 tons. The proposed amendment does not change the previously 
evaluated and currently acceptable heavy load handling practices in 
use at KPS. The number and types of lifts made using this crane in 
support of KPS plant operations will not significantly change from 
those contemplated during original plant licensing. Furthermore, the 
basic operations of the crane (i.e., hoisting and horizontal travel) 
will remain the same. Therefore, the proposed amendment does not 
create a new or different kind of accident from any previously 
evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The purpose of the proposed methodology is to determine the 
design loads (forces and moments), accelerations, and displacements 
on the crane and building support structure. These loads will 
subsequently be used to perform the structural analysis of these 
components to confirm that the design meets all applicable 
acceptance criteria using previously approved industry codes and 
standards for such analyses. If the stresses computed in the 
structural components as a result of a seismic event are less than 
the limits contained in these codes, the structural integrity of the 
crane is maintained, and a suspended load will remain suspended 
during a seismic event. Meeting these code limits maintains an 
acceptable margin of safety for the individual components and the 
crane as a whole. Therefore, the proposed amendment does not involve 
a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
its own analysis, proposes that the three standards of 10 CFR 50.92(c) 
are satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., Counsel for Dominion Energy Kewaunee, Inc., 
120 Tredegar Street, Richmond, VA 23219.
    NRC Branch Chief: Lois M. James.
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power 
Station (KPS), Kewaunee County, Wisconsin
    Date of amendment request: July 16, 2008.
    Description of amendment request: The proposed amendment would 
align the Technical Specifications (TS) with the results of an 
evaluation performed according to Westinghouse Nuclear Safety Advisory 
Letter NSAL-07-7, ``Short-Term Recriticality During a PWR [pressurized-
water reactor] Large-Break LOCA [loss-of-coolant accident].'' NSAL-07-7 
advised that the potential exists for recriticality to occur during a 
large-break LOCA in the reflood stage after a LOCA. Westinghouse 
determined that Kewaunee is not susceptible to the subject issue based 
on the current KPS practice of maintaining safety injection (SI) 
accumulator boron concentration at or above 2500 ppm. However, to 
ensure that the KPS TS are conservative with respect to the results of 
NSAL-07-7, the licensee proposed to raise the minimum required boron 
concentration for the SI accumulators specified by the TS from the 
current 1900 parts-per-million (ppm) to 2400 ppm.
    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 reproduced below:

    (1) Does the proposed amendment involve a significant increase 
in the probability or consequences of an accident previously 
evaluated?
    Response: No.
    Increasing the minimum required boron concentration in the SI 
accumulators does not add, delete, or modify any Kewaunee systems, 
structures, or components (SSCs). The SI accumulators and their 
contents are not accident initiators. Rather, they are designed for 
accident mitigation. The effects of an increase in the minimum SI 
accumulator boron concentration from 1900 ppm to 2400 ppm are 
bounded by previous evaluations and determined to be acceptable. 
Thus, the proposed increase in minimum SI accumulator boron 
concentration has no adverse effect on the ability of the plant to 
mitigate the effects of design[-]basis accidents.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    (2) Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    Increasing the minimum required boron concentration in the SI 
accumulators does not change the design function of the SI 
accumulators or the SSCs designed to deliver borated water from the 
SI accumulators to the core. Increasing the minimum required boron 
concentration in the SI accumulators does not create any credible 
new failure mechanisms or malfunctions for plant equipment or the 
nuclear fuel. The reactivity control function of the borated water 
in the SI accumulators is not being changed.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    An evaluation has been performed that shows that maintaining 
boron concentration at a minimum of 2400 ppm is sufficient to assure 
that acceptable results for design[-]basis accident analyses will be 
maintained considering the reactivity of the core. Increasing the 
minimum boron concentration in the Sl accumulator from 1900 ppm to a 
minimum of 2400 ppm increases the margin of safety in the Kewaunee 
safety analyses, since additional post-accident negative reactivity 
will be available to the core. This additional negative reactivity 
compensates for the potential for recriticality occurring during the 
short-term reflood period during the large[-]break loss-of-coolant 
accident. Additionally, the proposed new minimum boron concentration 
of 2400 ppm is within the range required by current safety analyses 
(i.e., 1900 ppm to 2625 ppm), and well below the currently 
acceptable maximum boron concentration of 2625 ppm.
    The proposed amendment does not result in altering or exceeding 
a design basis or safety limit for the plant. All current fuel 
design criteria will continue to be satisfied, and the safety 
analysis of record, including evaluations of the radiological 
consequences of design[-]basis accidents, will remain applicable.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and determines 
that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, 
the NRC staff proposes to determine that the amendment request involves 
no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., Counsel for Dominion Energy Kewaunee, Inc., 
120 Tredegar Street, Richmond, VA 23219.
    NRC Branch Chief: Lois M. James.
Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi
    Date of amendment request: June 30, 2008.
    Description of amendment request: The proposed amendment would: (1) 
Revise the Technical Specification (TS) surveillance requirement (SR 
3.1.3.2) frequency in TS 3.1.3, ``Control Rod OPERABILITY,'' (2) 
clarify the requirement to fully insert all insertable control rods for 
the limiting condition for operation (LCO) in TS 3.3.1.2, Required 
Action E.2, ``Source Range Monitoring Instrumentation,'' and (3) revise 
Example 1.4-3 in Section 1.4 ``Frequency'' to clarify the applicability 
of the 1.25 surveillance test interval extension. The licensee is 
proposing to adopt the approved Technical Specification Task Force 
(TSTF) change traveler TSTF-475, Revision 1, ``Control Rod Notch 
Testing Frequency.''

[[Page 50360]]

    The NRC staff issued a ``Notice of Availability of Model 
Application Concerning Technical Specification Improvement To Revise 
Control Rod Notch Surveillance Frequency, Clarify SRM Insert Control 
Rod Action, and Clarify Frequency Example'' associated with TSTF-275, 
Revision 1, in the Federal Register on November 13, 2007 (72 FR 63935). 
The notice included a model safety evaluation, a model no significant 
hazards consideration (NSHC) determination and a model license 
amendment request, using the consolidated line item improvement 
process. In its application dated June 30, 2008, the licensee affirmed 
the applicability of the model NSHC determination which is presented 
below:
    Basis for proposed NSHC determination: As required by 10 CFR 
50.91(a), an analysis of the issue of NSHC determination 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 generically implements TSTF-475, Revision 1, 
``Control Rod Notch Testing Frequency and SRM Insert Control Rod 
Action.'' TSTF-475, Revision 1 modifies NUREG-1433 (BWR/4) and NUREG-
1434 (BWR/6) STS [(Standard Technical Specifications)]. The changes: 
(1) Revise TS testing frequency for surveillance requirement (SR) 
3.1.3.2 in TS 3.1.3, ``Control Rod OPERABILITY,'' (2) clarify the 
requirement to fully insert all insertable control rods for the 
limiting condition for operation (LCO) in TS 3.3.1.2, Required Action 
E.2, ``Source Range Monitoring Instrumentation'' (NUREG-1434 only), and 
(3) revise Example 1.4-3 in Section 1.4 ``Frequency'' to clarify the 
applicability of the 1.25 surveillance test interval extension. The 
consequences [and probability] of an accident after adopting TSTF-475, 
Revision 1 are no different than the consequences [and probability] of 
an accident prior to adoption. Therefore, this 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 involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. The proposed 
change will not introduce new failure modes or effects and will not, in 
the absence of other unrelated failures, lead to an accident whose 
consequences exceed the consequences of accidents previously analyzed. 
Thus, this change does not create the possibility of a new or different 
kind of accident from any accident.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety
    TSTF-475, Revision 1 will: (1) [revise the TS SR 3.1.3.2 frequency 
in TS 3.1.3, ``Control Rod OPERABILITY,'' (2) clarify the requirement 
to fully insert all insertable control rods for the limiting condition 
for operation (LCO) in TS 3.3.1.2, ``Source Range Monitoring 
Instrumentation,'' and (3)] revise Example 1.4-3 in Section 1.4 
``Frequency'' to clarify the applicability of the 1.25 surveillance 
test interval extension. [The GE Nuclear Energy Report, ``CRD Notching 
Surveillance Testing for Limerick Generating Station,'' dated November 
2006, concludes that extending the control rod notch test interval from 
weekly to monthly is not expected to impact the reliability of the 
scram system and that the analysis supports the decision to change the 
surveillance frequency.] Therefore, the proposed changes in TSTF-475, 
Revision 1 [* * *] do not involve a significant reduction in a margin 
of safety.
    The NRC staff has reviewed the analysis adopted by the licensee 
and, based on this review, it appears that the three standards of 10 
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves NSHC.
    Attorney for licensee: Terence A. Burke, Associate General 
Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Acting Branch Chief: Mohan C. Thadani.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois
    Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 
and 2, Ogle County, Illinois
    Date of amendment request: June 26, 2008.
    Description of amendment request: The proposed amendments would 
revise technical specification (TS) surveillance requirements (SRs) 
3.8.1.7, 3.8.1.12, 3.8.1.15, and 3.8.1.20, to clarify the requirements 
for the start time test performed by these SRs. The current requirement 
is to have the diesel generator (DG) within the voltage and frequency 
limits less than or equal to 10 seconds after the start signal. The 
proposed change is to have the DG above the minimum voltage and 
frequency within 10 seconds and verified to be within the voltage and 
frequency limits at steady state conditions. The change is consistent 
with TS Task Force (TSTF) Standard TS Change Traveler, TSTF-163, 
``Minimum vs. Steady State Voltage and Frequency,'' Revision 2.
    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 revises surveillance requirements to clarify 
what voltage and frequency limits are applicable during the 
transient and steady state portions of the DG start testing.
    The revised requirements do not affect the function of the DGs. 
The DGs and their associated emergency loads are accident mitigating 
features whose failure modes could not act as accident initiators or 
precursors. The proposed change does not impact the physical 
configuration or function of plant structures, systems, or 
components (SSCs) or the manner in which SSCs are operated, 
maintained, modified, or inspected. The proposed change does not 
impact the initiators or assumptions of analyzed events, nor does it 
impact the mitigation of accidents or transient events.
    The proposed change does not affect the design of the DGs, the 
operational characteristics of the DGs, the interfaces between the 
DGs and other plant systems, the function, or the reliability of the 
DGs. Thus, the DGs will be capable of performing their accident 
mitigation function and there is no impact to the radiological 
consequences of any accident analysis.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises surveillance requirements to clarify 
what voltage and frequency limits are applicable during the 
transient and steady state portions of DG testing.
    The function of the DGs is not altered by this change. 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 introduce a

[[Page 50361]]

new accident initiator, accident precursor, or malfunction 
mechanism.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises surveillance requirements to clarify 
what voltage and frequency limits are applicable during the 
transient and steady state portions of DG testing.
    The margin of safety is related to the confidence in the ability 
of the fission product barriers to perform their design functions 
during and following an accident situation. These barriers include 
the fuel cladding, the reactor coolant system, and the containment 
system. The proposed change does not directly affect these barriers, 
nor does it involve any adverse impact on the DGs which serve to 
support these barriers in the event of an accident concurrent with a 
loss of offsite power. The proposed change doesn't affect the DG's 
capabilities to provide emergency power to plant equipment that 
mitigate the consequences of the accident.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The Nuclear Regulatory Commission (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. Bradley J. Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Russell Gibbs.
FPL Energy Seabrook LLC, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire
    Date of amendment request: February 8, 2008.
    Description of amendment request: The amendment proposes to change 
the Technical Specifications to delete Surveillance Requirement 
4.6.3.1, which specifies post-maintenance testing requirements for 
containment isolation valves (CIVs).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed amendment to the technical specifications, which is 
consistent with NUREG-1431, ``Standard Technical Specifications, 
Westinghouse Plants,'' removes the surveillance requirement related 
to post-maintenance testing of containment isolation valves (CIVs). 
Surveillance requirements are not initiators of accidents; 
consequently, the proposed change does not significantly increase 
the probability of an accident previously evaluated. The proposed 
change does not alter the requirements regarding operability of 
CIVs, and appropriate testing will continue to confirm the 
operability of these valves following maintenance activities. The 
CIVs will continue to be tested in a manner and at a frequency that 
demonstrates they remain capable of performing their intended safety 
function. As a result, the proposed amendment does not significantly 
affect the consequences of an accident previously evaluated. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed change does not introduce any new accident 
scenarios, failure mechanisms, or single failures. The change does 
not add new equipment to the plant, does not modify or remove 
existing equipment, and does not significantly change the operation 
of the plant. The ability of any operable equipment to perform its 
specified safety function is unaffected by this change. Therefore, 
the proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The proposed change does not alter the initial conditions or 
results of any accident analyses. The operability requirements, 
performance, and design of the CIVs are unchanged with this proposed 
change. The CIVs will continue to meet the design bases for the 
containment isolation system as described in the Seabrook Station 
[updated final safety analysis report]. The proposed amendment will 
minimize unnecessary testing of CIVs. Therefore, the proposed change 
does not involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis, and based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. S. Ross, Florida Power & Light Company, 
P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: H. Chernoff.
GPU Nuclear, Inc., Docket No. 50-320, Three Mile Island Nuclear 
Station, Unit 2, Dauphin County, Pennsylvania
    Date of amendment request: June 11, 2008.
    Description of amendment request: The amendment application 
proposes to delete Technical Specification (TS) 6.5, ``Review and 
Audit.'' Specifically, the proposed change would delete TS 6.5.1, 
``Technical Review and Control'' and TS 6.5.3, ``Audits,'' which will 
be implemented by the current and proposed changes to the GPU Nuclear 
Post-Defueling Monitored Storage Quality Assurance Plan for Three Mile 
Island Unit 2 (PDMS QAP). The proposed change would also delete TS 
6.5.2, ``Independent Safety Review Function,'' with no replacement.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated? 
No.
    No physical changes to the TMI-2 Facility will occur as a result 
of this proposed amendment. The proposed changes will not alter the 
physical design or operational procedures associated with any plant 
structure, system, or component. As such, the change is 
administrative in nature and does not affect initiators of analyzed 
events or assumed mitigation of accidents.
    The proposed changes involve the deletion of several 
administrative requirements from the Technical Specifications (TS). 
The TS requirements involve Technical Review and Control and Audits 
that are now controlled under the TMI-2 Post Defueling Monitored 
Storage Quality Assurance Plan (PDMS QAP).
    In accordance with the guidance provided in NRC Administrative 
Letter 95-06, ``Relocation of Technical Specification Administrative 
Controls related to Quality Assurance,'' the proposed changes are an 
acceptable method for removing technical specification quality 
assurance requirements.
    Therefore, 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? No.
    The proposed changes are administrative in nature. The proposed 
changes do not alter the physical design, safety limits, or safety 
analysis assumptions associated with the operation of the plant. 
Accordingly, the changes do not introduce any new accident 
initiators, nor do they reduce or adversely affect the capabilities 
of any plant structure, system, or component to perform their safety 
function.
    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? No.

[[Page 50362]]

    The proposed changes conform to NRC regulatory guidance 
regarding the content of plant Technical Specifications. The 
guidance is presented in Administrative Letter 95-06 and NUREG-1430. 
The relocation of these administrative requirements to the PDMS QAP 
will not reduce the quality assurance commitments as accepted by the 
NRC, nor reduce administrative controls essential to the safe 
operation of the plant. Future changes to these administrative 
requirements will be performed in accordance with NRC regulation 10 
CFR 50.54(a), consistent with the guidance identified above. 
Accordingly, the replacement of TS requirements by existing proposed 
TMI-2 PDMS QAP requirements results in an equivalent level of 
regulatory control.
    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.
    NRC Branch Chief: Andrew Persinko.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, 
Alabama
    Date of amendment request: March 27, 2008.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) requirements related to control 
building envelope habitability in TS Section 3.7.3 Control Room 
Emergency Ventilation (CREV) System, and add TS Section 5.5.13, Control 
Building Envelope Habitability Program, to the Administrative Section 
of the TSs. The licensee has included conforming technical changes to 
the TS Bases. The proposed revision to the Bases also includes 
editorial and administrative changes to reflect applicable changes to 
the corresponding TS Bases, which were made to improve clarity, conform 
with the latest information and references, correct factual errors, and 
achieve more consistency with the standard TS NUREGs. The proposed 
revision to the TSs and associated Bases is similar to TS Task Force 
Traveler No. TSTF-448, Revision 3. However the references to chemical 
and smoke hazards are not included in the proposed revision to TS 
Section 3.7.3, TS Section 5.5.13 and TS Bases 3.7.3, as the CREV System 
was not designed to protect the control room envelope (CRE) occupants 
from these hazards and no toxic gas detectors are provided to initiate 
a CRE isolation.
    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 (NSHC) through incorporation by reference of the NSHC 
determination published in the Federal Register Notice dated January 
17, 2007 (73 FR 2022), which is presented below:

    1. Does the proposed Technical Specification change involve a 
significant increase in the probability or consequences of an 
accident previously evaluated?
    No. The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change revises the TS for the CRE emergency 
ventilation system, which is a mitigation system designed to 
minimize unfiltered air leakage into the CRE and to filter the CRE 
atmosphere to protect the CRE occupants in the event of accidents 
previously analyzed. An important part of the CRE emergency 
ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated. Therefore, the probability of any accident 
previously evaluated is not increased. Performing tests to verify 
the operability of the CRE boundary and implementing a program to 
assess and maintain CRE habitability ensure that the CRE emergency 
ventilation system is capable of adequately mitigating radiological 
consequences to CRE occupants during accident conditions, and that 
the CRE emergency ventilation system will perform as assumed in the 
consequence analyses of design basis accidents. Thus, the 
consequences of any accident previously evaluated are not increased. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed Technical Specification change create the 
possibility of a new or different kind of accident from any accident 
previously evaluated?
    No. The proposed change does not impact the accident analysis. 
The proposed change does not alter the required mitigation 
capability of the CRE emergency ventilation system, or its 
functioning during accident conditions as assumed in the licensing 
basis analyses of design basis accident radiological consequences to 
CRE occupants. No new or different accidents result from performing 
the new surveillance or following the new program. The proposed 
change does not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed) or a 
significant change in the methods governing normal plant operation. 
The proposed change does not alter any safety analysis assumptions 
and is consistent with current plant operating practice. Therefore, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does the proposed Technical Specification change involve a 
significant reduction in a 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.

    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Thomas H. Boyce.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas
    Date of amendment request: July 10, 2008.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) requirements consistent with 
the Technical Specification Task Force (TSTF) Standard Technical 
Specification Change Traveler, TSTF-419, Revision 0, ``Revise PTLR 
[Pressure and Temperature Limits Report] Definition and References in 
ISTS [Improved Standard TS] 5.6.6, RCS [Reactor Coolant System] PTLR.'' 
The proposed change would reference only the Topical Report (TR) number 
and title in TS 5.6.6, ``Reactor Coolant System (RCS) PRESSURE AND 
TEMPERATURE LIMITS REPORT (PTLR).'' This would allow the use of 
currently approved TRs to determine the pressure and temperature limits 
in the PTLR without having to submit an amendment to the Operating 
License. The change would not alter (1) the U.S. Nuclear Regulatory 
Commission (NRC) reviewed and approved analytical methods used to 
determine the pressure and temperature limits or Low Temperature 
Overpressure Protection (LTOP) System setpoints, or (2) the requirement 
to use NRC-approved analytical methods to determine the limits or 
setpoints.

[[Page 50363]]

    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 (NSHC), which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to reference the Topical Report number and 
title do not alter the use of the analytical methods used to 
determine the P/T [pressure/temperature] limits or LTOP System 
setpoints that have been reviewed and approved by the NRC. This 
method of referencing Topical Reports would allow the use of current 
[NRC-approved] Topical Reports to support limits in the PTLR without 
having to submit an amendment to the Operating License. 
Implementation of revisions to Topical Reports would still receive 
regulatory reviews and where required receive NRC review and 
approval. The proposed changes to add ``LTOP arming'' into TS 
5.6.6a. as a RCS pressure and temperature limit established and 
documented in the PTLR and deletion of ``and Cold Overpressure 
Mitigation System'' from TS 5.6.6b are administrative changes for 
consistency. 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. Therefore, 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 to reference the Topical Report number and 
title do not alter the use of the analytical methods used to 
determine the P/T limits or LTOP System setpoints that have been 
reviewed and approved by the NRC. This method of referencing Topical 
Reports would allow the use of current [NRC-approved] Topical 
Reports to support limits in the PTLR without having to submit an 
amendment to the Operating License. Implementation of revisions to 
Topical Reports would still receive regulatory reviews and where 
required receive NRC review and approval. The proposed changes to 
add ``LTOP arming'' into TS 5.6.6a. as a RCS pressure and 
temperature limit established and documented in the PTLR and 
deletion of ``and Cold Overpressure Mitigation System'' from TS 
5.6.6b are administrative changes for consistency. The changes do 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a change in the 
methods governing normal plant operation. In addition, the changes 
do not impose any new or different requirements or eliminate any 
existing requirements [except that NRC-approved TRs can be used 
without an amendment]. The changes do not alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumptions and current plant operating practice. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    Response: No.
    The proposed changes to reference the Topical Report number and 
title do not alter the use of the analytical methods used to 
determine the P/T limits or LTOP System setpoints that have been 
reviewed and approved by the NRC. This method of referencing Topical 
Reports would allow the use of current Topical Reports to support 
limits in the PTLR without having to submit an amendment to the 
Operating License. Implementation of revisions to Topical Reports 
would still receive regulatory reviews and where required receive 
NRC review and approval. The proposed changes to add ``LTOP arming'' 
into TS 5.6.6a. as a RCS pressure and temperature limit established 
and documented in the PTLR and deletion of ``and Cold Overpressure 
Mitigation System'' from TS 5.6.6b are administrative changes for 
consistency. The proposed changes do not alter the manner in which 
safety limits, limiting safety system settings or limiting 
conditions for operation are determined. The setpoints at which 
protective actions are initiated are not altered by the proposed 
changes. Sufficient equipment remains available to actuate upon 
demand for the purpose of mitigating an analyzed event. 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 NSHC.
    Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief (Acting): Jack N. Donohew.

Notice of Issuance of Amendments to Facility Operating Licenses

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

[[Page 50364]]

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland
    Date of application for amendments: June 29, 2007, as supplemented 
by letters dated March 12 and June 11, 2008.
    Brief description of amendments: The amendments modify Technical 
Specification (TS) 3.7.8, ``Control Room Emergency Ventilation System 
(CREVS),'' and introduce TS 5.5.17, ``Control Room Envelope 
Habitability Program,'' consistent with Technical Specification Task 
Force-448, Revision 3, ``Control Room Habitability.''
    Date of issuance: July 29, 2008.
    Effective date: These license amendments are effective as of the 
date of its issuance and shall be implemented within 60 days following 
completion of the installation and testing of the plant modifications 
described in Amendment Nos. 281 and 258 issued on August 29, 2007.
    Amendment Nos.: 287, 264.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the License and Technical Specifications.
    Date of initial notice in Federal Register: August 14, 2007 (72 FR 
45456)
    The letters dated March 12 and June 11, 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 as 
published in the Federal Register.
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated July 29, 2008.
    No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina
    Date of application for amendments: July 17, 2007, as supplemented 
by letters February 27 and July 9, 2008.
    Brief Description of amendments: The amendments revise Brunswick 
Steam Electric Plant (BSEP), Units 1 and 2 technical specifications 
(TS) to adopt NRC-approved industry Technical Specifications Task Force 
(TSTF) standard TS change traveler, TSTF-448, Revision 3, ``Control 
Room Habitability.'' This technical specification change was made 
available by the U.S. Nuclear Regulatory Commission on January 17, 2007 
(72 FR 2022) as part of the consolidated line item improvement process. 
The amendments modify the BSEP technical specification requirements 
regarding control room envelope habitability in TS 3.7.3, ``Control 
Room Emergency Ventilation (CREV) System,'' and TS Section 5.5, 
``Programs and Manuals.''
    Date of issuance: July 25, 2008.
    Effective date: Date of issuance, to be implemented within 180 
days.
    Amendment Nos. 248 and 276.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the technical specifications and add a license condition.
    Date of initial notice in Federal Register: (73 FR 29161). The 
supplemental letters dated February 27 and July 9, 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 as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 25, 2008.
    No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket Nos. 50-336 and 50-423, 
Millstone Power Station, Unit Nos. 2 and 3, New London County, 
Connecticut
    Date of application for amendment: July 2, 2007, as supplemented on 
May 5, 2008.
    Brief description of amendment: The amendment modifies the 
Technical Specification (TS) 4.0.5 to reference the American Society of 
Mechanical Engineers (ASME) Code of Operation and Maintenance of 
Nuclear Power Plants (OM Code) instead of Section XI of the ASME Boiler 
and Pressure Vessel Code. Specifically the amendment updates the 
inservice testing (IST) of ASME Code Class 1, 2, and 3 pumps and valves 
to reflect the requirements in the ASME OM Code. The amendment also 
extends the TS Surveillance Requirement 4.0.2 25 percent time extension 
to other normal and accelerated frequencies specified as 2 years or 
less in the IST program. In addition, the ISI requirement in TS 4.0.5 
is being removed and the reference to the ISI requirement is being 
updated in the snubbers' TS surveillance frequency.
    Date of issuance: July 31, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days from the date of issuance.
    Amendment No.: 304 and 241.
    Renewed Facility Operating License Nos. DPR-65 and NPF-49: 
Amendment revised the License and Technical Specifications.
    Date of initial notice in Federal Register: December 4, 2007 (72 FR 
68210) The supplement dated May 5, 2008, clarified the application, did 
not expand the scope of the application as originally noticed, and did 
not change the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 31, 2008.
    No significant hazards consideration comments received: No.
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 application for amendments: July 30, 2007, as supplement 
May 27, 2008, and June 23, 2008.
    Brief description of amendments: The amendments revised the 
Technical Specifications to allow single supply header operation of the 
buried nuclear service water (RN) system piping for up to 30 days only 
during preplanned maintenance of the buried RN system piping.
    Date of issuance: July 30, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 243 and 237.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the licenses and the technical specifications.
    Date of