[Federal Register Volume 73, Number 19 (Tuesday, January 29, 2008)]
[Notices]
[Pages 5215-5235]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E8-1300]


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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 January 3, to January 16, 2008. The last 
biweekly notice was published on January 15, 2008 (73 FR 2546).

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

[[Page 5216]]

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 
[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

[[Page 5217]]

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].

    Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 
2, and 3, Maricopa County, Arizona.
    Date of amendment request: November 14, 2007.
    Description of amendment request: The proposed amendments would 
modify the Technical Specifications (TS) by adding Limiting Condition 
for Operation (LCO) 3.0.8 on the inoperability of snubbers using the 
Consolidated Line Item Improvement Process (CLIIP). The proposed 
amendments would also make conforming changes to TS LCO 3.0.1. This 
request is consistent with NRC-approved Industry/Technical 
Specification Task Force (TSTF) Traveler No. 372, Revision 4, 
``Addition of LCO 3.0.8, Inoperability of Snubbers.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on November 24, 2004 (69 FR 68412), on possible 
license amendments adopting TSTF-372 using the NRC's 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 May 4, 2005 (70 FR 23252), which included the resolution of public 
comments on the model SE. The May 4, 2005, notice of availability 
referenced the November 24, 2004, notice. 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[s] [do] not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed change[s] [allow] a delay time for entering a 
supported system technical specification (TS) when the inoperability 
is due solely to an inoperable snubber if risk is assessed and 
managed. The postulated seismic event requiring snubbers is a low-
probability occurrence and the overall TS system safety function 
would still be available for the vast majority of anticipated 
challenges. Therefore, the probability of an accident previously 
evaluated is not significantly increased, if at all. The 
consequences of an accident while relying on allowance provided by 
proposed LCO 3.0.8 are no different than the consequences of an 
accident while relying on the TS required actions in effect without 
the allowance provided by proposed LCO 3.0.8. Therefore, the 
consequences of an accident previously evaluated are not 
significantly affected by [these] change[s]. The addition of a 
requirement to assess and manage the risk introduced by [these] 
change[s] will further minimize possible concerns. Therefore, 
[these] change[s] [do] not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Criterion 2--The proposed change[s] [do] not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed change[s] [do] not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Allowing delay times for entering [a] supported system TS when 
inoperability is due solely to inoperable snubbers, if risk is 
assessed and managed, will not introduce new failure modes or 
effects and will not, in the absence of other unrelated failures, 
lead to an accident whose consequences exceed the consequences of 
accidents previously evaluated. The addition of a requirement to 
assess and manage the risk introduced by [these] change[s] will 
further minimize possible concerns. Thus, [these] change[s] [do] not 
create the possibility of a new or different kind of accident from 
an accident previously evaluated.
    Criterion 3--The proposed change[s] [do] not involve a 
significant reduction in the margin of safety.
    The proposed change[s] [allow] a delay time for entering a 
supported system TS when the inoperability is due solely to an 
inoperable snubber, if risk is assessed and managed. The postulated 
seismic event requiring snubbers is a low-probability occurrence and 
the overall TS system safety function would still be available for 
the vast majority of anticipated challenges. The risk impact of the 
proposed TS changes was assessed following the three-tiered approach 
recommended in [NRC] RG [Regulatory Guide] 1.177. A bounding risk 
assessment was performed to justify the proposed TS changes. This 
application of LCO 3.0.8 is predicated upon the licensee's 
performance of a risk assessment and the management of plant risk[, 
which is required by the proposed TS 3.0.8]. The net change to the 
margin of safety is insignificant. Therefore, [these] change[s] [do] 
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: Michael G. Green, Senior Regulatory Counsel, 
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695, 
Phoenix, Arizona 85072-2034
    NRC Branch Chief: Thomas G. Hiltz.

    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 amendments request: September 29, 2007, as supplemented on 
December 7, 2007.
    Description of amendments request: The amendment would revise the 
Technical Specification (TS) Administrative Controls section pertaining 
to the American Society of Mechanical Engineers (ASME) Boiler and 
Pressure Vessel Code (Code) requirements for inservice testing of pumps 
and valves. The changes are based on Technical Specification Task Force 
(TSTF) Traveler TSTF-479, ``Changes to Reflect Revision of 10 CFR 
50.55a,'' as modified by TSTF-497, ``Limit Inservice Testing Program SR 
3.0.2 Application to Frequencies of 2 Years or Less.''
    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 TS 5.5.6, ``Inservice Testing 
Program,'' for consistency with the requirements of 10 CFR 
50.55a(f)(4) regarding the inservice testing of pumps and valves 
which are classified as ASME Code Class 1, Class 2, and Class 3. 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

[[Page 5218]]

events. The proposed change does not involve the addition or removal 
of any equipment, or any design changes to the facility. Therefore, 
this proposed change does not involve an 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 TS 5.5.6, ``Inservice Testing 
Program,'' for consistency with the requirements of 10 CFR 
50.55a(f)(4) regarding the inservice testing of pumps and valves 
which are classified as ASME Code Class 1, Class 2, and Class 3. 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 involve a 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 offsite and there is no increase in individual or 
cumulative occupational exposure. Therefore, the proposed change 
does not create the possibility of an accident of a different kind 
than previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises TS 5.5.6, ``Inservice Testing 
Program,'' for consistency with the requirements of 10 CFR 
50.55a(f)(4) regarding the inservice testing of pumps and valves 
which are classified as ASME Code Class 1, Class 2, and Class 3. 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 the methods governing normal plant operation. 
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 function of the affected pumps and valves will be 
maintained. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.

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

    Duke Power Company LLC, et al., Docket Nos. 50-413 and 50-414, 
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina.
    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: November 12, 2007.
    Description of amendment request: The amendments would approve 
proposed changes to the licensing bases and final updated safety 
analysis report for both the Catawba Nuclear Power Station, Units 1 and 
2, and the McGuire Nuclear Power Station, Units 1 and 2, concerning 
Revision 1 to DPC-NE-1005-P, Nuclear Design Methodology Using CASMO-4/
SlMULATE-3 MOX.
    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 UFSAR change to allow the use of the CASMO-4/
SIMULATE-3 MOX reload design software to analyze reactor cores with 
fuel containing gadolinia does not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The CASMO-4 and SIMULATE-3 MOX codes are used to perform reactivity 
and power distribution calculations to develop power distribution 
limits and provide confirmation of reactivity and power distribution 
input assumptions used in the evaluation of UFSAR Chapter 15 
accidents. The SIMULATE-3 MOX code is also used to confirm the 
acceptability of thermal limits at post accident conditions. Since 
the CASMO-4/SIMULATE-3 MOX software is not used in the operation of 
any plant equipment, the probability of an accident previously 
evaluated in the UFSAR is not increased.
    The benchmark calculations performed in Revision 1 to DPC-NE-
1005-P verified the acceptability of the CASMO-4/SIMULATE-3 MOX 
codes for performing reload design calculations for reactor cores 
containing gadolinia. These calculations confirmed the accuracy of 
the codes and developed a methodology for calculating power 
distribution uncertainties for use in reload design calculations. 
The use of power distribution uncertainties applicable to gadolinia 
core designs in conjunction with predicted peaking factors ensures 
that thermal accident acceptance criteria are satisfied.
    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.
    The extension of the reload design software to perform reload 
design calculations for reactor cores containing gadolinia will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated. The CASMO-4/SIMULATE-3 MOX 
software is not installed in any plant equipment and therefore the 
software is incapable of initiating an equipment malfunction that 
would result in a new or different type of accident from any 
previously evaluated. The evaluation of UFSAR accidents and the 
associated acceptance criteria for these accidents remains 
unchanged.
    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 extension of the CASMO-4/SIMULATE-3 MOX reload design 
software to perform reload design calculations for reactor cores 
containing gadolinia will not involve a significant reduction in a 
margin of safety.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their design function during 
and following an accident. These barriers include the fuel cladding, 
the reactor coolant system and the containment system. The reload 
design process assures the acceptability of thermal limits under 
normal, transient, and accident conditions. The CASMO-4/SIMULATE-3 
MOX reload design software was qualified for the analysis of reactor 
cores containing gadolinia in Revision 1 to DPC-NE-1005-P and a 
methodology for developing appropriate power distribution 
uncertainties for application in reload design analyses was 
developed. The use of these uncertainties for analysis of reload 
cores with gadolinia ensures that design and safety limits are 
satisfied such that the fission product barriers perform their 
design function.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

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

    Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim 
Nuclear Power Station, Plymouth County, Massachusetts.
    Date of amendment request: November 29, 2007.

[[Page 5219]]

    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) requirements related to control 
room envelope habitability in TS 3.7.B.2 ``Control Room High Efficiency 
Air Filtration System (CRHEAFS)'' and TS Section 5.5 ``Administrative 
Controls--Programs and Manuals'' consistent with Technical 
Specification Task Force (TSTF)-448, Revision 3.
    The availability of TS improvement was announced in the Federal 
Register on January 17, 2007 (72 FR 2022), including a model safety 
evaluation and model no significant hazards consideration (NSHC) 
determination, as part of the consolidated line item improvement 
process.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration 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 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.

    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., 400 Hamilton Avenue, White 
Plains, NY 10601.

    NRC Branch Chief: Mark G. Kowal.
    Entergy Operations Inc., Docket No. 50-382, Waterford Steam 
Electric Station, Unit 3, St. Charles Parish, Louisiana.
    Date of amendment request: January 2, 2008.
    Description of amendment request: The proposed amendment revises 
the action requirements for certain inoperable containment isolation 
valves in Technical Specification 3/4.6.3, ``Containment Isolation 
Valves,'' to increase the allowed outage time from 4 hours to 72 hours.
    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 modifies existing action requirements for 
inoperable containment isolation valves. The condition evaluated, 
the Action requirements and the associated allowed outage times do 
not impact initiating conditions for any accident previously 
evaluated. Containment integrity will continue to be maintained by 
the closed system when the proposed actions are implemented. The new 
action requirement provides appropriate remedial actions to be taken 
in response to an inoperable containment isolation valve in a closed 
system while minimizing the risk associated with continued 
operation. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve any changes to plant 
equipment or system design functions. The specification for 
containment isolation valves provides controls for maintaining the 
containment pressure boundary. The new action requirement and 
surveillance requirement are sufficient to ensure that the 
containment isolation function is maintained. No new accident 
initiators are introduced 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. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The new action requirement does not involve a significant 
reduction in the margin of safety. The proposed action for an 
inoperable containment isolation valve in a closed system minimizes 
the risk of continued operation under the specified conditions, 
considering the reliability of the closed system (i.e., passive 
barrier), a reasonable time for repairs or replacement of the 
isolation feature, and that 72 hours is typically provided for 
losing one train of redundancy throughout the NUREGs, and the low 
probability of a design basis accident occurring during the allowed 
outage time period (reference TSTF [Technical Specifications Task 
Force ]-30). Should the penetration required to be isolated, 
Technical Specification 3.6.1.1 provides the surveillance 
requirement to verify at least once every 31 days that the affected 
penetration flow path is isolated if the penetration is not capable 
of being closed by operable containment isolation valves. Therefore, 
the proposed change does not involve a significant reduction in a 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears

[[Page 5220]]

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: Terence A. Burke, Associate General 
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.

    NRC Branch Chief: Thomas G. Hiltz.

    FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold 
Energy Center, Linn County, Iowa.
    Date of amendment request: November 14, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 5.5.12, ``Primary Containment 
Leakage Rate Testing Program,'' to allow use of the requirements of 
American Society of Mechanical Engineers (ASME) Boiler and Pressure 
Vessel Code (the Code), Section XI, Subsection IWE for visual 
examination of the steel containment. This license amendment request is 
consistent with NRC approved Technical Specification Task Force (TSTF) 
Traveler number TSTF-343, Revision 1, ``Containment Structural 
Integrity.''
    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 the TS administrative controls 
programs for consistency with the requirements of 10 CFR 
50.55a(g)(4) for components classified as Code Class MC. The 
proposed change affects the frequency of visual examinations that 
will be performed for the containment. The frequency of visual 
examinations of the containment has no relationship to or adverse 
impact on the probability of any of the initiating events assumed in 
the accident analyses. The proposed change would allow visual 
examinations that are performed pursuant to NRC approved ASME 
Section Xl Code requirements (except where relief has been granted 
by the NRC) to meet the intent of visual examinations required by 
Regulatory Guide 1.163, without requiring additional visual 
examinations pursuant to the Regulatory Guide. The intent of early 
detection of deterioration will continue to be met by the more 
rigorous requirements of the Code required visual examinations. As 
such, the safety function of the containment as a fission product 
barrier is maintained. The proposed change does not impact any 
accident initiators or analyzed events or assumed mitigation of 
accident or transient events. It does not involve the addition or 
removal of any equipment, or any design changes to the facility.
    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 the TS administrative controls 
programs for consistency with the requirements of 10 CFR 
50.55a(g)(4) for components classified as Code Class MC. The change 
affects the frequency of visual examinations that will be performed 
for the containment. 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 safety function of the containment as a 
fission product barrier is maintained. 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, 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 the Improved Standard Technical 
Specification Administrative Controls program requirements for 
consistency with the requirements of 10 CFR 50, paragraph 55a(g)(4) 
for components classified as Code Class MC. The change affects the 
frequency of visual examinations that will be performed for the 
containment. The safety function of the containment as a fission 
product barrier will be maintained.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Marjan Mashhadi, Florida Power & Light 
Company, 801 Pennsylvania Avenue, Suite 220, Washington, DC 20004.
    NRC Acting Branch Chief: Cliff Munson.

    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 amendment request: December 29, 2007.
    Description of amendment request: The amendment would revise the 
Point Beach Nuclear Plant (PBNP) Units 1 and 2 Technical Specifications 
(TS) requirement for the completion time (CT) of TS 3.7.5.C. This 
revision would allow two separate one-time extensions of the CT for TS 
3.7.5.C from seven days to 16 days; one extension for each of the 
train-specific motor-driven auxiliary feedwater (MDAFW) pumps.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The results of the Technical Evaluation (Section 3.0) [of the 
application] demonstrate that, with the requested change, the 
increase in the probability of an accident previously evaluated fall 
within the guidance in RG 1.177 [Regulatory Guide 1.177, An Approach 
for Plant-Specific, Risk-Informed Decisionmaking: Technical 
Specifications]. Therefore, the risk impact of the proposed CT 
extensions is small.
    The ability of the AFW [auxiliary feedwater] system to deliver 
the required flow to mitigate design basis accidents is maintained. 
The ability to isolate AFW flow to or steam supply from the affected 
steam generator during design basis accidents is unaffected by this 
requested change. The applicable radiological analyses remain 
bounding.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The requested change to extend the CT of TS 3.7.5.C from 7 days 
to 16 days to replace a MDAFW pump and motor will not create the 
possibility of a new or different kind of accident. Two unit-
specific TDAFW pump systems and one MDAFW pump system will remain 
OPERABLE and capable of performing the AFW system function. Prior to 
taking the MDAFW pump out of service for pump and motor replacement, 
both unit-specific turbine-driven auxiliary feedwater (TDAFW) pump 
systems and the other MDAFW pump system will be demonstrated 
OPERABLE. To ensure that the redundant AFW pump systems remain 
OPERABLE, risk management actions will be taken that include 
protecting the redundant operable AFW pump systems.
    To manage the fire risk due to a MDAFW pump being inoperable, 
compensatory measures will be initiated to monitor and ensure that 
combustible loading, work activities, and other activities that 
could

[[Page 5221]]

increase the likelihood of a fire are minimized. An initial baseline 
and weekly thermography of potential fire initiators will be 
performed to detect degrading operating equipment. No new failure 
will be created.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The ability of the AFW system to deliver the required flow to 
mitigate design basis accidents will be maintained. The ability to 
isolate AFW flow to or steam supply from the affected steam 
generator during design basis accidents is unaffected by this 
requested change. The applicable radiological analyses remain 
bounding. No significant reduction in a margin of safety will occur.

    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: Antonio Fernandez, Esquire, Senior Attorney, 
FPL Energy Point Beach, LLC, P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Acting Branch Chief: Cliff Munson.

    FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, 
Perry Nuclear Power Plant, Unit No. 1, Lake County, Ohio.
    Date of amendment request: September 5, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TS) 3.6.1, 3.6.4, and 3.6.5 to relax 
the position verification requirements for primary containment 
isolation devices, secondary containment isolation devices, and drywell 
isolation devices that are locked, sealed, or otherwise secured. These 
changes are based on TS Task Force (TSTF) change traveler TSTF-45 
(Revision 2) and TSTF-269 (Revision 2), which have been approved 
generically for the Boiling Water Reactor (BWR) Standard Technical 
Specifications, NUREG-1434 (BWR/6).
    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 change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change will revise the position verification 
requirements for manual containment and drywell isolation devices 
that are locked, sealed, or otherwise secured in the closed 
position. Revising the verification requirements will not introduce 
any physical changes or result in the equipment being operated in a 
new or different manner. All systems, structures, and components 
previously required for mitigation of a transient remain capable of 
performing their designed functions. Furthermore, although the 
proposed change would revise the position verification requirements, 
no physical change is being made to the assumed position of the 
valves for accident analysis. Therefore, this change does not 
involve a significant increase to the probability or consequences of 
an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    No new accident scenarios or failure mechanisms are introduced 
as a result of this proposed change. The proposed amendment would 
revise the position verification requirements but not alter any 
valve positions. With no changes to the plant lineup, no new or 
different accidents are possible. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed amendment revises the position verification 
requirements for manual containment and drywell isolation valves 
that are locked, sealed, or otherwise secured in the closed 
position. The revised position verification requirements have no 
adverse effects on any safety-related system or component and do not 
challenge the performance or integrity of any safety-related system. 
Additionally, position verification does not alter the actual valve 
positions, introduce any physical changes, or reduce the ability of 
the valve to control leakage rates during design basis radiological 
accidents. Therefore, the proposed amendment does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy 
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Russell Gibbs.

    FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, 
Perry Nuclear Power Plant, Unit No. 1, Lake County, Ohio.
    Date of amendment request: September 18, 2007.
    Description of amendment request: The proposed license amendment 
would modify technical specification (TS) requirements related to 
control room envelope habitability in accordance with Technical 
Specification Task Force (TSTF) Change Traveler TSTF-448, Revision 3, 
per the consolidated line item improvement process (CLIIP).
    The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice 
of opportunity for comment in the Federal Register on October 17, 2006 
(71 FR 61075-61084), on possible amendments concerning the CLIIP, 
including a model safety evaluation and a model no significant hazards 
consideration 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), 
as part of the CLIIP. In its application dated September 18, 2007, the 
licensee affirmed the applicability of the following determination.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    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

[[Page 5222]]

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 an 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 the analysis adopted by the licensee 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 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy 
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Russell Gibbs.

    Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit 3 Nuclear Generating Plant (CR-3), Citrus County, Florida.
    Date of amendment request: July 31, 2007.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TSs) to impose more restrictive 
voltage and frequency limits during surveillance testing of the 
emergency diesel generators.
    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 not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The LAR [license amendment request] proposes to provide more 
restrictive steady state voltage and frequency limits for the 
Emergency Diesel Generators (EDGs). The voltage band is going from a 
range of greater than or equal to 3933 VAC [volts, alternating 
current] but less than or equal to 4400 VAC to greater than or equal 
to 4077 VAC but less than or equal to 4243 VAC. The proposed limits 
are plus or minus 2% around the nominal safety-related bus voltage 
of 4160 VAC. The Frequency Limits are going from a 2% tolerance band 
to a 1% tolerance band around the nominal frequency of 60 Hz (59.4 
to 60.6 Hz), for fast starts and emergency starts of the EDGs. These 
acceptance limits are specifically for steady state conditions 
following a fast start of the EDGs.
    Slow starts will also have a more restrictive frequency band, 
but it will be slightly larger than for fast starts. The reason for 
this difference is based on the speed control circuitry for the EDG. 
The EDG has an electro-mechanical component in the slow start 
circuitry that is not present in the fast start circuitry. The 
proposed slow start limits are plus or minus 1.5% (59.1 Hz to 60.9 
Hz). The voltage limits for a slow start will be the same as for a 
fast start.
    The EDGs are a safety related system that functions to mitigate 
the impact of an accident with a concurrent loss of offsite power. A 
loss of offsite power is typically a significant contributor to 
postulated plant risk and, as such, onsite AC generators have to be 
maintained available and reliable in the event of a loss of offsite 
power event. The EDGs are not initiators for any analyzed accident, 
therefore; the probability for an accident that was previously 
evaluated is not increased by this change. The revised, voltage and 
frequency limits will ensure the EDGs will remain capable of 
performing their design function.
    The consequences of an accident refer to the impact on both the 
plant personnel and the public from any radiological release 
associated with the accident. The EDG supports equipment that is 
supposed to preclude any radiological release. More restrictive 
voltage and frequency limits for the output of the EDG restores 
design margin, and provides assurance that the equipment supplied by 
the EDG will operate correctly and within the assumed timeframe to 
perform their mitigating functions.
    Until the proposed CR-3 ITS [improved TS] EDG voltage and 
frequency limits are approved, administratively controlled limits 
have been established in accordance with Administrative Letter 98-10 
to ensure all EDG mitigation functions will be performed in the 
event of a loss of offsite power. These administrative limits have 
been determined as acceptable and have been incorporated into the 
Surveillance test procedures under the provisions of 10 CFR 50.59. 
Periodic testing has been performed with acceptable results. Since 
EDGs are mitigating components and are not initiators for any 
analyzed accident, no increased probability of an accident can 
occur. Since administrative limits will ensure the EDGs will perform 
as designed, consequences will not be significantly affected.
    (2) Does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    Administrative voltage limits were established using verified 
design calculations and the guidance of NRC Administrative Letter 
98-10. These administrative limits will ensure the EDGs will perform 
as designed. No new configuration is established by this change. The 
administrative limits for the EDG frequency were determined to be 
sufficient to account for measurement and other uncertainties.
    The proposed amendment will place the administrative limits into 
the CR-3 ITS. The more restrictive voltage and frequency limits will 
provide additional assurance that the EDG can provide the necessary 
power to supply the required safety-related loads during an analyzed 
accident.
    The proposed voltage and frequency ITS limits restore the EDG 
capability to those analyzed. No new configuration is established. 
Therefore, no new or different kind of accident from any previously 
evaluated can be created.
    (3) Does not involve a significant reduction in a margin of 
safety.
    The LAR proposes to provide more restrictive steady state 
voltage and frequency limits for the EDGs. The change in the 
acceptance criteria for specific surveillance testing provides 
assurance that the EDGs will be capable of performing their design 
function. Previous test history has shown that the new limits are 
well within the capability of the EDGs and are repeatable. The 
frequency ``as left'' setting will be adjusted such that it remains 
within a tight band and this assures the ``as found'' setting will 
be in the acceptable band. The requirement to adjust the as left 
frequency setting as well as the limitations on the frequency as 
left tolerance have been proceduralized to assure the requirement is 
satisfied.
    The proposed ITS limits on voltage and frequency will assure the 
EDG will be able to perform all design function assumed in the 
accident analyses. Administrative limits are in place to ensure 
these parameters remain within analyzed limits. As such, the 
proposed change does not involve a significant reduction in a margin 
of safety.

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

[[Page 5223]]

    NRC Branch Chief: Thomas H. Boyce.

    Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit 3 Nuclear Generating Plant, Citrus County, Florida.
    Date of amendment request: October 25, 2007.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TSs) by relocating references to 
specific American Society for Testing and Materials (ASTM) standards 
for fuel oil testing to licensee-controlled documents. The proposed 
change is based on TS Task Force (TSTF) Traveler TSTF-374, ``Revision 
to TS 5.5.13 and Associated Bases for Diesel Fuel Oil,'' and was 
submitted using the Consolidated Line Item Improvement Process (CLIIP). 
Some changes included in TSTF-374, such as the addition of alternate 
criteria to the ``clear and bright'' acceptance test for new fuel oil, 
were not included in the application because they are already part of 
the licensing basis for Crystal River Unit 3.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on February 22, 2006 (71 FR 9179), on possible 
amendments to revise plant-specific TSs in accordance with TSTF-374, 
including a model safety evaluation and model No Significant Hazards 
Consideration (NSHC) Determination, using the CLIIP. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on April 21, 2006 (71 FR 20735). The licensee affirmed the 
applicability of the model NSHC determination in its application dated 
October 25, 2007.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of any accident previously 
evaluated?
    Response: No.
    The proposed changes relocate the specific ASTM standard 
references from the Administrative Controls Section of TS to a 
licensee-controlled document. Requirements to perform testing in 
accordance with applicable ASTM standards are retained in the TS as 
are requirements to perform surveillances of both new and stored 
diesel fuel oil. Future changes to the licensee-controlled document 
will be evaluated pursuant to the requirements of 10 CFR 50.59, 
``Changes, tests and experiments,'' to ensure that such changes do 
not result in more than a minimal increase in the probability or 
consequences of an accident previously evaluated. In addition, the 
``clear and bright'' test used to establish the acceptability of new 
fuel oil for use prior to addition to storage tanks has been 
expanded to recognize more rigorous testing of water and sediment 
content. Relocating the specific ASTM standard references from the 
TS to a licensee-controlled document and allowing a water and 
sediment content test to be performed to establish the acceptability 
of new fuel oil will not affect nor degrade the ability of the 
emergency diesel generators (DGs) to perform their specified safety 
function. Fuel oil quality will continue to meet ASTM requirements.
    The proposed changes do 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 do not adversely 
affect the ability of structures, systems, and components (SSCs) to 
perform their intended safety 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 any accident previously evaluated. 
Further, the proposed changes do not increase the types and amounts 
of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures. Therefore, the changes do not involve a 
significant increase in the probability or consequences of any 
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 changes relocate the specific ASTM standard 
references from the Administrative Controls Section of TS to a 
licensee-controlled document. In addition, the ``clear and bright'' 
test used to establish the acceptability of new fuel oil for use 
prior to addition to storage tanks has been expanded to allow a 
water and sediment content test to be performed to establish the 
acceptability of new fuel oil. 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. The requirements retained in the TS continue to require 
testing of the diesel fuel oil to ensure the proper functioning of 
the DGs.
    Therefore, the changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes relocate the specific ASTM standard 
references from the Administrative Controls Section of TS to a 
licensee-controlled document. Instituting the proposed changes will 
continue to ensure the use of applicable ASTM standards to evaluate 
the quality of both new and stored fuel oil designated for use in 
the emergency DGs. Changes to the licensee-controlled document are 
performed in accordance with the provisions of 10 CFR 50.59. This 
approach provides an effective level of regulatory control and 
ensures that diesel fuel oil testing is conducted such that there is 
no significant reduction in a margin of safety.
    The ``clear and bright'' test used to establish the 
acceptability of new fuel oil for use prior to addition to storage 
tanks has been expanded to allow a water and sediment content test 
to be performed to establish the acceptability of new fuel oil. The 
margin of safety provided by the DGs is unaffected by the proposed 
changes since there continue to be TS requirements to ensure fuel 
oil is of the appropriate quality for emergency DG use. The proposed 
changes provide the flexibility needed to improve fuel oil sampling 
and analysis methodologies while maintaining sufficient controls to 
preserve the current margins 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: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Branch Chief: Thomas H. Boyce.

    Indiana Michigan Power Company (I&M), Docket No. 50-315, Donald C. 
Cook Nuclear Plant, Unit 1 (DCCNP-1), Berrien County, Michigan.
    Date of amendment request: December 27, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TS) Section 3.4.1, ``RCS [Reactor 
Coolant System] Pressure, Temperature, and Flow Departure from Nucleate 
Boiling (DNB) Limits,'' to increase the minimum reactor coolant system 
(RCS) flow rate from 341,100 to 354,000 gallons per minute. The new 
analysis is performed using the NRC-approved methodology set forth in 
Westinghouse Topical Report WCAP-16009-P-A, ``Realistic Large-Break 
LOCA [Loss-of-Coolant Accident] Evaluation Methodology Using the 
Automated Statistical Treatment of Uncertainty Method (ASTRUM)''; the 
licensee proposed to endorse this methodology by a revision of Section 
5.6.5, ``Core Operating Limits Report (COLR).''
    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 has performed its own analysis, which is 
presented below:


[[Page 5224]]


    1. Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    No. The proposed amendment would revise the subject TS sections 
to endorse a change in licensing basis, which involves use of an 
NRC-approved large break LOCA analysis methodology as set forth in 
Topical Report WCAP-16009-P-A, and to increase the required RCS flow 
rate. This change in licensing basis does not result in modification 
of plant design or method of operation that could change initiators 
of previously analyzed accidents. Further, this change does not 
modify the design performance of structures, systems, and 
components, relied upon to mitigate previously analyzed accidents. 
Thus, DCCNP-1 will continue to operate as before, resulting in no 
significant increase of the probability of occurrence of any 
accident previously analyzed, and no significant increase in 
consequences should any of the previously analyzed accidents occur.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The proposed TS and licensing basis changes would support a 
modification permitting four-loop injection of the low-head safety 
injection system, an accident-mitigating system. Accident-mitigating 
systems are not identified as accident initiators in previously 
analyzed accidents. There is no modification of other structure, 
system, or component, and no change to reactor protection system or 
engineered safeguards feature actuating system setpoints. 
Accordingly, no new transient or accident event would result due to 
modification of the low-head safety injection system. In addition, 
employing the ASTRUM methodology in an analysis does not create any 
new failure modes that could lead to a different kind of accident. 
Therefore, the proposed amendment does not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. Margins of safety are established in the design of 
components, the configuration of components to meet certain 
performance parameters, and in the models and associated assumptions 
used to analysis the system's performance. The subject system will 
continue to perform the same accident-mitigating function to the 
same level of reliability as defined in the DCCNP-1 Updated Safety 
Analysis Report. The analysis model to be endorsed by the revised TS 
is an NRC-approved methodology which will continue to show that 
DCCNP-1 operates with the same margin of safety. Therefore, the 
proposed amendment 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 its own analysis, it appears that the 
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC 
staff proposes to determine that the proposed amendment involves no 
significant hazards consideration.
    Attorney for licensee: Kimberly A. Harshaw, Esquire, One Cook 
Place, Bridgman, MI 49106.
    NRC Acting Branch Chief: Cliff Munson.

    Indiana Michigan Power Company (I&M), Docket Nos. 50-315 and 50-
316, Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, 
Michigan.
    Date of amendment request: December 27, 2007.
    Description of amendment request: The proposed amendment would 
modify Technical Specifications (TS) requirements related to control 
room envelope habitability in TS Section 3.7.10, ``Control Room 
Emergency Ventilation (CREV) System,'' and Section 5.5, ``Programs and 
Manuals.'' The proposed changes are consistent with Technical 
Specification Task Force (TSTF) Standard Technical Specifications (STS) 
change TSTF-448, Revision 3.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) Part 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration (NSHC) by 
referencing the NRC staff's model NSHC analysis published on January 
17, 2007 (72 FR 2022). The NRC staff's model NSHC analysis is 
reproduced 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 the licensee's referenced analysis, and 
has found that the three standards of 10 CFR 50.92(c) are satisfied. 
Therefore, the NRC staff proposes to determine that the proposed 
amendment involves no significant hazards consideration.
    Attorney for licensee: Kimberly A. Harshaw, Esquire, One Cook 
Place, Bridgman, MI 49106.
    NRC Acting Branch Chief: Cliff Munson.

    Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska.
    Date of amendment request: November 19, 2007.
    Description of amendment request: The proposed changes to the 
license and Technical Specifications reflect an increase in the rated 
thermal power from 2381 to 2419 megawatts thermal

[[Page 5225]]

(1.62 percent increase) based upon increased feedwater flow measurement 
accuracy to be achieved by using high accuracy ultrasonic flow 
measurement instrumentation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The comprehensive analytical efforts performed to support the 
proposed uprate conditions included a review and evaluation of 
components and systems that could be affected by this change. 
Evaluation of accident analyses confirmed the effects of the 
proposed uprate are bounded by the current dose analyses. All 
systems will function as designed, and all performance requirements 
for these systems have been evaluated and found acceptable.
    The primary loop components (reactor vessel, reactor intemals, 
control rod drive housings, piping and supports, recirculation 
pumps, etc.) continue to comply with their applicable structural 
limits and will continue to perform their intended design functions. 
Thus, there is no increase in the probability of a structural 
failure of these components.
    All of the Nuclear Steam Supply Systems (NSSS) will still 
perform their intended design functions during normal and accident 
conditions. The balance of plant (BOP) systems and components 
continue to meet their applicable structural limits and will 
continue to perform their intended design functions. Thus, there is 
no increase in the probability of a structural failure of these 
components. All of the NSSS/BOP interface systems will continue to 
perform their intended design functions. The safety relief valves 
and containment isolation valves meet design sizing requirements at 
the uprated power level.
    Because the integrity of the plant will not be affected by 
operation at the uprated condition, NPPD [Nebraska Public Power 
District] has concluded that all structures, systems, and components 
required to mitigate a transient remain capable of fulfilling their 
intended functions. The reduced uncertainty in the flow input to the 
core thermal power uncertainty measurement allows a majority of the 
current safety analyses to be used, with small changes to the core 
operating limits, to support operation at a core power of 2419 MWt 
[mmegawatts thermal]. Other analyses performed at a nominal power 
level have either been evaluated or re-performed for the 1.62% 
increased power level. The results demonstrate that acceptance 
criteria of the applicable analyses continues to be met at the 1.62% 
uprate conditions. As such, all CNS [Cooper Nuclear Station] USAR 
[updated safety analysis report] Chapter 14 accident analyses 
continue to demonstrate compliance with the relevant event 
acceptance criteria. The analyses performed to assess the effects of 
mass and energy releases remain valid. The source terms used to 
assess radiological consequences have been reviewed and determined 
to bound operation at the 1.62% uprated condition.
    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.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed changes. 
All systems, structures, and components previously required for the 
mitigation of a transient remain capable of fulfilling their 
intended design functions. The proposed changes have no adverse 
effects on any safety-related system or component and do not 
challenge the performance or integrity of any safety-related system. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Operation at the uprated power condition does not involve a 
significant reduction in a margin of safety. Analyses of the primary 
fission product barriers have concluded that relevant design 
criteria remain satisfied, both from the standpoint of the integrity 
of the primary fission product barrier, and from the standpoint of 
compliance with the required acceptance criteria. As appropriate, 
all evaluations have been performed using methods that have either 
been reviewed or approved by the Nuclear Regulatory Commission, or 
that are in compliance with regulatory review guidance and 
standards. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Branch Chief: Thomas G. Hiltz.

    Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket Nos. 50-220 
and 50-410, Nine Mile Point Nuclear Station Unit Nos. 1 (NMP1) and 2 
(NMP2), Oswego County, New York.
    Date of amendment request: December 20, 2007.
    Description of amendment request: The proposed amendment would 
revise NMPI Technical Specification (TS) 6.3, ``Unit Staff 
Qualifications,'' and NMP2 TS 5.3, ``Unit Staff Qualifications,'' to 
update requirements that have been superseded due to the accreditation 
of the NMPNS licensed operator training program and due to promulgation 
of the revised Title 10 of the Code of Federal Regulations (10 CFR), 
Part 55, ``Operators' Licenses,'' which became effective on May 26, 
1987 (52 FR 9453). Additionally, the proposed amendment would revise 
NMP1 TS 6.3 by eliminating the qualification requirement exceptions 
listed for the position of Manager Operations, and previously approved 
by the Nuclear Regulatory Commission (NRC) staff. The position of 
Manager Operations will meet the minimum qualification requirements as 
required in American National Standard Institute (ANSI) Standard NI8.1-
1971, ``American National Standard for Selection and Training of 
Nuclear Power Plant Personnel.''
    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 Technical Specifications change to the licensed 
operator qualification requirements is an administrative change to 
revise the present operator qualification program to the more 
current National Academy for Nuclear Training (NANT) guidelines for 
initial training and qualification of licensed operators. The change 
conforms to the current requirements of 10 CFR [Part] 55, 
``Operators' Licenses.''
    Although the licensed operator qualification and training 
program may have an indirect impact on accidents previously 
evaluated, the NRC considered this impact during the rulemaking 
process, and by promulgation of the revised 10 CFR [Part] 55 rule, 
concluded that this impact remains acceptable as long as the 
licensed operator training program is accredited and is based on a 
systems approach to training. NMPNS's licensed operator training 
program is accredited by the Institute of Nuclear Power Operation 
(INPO) and is based on a systems approach to training.
    The proposed Technical Specifications amendment to re-establish 
a previously revised commitment to administer the standards of ANSI 
N18.1-1971 for the position of Manager Operations is also an 
administrative change. The change does not alter the manner in which 
the plant systems are operated.
    Therefore, the proposed changes do not involve a significant 
increase in probability

[[Page 5226]]

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 TS change to clarify the current requirements for 
licensed operator qualification and the licensed operator training 
program are administrative changes, and conform to the requirements 
of 10 CFR [Part] 55. The TS requirements for all other unit staff 
qualifications remain unchanged.
    Although licensed operator qualification and training may have 
an indirect impact on the possibility of a new or different kind of 
accident from any accident previously evaluated, the NRC considered 
this impact during the rule making process, and by promulgation of 
the revised rule, concluded that this impact remains acceptable as 
long as the licensed operator training program is accredited and 
based on a systems approach to training. As previously noted, NMPNS 
licensed operator training program is accredited by INPO and is 
based on a systems approach to training.
    The proposed TS change to delete a previously approved exception 
to the qualification requirements contained in ANSI N18.1-1971 for 
the position of Manager Operations is also an administrative change.
    None of the precursors of previously evaluated accidents are 
affected by these changes, and no new failure modes 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 TS change to update the current requirements 
applicable to licensed operator qualification and the licensed 
operator training program are administrative changes. The change is 
consistent with the requirements of 10 CFR [Part] 55. The TS 
qualification requirements for all other unit staff remain 
unchanged.
    Licensed operator qualification and training can have an 
indirect impact on a margin of safety. However, the NRC considered 
this impact during the rule making process, and by promulgation of 
the revised 10 CFR [Part] 55, determined that this impact remains 
acceptable when licensees maintain a licensed operator training 
program that is accredited and based on a systems approach to 
training. As previously noted, the NMPNS licensed operator training 
program is accredited by INPO and is based on a systems approach to 
training.
    The NRC has concluded, as stated in NUREG-1262, ``Answers to 
Questions at Public Meetings Regarding Implementation of Title 10, 
Code of Federal regulations, Part 55 on Operators' Licenses,'' that 
the standards and guidelines applied by INPO in their training 
accreditation program are equivalent to those put forth or endorsed 
by the NRC. As a result, maintaining an INPO accredited, systems 
approach based licensed operator training program is equivalent to 
maintaining an NRC approved licensed operator training program which 
conforms with applicable NRC Regulatory Guides or NRC endorsed 
industry standards. The margin of safety is maintained by virtue of 
maintaining an INPO accredited licensed operator training program.
    In addition, the NRC has published NRC Regulatory Issue Summary 
2001-01, ``Eligibility of Operator License Applicants,'' dated 
January 18, 2001, ``to familiarize addressees with the NRC's current 
guidelines for the qualification and training of reactor operator 
and senior operator license applicants.'' This document again 
acknowledges that the INPO National Academy for Nuclear Training 
(NANT) guidelines for education and experience, outline acceptable 
methods for implementing the NRC 's regulations in this area.
    The proposed Technical Specifications change to re-establish a 
previously revised plant commitment to administer the standards of 
ANSI N18.1-1971 for the position of Manager Operations is an 
administrative change.
    The proposed changes do not involve a physical modification of 
the plant or involve any changes to the methods in which plant 
systems are operated. The changes do not, in themselves, adversely 
affect any physical barrier which could contribute to the release of 
radiation to plant personnel or to the public.
    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: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006.
    NRC Branch Chief: Mark G. Kowal.

    Nuclear Management Company, LLC, Docket No. 50-282, Prairie Island 
Nuclear Generating Plant, (PINGP) Unit 1, Goodhue County, Minnesota.
    Date of amendment request: August 16, 2007.
    Description of amendment request: The proposed amendment would 
require PINGP monthly Emergency Diesel Generators (EDGS) load test (SR 
3.8.1.3) to be performed at or above 90 percent of the diesel 
generator's continuous power rating. This fulfills the commitment made 
in the supplement to license amendment request for extension of 
Technical Specification (TS) 3.8.1, ``AC Sources-Operating,'' Emergency 
Diesel Generator Completion Time (TAC Nos. MC9001 and MC9002), dated 
May 10, 2007, Agencywide Documents Access and Management System 
Accession No. ML071310108.
    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.
    This license amendment request proposes Technical Specification 
Surveillance Requirement changes which will increase the monthly 
test load for the Unit 1 emergency diesel generators to a load 
greater than 90% of their continuous rated load which is consistent 
with the guidance of Regulatory Guide 1.9, ``Application and Testing 
of Safety-Related Diesel Generators in Nuclear Power Plants'', 
Revision 4.
    The emergency diesel generators are not accident initiators and 
therefore, these changes do not involve a significant increase [in] 
the probability of an accident. The proposed changes increase the 
test load requirements, are consistent with current regulatory 
guidance for testing emergency diesel generators, and will continue 
to assure that this equipment performs its design function. Thus 
these changes do not involve a significant increase in the 
consequences of an accident.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    This license amendment request proposes Technical Specification 
Surveillance Requirement changes which will increase the monthly 
test load for the Unit 1 emergency diesel generators to a load 
greater than 90% of their continuous rated load which is consistent 
with the guidance of Regulatory Guide 1.9, ``Application and Testing 
of Safety-Related Diesel Generators in Nuclear Power Plants'', 
Revision 4.
    The changes proposed for the emergency diesel generators do not 
change any system operations or maintenance activities. Testing 
requirements will be revised and will continue to demonstrate that 
the Limiting Conditions for Operation are met and the system 
components are functional. The revised test load is consistent with 
current plant procedures and practices. These changes do not create 
new failure modes or mechanisms and no new accident precursors are 
generated.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    This license amendment request proposes Technical Specification 
Surveillance Requirement changes which will increase the

[[Page 5227]]

monthly test load for the Unit 1 emergency diesel generators to a 
load greater than 90% of their continuous rated load which is 
consistent with the guidance of Regulatory Guide 1.9, ``Application 
and Testing of Safety-Related Diesel Generators in Nuclear Power 
Plants'', Revision 4.
    Current plant procedures require the Unit 1 emergency diesel 
generators to be load tested above 90% of their continuous rated 
load each month. This license amendment request proposes to make 
testing above 90% of the Unit 1 emergency diesel generator's 
continuous rated load a Technical Specification requirement. Since 
this change is an increase in the test requirements and the change 
is consistent with current regulatory guidance, this change does not 
involve a significant reduction in a margin of safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

    Omaha Public Power District, Docket No. 50-285, Fort Calhoun 
Station, Unit No. 1, Washington County, Nebraska.
    Date of amendment request: October 19, 2007, as supplemented by 
letter dated December 12, 2007.
    Description of amendment request: The proposed amendment modifies 
Technical Specification (TS) 3.6(3), ``Containment Recirculating Air 
Cooling and Filtering System.'' The licensee has determined that 
emergency mode (remotely operated) dampers in the containment air 
cooling and filtering system (CACFS) can be maintained in their 
accident positions permanently in all plant operating modes. 
Surveillance Requirement (SR) 3.6.3.a for testing the CACFS emergency 
mode (remotely operated) dampers each refueling outage will be deleted 
and be replaced with an SR to verify that the emergency mode dampers 
are in their accident positions. The licensee also proposes to delete 
the SR of TS 3.6(3)b to exercise the remotely operated (emergency mode) 
dampers at intervals not to exceed 3 months.
    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 containment air cooling and filtering system (CACFS) is not 
an initiator of any accident previously evaluated at the Fort 
Calhoun Station (FCS). The CACFS is an accident mitigation system. 
The current licensing basis function of the CACFS is to limit the 
containment pressure rise by providing a means for cooling the 
containment following a main steam line break (MSLB) design basis 
accident (DBA).
    The CACFS face and bypass dampers will be aligned to their 
accident positions permanently causing the CACFS to operate in 
filtered air mode. Surveillance testing has shown that operating the 
system in this alignment over long periods does not jeopardize 
filter performance. Over the lifetime of the plant, the differential 
pressures measured across the combined high efficiency particulate 
air (HEPA) and charcoal filter banks have met test acceptance 
criteria.
    With the dampers aligned to their accident positions 
permanently, the removal of TS requirements to check and exercise 
the dampers does not adversely affect the function of the CACFS. 
Each refueling outage, the dampers will be verified to be in their 
accident positions.
    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.
    The CACFS was designed to remove heat released to the 
containment atmosphere during a DBA to the extent necessary to 
maintain the containment structure below its design pressure. The 
face and bypass dampers will be aligned in their accident positions 
permanently, and the air supply, power, and ventilation isolation 
actuation signal to these dampers will be removed. Thus, the dampers 
will no longer have an active function and will not be required to 
change position under accident conditions.
    Each refueling outage, the dampers will be verified to be in 
their accident positions. The CACFS will continue to operate as 
before except that filter bypass mode will be unavailable. 
Surveillance testing has shown that the filters are capable of long-
term operation in filtered air mode without degrading their ability 
to respond to a DBA loss-of-coolant accident (LOCA).
    No credible new failure mechanisms, malfunctions, or accident 
initiators not previously considered in the design and licensing 
basis are created and none of the initial condition assumptions of 
any accident evaluated in the safety analysis are impacted.
    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.
    The containment building and associated penetrations are 
designed to withstand an internal pressure of 60 psig [pounds per 
square inch gauge] at 305 [deg]F [degrees Fahrenheit], including all 
thermal loads resulting from the temperature associated with this 
pressure, with a leakage rate of 0.1 percent by weight or less of 
the contained volume per 24 hours. The CACFS is credited for 
maintaining containment pressure and temperatures within design 
limits. The air coolers are also credited for limiting peak 
containment pressure for an MSLB.
    The CACFS consists of two redundant trains, each train with one 
air cooling and filtering unit and one air cooling unit, for a total 
of four cooling units. In accordance with analyses completed for 
replacement of the FCS steam generators in 2006, operation of the 
CACFS will continue to be credited in the MSLB containment pressure 
analysis. The CACFS face and bypass dampers will be aligned to their 
accident positions permanently. Therefore, TS surveillance 
requirements to periodically check and exercise these dampers are 
unnecessary. Each refueling outage, the dampers will be verified to 
be in their accident positions.
    The containment heat removal licensing basis is not adversely 
affected by the proposed changes. The ability to maintain design 
limits for containment peak pressure and temperature, as well as 
long-term containment pressure and temperature, is preserved.
    Therefore, the proposed [change] [does] not involve a 
significant reduction in a margin of safety.

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

    Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, 
Diablo Canyon Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California..
    Date of amendment request: December 17, 2007.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3.5.2, ``ECCS [Emergency Core 
Cooling System]--Operating,'' and TS 3.6.6, ``Containment Spray and 
Cooling Systems.'' The Diablo Canyon Power Plant ECCS consists of three 
separate subsystems: centrifugal charging, safety injection, and 
residual heat removal. The proposed changes to TS 3.5.2 would add new 
required actions and extend the Completion Time (CT) of the ECCS from 
72 hours to 14 days.

[[Page 5228]]

Similarly, the proposed change to TS 3.6.6 involves extending the CT 
for one inoperable containment spray train from 72 hours to 14 days. 
These amendments are risk-informed licensing changes.
    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 changes increase the Emergency Core Cooling System 
(ECCS) completion time (CT) to 14 days when one subsystem of one 
ECCS train is inoperable. Similarly, the proposed changes also 
increase the containment spray (CS) system CT to 14 days when one CS 
train is inoperable. These proposed changes do not physically alter 
any plant structures, systems, or components, and are not accident 
initiators; therefore, there is no effect on the probability of 
accidents previously evaluated. When one or more ECCS trains is 
inoperable, the Technical Specifications (TS) still requires at 
least 100 percent of the ECCS flow equivalent to a single OPERABLE 
ECCS train available. Similarly, when one CS train is inoperable, 
the TS still requires the redundant CS train to be OPERABLE. 
Therefore, redundant system and subsystems are still able to perform 
their safety functions. Also the proposed changes do not affect the 
types or amounts of radionuclides released following an accident, or 
affect the initiation and duration of their release. Therefore the 
consequences of accidents previously evaluated, which rely on the 
ECCS and CS system to mitigate, are not significantly increased.
    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 accident from any accident previously evaluated?
    Response: No.
    There are no new failure modes or mechanisms created due to 
plant operation with an extended CT. Extended operation with one 
ECCS train with one subsystem inoperable or with one train of CS 
system inoperable does not involve any modification to the 
operational limits or physical design of the systems. There are no 
new accident precursors generated due to the extended CT.
    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 change[s] [are] based upon both a deterministic 
evaluation and a risk-informed assessment. The deterministic 
evaluation concluded that though one ECCS train is inoperable for a 
longer period of time, the availability of the redundant OPERABLE 
ECCS train can still perform its safety function. Similarly, though 
one train of the CS system is inoperable for a longer period of 
time, the redundant OPERABLE CS train can still perform its safety 
function by providing at least the minimum spray flow to the 
containment assumed in the accident analyses.
    The risk assessment performed to support this license amendment 
request concluded that the increase in plant risk is small and 
consistent with the NRC's Safety Goal Policy Statement, ``Use of 
Probabilistic Risk Assessment Methods in Nuclear Activities: Final 
Policy Statement,'' and guidance [contained in] of Regulatory Guides 
(RG) 1.174, ``An Approach for Using Probabilistic Risk Assessment in 
Risk-Informed Decisions on Plant-Specific Changes to the Licensing 
Basis,'' and RG 1.177, ``An Approach for Plant-Specific, Risk-
Informed Decisionmaking: Technical Specifications.''
    Together, the deterministic evaluation and the risk-informed 
assessment provide assurance that the ECCS and the CS system will 
still meet their design requirements with the longer CTs proposed.
    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 three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Jennifer Post, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Branch Chief: Thomas G. Hiltz.

    Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, 
Diablo Canyon Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California.
    Date of amendment request: December 26, 2007.
    Description of amendment request: The proposed amendments would 
modify the Technical Specification (TS) to establish more effective and 
appropriate action, surveillance, and administrative requirements 
related to ensuring the habitability of the control room envelope (CRE) 
in accordance with Nuclear Regulatory Commission (NRC)-approved TS Task 
Force (TSTF) Standard Technical Specification change traveler TSTF-448, 
Revision 3, ``Control Room Habitability.'' Specifically, the proposed 
amendments would modify TS 3.7.10, ``Control Room Ventilation System 
(CRVS),'' and would establish a CRE habitability program in TS Section 
5.5, ``Administrative Controls--Programs and Manuals.'' The NRC staff 
issued a ``Notice of Availability of Technical Specification 
Improvement to Modify Requirements Regarding Control Room Envelope 
Habitability Using the Consolidated Line Item Improvement Process'' 
associated with TSTF-448, Revision 3, in the Federal Register on 
January 17, 2007 (72 FR 2022). The notice included a model safety 
evaluation, a model no significant hazards consideration (NSHC) 
determination, and a model license amendment request. In its 
application dated December 26, 2007, the licensee affirmed the 
applicability of the model NSHC determination which is presented below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC adopted by the licensee 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

[[Page 5229]]

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 the analysis adopted by the licensee 
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 NSHC.
    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. 50-348 and 
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama.
    Date of amendment request: November 5, 2007.
    Description of amendment request: The proposed amendments would 
revise Facility Operating License No. NPF-2 and Facility Operating 
License No. NPF-8 for Farley Nuclear Plant (FNP), Units 1 and 2, 
specifically, TS Section 5.5.17, ``Containment Leakage Rate Testing 
Program,'' to resolve a timing conflict between the FNP, Unit 2 R20 
refueling outage schedule and the 15-year test date for the FNP, Unit 2 
Type A Containment Integrated Leak Rate Test (ILRT), which has a 
required completion date of March 2010. Although Unit 1 does not have a 
current timing conflict, a similar Unit 1 change is proposed for 
consistency.
    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 revision to Technical Specifications 5.5.17, 
``Containment Leakage Rate Testing Program,'' resolves a schedule 
conflict between the Farley Nuclear Plant (FNP) Unit 2 refueling 
outage and the fifteen (15) year Containment Integrated Leak Rate 
Test date that is currently stated in the FNP Technical 
Specifications. The previous Integrated Leakage Rate Tests were 
completed in March 1994 for FNP Unit 1 and March 1995 for FNP Unit 
2. A 15 year deferral, granted by Amendments No. 159 and No. 150, 
placed the next integrated leak rate testing for FNP Unit 1 in March 
2009 and FNP Unit 2 in March 2010. Due to minor variations in the 
refueling outage schedule, the current refueling outage for FNP Unit 
2 has been scheduled for April 3, 2010 (Spring 2010). The Type A 
testing will begin during the FNP Unit 2 refueling outage which is 
three days after the 15 year time period from the March 1995 date 
that is currently stated in the revised FNP Technical Specifications 
(TS). This proposed change will revise FNP TS section 5.5.17 to 
include the current refueling outage schedule R22 (Spring 2009) for 
Unit 1 and R20 (Spring 2010) for Unit 2. The proposed Technical 
Specification change does not involve a physical change to the plant 
or a change in the manner in which the plant is operated or 
controlled. The reactor containment is designed to provide an 
essentially leak tight barrier against the uncontrolled release of 
radioactivity to the environment for postulated accidents. As such, 
the reactor containment exists to ensure the plant's ability to 
mitigate the consequences of an accident, and does not involve the 
prevention or identification of any precursors of an accident. 
Therefore, the proposed Technical Specification change does not 
involve a significant increase in the probability of an accident 
previously evaluated.
    Type B and C containment leakage testing will continue to be 
performed at the frequency currently required by plant Technical 
Specifications. Industry experience has shown, as documented in 
NUREG-1493, that Type B and C containment leakage tests have 
identified a very large percentage of containment leakage paths and 
that the percentage of containment leakage paths that are detected 
only by Type A testing is very small. FNP test history listed in 
letter from Southern Nuclear Operating Company to the Nuclear 
Regulatory Commission dated April 4, 2002 supports this conclusion. 
The basis and the conclusions reached in the significant hazards 
evaluation provide in the original SNC amendment request for the 
ILRT interval extension remain valid and unchanged. Therefore, this 
change does not involve a significant increase in the 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.
    This proposed change will revise FNP TS section 5.5.17 to 
include the current refueling outage schedule of R22 for Unit 1 and 
R20 for Unit 2. The basis and the conclusions reached in the 
significant hazards evaluation provided in the original amendment 
request for the ILRT interval extension provided in the original 
amendment request for the ILRT interval extension remain valid and 
unchanged.
    The reactor containment and the testing requirements invoked to 
periodically demonstrate the integrity of the reactor containment 
exist to ensure the plant's ability to mitigate the consequences of 
an accident and do not involve the prevention or identification of 
any precursors of an accident. The proposed Technical Specification 
change does not involve a physical change to the plant or a change 
in the manner in which the plant is operated or controlled. 
Therefore, the proposed Technical Specification 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 decrease in a 
margin of safety?
    Response: No.
    This proposed change will revise FNP TS section 5.5.17 to 
include the current refueling outage schedule of R22 for Unit 1 and 
R20 for Unit 2. The basis and the conclusions reached in the 
significant hazards evaluation provided in the original amendment 
request for the ILRT interval extension remain valid and unchanged. 
The proposed Technical Specifications change does not involve a 
physical change to the plant or a change in the manner in which the 
plant is operated or controlled. The specific requirements and 
conditions of the Containment Leakage Rate Testing Program, as 
defined in Technical Specifications, exist to ensure that the degree 
of reactor containment structural integrity and leak tightness that 
is considered in the plant safety analysis is maintained. The 
overall containment leakage rate limit specified by Technical 
Specifications is maintained. Type B and C containment leakage 
testing will continue to be performed at the frequency currently 
required by plant Technical Specifications. Industry experience has 
shown, as documented in NUREG-1493, that Type B and C containment 
leakage tests have identified a very large percentage of containment 
leakage paths and that the percentage of containment leakage paths 
that are detected only by Type A testing is very small. FNP test 
history listed in a letter from Southern Nuclear Operating Company 
dated April 4, 2002 to the Nuclear Regulatory Commission supports 
this conclusion. Therefore, this change does not involve a 
significant reduction in a margin of safety.
    Based on the above, Southern Nuclear Operating Company concludes 
that the

[[Page 5230]]

proposed amendment does not involve a significant hazards 
consideration under the standards set forth in 10 CFR 50.92, and 
accordingly, a finding of ``no significant hazards consideration'' 
is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Branch Chief: John Stang, Acting Chief.

    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: January 9, 2008.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3.3.2, ``Engineered Safety Feature 
Actuation System (ESFAS) Instrumentation,'' Table 3.3.2-1, ``Engineered 
Safety Feature Actuation System Instrumentation,'' Function 7.b, and TS 
3.5.4, ``Refueling Water Storage Tank (RWST),'' Surveillance 
Requirement (SR) 3.5.4.2. The proposed change to TS 3.3.2 lowers the 
Nominal Trip setpoint and corresponding Allowable Value of the 
Refueling water Storage Tank (RWST) Level--Low Low at which the semi-
automatic switchover from the RWST to the containment emergency sump 
occurs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    TS 3.3.2, ``ESFAS Instrumentation,'' Table 3.3.2-1 (page 6 of 
7), ``Engineered Safety Feature Actuation System Instrumentation,'' 
Function 7.b:
    No. The previously analyzed accidents are initiated by the 
failure of plant structures, systems, or components. The proposed 
change that decreases the Allowable Value and Nominal Trip Setpoint 
(NTS) of the semi-automatic switchover to containment sump (RWST 
Level--Low Low) does not have a detrimental impact on the integrity 
of any plant structure, system, or component (SSC) that initiates an 
analyzed event. The change does not adversely affect the protective 
and mitigative capabilities of the plant, nor does the change impact 
the initiation or probability of occurrence of any accident. The 
SSCs will continue to perform their intended safety functions.
    The minimum containment sump pH used in calculating the 
radiological consequences for a LOCA remains bounding. The offsite 
and control room doses will continue to meet the requirements of 10 
CFR 100 (Reactor Site Criteria) and 10 CFR 50 Appendix A GDC 19 
(General Design Criteria--Control Room).
    The proposed AV and NTS for TS Table 3.3.2-1, Function 7.b were 
determined using an uncertainty methodology previously approved by 
the NRC for this application. These values provide adequate 
assurance that required protective and mitigative functions will be 
initiated as assumed in the transient and accident analyses. 
Therefore, there is no significant increase in the probability or 
consequences of an accident previously evaluated.
    TS 3.5.4, ``Refueling Water Storage Tank (RWST),'' SR 3.5.4.2:
    No. The proposed change that increases the RWST borated water 
volume does not have a detrimental impact on the integrity of any 
plant structure, system, or component that initiates an analyzed 
event. The RWST borated water volume is not an initiator of any 
accident previously evaluated. As a result, the probability of an 
accident previously evaluated is not affected.
    The proposed change does not alter or prevent the ability of 
structures, systems, and components from performing their intended 
safety functions to mitigate the consequences of an initiating event 
within the assumed acceptance limits. The impact on the containment 
flood level, equipment qualification, and containment sump pH 
remains within the limits assumed in the design and accident 
analyses. The proposed change does not affect the source term, 
containment isolation, or radiological release assumptions used in 
evaluating the radiological consequences of an accident previously 
evaluated. The proposed change is consistent with the safety 
analysis assumptions and resultant consequences.
    The proposed change will not alter the operation of, or 
otherwise increase the failure probability of, any plant equipment 
that initiates an analyzed accident. Therefore, the proposed change 
does not involve a significant increase in the probability of an 
accident previously evaluated.
    Based on the above discussions, the proposed TS changes do not 
involve an increase in the consequences of an accident previously 
evaluated.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    No. The proposed changes do not involve the use or installation 
of new equipment and the currently installed equipment will not be 
operated in a new or different manner. No new or different system 
interactions are created and no new processes are introduced. The 
proposed changes will not introduce any new failure mechanisms, 
malfunctions, or accident initiators not already considered in the 
design and licensing bases. The possibility of a new or different 
malfunction of safety-related equipment is not created. No new 
accident scenarios, transient precursors, or limiting single 
failures are introduced as a result of these changes. There will be 
no adverse effect or challenges imposed on any safety-related system 
as a result of these changes.
    Based on this evaluation, 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. The proposed changes to the semi-automatic switchover to the 
containment sump RWST Level--Low Low AV and NTS and to the required 
RWST minimum borated water volume 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. The proposed 
changes will not result in plant operation in a configuration 
outside of the design basis. The proposed changes do not alter or 
prevent the ability of structures, systems, and components from 
performing their intending function to mitigate the consequences of 
an initiating event within the applicable acceptance criteria.
    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. 
The minimum and maximum pH values remain bounding; therefore, the 
required amount of trisodium phosphate (TSP) remains unchanged. The 
impact on the containment flood level, equipment qualification, 
hydrogen produced by the corrosion of galvanized surfaces and zinc 
based paints, and chloride induced stress corrosion remains within 
the limits assumed in the design and accident analyses.
    There will be no effect on the manner in which the Safety Limits 
or Limiting Safety System Settings are determined, nor will there be 
any effect on those plant systems necessary to assure the 
accomplishment of protection functions. 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: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Branch Chief: John Stang, Acting Chief.

[[Page 5231]]

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].

    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 applications for amendments: February 27, 2007.
    Brief description of amendments: These amendments modify Technical 
Specification (TS) 4.2.1, ``Fuel Assemblies,'' to permit up to four 
lead fuel assemblies (LFAs) with advanced cladding material to be 
inserted into the Unit 1 core for operating cycle 19 which is scheduled 
to begin in April 2008. Two of the LFAs were manufactured by 
Westinghouse Electric Company and contain a limited number of fuel rods 
with advanced zirconium-based alloys. The other two LFAs were 
manufactured by AREVA with fuel rod cladding material classified as 
M5\TM\ alloy. These LFAs, which were originally inserted into the Unit 
2 core in April 2003, remained there for operating cycles 15 and 16 and 
were subsequently removed in April 2007. These amendments also modify 
TS 5.6.5, ``Core Operating Limits Report (COLR),'' for the Calvert 
Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, to include WCAP-15604-
NP, ``Limited Scope High Burnup Lead Test Assemblies,'' as an approved 
analytical method for extended LFA burnup limits.
    Date of issuance: December 20, 2007.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 283 and 260.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the License and Technical Specifications.
    Date of initial notice in Federal Register: April 24, 2007 (72 FR 
20377).
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated December 20, 2007.
    No significant hazards consideration comments received: No.

    Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 
50-318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland.
    Date of application for amendments: February 1, 2007, as 
supplemented by letter dated August 17, 2007.
    Brief description of amendments: These amendments revise 
Surveillance Requirement 3.5.2.8 in Technical Specification 3.5.2, 
``ECCS--Operating,'' to reflect the replacement of the containment 
recirculation sump suction inlet trash racks and screens with 
strainers. The containment recirculation sump suction inlet trash racks 
and screens are being replaced with a strainer design with 
significantly larger effective surface area in response to Nuclear 
Regulatory Commission Generic Letter 2004-02, ``Potential Impact of 
Debris Blockage on Emergency Recirculation during Design Basis 
Accidents at Pressurized-Water Reactors.''
    Date of issuance: December 27, 2007.
    Effective date: As of the date of issuance to be implemented within 
60 days following completion of the installation and testing of the 
plant modifications described in the licensee's letters dated February 
1 and August 17, 2007.
    Amendment Nos.: 284 and 261.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the License and Technical Specifications.
    Date of initial notice in Federal Register: March 13, 2007 (72 FR 
11385).
    The letter dated August 17, 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 these amendments is 
contained in a Safety Evaluation dated December 27, 2007.
    No significant hazards consideration comments received: No.

    Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina.
    Date of application for amendments: December 21, 2006.
    Brief Description of amendments: The amendments change the 
Technical Specifications (TSs) related to the reactor recirculation 
system flow balance.
    Date of issuance: December 17, 2007.
    Effective date: Date of issuance, to be implemented within 60 days.
    Amendment Nos.: 244 and 272.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
changed the TSs.
    Date of initial notice in Federal Register: March 13, 2007 (72 FR 
11385).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 17, 2007.
    No significant hazards consideration comments received: No.

    Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York.
    Date of application for amendment: July 17, 2007, as supplemented 
on August 13, 2007.

[[Page 5232]]

    Brief description of amendment: The proposed amendment would revise 
action and surveillance requirements related to control room envelope 
(CRE) habitability in Technical Specification (TS) Section 3.7.3 
``Control Room Emergency Ventilation Air Supply (CREVAS) System,'' and 
adds a new administrative controls program, TS Section 5.5.14, 
``Control Room Envelope Habitability Program.'' In addition, the 
proposed amendment adds a license condition which specifies the 
schedule for performing the new surveillance and assessment 
requirements for the Control Room Envelope Habitability Program, and 
corrects a typographical error in Appendix C of the license. The 
changes are consistent with NRC-approved Revision 3 to Technical 
Specifications Task Force (TSTF) Improved Standard Technical 
Specifications Change Traveler, TSTF-448, ``Control Room 
Habitability.'' TSTF-448, Revision 3 is a proposal to establish more 
effective and appropriate action, surveillance, and administrative TS 
requirements related to ensuring the habitability of the control room 
envelope.
    Date of issuance: January 3, 2008.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 289.
    Facility Operating License No. DPR-59: The amendment revised the 
License and the Technical Specifications.
    Date of initial notice in Federal Register: September 11, 2007 (72 
FR 51854).
    The August 13, 2007, supplement provided additional information 
that clarified the application, did not expand the scope of the 
application as originally noticed, and did not change the NRC staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 3, 2008.
    No significant hazards consideration comments received: No.

    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 application for amendments: January 11, 2007, as 
supplemented by letters dated August 9, and September 28, 2007.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TS) to support replacement of the steam 
generators (SGs). They revise TS 3.3.2, ``Engineered Safety Feature 
Actuation System (ESFAS) Instrumentation,'' TS 5.5.9, ``Steam Generator 
(SG) Program,'' and TS 5.6.10, ``Steam Generator (SG) Tube Inspection 
Report.''
    Date of issuance: January 8, 2008.
    Effective date: As of its date of issuance and shall be implemented 
prior to entry into Mode 4 following the 14th refueling outage for Unit 
2 and prior to entry into Mode 4 following the 15th refueling outage 
for Unit 1.
    Amendment Nos.: Unit 1--198; Unit 2--199.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: February 13, 2007 (72 
FR 6787).
    The supplemental letters dated August 9, and September 28, 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 January 8, 2008.
    No significant hazards consideration comments received: No.

    Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 
50-281, Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.
    Date of application for amendments: March 6, 2007.
    Brief Description of amendments: These amendments authorized 
revisions to the Updated Final Safety Analysis Report (UFSAR) to permit 
irradiation of the fuel assemblies beginning with Surry Power Station, 
Unit Nos. 1 and 2, improved fuel assemblies with ZIRLO (Westinghouse 
trademark) cladding to a lead rod average burnup of 62,000 MWD/MTU.
    Date of issuance: December 19, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days. The UFSAR changes shall be implemented in the next 
periodic update made in accordance with 10 CFR 50.71(e).
    Amendment Nos.: 257, 256.
    Renewed Facility Operating License Nos. DPR-32 and DPR-37: 
Amendments changed the licenses.
    Date of initial notice in Federal Register: March 27, 2007 (72 FR 
14309).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 19, 2007.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time

[[Page 5233]]

for public comment but less than 30 days, the Commission may provide an 
opportunity for public comment. If comments have been requested, it is 
so stated. In either event, the State has been consulted by telephone 
whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room (PDR), located at One White Flint North, Public File Area 01F21, 
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly 
available records will be accessible from the Agencywide Documents 
Access and Management System's (ADAMS) Public Electronic Reading Room 
on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems 
in accessing the documents located in ADAMS, contact the PDR Reference 
staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to [email protected].
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. Within 60 days after the date 
of publication of this notice, 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, and electronically on the Internet at the NRC Web 
site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are 
problems in accessing the document, contact the PDR Reference staff at 
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. If a 
request for a hearing or petition for leave to intervene is filed by 
the above date, the Commission or a presiding officer designated by the 
Commission or by the Chief Administrative Judge of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the Chief Administrative Judge of the Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion, which support the contention and on which the petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to those specific sources and 
documents of which the petitioner is aware and on which the petitioner 
intends to rely to establish those facts or expert opinion. The 
petition must include sufficient information to show that a genuine 
dispute exists with the applicant on a material issue of law or 
fact.\1\ Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner to relief. A petitioner/requestor 
who fails to satisfy these requirements with respect to at least one 
contention will not be permitted to participate as a party.
---------------------------------------------------------------------------

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

    Each contention shall be given a separate numeric or alpha 
designation within one of the following groups:
    1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the 
applications.
    2. Environmental--primarily concerns/issues relating to matters 
discussed or referenced in the environmental analysis for the 
applications.
    3. Miscellaneous--does not fall into one of the categories outlined 
above.
    As specified in 10 CFR 2.309, if two or more petitioners/requestors 
seek to co-sponsor a contention, the petitioners/requestors shall 
jointly designate a representative who shall have the authority to act 
for the petitioners/requestors with respect to that contention. If a 
petitioner/requestor seeks to adopt the contention of another 
sponsoring petitioner/requestor, the petitioner/requestor who seeks to 
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that 
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the 
petitioners/requestors with respect to that contention.
    Those permitted to intervene become parties to the proceeding, 
subject to any

[[Page 5234]]

limitations in the order granting leave to intervene, and have the 
opportunity to participate fully in the conduct of the hearing. Since 
the Commission has made a final determination that the amendment 
involves no significant hazards consideration, if a hearing is 
requested, it will not stay the effectiveness of the amendment. Any 
hearing held would take place while the amendment is in effect.
    A request for hearing or a petition for leave to intervene must be 
filed in accordance with the NRC E-Filing rule, which the NRC 
promulgated on 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 Viewer\TM\ to 
access the Electronic Information Exchange (EIE), a component of the E-
Filing system. The Workplace Forms Viewer\TM\ 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.

    Duke Power Company LLC, et al., Docket No. 50-413, Catawba Nuclear 
Station, Unit 1 York County, South Carolina.
    Date of amendment request: January 1, 2008, as supplemented January 
2, 2008.
    Description of amendment request: The amendment approved a one-time 
extension of the allowed outage time (AOT) for the 1B centrifugal 
charging (NV) pump beyond the 72 hours allowed by the Technical 
Specifications (TSs) up to a total of 240 hours as part of the 1B NV 
pump repair. In addition, the amendment approved a one-time extension 
for the auxiliary building filtered ventilation exhaust system 
(ABFVES), to have two ABFVES trains inoperable.
    Date of issuance: January 2, 2008.
    Effective date: January 2, 2008.
    Amendment No.: 239.
    Facility Operating License No. (NPF-68): Amendment revised the 
technical specifications and license.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, state consultation, and final NSHC 
determination are contained in a safety evaluation dated January 2, 
2008.
    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 Acting Branch Chief: John F. Stang, Acting.

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


[[Page 5235]]


    For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
 [FR Doc. E8-1300 Filed 1-28-08; 8:45 am]
BILLING CODE 7590-01-P