[Federal Register Volume 71, Number 49 (Tuesday, March 14, 2006)]
[Notices]
[Pages 13169-13188]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 06-2383]


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NUCLEAR REGULATORY COMMISSION


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

Background

    Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from February 16, 2006 to March 2, 2006. The last 
biweekly notice was published on February 28, 2006 (71 FR 10071).

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,

[[Page 13170]]

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 
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, 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, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one 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, [email protected]; 
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 
[email protected]. 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, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact

[[Page 13171]]

the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or by e-
mail to [email protected].

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, 
and 3, Maricopa County, Arizona

    Date of amendments request: February 14, 2006.
    Description of amendments request: The amendments would revise 
Technical Specifications (TS) 3.6.3 to allow a blind flange to be used 
for containment isolation in each of the two flow paths of the 42 inch 
refueling purge valves in Modes 1 through 4 without remaining in TS 
3.6.3 Condition D.
    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 probability of an accident previously evaluated would not be 
affected by the proposed changes to allow the use of blind flanges 
for containment isolation in each of the two 42 inch refueling purge 
valve flow paths. The blind flanges are passive components that 
could not initiate an accident.
    The consequences of an accident previously evaluated would not 
be increased because the blind flanges would provide containment 
isolation assumed in the accident analyses instead of the 42 inch 
refueling purge valves. The blind flanges are passive devices not 
susceptible to an active failure or malfunction that could result in 
a loss of isolation or leakage that exceeds limits assumed in the 
safety analysis. The blind flanges are leak rate tested in 
accordance with the containment leakage rate testing program that is 
required by TS surveillance requirement (SR) 3.6.1.1 and TS 5.5.16. 
The blind flanges are sealed using two separate concentric O-rings 
and are leak rate tested after installation by pressurizing the 
space between the O-rings through a test connection and measuring 
the leakage. In addition, the outboard 42 inch refueling purge valve 
packing leakage is measured by pressurizing the stuffing box through 
the leak off line after flange installation and after any 
maintenance on the packing. The sum of the individual leakage rates 
is compared to the acceptance criteria. The blind flanges are 
verified to be in position at a frequency of 31 days in accordance 
with TS SR 3.6.3.3.
    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 accident from any accident previously evaluated?
    Response: No.
    A new or different kind of accident from any accident previously 
evaluated would not be created by the proposed changes to allow the 
use of blind flanges for containment isolation in each of the two 42 
inch refueling purge valve flow paths. The blind flanges are passive 
components that could not create an accident.
    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.
    No margin of safety is affected by the proposed changes to allow 
the use of blind flanges for containment isolation in each of the 
two 42 inch refueling purge valve flow paths. The blind flanges 
would provide containment isolation assumed in the accident analyses 
instead of the 42 inch refueling purge valves. The blind flanges are 
passive devices not susceptible to an active failure or malfunction 
that could result in a loss of isolation or leakage that exceeds 
limits assumed in the safety analysis. The blind flanges are leak 
rate tested in accordance with the containment leakage rate testing 
program that is required by TS SR 3.6.1.1 and TS 5.5.16. The blind 
flanges are leak rate tested after installation by pressurizing the 
space between the O-rings through a test connection and measuring 
the leakage. In addition, the outboard 42 inch refueling purge valve 
packing leakage is measured by pressurizing the stuffing box through 
the leak off line after flange installation and after any 
maintenance on the packing. The sum of the individual leakage rates 
is compared to the acceptance criteria. The blind flanges are 
verified to be in position at a frequency of 31 days in accordance 
with SR 3.6.3.3.
    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 
that review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Kenneth C. Manne, Senior Attorney, Arizona 
Public Service Company, P.O. Box 52034, Mail Station 7636, Phoenix, 
Arizona 85072-2034.
    NRC Branch Chief: David Terao.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: January 31, 2006.
    Description of amendment request: The proposed amendment would 
address an inconsistency that was inadvertently introduced during 
conversion to improved technical specifications (TSs) when ``1 per 
room'' replaced ``2'' as the required channels per trip system for the 
reactor water cleanup (RWCU) area ventilation differential 
temperature--high isolation function.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change clarifies the requirement to maintain 
isolation capability for the RWCU Area Ventilation Differential 
Temperature--High isolation instrumentation by addition of a note to 
TS 3.3.6.1 Condition B, modification of TS 3.3.6.1 Surveillance 
Requirements Notes, and by clarifying the number of instruments 
required to be available in TS Table 3.3.6.1-1, ``Primary 
Containment Isolation Instrumentation,'' Function 5.c, by the 
addition of note (d). This ensures, during surveillance testing and 
normal operation, there will always be at least one instrument 
monitoring for a small leak in all RWCU locations. No changes in 
operating practices or physical plant equipment are created as a 
result of this change. 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 type of accident from any accident previously evaluated?
    Response: No.
    The proposed change clarifies the requirement to maintain 
isolation capability for the RWCU Area Ventilation Differential 
Temperature--High isolation instrumentation by addition of a note to 
TS 3.3.6.1 Condition B, modification of TS 3.3.6.1 Surveillance 
Requirements Notes, and by clarifying the number of instruments 
required to be available in TS Table 3.3.6.1-1, ``Primary 
Containment Isolation Instrumentation,'' Function 5.c, by the 
addition of note (d). This ensures, during surveillance testing and 
normal operation, there will always be at least one instrument 
monitoring for a small leak in all RWCU locations. No physical 
change in plant equipment will result from this proposed change. 
Therefore, the proposed change does not create the possibility of a 
new or different type of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change clarifies the requirement to maintain 
isolation capability for the RWCU Area Ventilation Differential 
Temperature--High isolation

[[Page 13172]]

instrumentation by addition of a note to TS 3.3.6.1 Condition B, 
modification of TS 3.3.6.1 Surveillance Requirements Notes, and by 
clarifying the number of instruments required to be available in TS 
Table 3.3.6.1-1, ``Primary Containment Isolation Instrumentation,'' 
Function 5.c, by the addition of note (d). This ensures, during 
surveillance testing and normal operation, there will always be at 
least one instrument monitoring for a small leak in all RWCU 
locations. 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 G. Pettinari, Legal Department, 688 
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279.
    NRC Branch Chief: Timothy J. Kobetz, Acting.

Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power 
Station, Kewaunee County, Wisconsin

    Date of amendment request: January 30, 2006.
    Description of amendment request: The license amendment request 
would modify the currently approved radiological accident analyses 
(RAA) and associated Technical Specifications (TS) to account for the 
difference between the control room emergency zone (CREZ) unfiltered 
in-leakage (UFI) assumed in the current RAA and the CREZ UFI that was 
measured during testing.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. There are no system, structural, or component (SSC) 
alterations due to these changes. The radiological accident analyses 
inputs modified by this request are not accident initiators and do 
not affect the frequency of occurrence of previously analyzed 
transients.
    The radiological accident analyses have demonstrated acceptable 
results using the revised inputs for all affected accidents. 
Further, the proposed changes do not alter or prevent the ability of 
structures, systems or components to perform their intended function 
to mitigate the consequences of accidents previously evaluated in 
the Updated Safety Analysis Report.
    Therefore, the changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. There are no physical changes to the plant SSCs and there is 
no adverse impact on component or system interactions due to the 
proposed changes. The modes of operation of the plant remain 
unchanged and the design functions of all the safety systems remain 
in compliance with the applicable safety analysis acceptance 
criteria. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The radiological accident analysis inputs modified by this 
request were incorporated into the revised radiological accident 
analyses. The revised radiological analyses satisfy all applicable 
acceptance criteria. There is no adverse effect on plant safety due 
to this proposed license amendment. Therefore, the 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: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    Acting NRC Branch Chief: T. Kobetz.

Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power 
Station, Kewaunee County, Wisconsin

    Date of amendment request: February 6, 2006.
    Description of amendment request: The proposed amendment adds a 
license condition to extend certain Technical Specification (TS) 
surveillance test intervals on a one-time basis to account for the 
effects of an extended forced outage in the spring of 2005.
    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 requested action is a one-time extension to the performance 
interval of a limited number of TS surveillance requirements. The 
performance of these surveillances, or the failure to perform these 
surveillances, is not a precursor to an accident. Performing these 
surveillances or failing to perform these surveillances does not 
affect the probability of an accident. Therefore, the proposed delay 
in performance of the surveillance requirements in this amendment 
request does not increase the probability of an accident previously 
evaluated.
    A delay in performing these surveillances does not result in a 
system being unable to perform its required function. In the case of 
this one-time extension request, the relatively short period of 
additional time that the systems and components will be in service 
before the next performance of the surveillance will not affect the 
ability of those systems to operate as designed. Therefore, the 
systems required to mitigate accidents will remain capable of 
performing their required function. No new failure modes have been 
introduced because of this action and the consequences remain 
consistent with previously evaluated accidents. Therefore, the 
proposed delay in performance of the surveillance requirements in 
this amendment request does not involve a significant increase in 
the consequences of an accident.
    Therefore, operation of the facility in accordance with the 
proposed license amendment would not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment does not involve a physical alteration of 
any system, structure, or component (SSC) or a change in the way any 
SSC is operated. The proposed amendment does not involve operation 
of any SSCs in a manner or configuration different from those 
previously recognized or evaluated. No new failure mechanisms will 
be introduced by the one-time surveillance requirement deferrals 
being requested.
    Thus, the proposed amendment does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment is a one-time extension of the 
performance interval of a limited number of TS surveillance 
requirements. Extending these surveillance requirements does not 
involve a modification of any TS Limiting Conditions for Operation. 
Extending these surveillance requirements does not involve a change 
to any limit on accident consequences specified in the license or 
regulations. Extending these surveillance requirements does not 
involve a change to how accidents are mitigated or a significant 
increase in the consequences of an accident. Extending these 
surveillance requirements does not involve a change in a methodology 
used to evaluate consequences of an accident. Extending these 
surveillance requirements does not involve a change in any operating 
procedure or process.

[[Page 13173]]

    The instrumentation and components involved in this request have 
exhibited reliable operation based on the results of the most recent 
performance of their 18-month surveillance requirements.
    Based on the limited additional period of time that the systems 
and components will be in service before the surveillances are next 
performed, as well as the operating experience that these 
surveillances are typically successful when performed, it is 
reasonable to conclude that the margins of safety associated with 
these surveillance requirements will not be affected by the 
requested extension.
    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: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    Acting NRC Branch Chief: T. Kobetz.

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

    Date of amendment request: December 19, 2005.
    Description of amendment request: The amendment proposes to revise 
the Technical Specifications (TS) to make the temporary changes to TS 
Table 3.3.8.1-1, previously approved by Amendment No. 147, permanent. 
TS Table 3.3.8.1-1 would be revised to delete the temporary note, 
correct the number of Required Channels per Division for the Loss of 
Power (LOP) time delay functions, and delete the requirement to perform 
Surveillance Requirement (SR) 3.3.8.1.2, the monthly Channel Functional 
Test, on certain LOP time delay functions.
    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 regarding the number of required channels 
per division for the LOP time delay functions are administrative in 
nature. The changes do not alter the instrumentation design or their 
physical configuration, and will not affect their operation or 
manner of control. The proposed changes correct an inconsistency 
between a TS Table and the RBS [River Bend Station, Unit 1] design 
basis. The TS required number of voltage sensors per division and 
associated channel components that monitor voltage conditions and 
provide the 4.16 kV bus undervoltage protection are unchanged.
    The exclusion of the time delay functions from the monthly 
Channel Functional Test is proposed because the test creates a loss 
of function for the LOP instrumentation and is, therefore, 
undesirable during unit operations. The test also introduces the 
potential for an unintended plan transient, so the elimination of 
the requirement reduces the potential for such transients.
    The channel functional test will continue to be performed every 
31 days for the sensor channels. In addition, the LOP time delay 
functions will continue to be functionally tested and calibrated 
every 18 months as required by SR 3.3.8.1.3 and SR 3.3.8.1.4. 
Therefore, the required LOP instrumentation will continue to be 
tested in a manner and at a frequency necessary to provide 
confidence that the instrumentation can perform its intended safety 
function.
    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 changes do not alter the instrumentation design or their 
physical configuration, and will not affect their operation or 
manner of control. The proposed TS changes do not introduce any new 
failure mechanisms, malfunctions, or accident initiators not 
considered in the design and licensing bases.
    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 have no affect on any safety analysis 
assumptions or methods of performing safety analyses. The changes do 
not adversely affect system OPERABILITY or design requirements and 
the equipment continues to be tested in a manner and at a frequency 
necessary to provide confidence that the equipment can perform its 
intended safety functions. [Regulation] 10 CFR 50.36(c)(3) requires 
the TS to include Surveillance Requirements relating to test, 
calibration, or inspection to assure that the necessary quality of 
systems and components is maintained, that facility operation will 
be within safety limits, and that the limiting conditions for 
operation will be met. The channel functional test will continue to 
be performed every 31 days for the sensor channels. In addition, the 
LOP time delay functions will continue to be functionally tested and 
calibrated every 18 months as required by SR 3.3.8.1.3 and SR 
3.3.8.1.4.
    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: Mark Wetterhahn, Esq., Winston & Strawn, 
1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: David Terao.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: January 26, 2006.
    Description of amendment request: The proposed amendment will 
modify Technical Specification (TS) requirements to support the 
implementation of Average Power Range Monitor (APRM), Rod Block 
Monitor, TS/Maximum Extended Operating Domain (ARTS/MEOD).
    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 revise thermal limit structure employed to 
comply with TS Section 3.2 LCOs [limiting conditions for operation]. 
The proposed changes will replace the flow-biased APRM scram and rod 
block trip setdown requirements with power and flow dependent 
adjustments to the Minimum Critical Power Ratio (MCPR) and Maximum 
Average Planar Linear Heat Generation Rate (MAPLHGR) or Linear Heat 
Generation Rate (LHGR) thermal limits. The adjustments to the 
thermal limits have been determined using NRC approved analytical 
methods as required by Technical Specifications 5.6.5.b and topical 
reports as specified in the Core Operating Limits Report (COLR). The 
proposed changes will not affect any accident initiating mechanism. 
Adjustments to thermal limits will be determined using NRC approved 
methodologies. The power and flow dependent adjustments will ensure 
that the MCPR safety limit will not be violated as a result of any 
anticipated operational occurrence (AOO), that the fuel thermal and 
mechanical design bases will be maintained, and that the 
consequences of the postulated loss of coolant accident (LOCA) will 
remain within acceptable limits. There are no changes to radioactive 
source terms or release pathways. Operation within the expanded 
operating domain has been evaluated and the affect on plant 
accidents was found to be

[[Page 13174]]

within acceptable parameters. The proposed changes do not result in 
any significant change in the availability of logic systems or 
safety-related systems themselves. Required protective functions 
will be maintained. The proposed changes do not degrade plant 
design, operation, or the performance of any safety system assumed 
to function in the accident analysis.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    2. [Does the proposed change] create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes do not introduce any new accident 
initiators or failure mechanisms because the changes and the affects 
on existing structures, systems and components have been evaluated 
and found to not have any adverse affects. The proposed changes 
eliminate the requirement for setdown of the flow-biased APRM scram 
and rod block trip setpoints or APRM adjustments under specified 
conditions and will substitute adjustments to the MCPR and MAPLHGR 
or LHGR thermal limits. Because the thermal limits will continue to 
be met, no transient event will escalate into a new or different 
type of accident due to the initial starting conditions permitted by 
the adjusted thermal limits.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident than those previously evaluated.
    3. [Does the proposed change] involve a significant reduction in 
a margin of safety?
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. There is no affect on the conclusions of 
any safety analysis. Replacement of the APRM setpoint requirement 
with power and flow dependent adjustments to the MCPR and MAPLHGR or 
LHGR thermal limits will continue to ensure that margins to the fuel 
cladding Safety Limit are preserved during operation at other than 
rated conditions. The fuel cladding safety limit will not be 
violated as a result of any anticipated operational occurrence. The 
flow and power dependent adjustments will be determined using NRC 
approved methodologies. The flow and power dependent adjustments 
will also ensure that all fuel thermal-mechanical design bases shall 
remain within the licensing limits. The proposed changes do not 
involve any increase in calculated off-site dose consequences. 
Operability of protective instrumentation and the associated systems 
is assured, and performance of equipment will not be significantly 
affected.
    Therefore, there is no significant reduction in the margin of 
safety as a result of the proposed changes.

    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 Branch Chief: Richard J. Laufer.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant (JAFNPP), Oswego County, New York

    Date of amendment request: January 26, 2006.
    Description of amendment request: The proposed license amendment 
replaces the existing Reactor Vessel Material Surveillance Program with 
the Boiling Water Reactor Vessel and Internals Project (BWRVIP) 
Integrated Surveillance Program (ISP).
    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 to the licensing basis continues to assure 
that applicable regulatory requirements are met and the same 
assurance of reactor pressure vessel integrity continues to be 
provided. The proposed change to the License and licensing basis 
follow the NRC Safety Evaluation approving the implementation of the 
ISP. The proposed change ensures that the reactor pressure vessel 
will continue to be operated within the design, operational, and 
testing limits.
    The proposed change does not modify the reactor coolant pressure 
boundary, (i.e., there are no changes in operating pressure, 
materials, or seismic loading). The proposed change does not 
adversely affect the integrity of the reactor coolant pressure 
boundary such that its function in the control of radiological 
consequences is affected.
    Therefore, the proposed change does not involve a significant 
increase in 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.
    The proposed change does not involve a modification to the 
design of plant structures, systems, or components. Thus, no new 
modes of operation are introduced by the proposed change. The 
proposed change will not create any failure mode not bounded by 
previously evaluated accidents.
    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 implementation of ISP has been previously approved 
by the NRC and found to provide an acceptable alternative to plant-
specific reactor vessel material surveillance programs. Operation of 
JAFNPP within the program ensures that the reactor vessel materials 
will continue to behave in a non-brittle manner, thereby preserving 
the original safety design bases.
    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 Branch Chief: Richard J. Laufer.

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: February 15, 2006.
    Description of amendment request: The proposed change will 
specifically credit the measurement tank weir flow instrumentation for 
the containment fan cooler condensate flow monitoring system in place 
of the one containment fan cooler condensate flow switch currently 
required by Technical Specification 3.4.5.1, ``Reactor Coolant System 
Leakage--Leakage Detection 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 Reactor Coolant System (RCS) leakage detections systems are 
passive monitoring systems; therefore, the proposed changes do not 
affect reactor operations or accident analyses and have no 
radiological consequences. The change maintains conservative 
restrictions on RCS leakage detections systems consistent with 
Regulatory Guide 1.45 [``Reactor Coolant Pressure Boundary Leakage 
Detection Systems''] and 10 CFR [Part] 50, Appendix A, General 
Design Criteri[on] 30.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

[[Page 13175]]

    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 introduces no new mode of plant operation or 
any plant modification. The RCS leakage detection instrumentation is 
not part of plant control instruments or engineered safety feature 
actuation circuits but is used solely for monitoring purposes. The 
change does not vary or affect any plant operating condition or 
parameter.
    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.
    There will be no adverse affects on margins of safety since more 
stringent requirements will be applied to the third method (CFC 
[Containment Fan Cooler] condensate flow monitoring) of detecting 
RCS leakage. The third required RCS leakage detection method will 
now be capable of detecting a one gallon per minute leak within one 
hour.
    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: N. S. Reynolds, Esquire, Winston & Strawn, 
1700 K Street NW., Washington, DC 20006-3817.
    NRC Branch Chief: David Terao.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 
and 2, Will County, Illinois

    Date of amendment request: October 3, 2005.
    Description of amendment request: The proposed amendments would 
revise the reactor coolant system pressure and temperature limits 
report (PTLR) requirements. Specifically, the amendment would revise 
the TS Section 1.1, ``Definitions,'' description of the PTLR by 
deleting reference to specifications containing limits in the PTLR; (2) 
revise the administrative controls TS 5.6.6, ``Reactor Coolant System 
(RCS) Pressure and Temperature Limits Report (PTLR),'' by requiring the 
NRC approval documents to be identified by date and topical reports to 
be identified by number and title in accordance with Industry/Technical 
Specification Task Force (TSTF) Standard Technical Specification Change 
Traveler, TSTF-419; ``Revise PTLR Definition and References in ISTS 
5.6.6, RC PTLR,'' and (3) add Westinghouse Electric Company, LLC, WCAP-
16143, ``Reactor Vessel Closure Head/Vessel Flange Requirements 
Evaluation for Byron/Braidwood Units 1 and 2,'' to the list of 
analytical methods provided in TS 5.6.6.
    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 proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to the definition of PTLR is considered to 
be an editorial change because the requirements of TS 5.6.6 continue 
to specify the Limiting Conditions for Operation that address 
operation within the P-T [pressure temperature] limits.
    The proposed changes to reference only the Topical Report number 
and title do not alter the use of the analytical methods used to 
determine the pressure temperature (P-T) limits or Low Temperature 
Overpressure Protection (LTOP) System setpoints that have been 
reviewed and approved by the NRC. This method of referencing Topical 
Reports would allow the use of current Topical Reports to support 
limits in the PTLR without having to submit an amendment to the 
operating license provided there is no change to the approved 
methodology. TS 5.6.6.b requires that the analytical methods used to 
determine the P-T limits be those previously reviewed and approved 
by the NRC. Implementation of revisions to Topical Reports would 
still be reviewed in accordance with 10 CFR 50.59, ``Changes, tests 
and experiments,'' and where required receive NRC review and 
approval.
    The use of WCAP-16143, following approval by the NRC, for 
generation of P-T limits will continue to ensure that reactor 
pressure vessel integrity is maintained under all conditions.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, or 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not alter or 
prevent the ability of structures, systems, and components (SSCs) 
from performing their intended function to mitigate the consequences 
of an initiating event within the assumed acceptance limits. The 
proposed changes do not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of an accident previously evaluated. 
The proposed changes do not increase the types or amounts of 
radioactive effluent that may be released offsite, nor significantly 
increase individual or cumulative occupational/public radiation 
exposures. The proposed changes are consistent with safety analysis 
assumptions and resultant consequences.
    Based on the above discussion, the proposed changes do 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 to the definition of PTLR is considered to 
be an editorial change because the requirements of TS 5.6.6 continue 
to specify the Limiting Conditions for Operation that address 
operation within the P-T limits.
    The proposed changes to reference only the Topical Report Number 
and title do not alter the use of the analytical methods used to 
determine the P-T limits or LTOP setpoints that have been reviewed 
and approved by the NRC. This method of referencing Topical Reports 
would allow the use of current Topical Reports to support limits in 
the PTLR without having to submit an amendment to the operating 
license provided there is no change to the approved methodology. TS 
5.6.6.b requires that the analytical methods used to determine the 
P-T limits be those previously reviewed and approved by the NRC. 
Implementation of revisions to Topical Reports would still be 
reviewed in accordance with 10 CFR 50.59 and where required receive 
NRC review and approval.
    The use of WCAP-16143, following approval by the NRC, for 
generation of P-T limits will continue to ensure that reactor 
pressure vessel integrity is maintained under all conditions.
    The proposed changes will allow the use of a new NRC-approved 
methodology for the calculation of P-T limits. However, the changes 
do not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) and do not introduce 
a new mode of plant operation. Safety functions associated with P-T 
limits and LTOP setpoints will continue to function as previously 
assumed in accident analyses.
    Based on this evaluation, 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 to the definition of PTLR is considered to 
be an editorial change because the requirements of TS 5.6.6 continue 
to specify the Limiting Conditions for Operation that address 
operation within the P-T limits. The proposed changes to reference 
only the Topical Report Number and title do not alter the use of the 
analytical methods used to determine the P-T limits or LTOP 
setpoints that have been reviewed and approved by the NRC. This 
method of referencing Topical Reports would allow the use of current 
Topical Reports to support limits in the PTLR without having to 
submit an amendment to the operating license provided there is no 
change to the approved

[[Page 13176]]

methodology. TS 5 .6.6.b requires that the analytical methods used 
to determine the P-T limits be those previously reviewed and 
approved by the NRC. Implementation of revisions to Topical Reports 
would still be reviewed in accordance with 10 CFR 50.59 and where 
required receive NRC review and approval.
    The P-T limits provide assurance that the reactor pressure 
vessel is maintained. The use of WCAP-16143, following approval by 
the NRC, for generation of P-T limits will continue to ensure that 
reactor pressure vessel integrity is maintained under all 
conditions.
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. Changes to setpoints at which protective 
actions are initiated that are allowed by the use of WCAP-16143 are 
evaluated in accordance with 10 CFR 50.59 and where required receive 
NRC review and approval. Sufficient equipment remains available to 
actuate upon demand for the purpose of mitigating an analyzed event.
    Based on this evaluation, 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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. J. Bradley Fewell, Assistant General 
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett 
Square, PA 19348.
    NRC Branch Chief: Mindy Landau, Acting.

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

    Date of amendment requests: January 25, 2006.
    Description of amendment requests: The proposed amendments would 
revise Technical Specification (TS) 1.1, ``Definitions,'' and TS 
3.4.16, ``RCS Specific Activity.'' The proposed amendments would 
replace the current TS 3.4.16 limit on reactor coolant system (RCS) 
gross specific activity with a new limit on RCS noble gas specific 
activity. The noble gas specific activity limit would be based on a new 
DOSE EQUIVALENT XE-133 definition (corresponding to the Xenon-133 
isotope) that would replace the current--AVERAGE DISINTEGRATION ENERGY 
definition. In addition, the current DOSE EQUIVALENT I-131 definition 
(corresponding to the Iodine-131 isotope) would be revised to allow the 
use of alternate, NRC-approved thyroid dose conversion factors.
    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 to add a new thyroid dose conversion factor 
reference to the definition of DOSE EQUIVALENT I-131, eliminate the 
definition of E--AVERAGE DISINTEGRATION ENERGY, add a new definition 
of DOSE EQUIVALENT XE-133, replace the Technical Specification (TS) 
3.4.16 limit on reactor coolant system (RCS) gross specific activity 
with a limit on noble gas specific activity in the form of a 
Limiting Condition for Operation (LCO) on DOSE EQUIVALENT XE-133, 
replace TS Figure 3.4.16-1 with a maximum limit on DOSE EQUIVALENT 
I-131, extend the Applicability of LCO 3.4.16, and make 
corresponding changes to TS 3.4.16 to reflect all of the above are 
not accident initiators and have no impact on the probability of 
occurrence for any design[-]basis accidents.
    The proposed changes will have no impact on the consequences of 
a design[-basis accident because they will limit the RCS noble gas 
specific activity to be consistent with the values assumed in the 
radiological consequence analyses. The changes will also limit the 
potential RCS iodine concentration excursion to the value currently 
associated with full power operation, which is more restrictive on 
plant operation than the existing allowable RCS iodine specific 
activity at lower power levels.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not alter any physical part of the plant 
nor do they affect any plant operating parameters besides the 
allowable specific activity in the RCS. The changes that impact the 
allowable specific activity in the RCS are consistent with the 
assumptions assumed in the current radiological consequence 
analyses.
    Therefore, the proposed changes do not create the possibility of 
a new or different accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The acceptance criteria related to the proposed changes involve 
the allowable control room and offsite radiological consequences 
following a design[-]basis accident. The proposed changes will have 
no impact on the radiological consequences of a design[-]basis 
accident because they will limit the RCS noble gas specific activity 
to be consistent with the values assumed in the radiological 
consequence analyses. The changes will also limit the potential RCS 
iodine specific activity excursion to the value currently associated 
with full power operation, which is more restrictive on plant 
operation than the existing allowable RCS iodine specific activity 
at lower power levels.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Branch Chief: David Terao.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 
No. 1, Fairfield County, South Carolina

    Date of amendment request: October 28, 2005.
    Description of amendment request: The amendment would revise the 
Virgil C. Summer Nuclear Station (VCSNS) Technical Specifications (TS) 
TS 3.8.1 to incorporate changes implementing requirements for an 
Alternate AC (AAC) power supply.
    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.
    The proposed change revised two action statements and relocated 
a surveillance requirement. The first AOT [allowable outage time] 
extension permits one EDG [emergency diesel generator] to be 
inoperable for up to 14 days, but the AAC [alternate alternating 
current] source will have to be available. This proposed change will 
be primarily used for scheduled preventative maintenance while the 
plant is online. If used for corrective maintenance, the AAC source 
will have to be capable of providing power within one hour, 
otherwise the existing 72-hour AOT would apply. This assures that 
adequate power remains available to the ESF buses to enable the 
plant to safely shut down, maintain a safe shutdown condition, and/
or mitigate the effects of a design basis accident.
    The second AOT extension provides an additional two hours to 
complete the

[[Page 13177]]

verification of supported equipment for operability. This additional 
time allows for a planned and systematic approach to performing this 
verification. Since there are other more immediate ways for the 
control room staff to be notified of the inoperable status of ESF 
[engineered safety feature] equipment, (annunciators, BISI, status 
lights), the TS requirement is not critical in knowing the status of 
the plant. Should some equipment be discovered inoperable, the 
extended AOT provides for some opportunity to restore the status to 
operable.
    The deletion of a surveillance requirement that requires 
performing a vendor recommended maintenance at a specific frequency 
does not impact the ability of the EDG to perform its intended 
function for the mission time assumed in the accident analysis. EDG 
maintenance will continue to be performed and controlled under 
station procedures. The risk associated with the maintenance will be 
assessed under the provisions of 10 CFR 50.65 [Requirements for 
monitoring the effectiveness of maintenance at nuclear power 
plants], section (a) 4. The TS frequency was initially established 
to coincide with refueling outages, the only time that one EDG could 
be inoperable for any extended time. However, multiple plants have 
extended the time between refueling outages to 24 months with no 
discernable impact on reliability or availability. In addition, the 
Fairbanks-Morse diesel engine owners group has evaluated the 
maintenance requirements and determined that the TS required 
frequency should be based on performance and inspection results, not 
an arbitrary period that coincides with the best opportunity to 
perform the work. The Maintenance Rule requires evaluation for 
additional corrective actions and increased monitoring for scoped 
systems if the reliability and/or availability fall below pre-
established criteria. This approach ensures appropriate actions in a 
timely manner are taken to ensure that equipment relied upon for 
accident mitigation is available when required.
    There are no changes in operational limits or physical design of 
the onsite electric power systems. The proposed changes do not 
change the function or operation of plant equipment or affect the 
response of the equipment if called upon to operate. The EDGs are 
not the initiators of previously evaluated accidents. The EDGs are 
designed to mitigate the consequences of accidents. The risk 
informed assessment that was performed concluded that the increase 
in plant risk is small and consistent with the guidance in 
Regulatory Guide 1.174, [``An Approach for Using Probabilistic Risk 
Assessment in Risk Informed Decisions on Plant-Specific Changes to 
the licensing Basis'']. This assessment considers the possibility of 
an accident occurring during the extended period that the EDG would 
be unavailable. The proposed changes allow for additional 
operational flexibility and will not cause a significant increase in 
the probability or consequences of an accident previously evaluated. 
In actuality, the installation and availability of the AAC will have 
an overall net reduction in core damage frequency.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No.
    The proposed change to extend the EDG AOT to 14 days is based 
upon the installation of an AAC power source and the significant 
reduction in core damage frequency that results. There are no 
significant changes in installed plant equipment or operation of 
safety related equipment. The accident analysis considered the 
credible accidents and bounded those that apply.
    The installation of the AAC and the extended AOT for one EDG to 
be inoperable remain bounded by previous evaluations.
    The AOT extension to provide additional time to perform the 
redundant equipment verification is based on the other methods 
available for the Control Room staff to be made aware of a change in 
ESF equipment status and the safety benefit of performing this 
verification in an unhurried manner. This verification has been 
extended by other plants, both those who have converted to ITS and 
those that have not. No plant modifications are required and 
operator training is unaffected. The verification process does not 
utilize any new or complex software and any new accident is bounded 
by a Loss of Site Power or Station Blackout analysis.
    The deletion of a surveillance requirement to perform the 
manufacturer's recommended inspection and maintenance is based on 
the recommendations from the vendor and the Fairbanks Morse owners 
group. The recommendation is to continue to perform the inspections 
and maintenance but the frequency should not be based on the 
refueling outage frequency. The effectiveness of the maintenance 
will be assured through monitoring under the Maintenance Rule 
program which would require evaluation and corrective actions should 
the EDG not meet its performance criteria for reliability and 
availability.
    The EDG performs a function of supplying power when the normal 
ESF sources are unavailable. This is a function that mitigates the 
effects of the event and the proposed changes cannot cause the 
possibility of an accident that was not previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    No.
    The proposed change to extend the EDG AOT to 14 days from the 
current 72 hours will assure that an alternative source of power for 
the ESF onsite distribution system is available and ready. The AAC 
and interfacing equipment are designed to maintain independence and 
separation, particularly during faulted conditions. The plant 
equipment will continue to respond per the design and analysis. The 
performance capability of the EDGs will not be affected. 
Installation of the AAC will have a net reduction in the core damage 
frequency. In addition, administrative controls will ensure that 
there are adequate compensatory measures that can and will be taken 
during extended EDG maintenance activities to reduce overall risk.
    The AOT extension to provide additional time to perform the 
redundant equipment verification for operability verification allows 
some time to discover a problem and make a minor repair prior to 
placing the plant in a shutdown transient. The types of corrective 
or preventative maintenance associated with an EDG will not change. 
Plant operating and emergency procedures will be enhanced with 
guidance on when to use the AAC and how to connect up to the ESF 
bus.
    The deletion of the periodic EDG inspection per the vendor's 
recommendation at a proscribed frequency provides significant 
flexibility in when to schedule the inspection and preventative 
maintenance. The activities would still be performed but the 
frequency would be based on equipment performance and owners group 
recommendation. The plant analysis only considers the availability 
of the EDG. The TS surveillances that assure the EDG remains 
operable remain in place at their current frequencies and the 
maintenance requirement will assure that the EDG receives sufficient 
maintenance to remain operable.
    Since the operation of the plant remains largely unaffected and 
the EDG or the AAC will supply power to the ESF equipment as needed, 
there is no 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: J. Hamilton Hagood, Jr., South Carolina 
Electric & Gas Company, Post Office Box 764, Columbia, South Carolina 
29218.
    NRC Section Chief: Evangelos C. Marinos.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 
No. 1, Fairfield County, South Carolina

    Date of amendment request: November 29, 2005.
    Description of amendment request: The proposed amendment would add 
requirements to TS 3/4.7.1.2 to assure continued operability of the 
Emergency Feedwater (EFW) System based on LER 1998-004-00, by including 
the newly installed six emergency feedwater system automatic isolation 
valves into the Surveillance Requirements to assure the capability for 
automatic isolation of EFW in the event of a faulted steam generator.
    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

[[Page 13178]]

consequences of an accident previously evaluated?
    No.
    The proposed change addresses necessary changes to the VCSNS 
[Virgil C. Summer Nuclear Station] Technical Specification (TS) 
4.7.1.2.b and 4.7.1.2.c.2 associated with the installation of six 
new automatic isolation valves in the EF[W] system.
    The only Final Safety Analysis Report (FSAR) analyzed accident 
for which the EF[W] system could contribute as an initiator would be 
minor secondary line break, as described in Section 15.3.2. The 
addition of isolation valves in the EF[W] piping to the steam 
generators [SGs] will not increase the likelihood of a pipe break, 
since the addition will be in accordance with the same codes and 
standards as the corresponding, existing portions of the system. 
Piping stress analyses have demonstrated the addition of these 
valves does not result in the need to postulate any additional pipe 
breaks.
    The accidents analyzed in the FSAR, which rely on EF[W system] 
to mitigate consequences, are loss of normal feedwater, loss of off-
site power, and major secondary system pipe ruptures. The addition 
of these automatic isolation valves will eliminate the need for 
operator action to manually close a flow control valve in response 
to a major secondary system line break. The elimination of operator 
manual action is accomplished by the addition of a new pneumatically 
operated isolation valve in series with each of the six existing 
flow control valves.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    No.
    This proposed change does not result in changes to actual 
operating pressures, flow rates, flow paths, or system interfaces. 
There are no alterations to system operability requirements. The 
existing system alarm set points are not affected, neither is the 
information available to the operators. The addition of six new 
isolation valves will not change system design criteria and the 
surveillance testing will be the same as for the existing flow 
control valves.
    This change does not introduce any new or different kind of 
failure mechanisms or limiting single failures. Piping analysis has 
concluded that no new pipe break locations or break sizes will 
result from this change. Equipment protection features are not 
impacted, the frequency of pump and valve operation remains the 
same. Independence and redundancy are actually improved.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No.
    The design basis for the EF[W] system is to assure the required 
flow and pressure to remove decay heat from the core under the worst 
postulated conditions. An additional function of the system is to 
isolate flow to a faulted SG within the time assumed in the safety 
analysis. The proposed change eliminates the need for operators to 
take actions to manually close the flow control valves in the event 
of a single failure.
    The proposed change will create a surveillance requirement for 
the new isolation valves that is the same as the existing flow 
control valves. The acceptance criteria will assure the operability 
of these valves. The design and installation of these isolation 
valves will maintain the requirements for independence, redundancy, 
separation and testability. The margins assumed in the safety 
analysis will be enhanced by this proposed change. Due to the 
automatic isolation capability, additional water will be available 
for the intact SGs and a reduced mass will be available to be 
released into the containment building. No credible single failure 
will be capable of preventing isolation of a faulted SG upon a high 
flow signal.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: J. Hagood Hamilton, South Carolina Electric 
& Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Branch Chief: Evangelos C. Marinos.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 
No. 1, Fairfield County, South Carolina

    Date of amendment request: November 29, 2005.
    Description of amendment request: This amendment revises Technical 
Specification (TS) 6.9.1.5 and TS 6.9.1.10 by eliminating the 
requirements to submit monthly operating reports and occupational 
radiation exposure reports. This consolidated line item improvement 
process (CLIIP) TS change was noticed in the Federal Register on June 
23, 2004, (69 FR 35067). In addition, the TSs are revised beyond the 
scope of the CLIIP by the deletion of the TS 6.9.15 requirement to 
report exceedence of coolant specific activity limits and an 
administrative change to a TS index page.
    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:

    SCE&G has reviewed the proposed no significant hazards 
consideration determination published on June 23, 2004 (69 FR 35067) 
as part of the CLIIP. SCE&G has concluded that the proposed 
determination presented in the notice is applicable to the VCSNS, 
and the determination is hereby incorporated by reference to satisfy 
the requirements of 10 CFR 50.91(a).
    The deletion of the additional paragraph in 6.9.1.5 is beyond 
the scope of the CLIIP and as such is beyond the scope of the no 
significant hazards consideration determination published on June 
23, 2004. Therefore the following evaluation has been performed.
    In accordance with the criteria set forth in 10 CFR 50.92, SCE&G 
has evaluated the proposed beyond scope Technical Specification 
change and determined it does not represent a significant hazards 
consideration. The following is provided to support this conclusion.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No.
    The proposed change is the deletion of a paragraph in the 
administrative controls section of the facility Technical 
Specifications. The paragraph identifies required information that 
was to be provided in a report to the staff in the event where the 
RCS specific activity exceeded TS limits. This report has been found 
to be un-necessary due to reporting requirements located in 10 CFR 
50.73 (exceeding a TS limit). Additionally, the TS limits are set 
such that there is very little risk to the health and safety of the 
public. Before the condition became significant, the NRC would have 
been notified due to the 10 CFR 50.73 requirement to report 
significant degradations in a principal fission product barrier.
    Deletion of an administrative controls paragraph that provides 
reporting requirements is not a precursor to an accident. No changes 
are being proposed to any installed plant equipment or procedures. 
The operating philosophy is unaffected and training is not impacted. 
Therefore, the proposed change does not involve a significant 
increase in 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?
    No.
    The proposed change is the deletion of a paragraph that was 
inserted per the guidance of Generic Letter 85-19. The staff was 
concerned that the reporting requirements prior to that time were 
too restrictive and relaxed them through the Generic Letter. Since 
that time, it was determined that specific reporting could be 
performed via requirements in 10 CFR 50.73. Exceeding the TS limit 
is now an uncommon condition as proper fuel management and 
fabrication techniques should preclude approaching the TS limit.
    Revising or even deleting a reporting requirement in the 
facility TS will not impact

[[Page 13179]]

how the plant is operated, how data is evaluated, or what 
instructions are located in operating and emergency procedures. No 
new equipment is being installed and no plant modifications are 
resulting from this proposed change. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    No.
    The proposed change to delete some specific reporting 
requirements in the Administrative Controls section of TS has no 
impact on any plant evaluation or analysis. No plant setpoints are 
impacted; no alarm or annunciator functions are affected. This 
change has been approved for other plants. 10 CFR 50.73 will still 
require reporting the condition should it ever occur. 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: J. Hagood Hamilton, South Carolina Electric 
& Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: Evangelos C. Marinos.

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: September 19, 2005.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Limiting Conditions for Operation 
(LCO) 3.3.1, ``Reactor Trip system (RTS) Instrumentation'' and TS 
Surveillance Requirements (SR) 3.2.4.2, ``Quadrant Power Tilt Ration 
(QPTR)'' to avoid confusion as to when a flux map for QPTR is required.
    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.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, or 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not alter or 
prevent the ability of structures, systems, and components (SSCs) 
from performing their intended function to mitigate the consequences 
of an initiating event within the assumed acceptance limits. The 
proposed changes do not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of an accident previously evaluated. 
Further, the proposed changes do not increase the types or amounts 
of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures. The proposed changes are consistent with safety 
analysis assumptions and resultant consequences. Therefore, the 
proposed changes do not 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 previously evaluated?
    No.
    The proposed changes do not result in a change in the manner in 
which the RTS and ESFAS provide plant protection. The RTS and ESFAS 
will continue to have the same set points after the proposed changes 
are implemented. There are no design changes associated with the 
license amendment.
    The changes do not involve a physical alteration of the plant 
(i.e., no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. In addition, 
the changes do not impose any new or different requirements or 
eliminate any existing requirements. The changes do not alter 
assumptions made in the safety analysis. The proposed changes are 
consistent with the safety analysis assumptions and current plant 
operating practice.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No.
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not impacted by these changes. Redundant RTS and ESFAS trains 
are maintained, and diversity with regard to the signals that 
provide reactor trip and engineered safety features actuation is 
also maintained. All signals credited as primary or secondary, and 
all operator actions credited in the accident analyses will remain 
the same. The proposed changes will not result in plant operation in 
a configuration outside the design basis.
    Therefore, the proposed changes do not involve a significant 
reduction in 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: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Branch Chief: Evangelos C. Marinos.

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station, Units 1 and 2, Somervell County, Texas

    Date of amendment request: December 16, 2005.
    Brief description of amendments: The proposed change would revise 
Technical Specifications (TSs) 3.3.2, ``ESFAS [Engineered Safety 
Features Actuation System] Instrumentation''; 3.5.2, ``ECCS [Emergency 
Core Cooling System]--Operating''; and 3.6.7, ``Spray Additive 
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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    None of the changes impact the initiation or probability of 
occurrence of any accident.
    The consequences of accidents evaluated in the FSAR [Final 
Safety Analysis Report] that could be affected by this proposed 
change are those involving the pressurization of the containment and 
associated flooding of the containment and recirculation of this 
fluid within the ECCS or the Containment Spray System (e.g., LOCAs 
[loss-of-coolant accidents]).
    Although the water level in the containment flood plain will be 
higher at the start of ECCS switchover, the maximum water levels 
observed for the duration of the accident are unchanged by the 
nominal setpoint changes.
    The increase in the minimum water delivered to containment by 
the RWST [Refueling Water Storage Tank] setpoint change will reduce 
the radiological consequences of LOCAs by diluting the radioiodine 
concentrations in the recirculating sump fluid which could be 
released by Engineered Safety Features (ESF) leakage. This increase 
in water will also reduce the maximum pH and its deleterious effects 
on equipment and sump performance.
    The increase in water level and the change in strainer design 
will significantly increase NPSH [net positive suction head] and 
headloss margins required to assure long term core cooling.
    The change to a minimum pH of 7.1 will not result in a 
significant increase in the radiological consequences of a LOCA as 
described below.

[[Page 13180]]

    The buffering agent will dissolve in the containment sump fluid 
resulting from these accidents raising the pH of the fluid, which 
would initially be greater than or equal to 4.0 but less than 7.0 
during the injection phase of containment spray operation. The 
equilibrium spray pH during the recirculation phase resulting from 
this change will be greater than or equal to 7.1. The pH range for 
the spray will be bounded by the water spray solution which is 
borated water with a maximum of 2600 ppm [parts per million] boron 
buffered to a final spray solution pH much less than the 10.5 as 
described in the current FSAR Section 3.11(B) for the postulated 
spray solution environment. The maximum pH is the limiting parameter 
for equipment qualification. Since the resulting pH level will be 
closer to neutral using the lower limit of 7.1, post-LOCA corrosion 
of containment components will not be increased. Post-LOCA hydrogen 
generation will be reduced. There will not be an adverse radiation 
dose effect on any safety-related equipment. Thus, the potential for 
failures of the ECCS or safety-related equipment following a LOCA 
will not be increased as a result of the proposed change.
    This modification affects the Containment Spray System which is 
intended to respond to and mitigate the effects of a LOCA. The 
chemical additive baskets serve a passive function to provide a 
buffering agent to neutralize the sump solution. Failure of a basket 
would not initiate an accident. The Containment Spray System will 
continue to function in a manner consistent with the plant design 
basis. There will be no degradation in the performance of nor an 
increase in the number of challenges to equipment assumed to 
function during an accident situation.
    As such, these Technical Specification revisions do not affect 
the probability of any event initiators. There will be no adverse 
changes to normal plant operating parameters, ESF actuation 
setpoints, or accident mitigation capabilities.
    The proposed change allows a passive Spray Additive System to 
replace the active Spray Additive System currently used to mitigate 
the consequences of an accident. By substituting a passive system 
for an active system, the probability of occurrence of a malfunction 
of equipment associated with the Spray Additive System will be 
reduced since the number of active components subject to malfunction 
is reduced. This TS surveillance change will maintain the 
equilibrium sump pH at greater than or equal to 7.1 to minimize 
chloride-induced stress corrosion cracking in austenitic stainless 
components important to safety located inside containment. 
Therefore, the proposed changes will not increase the probability of 
an accident or malfunction of equipment important to safety 
previously evaluated in the FSAR.
    The offsite and control room doses will continue to meet the 
requirements of 10 CFR [Part] 100; 10 CFR [Part] 50, Appendix A, GDC 
[General Design Criterion] 19; SRP [Standard Review Plan] 15.6.5.11; 
and SRP 6.4.11. The deletion of the active Spray Additive System and 
replacement with a sump pH control system using TSP-C [Trisodium 
Phosphate crystalline] will not increase the reported radiological 
consequences of a postulated LOCA. The proposed new pH control 
system will provide satisfactory retention of iodine in the sump 
water, as well as provide adequate pH control to minimize the 
potential of chloride-induced stress corrosion cracking of 
austenitic stainless steel components.
    The baskets which will contain the trisodium phosphate are 
seismically designed and located in the post-accident flood plane 
area to ensure mixing with the recirculating fluid. The consequences 
of a malfunction of any piece of equipment associated with the 
Containment Spray System would not be affected by the change from an 
active Spray Additive System to a passive system. The consequences 
of a failure in the active Spray Additive System are eliminated by 
this passive system. The proposed changes do not increase the 
malfunction of equipment important to safety previously evaluated in 
the FSAR. Therefore, the proposed 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 accident previously evaluated?
    Response: No.
    The changes to the new Containment Spray Additive System are 
essentially a passive system, i.e., no operator or automatic action 
of electrical devices is required to actuate the system. There are 
no electrical components being added whose failure could prevent the 
new system from functioning. The only new components being added are 
the storage baskets for the chemical buffering agent. Seismic 
requirements have been included in the design to ensure the 
structural integrity of the baskets will be maintained during a 
seismic event.
    No new accident scenarios, transient precursors, or limiting 
single failures are introduced as a result of these changes. There 
will be no adverse effect or challenges imposed on any safety-
related system as a result of these changes. The use of dry sodium 
phosphates is allowed for adjustment of the post-LOCA sump solution 
pH as discussed in SRP 6.1.1. The quantity of trisodium phosphate or 
any other buffering agent chosen will provide a minimum equilibrium 
sump pH of 7.1 following dissolution and mixing. Therefore, the 
possibility of a new or different type of accident is not created.
    There are no changes which would cause the malfunction of 
safety-related equipment, assumed to be operable in the accident 
analyses, as a result of the proposed Technical Specification 
changes. No new equipment performance burdens are imposed; however, 
there is the potential for an unlikely, but possible, event in which 
an initially concentrated solution of buffering agent could be 
transported to the stagnant volume of an inactive sump during 
blowdown and pool fill. This situation would be short-lived since, 
as the recirculated sump fluid is cooled in the RHR [residual heat 
removal] heat exchangers, sufficient buoyancy-driven circulation 
within containment will result to displace the stagnant solution and 
eventually yield a uniform, equilibrium solution. In the current 
design, all of the chemical additive is delivered to the 
recirculation sump even in the event of the worst single active 
failure. The possibility of a malfunction of safety-related 
equipment with a different result is not created. Therefore, the 
proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The RWST Low-Low nominal setpoint, in conjunction with the plant 
modifications, ensures that both the ECCS and Containment Spray 
Systems can be transferred from injection to recirculation without 
stopping the pumps and with no credit for containment overpressure. 
Analyses have been performed which show that, even with worst case 
single active failures, suction to the pumps would not be lost.
    The only function of the NaOH spray additive solution is to 
provide pH control of the post-accident containment recirculation 
sump water, since the borated water from the Refueling Water Storage 
Tank (RWST) used as the containment spray pump suction source during 
injection is sufficient to remove iodine from the containment 
atmosphere following a LOCA. The net effect on the pH control 
function of reducing the amount of NaOH or replacing NaOH with the 
chemical buffering agent TSP-C is that the equilibrium sump pH will 
be lowered to a minimum of 7.1. There will be no change to the 
current Technical Specification acceptance limits on RWST volume and 
boron concentration. The resulting equilibrium sump pH level from 
this change will be closer to neutral; therefore, the post-LOCA 
corrosion of containment components will not be increased.
    Because the long term pH will be maintained greater than or 
equal to 7.1, margin to minimize the potential for stress corrosion 
cracking is maintained.
    The radiological analysis as discussed in the technical analysis 
above, is shown not to be impacted. There will be no change to the 
DNBR [departure from nucleate boiling ratio] Correlation Limit, the 
design DNBR limits, or the safety analysis DNBR limits discussed in 
Bases Section 2.1.1. There will be no effect on the manner in which 
Safety Limits or Limiting Safety System Settings are determined nor 
will there be any effect on those plant systems necessary to assure 
the accomplishment of protection functions. There will be no adverse 
impact on DNBR limits, FQ, F-delta-H, LOCA PCT [peak 
cladding temperature], peak local power density, or any other margin 
of safety. Therefore the proposed change does not involve a 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff

[[Page 13181]]

proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Branch Chief: David Terao.

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station, Units 1 and 2, Somervell County, Texas

    Date of amendment request: December 16, 2005.
    Brief description of amendments: The amendment would revise the 
Technical Specifications (TS) to adopt NRC-approved Revision 4 to 
Technical Specification Task Force (TSTF) Standard Technical 
Specification Change Traveler, TSTF-449, ``Steam Generator Tube 
Integrity.'' The proposed amendment includes:

--Revised TS definition of Leakage,
--Revised TS 3.4.13, ``RCS [Reactor Coolant System] Operational 
Leakage,''
--Added new TS 3.4.17, ``Steam Generator (SG) Tube Integrity,''
--Revised TS 5.5.9, ``Steam Generator Program''
--Added new TS 5.6.9, ``Steam Generator Tube Inspection Report,'' and
--Revised TS 5.6.10, ``Steam Generator Tube Inspection Report'' (for 
existing Unit 1 SGs).

    The proposed changes are necessary in order to implement the 
guidance for the industry initiative on Nuclear Energy Institute (NEI) 
Report 97-06, ``Steam Generator Program Guidelines.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on March 2, 2005 (70 FR 10298), on possible amendments 
adopting TSTF-449, including a model safety evaluation and model no 
significant hazards consideration (NSHC) determination, using the 
consolidated line item improvement process. The NRC staff subsequently 
issued a notice of availability of the models for referencing in 
license amendment applications in the Federal Register on May 6, 2005 
(70 FR 24126). The licensee affirmed the applicability of the following 
NSHC determination in its application dated December 16, 2005.
    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 requires a SG Program that includes 
performance criteria that will provide reasonable assurance that the 
SG tubing will retain integrity over the full range of operating 
conditions (including startup, operation in the power range, hot 
standby, cooldown and all anticipated transients included in the 
design specification). The SG performance criteria are based on tube 
structural integrity, accident induced leakage, and operational 
LEAKAGE.
    A SGTR [steam generator tube rupture] event is one of the design 
basis accidents that are analyzed as part of a plant's licensing 
basis. In the analysis of a SGTR event, a bounding primary to 
secondary LEAKAGE rate equal to the operational LEAKAGE rate limits 
in the licensing basis plus the LEAKAGE rate associated with a 
double-ended rupture of a single tube is assumed.
    For other design basis accidents such as a MSLB [main steam line 
break], rod ejection, and reactor coolant pump locked rotor the 
tubes are assumed to retain their structural integrity (i.e., they 
are assumed not to rupture). These analyses typically assume that 
primary to secondary LEAKAGE for all SGs is 1 gallon per minute or 
increases to 1 gallon per minute as a result of accident induced 
stresses. The accident induced leakage criterion introduced by the 
proposed changes accounts for tubes that may leak during design 
basis accidents. The accident induced leakage criterion limits this 
leakage to no more than the value assumed in the accident analysis.
    The SG performance criteria proposed change to the TS identify 
the standards against which tube integrity is to be measured. 
Meeting the performance criteria provides reasonable assurance that 
the SG tubing will remain capable of fulfilling its specific safety 
function of maintaining reactor coolant pressure boundary integrity 
throughout each operating cycle and in the unlikely event of a 
design basis accident. The performance criteria are only a part of 
the SG Program required by the proposed change to the TS. The 
program, defined by NEI 97-06, Steam Generator Program Guidelines, 
includes a framework that incorporates a balance of prevention, 
inspection, evaluation, repair, and leakage monitoring. The proposed 
changes do not, therefore, significantly increase the probability of 
an accident previously evaluated.
    The consequences of design basis accidents are, in part, 
functions of the DOSE EQUIVALENT I-131 in the primary coolant and 
the primary to secondary LEAKAGE rates resulting from an accident. 
Therefore, limits are included in the plant technical specifications 
for operational leakage and for DOSE EQUIVALENT I-131 in primary 
coolant to ensure the plant is operated within its analyzed 
condition. The analysis of the limiting design basis accident 
assumes that primary to secondary leak rate after the accident is 1 
gallon per minute with no more than 150 gallons per day in any one 
SG, and that the reactor coolant activity levels of DOSE EQUIVALENT 
I-131 are at the TS values before the accident.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary coolant chemistry controls. The 
proposed approach updates the current TSs and enhances the 
requirements for SG inspections. The proposed change does not 
adversely impact any other previously evaluated design basis 
accident and is an improvement over the current TSs.
    Therefore, the proposed change does not affect the consequences 
of a SGTR accident and the probability of such an accident is 
reduced. In addition, the proposed changes do not affect the 
consequences of an MSLB, rod ejection, or a reactor coolant pump 
locked rotor event, or other previously evaluated accident.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Previously Evaluated

    The proposed performance based requirements are an improvement 
over the requirements imposed by the current technical 
specifications. Implementation of the proposed SG Program will not 
introduce any adverse changes to the plant design basis or 
postulated accidents resulting from potential tube degradation. The 
result of the implementation of the SG Program will be an 
enhancement of SG tube performance. Primary to secondary LEAKAGE 
that may be experienced during all plant conditions will be 
monitored to ensure it remains within current accident analysis 
assumptions.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary or secondary coolant chemistry 
controls. In addition, the proposed change does not impact any other 
plant system or component. The change enhances SG inspection 
requirements.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.

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

    The SG tubes in pressurized water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
isolate the radioactive fission products in the primary coolant from 
the secondary system. In summary, the safety function of an SG is 
maintained by ensuring the integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical condition of the tube. The proposed 
change does not affect tube design or operating environment. The 
proposed change is expected to result in an improvement in the tube 
integrity by implementing the SG Program to manage SG tube 
inspection,

[[Page 13182]]

assessment, repair, and plugging. The requirements established by 
the SG Program are consistent with those in the applicable design 
codes and standards and are an improvement over the requirements in 
the current TSs.
    For the above reasons, the margin of safety is not changed and 
overall plant safety will be enhanced by the proposed change to the 
TS.

    The NRC staff proposes to determine that the amendments request 
involves no significant hazards consideration.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Branch Chief: David Terao.

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station, Units 1 and 2, Somervell County, Texas

    Date of amendment request: December 16, 2005.
    Brief description of amendments: The proposed amendments would 
revise the Technical Specifications (TSs) consistent with the Nuclear 
Regulatory Commission (NRC)-approved Technical Specification Task Force 
(TSTF) Standard Technical Specification Change Traveller, TSTF-419, 
``Revise PTLR [Pressure and Temperature Limits Report] Definition and 
References in ISTS [improved Standard TS] 5.6.6.
    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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to reference the Topical Report number and 
title do not alter the use of the analytical methods used to 
determine the P/T [Pressure/Temperature] limits or LTOP [Low 
Temperature Overpressure Protection] setpoints that have been 
reviewed and approved by the NRC. This method of referencing Topical 
Reports would allow the use of current Topical Reports to support 
limits in the PTLR without having to submit an amendment to the 
operating license. Implementation of revisions to Topical Reports 
would still be reviewed in accordance with 10 CFR 50.59 and where 
required receive NRC review and approval. The proposed changes do 
not adversely affect accident initiators or precursors nor alter the 
design assumptions, conditions, or configuration of the facility or 
the manner in which the plant is operated and maintained. The 
proposed changes do not alter or prevent the ability of structures, 
systems, and components (SSCs) from performing their intended 
function to mitigate the consequences of an initiating event within 
the assumed acceptance limits. The proposed changes do not affect 
the source term, containment isolation, or radiological release 
assumptions used in evaluating the radiological consequences of an 
accident previously evaluated. Further, the proposed changes do not 
increase the types or amounts of radioactive effluent that may be 
released offsite, nor significantly increase individual or 
cumulative occupational/public radiation exposures. The proposed 
changes are consistent with safety analysis assumptions and 
resultant consequences.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes to reference the Topical Report number and 
title do not alter the use of the analytical methods used to 
determine the P/T limits or LTOP setpoints that have been reviewed 
and approved by the NRC. This method of referencing Topical Reports 
would allow the use of current Topical Reports to support limits in 
the PTLR without having to submit an amendment to the operating 
license. Implementation of revisions to Topical Reports would still 
be reviewed in accordance with 10 CFR 50.59 and where required 
receive NRC review and approval. The changes do not involve a 
physical alteration of the plant (i.e., no new or different type of 
equipment will be installed) or a change in the methods governing 
normal plant operation. In addition, the changes do not impose any 
new or different requirements or eliminate any existing 
requirements. The changes do not alter assumptions made in the 
safety analysis. The proposed changes are consistent with the safety 
analysis assumptions and current plant operating practice.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes to reference the Topical Report number and 
title do not alter the use of the analytical methods used to 
determine the P/T limits or LTOP setpoints that have been reviewed 
and approved by the NRC. This method of referencing Topical Reports 
would allow the use of current Topical Reports to support limits in 
the PTLR without having to submit an amendment to the operating 
license. Implementation of revisions to Topical Reports would still 
be reviewed in accordance with 10 CFR 50.59 and where required 
receive NRC review and approval. The proposed changes do not alter 
the manner in which safety limits, limiting safety system settings 
or limiting conditions for operation are determined. The setpoints 
at which protective actions are initiated are not altered by the 
proposed changes. Sufficient equipment remains available to actuate 
upon demand for the purpose of mitigating an analyzed event.
    Therefore, it is concluded that this change does not involve a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Branch Chief: David Terao.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: January 31, 2006.
    Description of amendment request: The proposed change would replace 
the current containment methodology with the methodology described in 
Topical Report DOM-NAF-3, ``GOTHIC Methodology for Analyzing the 
Response to Postulated Pipe Ruptures Inside Containment,'' increase the 
containment air partial pressure limits in Technical Specification (TS) 
3.8, ``Containment,'' revise the loss-of-coolant (LOCA) accident 
alternate source term (AST) analysis, and change the method of starting 
the recirculation spray (RS) 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. Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    No.
    The proposed changes include a physical alteration to the RS 
system to start the inside and outside RS pumps on RWST [Refueling 
Water Storage Tank] Level Low coincident with CLS [consequence 
limiting safeguards] High High containment pressure. The RS system 
is used for accident mitigation only, and changes in the operation 
of the RS system cannot have an impact on the probability of an 
accident. The other changes do not affect equipment and are not 
accident initiators. The RWST Level Low instrumentation will comply 
with all applicable regulatory requirements and design criteria 
(e.g., train separation, redundancy, single failure). Therefore, the 
design functions performed by the RS system are not changed.
    Delaying the start of the RS pumps affects long-term containment 
pressure and

[[Page 13183]]

temperature profiles. The environmental qualification of safety-
related equipment inside containment was confirmed to be acceptable, 
and accident mitigation systems will continue to operate within 
design temperatures and pressures. Delaying the RS pump start 
reduces the emergency diesel generator loading early during a design 
basis accident, and staggering the RS pump start avoids overloading 
on each emergency bus. The reduction in iodine removal efficiency 
during the delay period is offset by changes to other assumptions in 
the LOCA dose analysis. The net impact is a reduction in the 
predicted offsite doses and control room doses following a design 
basis LOCA.
    The UFSAR [Updated Final Safety Analysis Report] safety analysis 
acceptance criteria continue to be met for the proposed changes to 
the RS pump start method, the proposed TS containment air partial 
pressure limits, the implementation of the GOTHIC containment 
analysis methodology, and the changes to the LOCA dose consequences 
analyses. Based on this discussion, the proposed amendments do not 
increase the probability or consequence of an accident previously 
evaluated.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
identified?
    No.
    The proposed change alters the RS pump circuitry by initiating 
the start sequence with a new RWST Level Low signal instead of a 
timer after the CLS High High pressure setpoint is reached. The 
timers for the outside RS pumps will be used to sequence pump starts 
and preclude diesel generator overloading. The RS pump function is 
not changed. The RWST Level Low instrumentation will be included as 
part of the engineered safeguards features (ESF) instrumentation in 
the Surry TS and will be subject to the ESF surveillance 
requirements. The design of the RWST Level Low instrumentation 
complies with all applicable regulatory requirements and design 
criteria. The failure modes have been analyzed to ensure that the 
RWST Level Low circuitry can withstand a single active failure 
without affecting the RS system design functions. The RS system is 
an accident mitigation system only, so no new accident initiators 
are created.
    The remaining changes to the containment analysis methodology, 
the containment air partial pressures, and the LOCA AST analysis 
basis do not impact plant equipment design or function. Together, 
the changes assure that there is adequate margin available to meet 
the safety analysis criteria and that dose consequences are within 
regulatory limits. The proposed changes do not introduce failure 
modes, accident initiators, or malfunctions that would cause a new 
or different kind of accident. Therefore, the proposed changes do 
not create the possibility of a new or different kind of accident 
from any accident previously identified.
    3. Does the proposed license amendment involve a significant 
reduction in a margin of safety?
    No.
    The changes to the actuation of the RS pumps and the increased 
containment air partial pressure affect the containment response 
analyses and the LOCA dose analysis. Analyses have been performed 
that show the containment design basis limits are satisfied and the 
post-LOCA offsite and control room doses meet the required criteria 
for the proposed changes to the containment analysis methodology, 
the RS pump start method, the TS containment air partial pressure 
limits, and the LOCA AST bases. 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: Lillian M. Cuoco, Esq., Senior Counsel, 
Dominion Resources Services, Inc., Millstone Power Station, Building 
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
    NRC Branch Chief: Evangelos C. Marinos.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

    Date of application for amendment: February 25, 2005.
    Brief description of amendment: The amendment deleted Section 2.E 
of the Facility Operating License, which requires reporting of 
violations of the requirements in Section 2.C of the Facility Operating 
License.
    Date of Issuance: February 22, 2006.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 258.
    Facility Operating License No. DPR-16: The amendment revised the 
Facility Operating License.
    Date of initial notice in Federal Register: April 26, 2005 (70 FR 
21453).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated February 22, 2006.
    No significant hazards consideration comments received: No.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos. 
1, 2, and 3, Maricopa County, Arizona

    Date of application for amendments: March 4, 2005, as supplemented 
by letter dated January 25, 2006.
    Brief description of amendments: The proposed amendments deleted 
Section 2.F (2.G in Unit 3) of the Facility Operating Licenses, which 
requires reporting violations of the requirements

[[Page 13184]]

in Section 2.C of the Facility Operating License. The amendments also 
make administrative and editorial changes to the Technical 
Specifications (TSs). Changes to TS 1.4, ``Frequency,'' and TS 3.4.3, 
``RCS [Reactor Coolant System] Pressure and Temperature (P/T) Limits,'' 
correct editorial errors. The changes to TS 2.1.1, ``Reactor Core SLs 
[Safety Limits],'' and TS 3.3.1, ``Reactor Protective System (RPS) 
Instrumentation--Operating,'' remove the reference to departure from 
nucleate boiling ratios (DNBR) based on operating cycle, since only one 
of the listed DNBR values is now valid. TS 3.1.10, ``Special Test 
Exceptions (STE)--MODES 1 and 2,'' is changed to correct an 
inconsistency between the limiting condition for operation and the TS 
Bases. The changes to TS 3.7.2, ``Main Steam Isolation Valves 
(MSIVs),'' and TS 3.7.3, ``Main Feedwater Isolation Valves (MFIVs),'' 
correct the applicability for these specifications. The change to TS 
3.8.1, ``AC [Alternating Current] Sources--Operating,'' adds a note to 
a surveillance requirement. Changes to TS 3.8.4, ``DC [Direct Current] 
Sources--Operating,'' and TS 3.8.6, ``Battery Cell Parameters,'' remove 
the reference to AT&T batteries. The changes to TS 5.5.9, ``Steam 
Generator (SG) Tube Surveillance Program,'' correct the reference for 
NRC notification.
    Date of issuance: February 28, 2006.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days of the date of issuance.
    Amendment Nos.: Unit 1--158, Unit 2 --158, Unit 3--158.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Facility Operating Licenses and the Technical 
Specifications.
    Date of initial notice in Federal Register: May 10, 2005 (70 FR 
24647).
    The January 25, 2006, supplemental letter provided additional 
clarifying information, did not expand the scope of the application as 
originally noticed, and did not change the NRC staff's original 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 28, 2006.
    No significant hazards consideration comments received: No.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of application for amendment: October 31, 2005.
    Brief description of amendment: The amendment modified requirements 
by adding to the technical specifications a Limiting Condition for 
Operation (LCO) 3.0.8 that provides 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. In addition, a 
change to LCO 3.0.1 was required to reference the addition of LCO 
3.0.8.
    Date of issuance: February 15, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 172.
    Facility Operating License No. NPF-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 6, 2005 (70 FR 
72670).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 15, 2006.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: March 8, 2005, as supplemented by letter 
dated January 17, 2006.
    Brief description of amendment: The amendment allows a one-time 
extension of an additional 4 months beyond the 5-year extension already 
granted by the staff to the nominal 10-year interval of the test 
interval for the next Appendix J, Type A test.
    Date of issuance: February 9, 2006.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 150.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 29, 2005 (70 FR 
15942). The supplement dated January 17, 2006, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 9, 2006.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1 (GGNS), Claiborne 
County, Mississippi

    Date of application for amendment: March 30, 2005, as supplemented 
by letter dated November 21, 2005.
    Brief description of amendment: The amendment incorporated the 
following U.S. Nuclear Regulatory Commission (NRC)-approved Technical 
Specification Task Force (TSTF) changes that apply to the Boiling Water 
Reactor/6 Improved Standard Technical Specifications into GGNS 
Technical Specifications (TSs):

------------------------------------------------------------------------
                                                           TS section
           TSTF No.                   Description           affected
------------------------------------------------------------------------
TSTF-046, Rev. 1..............  Clarify the             Surveillance
                                 Containment Isolation   Requirement
                                 Valve surveillance to   (SR) 3.6.1.3.4,
                                 apply only to           SR 3.6.4.2.2,
                                 automatic isolation     SR 3.6.5.3.3.
                                 valves.
TSTF-222, Rev. 1..............  Control Rod Scram Time  SR 3.1.4.1, SR
                                 Testing.                3.1.4.4.
TSTF-264, Rev. 0..............  Delete flux monitors    SR 3.3.1.1.5, SR
                                 specific overlap SRs.   3.3.1.1.6,
                                                         Table 3.3.1.1-
                                                         1.
TSTF-275, Rev. 0..............  Clarify requirements    Table 3.3.5.1-1,
                                 for Diesel Generator    Footnote (a).
                                 (DG) start signal on
                                 Reactor Pressure
                                 Vessel (RPV) Level--
                                 Low, Low, Low during
                                 RPV cavity flood-up.
TSTF-276, Rev. 2..............  Revise DG full load      SR 3.8.1.9, SR
                                 rejection test.         3.8.1.10, SR
                                                         3.8.1.14.
TSTF-300, Rev. 0..............  Eliminate DG Loss of    SR 3.8.2.1.
                                 Coolant Accident
                                 (LOCA) Start SRs
                                 while in shutdown
                                 when Emergency Core
                                 Cooling System is not
                                 required.
TSTF-322, Rev. 2..............  Secondary Containment   SR 3.6.4.1.3, SR
                                 Integrity SRs.          3.6.4.1.4.
TSTF-400, Rev. 1..............  Clarify SR on bypass    SR 3.8.1.13.
                                 of DG automatic trips.

[[Page 13185]]

 
TSTF-416, Rev. 0..............  SR 3.5.1.2 Notation...  Limiting
                                                         Condition for
                                                         Operation (LCO)
                                                         3.5.1, SR
                                                         3.5.1.2, LCO
                                                         3.5.2, SR
                                                         3.5.2.4.
------------------------------------------------------------------------

    The amendment also granted delayed performance of the modified SRs 
for DG 12 until the next regularly scheduled performance rather than 
immediately upon implementation of this amendment, which is still 
consistent with NRC-approved TSTF changes. Those SRs are SR 3.8.1.9, SR 
3.8.1.10, and SR 3.8.1.14.
    Date of issuance: February 2, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance, with the exception of SR 3.8.1.9, SR 
3.8.1.10, and SR 3.8.1.14.
    Amendment No: 169.
    Facility Operating License No. NPF-29: The amendment revises the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: May 24, 2005 (70 FR 
29791). The supplemental letter dated November 21, 2005, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 2, 2006.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-237, Dresden Nuclear 
Power Station, Unit 2, Grundy County, Illinois

    Date of application for amendment: February 25, 2005.
    Brief description of amendment: The amendment deleted the reporting 
requirement in the Renewed Facility Operating License related to 
reporting violations of other requirements in the operating license.
    Date of issuance: February 17, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 210.
    Facility Operating License No. DPR-19: The amendments revised the 
Facility Operating License.
    Date of initial notice in Federal Register: April 26, 2005 (70 FR 
21456).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 17, 2006.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: April 13, 2005, as supplemented 
by letter dated December 22, 2005.
    Brief description of amendments: The amendment extended the 
completion time (CT) for Required Action A.1, ``Restore Residual Heat 
Removal Service Water subsystem to OPERABLE status,'' associated with 
Technical Specification (TS) Section 3.7.1 from 7 days to 10 days; 
established a 6-day (for Division 2 core standby cooling system (CSCS) 
maintenance) or 10-day (for Division 1 CSCS maintenance) CT for TS 
Section 3.7.2 when one or more required diesel generator cooling water 
subsystem(s) are inoperable. The Nuclear Regulatory Commission (NRC) 
staff is granting this amendment request with respect to TS Sections 
3.7.1 and 3.7.2 only. In the original submittal, the licensee also 
requested an extension of the CT for required Action C.4, ``Restore 
required Diesel Generator (DG) to OPERABLE status,'' associated with TS 
3.8.1 from 72 hours to 6 days; and extension of the CT for required 
Action F.1, ``Restore one required Diesel Generator (DG) to OPERABLE 
status,'' associated with TS 3.8.1 from 2 hours to 6 days. The NRC 
staff needs additional information from the licensee in order to 
complete its review and grant this portion of the amendment request. 
The staff will address the requests to extend CTs for TS 3.8.1 in a 
separate safety evaluation and license amendment, if granted.
    Date of issuance: February 23, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 175/161
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 7, 2005 (70 FR 
33213).
    The December 22, 2005, supplement, contained clarifying information 
and did not change the NRC staff's initial proposed finding of no 
significant hazards consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 23, 2006.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-352, Limerick Generating 
Station, Unit 1, Montgomery County, Pennsylvania

    Date of application for amendment: December 14, 2005, as 
supplemented by letter dated February 13, 2006.
    Brief description of amendment: The amendment modifies the 
Technical Specifications (TSs) to incorporate a revised Single Loop 
Operation Safety Limit Minimum Critical Power Ratio due to the cycle-
specific analysis.
    Date of issuance: March 1, 2006.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 183.
    Facility Operating License No. NPF-39 This amendment revised the 
TSs.
    Date of initial notice in Federal Register: January 17, 2006 (71 FR 
2590). The supplement dated February 13, 2006, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 1, 2006.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: December 17, 2004.
    Brief description of amendments: The amendments revised Appendix B, 
Environmental Protection Plan (non-radiological), of the Limerick 
Generating Station Operating Licenses.
    Date of issuance: February 17, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 180 and 142.
    Facility Operating License Nos. NPF-39 and NPF-85: The amendments

[[Page 13186]]

revised the Environmental Protection Plan.
    Date of initial notice in Federal Register: April 12, 2005 (70 FR 
19112).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 17, 2006.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: February 25, 2005.
    Brief description of amendments: The proposed amendment would 
delete the sections of the Facility Operating Licenses that require 
reporting of violations of the requirements in Section 2.C of the 
Facility Operating Licenses.
    Date of issuance: February 17, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 181 and 143.
    Facility Operating License Nos. NPF-39 and NPF-85: The amendments 
revised the Technical Specifications/license.
    Date of initial notice in Federal Register: April 26, 2005 (70 FR 
21457).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 17, 2006.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Unit Nos. 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: December 21, 2005.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) by relocating the Pressure Isolation 
Valve Table to the Technical Requirements Manual.
    Date of issuance: February 17, 2006.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment Nos.: 182 and 144.
    Facility Operating License Nos. NPF-39 and NPF-85. These amendments 
revised the TSs.
    Date of initial notice in Federal Register: January 17, 2006 (71 FR 
2590).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 17, 2006.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2), 
Beaver County, Pennsylvania

    Date of application for amendments: June 2, 2004, as supplemented 
February 11, May 12, October 31, and November 14, 2005.
    Brief description of amendments: These amendments approve 
conversion of the BVPS-1 and 2 containments from subatmospheric to 
atmospheric operating conditions. The proposed changes also approves 
the Modular Accident Analysis Program--Design Basis Accident (MAAP-DBA) 
computer code for the BVPS-1 and 2 containment integrity analysis and 
changes to mass and energy calculation methodologies.
    Date of issuance: February 6, 2006.
    Effective date: For BVPS-1, the amendment is effective as of the 
date of its issuance and shall be implemented prior to Mode 4 entry 
during startup from 1R17 which begins on or about February 10, 2006. 
For BVPS-2, the amendment is effective as of the date of its issuance 
and shall be implemented prior to Mode 4 entry during startup from 2R12 
which begins October 2006.
    Amendment Nos.: 272 and 154.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 20, 2004 (69 FR 
43462).
    The supplements dated February 11, May 12, October 31, and November 
14, 2005, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 6, 2006.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2), 
Beaver County, Pennsylvania

    Date of application for amendments: February 11, 2005, as 
supplemented August 8, 2005.
    Brief description of amendments: The amendments approved the 
adoption of the Relaxed axial offset control (RAOC) and FQ 
surveillance methodologies in accordance with NRC-approved Topical 
Report WCAP-10216-P-A, ``Relaxation of Constant Axial Offset Control--
FQ Surveillance Technical Specification.'' TS 3.2.1, ``Axial 
Flux Difference (AFD),'' and TS 3.2.2, ``Heat Flux Hot Channel Factor--
FQ(Z),'' were revised to adopt the RAOC calculational 
procedure of NUREG-1431, ``Standard Westinghouse Technical 
Specifications for Westinghouse Plants,'' Revision 3, June 2004. 
Changes to TS 3.2.3, ``Nuclear Enthalpy Hot Channel Factor--
FN[Delta]H,'' TS 3.2.4, ``Quadrant Power Tilt 
Ratio (QPTR),'' TS 3.3.1, ``Reactor Trip System Instrumentation (Table 
4.3-1, Note 3),'' and TS 6.9.5, ``Core Operating Limits Report 
(COLR),'' were made to provide consistency with the changes made to TSs 
3.2.1 and 3.2.2.
    Date of issuance: February 27, 2006.
    Effective date: Prior to entry into Mode 4 upon restart from the 
spring 2006 refueling outage which begins on or about February 10, 
2006, for BVPS-1 and prior to entry into Mode 4 from startup following 
the fall 2006 refueling outage which begins in October 2006, for BVPS-
2.
    Amendment Nos.: 274 and 155.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 26, 2005 (70 FR 
21457). The supplement dated August 8, 2005, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 27, 2006.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of application for amendment: April 20, 2005.
    Brief description of amendment: The changes revised the Technical 
Specifications (TSs) to replace plant-specific position titles with 
generic position titles. Also, the changes deleted TS 6.7, ``Safety 
Limit Violations or Protective Limit Violation,'' and included a change 
to TS 2.1.2, ``Reactor Core,'' associated with the deletion of TS 6.7. 
Additionally, the changes relocated to the Davis-Besse Nuclear Power 
Station Updated Safety Analysis Report the Process Control Program

[[Page 13187]]

requirements from TS 6.8, ``Procedures and Programs,'' and from TS 
6.14, ``Process Control Program (PCP).'' Associated with this change, 
TS Definition 1.30, ``Process Control Program,'' was deleted. Also, TS 
6.15, ``Offsite Dose Calculation Manual (ODCM),'' was modified to 
eliminate the requirement that changes to the ODCM be reviewed and 
accepted by the Plant Operations Review Committee (PORC). These changes 
to administrative requirements also eliminated the need to propose 
additional changes in the future to plant-specific position/
organizational titles. The changes are consistent with NUREG-1430, 
``Standard Technical Specifications--Babcock and Wilcox Plants,'' 
Revision 3, dated June 2004. Lastly, the changes revised in the TSs the 
title ``Industrial Security Plan'' to ``Physical Security Plan.''
    Date of issuance: February 7, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 272.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 24, 2005 (70 FR 
29795).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 7, 2006.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: March 28, 2005.
    Description of amendment request: The amendment revised the 
Seabrook Station, Unit No. 1, Technical Specifications (TSs) 
Surveillance Requirement 4.1.1.3, ``Moderator Temperature 
Coefficient,'' to allow the option of not measuring the moderator 
temperature coefficient within 7 effective full-power days of reaching 
an equilibrium boron concentration of 300 parts per million. This 
option is available only if the conditions described in WCAP-13749-P-A, 
``Safety Evaluation Supporting the Conditional Exemption of the Most 
Negative Moderator Temperature Coefficient Measurement'' have been met.
    Date of issuance: February 17, 2006.
    Effective date: As of its date of issuance, and shall be 
implemented within 90 days.
    Amendment No.: 107.
    Facility Operating License No. NPF-86: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: May 10, 2005 (70 FR 
24652).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 17, 2006.
    No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket Nos. 50-388, Susquehanna Steam Electric 
Station, Unit 2 (SSES-2), Luzerne County, Pennsylvania

    Date of application for amendment: January 28, 2005.
    Brief description of amendment: The amendment revises the SSES-2 
Technical Specification (TS) Table 3.3.5.1-1, ``Emergency Core Cooling 
System Instrumentation,'' Function 3.e, `` High Pressure Coolant 
Injection (HPCI) System,'' to change Condition ``D'' to ``C'' as the 
condition to reference from Required Action A.1. This is an editorial 
revision to correct a typographical error that had been present since 
the conversion to the Improved TSs in July 1998.
    Date of issuance: February 6, 2006.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 206.
    Facility Operating License No. NPF-22: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 10, 2005 (70 FR 
24654).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 2, 2006.
    No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, 
Pennsylvania

    Date of application for amendments: February 7, 2005
    Brief description of amendments: The amendments change the SSES 1 
and 2 Technical Specifications (TSs) for ``Secondary Containment,'' 
limiting condition for operation 3.6.4.1, by revising the frequency 
note applicable to Surveillance Requirements (SR) 3.6.4.1.4 and SR 
3.6.41.5. The revised note requires each zone configuration be tested 
at least once every 60 months.
    Date of issuance: February 2, 2006.
    Effective date: As of the date of issuance, and shall be 
implemented within 90 days.
    Amendment Nos.: 229 and 205.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 24, 2005 (70 FR 
29799).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 2, 2006.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

    Date of application for amendments: April 26, 2004, as supplemented 
by letters dated September 16, 2004, September 23, 2004, February 25, 
2005, and June 13, 2005.
    Brief description of amendments: These amendments revised the 
Technical Specifications to incorporate a full-scope application of an 
alternate source term methodology in accordance with 10 CFR 50.67.
    Date of issuance: February 17, 2006.
    Effective date: As of the date of issuance, to be implemented with 
90 days.
    Amendment Nos.: 271 and 252.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 22, 2004 (69 FR 
34705). The supplements did not effect the scope of changes discussed 
in the original no significant hazards determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 17, 2006.
    No significant hazards consideration comments received: No.

R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of application for amendment: April 29, 2005, as supplemented 
on September 19, 2005.
    Brief description of amendment: The amendment revised the Technical 
Specifications to incorporate the relaxed axial offset control and heat 
flux hot channel (FQ) surveillance methodologies. These methodologies 
are used to reduce operator action required to maintain conformance 
with power distribution control requirements and to increase the 
ability to return to power after a plant trip or transient. The changes 
are consistent with Westinghouse Electric Company Report WCAP-10216-P-
A, ``Relaxation of Constant Axial Offset Control/FQ 
Surveillance Technical Specification.''

[[Page 13188]]

    Date of issuance: February 15, 2006.
    Effective date: As of the date of issuance to be implemented prior 
to startup following the fall 2006 refueling outage.
    Amendment No.: 94.
    Renewed Facility Operating License No. DPR-18: Amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: June 7, 2005 (70 FR 
33220).
    The September 19, 2005, letter provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the staff's original proposed 
no significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 15, 2006.
    No significant hazards consideration comments received: No.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego 
County, California

    Date of application for amendments: July 15, 2005, and as 
supplemented by letter dated January 20, 2006.
    Brief description of amendments: The amendments are for the San 
Onofre Nuclear Generating Station (SONGS), Units 2 and 3, operating 
licenses, but they involved Unit 1, which is not an operating nuclear 
plant and is in the process of being decommissioned. The amendments 
revised License Condition 2.B.(6) for both SONGS, Units 2 and 3, by (1) 
deleting the sentence ``Transshipment of Unit 1 fuel between Units 1 
and [2 or 3] shall be in accordance with SCE [Southern California 
Edison Company] letters to U.S. Nuclear Regulatory Commission dated 
March 11, March 18 and March 23, 1988, and in accordance with the 
Quality Assurance requirements of 10 CFR Part 71'' and (2) adding the 
phrase ``and by the decommissioning of San Onofre Nuclear Generating 
Station Unit 1'' to the remaining sentence in the license condition. 
This change recognized that Unit 1 is now in the stage of 
decommissioning and that in the future any radioactive waste water 
produced in the further decommissioning of Unit 1 would be released 
from the San Onofre site by transferring the waste water from Unit 1 to 
Units 2 and 3. The processing (if required) and discharging of this 
waste water would be using the Units 2 and 3 radioactive waste system 
and ocean outfall discharge line.
    Date of issuance: February 28, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: Unit 2--202; Unit 3--193.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: September 13, 2005 (70 
FR 54089).
    The supplement dated January 20, 2006, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 28, 2006.
    No significant hazards consideration comments received: No.

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 amendments request: November 2, 2005.
    Brief Description of amendments: The amendments modify technical 
specifications (TS) to adopt the provisions of Industry/TS Task Force 
(TSTF) change TSTF-359, ``Increased Flexibility in Mode Restraints.''
    Date of issuance: February 22, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: 170 and 163.
    Renewed Facility Operating License Nos. NPF-2 and NPF-8: Amendments 
revise the Technical Specifications.
    Date of initial notice in Federal Register: December 20, 2005 (70 
FR 75498).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 22, 2006.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 7th day of March, 2006.

    For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 06-2383 Filed 3-13-06; 8:45 am]
BILLING CODE 7590-01-P