Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 5215-5235 [E8-1300]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 73, Number 19 (Tuesday, January 29, 2008)]
[Notices]
[Pages 5215-5235]
From the Federal Register Online via the Government Printing 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, https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request 
for a hearing or petition for leave to intervene is filed within 60 
days, the Commission or a presiding officer designated by the 
Commission or by the Chief Administrative Judge of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the Chief Administrative Judge of the Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's

[[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 
HEARINGDOCKET@NRC.GOV, or by calling (301) 415-1677, to request (1) a 
digital ID certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and/or (2) 
creation of an electronic docket for the proceeding (even in instances 
in which the petitioner/requestor (or its counsel or representative) 
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms ViewerTM 
to access the Electronic Information Exchange (EIE), a component of the 
E-Filing system. The Workplace Forms ViewerTM is free and is 
available at https://www.nrc.gov/site-help/e-submittals/install-
viewer.html. Information about applying for a digital ID certificate is 
available on NRC's public Web site at https://www.nrc.gov/site-help/e-
submittals/apply-certificates.html.
    Once a petitioner/requestor has obtained a digital ID certificate, 
had a docket created, and downloaded the EIE viewer, it can then submit 
a request for hearing or petition for leave to intervene. Submissions 
should be in Portable Document Format (PDF) in accordance with NRC 
guidance available on the NRC public Web site at https://www.nrc.gov/
site-help/e-submittals.html. A filing is considered complete at the 
time the filer submits its documents through EIE. To be timely, an 
electronic filing must be submitted to the EIE system no later than 
11:59 p.m. Eastern Time on the due date. Upon receipt of a 
transmission, the E-Filing system time-stamps the document and sends 
the submitter an e-mail notice confirming receipt of the document. The 
EIE system also distributes an e-mail notice that provides access to 
the document to the NRC Office of the General Counsel and any others 
who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically may seek assistance through the 
``Contact Us'' link located on the NRC Web site at https://www.nrc.gov/
site-help/e-submittals.html or by calling the NRC technical help line, 
which is available between 8:30 a.m. and 4:15 p.m., Eastern Time, 
Monday through Friday. The help line number is (800) 397-4209 or 
locally, (301) 415-4737.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file a motion, in accordance 
with 10 CFR 2.302(g), with their initial paper filing requesting 
authorization to continue to submit documents in paper format. Such 
filings must be submitted by: (1) First-class mail addressed to the 
Office of the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff; or (2) courier, express mail, or expedited 
delivery service to the Office of the Secretary, Sixteenth Floor, One 
White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852, 
Attention: Rulemaking and Adjudications Staff. Participants filing a 
document in this manner are responsible for serving the document on all 
other participants. Filing is considered complete by first-class mail 
as of the time of deposit in the mail, or by courier, express mail, or 
expedited delivery service upon depositing the document with the 
provider of the service.
    Non-timely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission, the presiding 
officer, or the Atomic Safety and Licensing Board that the petition 
and/or request should be granted and/or the contentions should be 
admitted, based on a balancing of the factors specified in 10 CFR 
2.309(c)(1)(i)-(viii). To be timely, filings must be submitted no later 
than 11:59 p.m. Eastern Time on the due date.
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
https://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant 
to an order of the Commission, an Atomic Safety and Licensing Board, or 
a Presiding Officer. Participants are requested not to include

[[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, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

    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 
cr