Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 67744-67757 [05-22002]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 70, Number 215 (Tuesday, November 8, 2005)]
[Notices]
[Pages 67744-67757]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 05-22002]


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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 October 14, 2005 to October 27, 2005. The 
last biweekly notice was published on October 25, 2005 (70 FR 61655).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination. Within 60 days after the date of publication of this 
notice, the licensee may file a request for a hearing with respect to 
issuance of the amendment to the subject facility operating license and 
any person whose interest may be affected by this proceeding and who 
wishes to participate as a party in the proceeding must file a written 
request for a hearing and a petition for leave to intervene.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives 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 
01F21, 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, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR part 2. Interested persons should 
consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, 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

[[Page 67745]]

Panel, will rule on the request and/or petition; and the Secretary or 
the Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, HearingDocket@nrc.gov; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, 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.

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

    Date of amendment request: August 18, 2005.
    Description of amendment request: The amendment will allow the use 
of fire-resistive electrical cable, which has been demonstrated to 
provide an equivalent level of protection as would be provided by 3-
hour and 1-hour rated electrical cable raceway fire barriers, for the 
protection of safe shutdown electrical cable.
    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.
    Operation of HNP in accordance with the proposed amendment does 
not increase the probability or consequences of accidents previously 
evaluated. The Final Safety Analysis Report (FSAR) documents the 
analyses of design basis accidents (DBA) at HNP. Any scenario or 
previously analyzed accidents that result in offsite dose were 
evaluated as part of this analysis. The proposed amendment does not 
adversely affect accident initiators nor alter design assumptions, 
conditions, or configurations of the facility. The proposed 
amendment does not adversely affect the ability of structures, 
systems, or components (SSCs) to perform their design function. SSCs 
required to safely shut down the reactor and to maintain it in a 
safe shutdown condition remain capable of performing their design 
functions.
    The purpose of this amendment is to assure that redundant trains 
of safe shutdown (SSD) control circuits remain protected from damage 
in the event of a postulated fire. The proposed amendment revises 
the Final Safety Analysis Report (FSAR) to use three-hour fire-
resistive electrical cable, which has been demonstrated to provide 
an equivalent level of protection as would be provided by three-hour 
and one-hour rated electrical cable raceway fire barriers, for the 
protection of

[[Page 67746]]

SSD electrical cables. Based on the above, SSD control circuit 
protection is maintained by this amendment.
    Therefore, this amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Operation of HNP in accordance with the proposed amendment does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated. The FSAR documents the 
analyses of design basis accidents (DBA) at HNP. Any scenario or 
previously analyzed accidents that result in offsite dose were 
evaluated as part of this analysis. The proposed amendment does not 
change or affect any accident previously evaluated in the FSAR, and 
no new or different scenarios are created by the proposed amendment. 
The proposed amendment does not adversely affect accident initiators 
nor alter design assumptions, conditions, or configurations of the 
facility. The proposed amendment does not adversely affect the 
ability of SSCs to perform their design function. SSCs required to 
safely shut down the reactor and to maintain it in a safe shutdown 
condition remain capable of performing their design functions.
    The purpose of this amendment is to assure that redundant trains 
of Safe Shutdown (SSD) control circuits remain protected from damage 
in the event of a postulated fire. The proposed amendment revises 
the Final Safety Analysis Report (FSAR) to use three-hour fire-
resistive electrical cable, which has been demonstrated to provide 
an equivalent level of protection as would be provided by three-hour 
and one-hour rated electrical cable raceway fire barriers, for the 
protection of SSD electrical cables. Based on the above, SSD control 
circuit protection is maintained by this amendment.
    Therefore, this amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Operation of HNP in accordance with the proposed amendment does 
not involve a significant reduction in a margin of safety. The 
proposed amendment does 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 this change. The proposed amendment does not adversely 
affect existing plant safety margins or the reliability of equipment 
assumed to mitigate accidents in the FSAR. The proposed amendment 
does not adversely affect the ability of SSCs to perform their 
design function. SSCs required to safely shut down the reactor and 
to maintain it in a safe shutdown condition remain capable of 
performing their design functions.
    The purpose of this amendment is to assure that redundant trains 
of Safe Shutdown (SSD) control circuits remain protected from damage 
in the event of a postulated fire. The proposed amendment revises 
the Final Safety Analysis Report (FSAR) to use three-hour fire-
resistive electrical cable, which has been demonstrated to provide 
an equivalent level of protection as would be provided by three-hour 
and one-hour rated electrical cable raceway fire barriers, for the 
protection of SSD electrical cables. Based on the above, SSD control 
circuit protection is maintained by this amendment.
    Therefore, this amendment does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: September 1, 2005.
    Description of amendment request: The amendment will add Technical 
Specification (TS) 3.7.14, ``Fuel Storage Pool Boron Concentration'' 
and revise TS 5.6, ``Fuel Storage.'' The proposed changes are related 
to requirements for ensuring adequate subcriticality margin in the 
spent fuel storage pools. TS 5.6.1 is being revised to include the 
design requirements for dry storage of new fuel.
    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 changes do not modify the facility. The accident 
previously analyzed for the spent fuel pool is a fuel handling 
accident. The proposed change applies administrative controls for 
maintaining the required boron concentration in the spent fuel 
storage pools, revises acceptance criteria and storage arrangements 
for fuel storage in PWR [pressurized-water reactor] ``flux trap'' 
style racks and adds acceptance criteria for dry storage of new fuel 
to the Technical Specifications. The controls on spent fuel pool 
boron and dry storage of new fuel have previously been implemented 
but are being added to the Technical Specifications as requirements. 
The proposed change applies new acceptance criteria for criticality 
safety of fuel storage in PWR ``flux trap'' style racks in Pools 
``A'' and ``B.'' The new acceptance criteria require new 
administrative controls on the placement of fuel in Pools ``A'' and 
``B.'' Similar administrative controls have previously been placed 
on fuel stored in Pools C and D. These changes will eliminate the 
dependence on Boraflex in the PWR ``flux trap'' style storage racks. 
These changes do not impact the probability of having a fuel 
handling accident and do not impact the consequences of a fuel 
handling accident.
    Therefore, this amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    No change is being made to the acceptance criteria of the dry 
storage of new fuel. These criteria are being added to Technical 
Specification Section 5.6.1. Detailed analyses have been performed 
to ensure a criticality accident in Pools ``A'' and ``B'' is not a 
credible event. The events that could lead to a criticality accident 
are not new. These events include a fuel mis-positioning event, a 
fuel drop event, and a boron dilution event. The proposed changes do 
not impact the probability of any of these events. The detailed 
criticality analyses performed demonstrate that criticality would 
not occur following any of these events. For the more likely event, 
such as a fuel mis-positioning event, the acceptance criteria for 
keff remains less than or equal to 0.95. For the unlikely 
event that the spent fuel storage pool boron concentration was 
reduced to zero, keff remains less than 1.0.
    Therefore, a criticality accident remains ``not credible,'' and 
this amendment does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Incorporation of acceptance criteria for dry storage of new fuel 
into TS 5.6.1 does not involve a reduction in the margin of safety. 
The new fuel storage condition continues to meet keff <= 
0.95 during normal conditions and keff <= 0.98 under 
optimal moderation conditions.
    The proposed changes for storage of new and irradiated fuel in 
Pools ``A'' and ``B'' continue to provide the controls necessary to 
ensure a criticality event could not occur in the spent fuel storage 
spool. The acceptance criteria are consistent with the acceptance 
criteria specified in 10 CFR 50.68, which provide an acceptable 
margin of safety with regard to the potential for a criticality 
event.
    Therefore, this amendment does not involve a significant 
reduction in a margin of safety.


[[Page 67747]]


    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 Section Chief: Michael L. Marshall, Jr.

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

    Date of amendment request: September 26, 2005.
    Description of amendment request: The proposed amendment will 
revise the analysis method used for the large-break loss-of-coolant 
accident (LBLOCA) by incorporating the use of a new approach (ASTRUM) 
for the treatment of parameter uncertainties. The new approach is 
described in Westinghouse Topical Report WCAP-16009-P-A, approved by 
the NRC on November 5, 2004.
    Changes to the Technical Specifications to reflect the proposed use 
of ASTRUM in LBLOCA analysis consist of revisions to the list of 
references provided in Technical Specification Section 5.6.5, Core 
Operating Limits Report.
    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 the analysis methodology used to 
account for the variation in parameters that are used for the safety 
analysis of the LBLOCA. This proposed change has no effect on the 
design or operation of plant equipment. Use of the new methodology 
will revise the results of the current analysis, but there will be 
no change in initiating events for this accident scenario or the 
ability of the plant equipment or plant operators to respond.
    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 modifications to existing 
plant equipment or the installation of any new equipment. The 
proposed change only affects the analysis methodology that is used 
to evaluate the response of existing plant equipment to the LBLOCA 
scenario. Plant operating and emergency procedures that are in place 
for the LBLOCA scenario are also not being changed by this proposed 
amendment. This proposed change does not create new failure modes or 
malfunctions of plant equipment nor is there a new credible failure 
mechanism.
    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 proposed license amendment revises the analysis methodology 
which is used to assess the impact of the LBLOCA scenario with 
respect to established acceptance criteria. Margins of safety for 
LBLOCA include quantitative limits for fuel performance established 
in 10 CFR 50.46. These acceptance criteria and the associated 
margins of safety are not being changed. The evaluation of the 
LBLOCA scenario, using the proposed new methodology must still meet 
the existing established acceptance criteria.
    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 Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: Richard J. Laufer.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of amendment request: April 4, 2005.
    Description of amendment request: The proposed amendments would 
revise the maximum and minimum allowable values for the degraded 
voltage function of the 4160 volt essential service system (ESS) bus 
under-voltage 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 proposed changes revise the Technical Specifications (TS) 
maximum and minimum allowable values for the degraded voltage 
protection function and implement the use of automatic load tap 
changers (LTCs) on transformers that provide power to safety-related 
equipment. The only accident previously evaluated for which the 
probability is potentially affected by these changes is the loss of 
offsite power (LOOP). An allowable value for the degraded voltage 
protection function that is too high could cause the emergency buses 
to transfer to the emergency diesel generators (EDG) and thus 
increase the probability of a LOOP. The allowable value for the 
degraded voltage protection function has been revised in accordance 
with an NRC-approved setpoint methodology and will continue to 
ensure that the degraded voltage protection function actuates when 
required, but does not actuate prematurely to cause a LOOP.
    A failure of an LTC while in automatic operation mode that 
results in decreased voltage to the ESS buses could also cause a 
LOOP. This could occur in two ways. A failure of the LTC controller 
that results in rapidly decreasing the voltage to the emergency 
buses is the most severe failure mode. However, a backup controller 
is provided with the LTC that makes this failure highly unlikely. A 
failure of the LTC controller to respond to decreasing grid voltage 
is less severe, since grid voltage changes occur slowly. In both of 
the above potential failure modes, operators will take manual 
control of the LTC to mitigate the effects of the failure. Thus, the 
probability of a LOOP is not significantly increased.
    The proposed changes will have no effect on the consequences of 
a LOOP, since the EDGs provide power to safety related equipment 
following a LOOP. The EDGs are not affected by the proposed changes.
    The probability of other accidents previously evaluated is not 
affected, since the proposed changes do not affect the way plant 
equipment is operated and thus do not contribute to the initiation 
of any of the previously evaluated accidents. The only way in which 
the consequences of other previously evaluated accidents could be 
affected is if a failure of the LTC while in automatic operation 
mode caused a sustained high voltage which resulted in damage to 
safety related equipment that is used to mitigate an accident. 
Damage due to over-voltage is time-dependent. Since the LTC is 
equipped with a backup controller, and since operator action is 
available to prevent a sustained high voltage condition from 
occurring, damage to safety related equipment is extremely unlikely, 
and thus the consequences of these accidents are not significantly 
increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of

[[Page 67748]]

accident from any accident previously evaluated?
    Response: No.
    The proposed changes involve functions that provide offsite 
power to safety related equipment for accident mitigation. Thus, the 
proposed changes potentially affect the consequences of previously 
evaluated accidents (as addressed in Question 1), but do not result 
in any new mechanisms that could initiate damage to the reactor and 
its principal safety barriers (i.e., fuel cladding, reactor coolant 
system, or primary containment).
    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 proposed changes do not affect the inputs or assumptions of 
any of the analyses that demonstrate the integrity of the fuel 
cladding, reactor coolant system, or containment during accident 
conditions. The allowable values for the degraded voltage protection 
function have been revised in accordance with an NRC-approved 
setpoint methodology and will continue to ensure that the degraded 
voltage protection function actuates when required, but does not 
actuate prematurely to cause a LOOP. Automatic operation of the LTC 
increases margin by reducing the potential for transferring to the 
EDGs during an event.
    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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Gene Y. Suh.

FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire

    Date of amendment request: September 22, 2005.
    Description of amendment request: The proposed amendment would 
revise the Seabrook Station, Unit No. 1 operating license and Technical 
Specifications to increase the licensed rated power level by 1.7 
percent from 3587 megawatts thermal (MWt) to 3648 MWt. Basis for 
proposed no significant hazards consideration determination: As 
required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration, which is presented 
below:

    1. The proposed change will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Seabrook Station performed evaluations of the Nuclear Steam 
Supply System (NSSS) and balance of plant systems, components, and 
analyses that could be affected by the proposed change. A power 
uncertainty calculation was performed, and the effect of increase 
core thermal power by 1.7 percent to 3648 MWt on the Seabrook 
Station design and licensing basis was evaluated. The result of the 
evaluations determined that all systems and components continue to 
be capable of performing their design function at the MUR 
[measurement uncertainty recapture] core power level of 3648 MWt. An 
evaluation of the accident analyses demonstrates that the applicable 
analyses acceptance criteria continue to be met. No accident 
initiators are affected by the MUR power uprate and no challenges to 
any plant safety barriers are created by the proposed change.
    The proposed change does not affect the release paths, the 
frequency of release, or the analyzed source term for any accidents 
previously evaluated in the Seabrook Station Updated Final Safety 
Analysis Report (UFSAR). Systems, structures, and components 
required to mitigate transients continue to be capable of performing 
their design functions, and thus were found acceptable. The reduced 
uncertainty in the feedwater flow input to the power calorimetric 
measurement ensures that applicable accident analyses acceptance 
criteria continue to be met, to support operation at the MUR core 
power level of 3648 MWt. Analyses performed to assess the effects of 
mass and energy remain valid. The source term used to assess 
radiological consequences [has] been reviewed and determined to 
bound operation at the MUR core power level.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    No new accident scenarios, failure mechanisms, or single 
failures are introduced as a result of the proposed change. The 
installation of the Caldon LEFM CheckPlusTM System has 
been analyzed, and failures of the system will have no adverse 
effect on any safety-related system or any systems, structures, and 
components required for transient mitigation. Systems, structures, 
and components previously required for the mitigation of a transient 
continue to be capable of fulfilling their intended design 
functions. The proposed change has no adverse affect on any safety-
related system or component and does not change the performance or 
integrity of any safety-related system.
    The proposed change does not adversely affect any current system 
interfaces or create any new interfaces that could result in an 
accident or malfunction of a different kind than previously 
evaluated. Operating at a core power level of 3648 MWt does not 
create any new accident initiators or precursors. The reduced 
uncertainty in the feedwater flow input to the power calorimetric 
measurement ensures that applicable accident analyses acceptance 
criteria continue to be met, to support operation at the MUR core 
power level of 3648 MWt. Credible malfunctions continue to be 
bounded by the current accident analyses of record or evaluations 
that demonstrate that applicable criteria continue to be met.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. The proposed change will not involve a significant reduction 
in a margin [of] safety.
    The margins of safety associated with the MUR are those 
pertaining to core thermal power. These include those associated 
with the fuel cladding, Reactor Coolant System pressure boundary, 
and containment barriers. An engineering evaluation of the 1.7 
percent increase in core thermal power from 3587 MWt to 3648 MWt was 
performed. The current licensing bases analyzed core power is 3659 
MWt. The analyzed core power level of 3659 MWt bounds the NSSS 
thermal and hydraulic parameters at the MUR core power level of 3648 
MWt. The NSSS systems and components were evaluated at the MUR core 
power level and it was determined that the NSSS systems and 
components continue to operate satisfactorily at the MUR power 
level. The NSSS accident analyses were evaluated at the MUR core 
power level of 3648 MWt. In all cases, the accident analyses at the 
MUR core power level of 3648 MWt were bounded by the current 
licensing bases analyzed core power level of 3659 MWt. As such, the 
margins of safety continue to be bounded by the current analyses of 
record for this change.
    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: M. S. Ross, Florida Power & Light Company, 
P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Section Chief: Darrell J. Roberts.

FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire

    Date of amendment request: September 29, 2005.
    Description of amendment request: The proposed amendment would 
revise the Seabrook Station, Unit No. 1, Technical Specifications (TSs) 
to permit a one-time, six-month extension to the currently approved 15-
year test interval for the containment integrated leak rate test.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 67749]]

As required by 10 CFR 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration, which is 
presented below:

    1. The proposed change [does] not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The probability or consequences of accidents previously 
evaluated in the UFSAR [updated final safety analysis report] are 
unaffected by this proposed change. There is no change to any 
equipment response or accident mitigation scenario, and this change 
results in no additional challenges to fission product barrier 
integrity. The proposed change does not alter the design, 
configuration, operation, or function of any plant system, 
structure, or component. As a result, the outcomes of previously 
evaluated accidents are unaffected. The proposed extension to the 
containment integrated leak rate test (ILRT) interval does not 
involve a significant increase in consequences because, as discussed 
in NUREG 1493, Performance Based Containment Leak Rate Test Program, 
Type B and C tests identify the vast majority (greater than 95 
percent) of all potential leakage paths. Further, ILRTs identify 
only a few potential leakage paths that cannot be identified through 
Type B and C testing, and leaks found by Type A testing have been 
only marginally greater than existing requirements. In addition, 
periodic inspections ensure that any significant containment 
degradation will not go undetected. Therefore, the proposed change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed change [does] not create the possibility of a 
new or different kind of accident from any [accident] previously 
evaluated.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed change. 
The proposed change does not challenge the performance or integrity 
of any safety-related system. The proposed change neither installs 
or removes any plant equipment, nor alters the design, physical 
configuration, or mode of operation of any plant structure, system, 
or component. No physical changes are being made to the plant, so no 
new accident causal mechanisms are being introduced. The proposed 
change only changes the frequency of performing the ILRT; however, 
the test implementation and acceptance criteria are unchanged. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. The proposed change [does] not involve a significant 
reduction in a margin of safety.
    The margin of safety associated with the acceptance criteria of 
any accident is unchanged. The proposed change will have no affect 
on the availability, operability, or performance of the safety-
related systems and components. The proposed change does not alter 
the design, configuration, operation, or function of any plant 
system, structure, or component. The ability of any operable 
structure, system, or component to perform its designated safety 
function is unaffected by this change. NUREG 1493 concluded that 
reducing the frequency of ILRTs to 20 years resulted in an 
imperceptible increase in risk. Also, inspections of containment, 
required by the ASME code [American Society of Mechanical Engineers 
Boiler and Pressure Vessel Code] and the maintenance rule, ensure 
that containment will not degrade in a manner that is only 
detectable by Type A (ILRT) testing. Therefore, the margin of safety 
as defined in the TS is not reduced and 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: M. S. Ross, Florida Power & Light Company, 
P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Section Chief: Darrell J. Roberts.

FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire

    Date of amendment request: September 29, 2005.
    Description of amendment request: The proposed amendment would 
revise the Seabrook Station, Unit No. 1 Technical Specifications to 
permit a change in the steam generator tube inspection requirements to 
include a sampling of the bulges and over-expansions for portions of 
the steam generator tubes within the hot leg tubesheet region.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The previously analyzed accidents are initiated by the failure 
of plant structures, systems, or components. The proposed changes 
that alter the steam generator inspection criteria do not have a 
detrimental impact on the integrity of any plant structure, system, 
or component that initiates an analyzed event. The proposed changes 
will not alter the operation of, or otherwise increase the failure 
probability of any plant equipment that initiates an analyzed 
accident.
    Of the applicable accidents previously evaluated, the limiting 
transients with consideration to the proposed changes to the steam 
generator tube inspection criteria, are the steam generator tube 
rupture (SGTR) event and the steam line break (SLB) accident.
    During the SGTR event, the required structural integrity margins 
of the steam generator tubes will be maintained by the presence of 
the steam generator tubesheet area. Tube rupture in tubes with 
cracks in the tubesheet is precluded by the constraint provided by 
the tubesheet. This constraint results from the hydraulic expansion 
process, thermal expansion mismatch between the tube and tubesheet 
and from the differential pressure between the primary and secondary 
side. Based on this design, the structural margins against burst, as 
discussed in Regulatory Guide (RG) 1.121, ``Bases for Plugging 
Degraded PWR [pressurized-water reactor] Steam Generator Tubes,'' 
are maintained for both normal and postulated accident conditions.
    At normal operating pressures, leakage from primary water stress 
corrosion cracking (PWSCC) below the proposed limited inspection 
depth is limited by both the tube-to-tubesheet crevice and the 
limited crack opening permitted by the tubesheet constraint. 
Consequently, negligible normal operating leakage is expected from 
cracks within the tubesheet region. The consequences of an SGTR 
event are affected by the primary-to-secondary leakage flow during 
the event. Primary-to-secondary leakage flow through a postulated 
ruptured tube is not affected by the proposed changes since the 
tubesheet enhances the tube integrity in the region of the hydraulic 
expansion by precluding tube deformation beyond its initial 
hydraulically-expanded outside diameter.
    Furthermore, the proposed changes do not affect other systems, 
structures, components or operational features. Therefore, the 
proposed changes result in no significant increase in the 
probability of the occurrence of a SGTR accident.
    The probability of a[n] SLB accident is unaffected by the 
potential failure of a steam generator tube as this failure is not 
an initiator for a[n] SLB accident.
    The consequences of a[n] SLB accident are also not significantly 
affected by the proposed changes. During a[n] SLB accident, the 
reduction in pressure above the tubesheet on the shell side of the 
steam generator creates an axially uniformly distributed load on the 
tubesheet due to the reactor coolant system pressure on the 
underside of the tubesheet. The resulting bending action constrains 
the tubes in the tubesheet thereby restricting primary-to-secondary 
leakage below the midplane.
    Primary-to-secondary leakage from tube degradation in the 
tubesheet area during the limiting accident (i.e., a[n] SLB) is 
limited by flow restrictions resulting from the crack and tube-to-
tubesheet contact pressures that provide a restricted leakage path 
above the indications and also limit the degree of potential crack 
face opening as compared to free span indications. The primary-to-
secondary leak rate during postulated SLB accident conditions would 
be expected to be less than that during normal operation for 
indications near the bottom of the tubesheet (i.e., including 
indications in the tube end welds). This conclusion is based on the

[[Page 67750]]

observation that while the driving pressure causing leakage 
increases by approximately a factor of (two) 2, the flow resistance 
associated with an increase in tube-to-tubesheet contact pressure, 
during a[n] SLB accident, increases by approximately a factor of 
2.5. While such a leakage decrease is logically expected, the 
postulated accident leak rate could be conservatively bounded by 
twice the normal operating leak rate even if the increase in contact 
pressure is ignored. Since normal operating leakage (spiking) is 
limited to less that 0.104 gpm (150 gpd) for continued power 
operation per station operating procedure OS 1227.02, ``Steam 
Generator Tube Leak,'' the associated accident condition leak rate, 
assuming all leakage to be from lower tube sheet indications, would 
be bound by 0.208 gpm (twice normal operating leak rate). This value 
is well within the assumed accident leakage rate of 0.347 gpm 
discussed in the Seabrook Station Updated Safety Analysis Report, 
Section 15.1.5 ``Steam System Piping Failure.'' Hence it is 
reasonable to omit any consideration of inspection of the tube, tube 
end weld, bulges / overexpansions or other anomalies below 17 inches 
from the top of the hot leg tubesheet. Therefore, the consequences 
of a[n] SLB accident remain unaffected.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any [accident] previously 
evaluated.
    The proposed changes do not introduce any new equipment, create 
new failure modes for existing equipment, or create any new limiting 
single failures. Plant operation will not be altered, and all safety 
functions will continue to perform as previously assumed in accident 
analyses. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The proposed changes maintain the required structural margins of 
the steam generator tubes for both normal and accident conditions. 
Nuclear Energy Institute (NEI) 97-06, ``Steam Generator Program 
Guidelines,'' and NRC Regulatory Guide (RG) 1.121, ``Bases for 
Plugging Degraded PWR Steam Generator Tubes,'' are used as the bases 
in the development of the limited hot leg tubesheet inspection depth 
methodology for determining that steam generator tube integrity 
considerations are maintained within acceptable limits. RG 1.121 
describes a method acceptable to the NRC for meeting General Design 
Criteria (GDC) 14, ``Reactor Coolant Pressure Boundary,'' GDC 15, 
``Reactor Coolant System Design,'' GDC 31, ``Fracture Prevention of 
Reactor Coolant Pressure Boundary,'' and GDC 32, ``Inspection of 
Reactor Coolant Pressure Boundary,'' by reducing the probability and 
consequences of a SGTR. RG 1.121 concludes that by determining the 
limiting safe conditions for tube wall degradation the probability 
and consequences of a SGTR are reduced. RG 1.121 uses safety factors 
on loads for tube burst that are consistent with the requirements of 
Section III of the American Society of Mechanical Engineers (ASME) 
Code.
    For axially oriented cracking located within the tubesheet, tube 
burst is precluded due to the presence of the tubesheet. For 
circumferentially oriented cracking, Westinghouse evaluation LTR-
CDME-05-170, ``Limited Inspection of the Steam Generator Tube 
Portion Within the Tubesheet at Seabrook Generating Station,'' 
defines a length of degradation-free expanded tubing that provides 
the necessary resistance to tube pullout due to the pressure induced 
forces, with applicable safety factors applied. Application of the 
limited hot leg tubesheet inspection criteria will preclude 
unacceptable primary-to-secondary leakage during all plant 
conditions. The methodology for determining leakage provides for 
large margins between calculated and actual leakage values in the 
proposed limited hot leg tubesheet inspection depth criteria.
    Therefore, the proposed changes do not involve a significant 
reduction in any 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: M. S. Ross, Florida Power & Light Company, 
P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Section Chief: Darrell J. Roberts.

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: September 27, 2005.
    Description of amendment request: The amendments proposed by 
Southern Nuclear Operating Company would revise the Technical 
Specifications (TS) to eliminate the Power Range Neutron Flux-High 
Negative Rate Reactor Trip function, based on the approved methodology 
contained in Westinghouse Topical Report WCAP-11394-P-A, ``Methodology 
for the Analysis of the Dropped Rod Event.'' The changes will allow the 
elimination of a trip circuitry that is not credited in the Farley 
Nuclear Plant safety analysis, and which can result in an unnecessary 
reactor trip. These changes will be implemented sequentially, 
concurrent with each unit's refueling outage during which the design 
change is implemented. Additionally, this amendment request deletes TS 
Bases text associated with an unconservative local Departure from 
Nucleate Boiling Ratio.
    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?
    No. The proposed changes do not significantly increase the 
probability or consequences of an accident previously evaluated in 
the Updated Final Safety Analysis Report (UFSAR). All of the safety 
analyses have been evaluated for impact due to this change. The 
elimination of the Power Range Neutron Flux-High Negative Rate 
Reactor Trip function and the elimination of text in the TS 
[Technical Specifications] Bases for LC0 3.3.1, page B 3.3.1-1 1, 
associated with an unconservative local DNBR [departure from 
nucleate boiling ratio], does not affect the dropped RCCA [Rod 
Cluster Control Assembly] analyses nor any other analyses, since it 
is not credited in any of the safety analyses; therefore, the 
probability of an accident has not been increased. All dose 
consequences have been evaluated with respect to the proposed 
changes, there is no impact due to the proposed change, and all 
acceptance criteria continue to be met. Therefore, these changes do 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    No. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident already evaluated in 
the UFSAR. No new accident scenarios, failure mechanisms or limiting 
single failures are introduced as result of the proposed changes. 
The changes have no adverse effects on any safety-related system. 
Therefore, all accident analyses criteria continue to be met and 
these changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    No. The proposed changes do not involve a significant reduction 
in a margin of safety. The dropped RCCA(s) event does not credit the 
Power Range Neutron Flux-High Negative Rate Reactor Trip function. 
The conclusion presented in the UFSAR Section 15.2.3.3 that the DNBR 
design basis is met for a dropped RCCA(s) event remains valid for 
the proposed changes, which are based on the NRC approved 
methodology contained in CAP-11394-PA. Additionally, WCAP-11394-P-A 
indicates that the analysis for a dropped rod event envelops a 
multiple rod drop accident at high power levels, and that such an 
accident will not result in an unconservative local DNBR. All 
applicable acceptance criteria continue to be met. Therefore, the 
proposed changes do not involve a significant reduction in a margin 
of safety.


[[Page 67751]]


    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 Section Chief: Evangelos C. Marinos.

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: October 6, 2005.
    Description of amendment request: The amendments proposed by 
Southern Nuclear Operating Company (SNC) would revise the Technical 
Specifications (TS) to support a revision to the Best Estimate Loss of 
Coolant Accident (BELOCA) for Farley Nuclear Plant (FNP). The NRC 
recently approved a new Westinghouse BELOCA methodology, Automated 
Statistical Treatment of Uncertainty Method (ASTRUM). ASTRUM was 
submitted in WCAP-16009-P. The NRC issued a Safety Evaluation Report in 
a letter dated November 5, 2004. Westinghouse issued WCAP-16009-P-A in 
January 2005. SNC has completed the analysis for FNP and the enclosed 
proposed amendment is to incorporate a reference to WCAP-16009-P-A in 
TS 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, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No physical plant changes are being made as a result of using 
the Westinghouse Best Estimate Large Break LOCA [Loss of Coolant 
Accident] (BELOCA) analysis methodology. The proposed TS changes 
simply involve updating the references in TS 5.6.5.b, Core Operating 
Limits Report (COLR), to reference the Westinghouse BELOCA analysis 
methodology. The plant conditions assumed in the analysis are 
bounded by the design conditions for all equipment in the plant; 
therefore, there will be no increase in the probability of a LOCA. 
The consequences of a LOCA are not being increased, since the 
analysis has shown that the Emergency Core Cooling System (ECCS) is 
designed such that its calculated cooling performance conforms to 
the criteria contained in 10 CFR 50.46, ``Acceptance criteria for 
emergency core cooling systems for light-water nuclear power 
reactors.'' No other accident consequence is potentially affected by 
this change.
    All systems will continue to be operated in accordance with 
current design requirements under the new analysis, therefore no new 
components or system interactions have been identified that could 
lead to an increase in the probability of any accident previously 
evaluated in the Updated Final Safety Analysis Report (UFSAR). No 
changes were required to the Reactor Protection System (RPS) or 
Engineering Safety Features (ESF) setpoints because of the new 
analysis methodology.
    Therefore, it is concluded that this change does not 
significantly increase 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?
    There are no physical changes being made to the plant as a 
result of using the Westinghouse Best Estimate Large Break LOCA 
analysis methodology. No new modes of plant operation are being 
introduced. The configuration, operation and accident response of 
the structures or components are unchanged by utilization of the new 
analysis methodology. Analyses of transient events have confirmed 
that no transient event results in a new sequence of events that 
could lead to a new accident scenario. The parameters assumed in the 
analysis are within the design limits of existing plant equipment.
    In addition, employing the Westinghouse Best Estimate Large 
Break LOCA analysis methodology does not create any new failure 
modes that could lead to a different kind of accident. The design of 
all systems remains unchanged and no new equipment or systems have 
been installed which could potentially introduce new failure modes 
or accident sequences. No changes have been made to any RPS or ESF 
actuation setpoints.
    Based on this review, it is concluded that no new accident 
scenarios, failure mechanisms or limiting single failures are 
introduced as a result of the proposed changes.
    Therefore, the proposed TS 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?
    It has been shown that the analytic technique used in the 
Westinghouse Best Estimate Large Break LOCA analysis methodology 
realistically describes the expected behavior of the reactor system 
during a postulated LOCA. Uncertainties have been accounted for as 
required by 10 CFR 50.46. A sufficient number of LOCAs with 
different break sizes, different locations, and other variations in 
properties have been considered to provide assurance that the most 
severe postulated LOCAs have been evaluated. The analysis has 
demonstrated that all acceptance criteria contained in 10 CFR 50.46 
paragraph b continue to be satisfied.
    Therefore, it is concluded that this change does not involve a 
significant reduction in the margin of safety.

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

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 27, 2005.
    Description of amendment request: The proposed amendments would 
revise Technical Specifications Limiting Conditions for Operations 
3.3.1, 3.3.2, 3.3.6, and 3.3.8, by extending the Surveillance Test 
Intervals for the Reactor Protection System.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the Proposed Change Involve a Significant Increase in 
the Probability or Consequences of an Accident Previously Evaluated?
    The proposed changes to the Completion Time, bypass test time, 
and Surveillance Frequencies reduce the potential for inadvertent 
reactor trips and spurious actuations and, therefore, do not 
increase the probability of any accident previously evaluated. The 
proposed changes to the allowed Completion Time, bypass test time, 
and Surveillance Frequencies do not change the response of the plant 
to any accidents and have an insignificant impact on the reliability 
of the reactor trip system and engineered safety feature actuation 
system (RTS and ESFAS) signals. The RTS and ESFAS will remain highly 
reliable, and the proposed changes will not result in a significant 
increase in the risk of plant operation. This is demonstrated by 
showing that the impact on plant safety as measured by core damage 
frequency (CDF) is less than 1.01E-06 per year and the impact on 
large early release frequency (LERF) is less than 1.0E-07 per year. 
In addition, for the Completion Time change, the incremental 
conditional core damage probabilities (ICCDP) and incremental 
conditional large early release probabilities (ICLERP) are less than 
5.0E-08. These changes meet the

[[Page 67752]]

acceptance criteria in Regulatory Guides 1.174 and 1.177. Therefore, 
since the RTS and ESFAS will continue to perform their functions 
with high reliability as originally assumed, and the increase in 
risk a