Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 53715-53726 [E6-14938]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 71, Number 176 (Tuesday, September 12, 2006)]
[Notices]
[Pages 53715-53726]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E6-14938]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from August 18, 2006 to August 31, 2006. The last 
biweekly notice was published on August 29, 2006 (71 FR 51222).

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

[[Page 53716]]

proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
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, https://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, HearingDocket@nrc.gov; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-

[[Page 53717]]

mail to OGCMailCenter@nrc.gov. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the attorney for 
the licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power 
Station, Unit No. 3 New London County, Connecticut

    Date of amendment request: June 14, 2006.
    Description of amendment request: The proposed amendment will 
permit Millstone Power Station, Unit 3 a one-time, 5-year extension, to 
Type A testing, of a surveillance requirement referenced in Technical 
Specification (TS) 4.6.1, relevant to the containment structure. TS 
4.6.1 specifies performance of an integrated leak rate test at a 
frequency of up to 10 years with allowance for a 15-month extension.
    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:

    Criterion 1:
    Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed extension to Type A testing cannot increase the 
probability of an accident previously evaluated since extension of 
the containment Type A testing is not a physical plant modification 
that could alter the probability of accident occurrence nor, is it 
an activity or modification that by itself could lead to equipment 
failure or accident initiation.
    The proposed one-time, five-year extension to Type A testing 
does not result in a significant increase in the consequences of an 
accident as documented in NUREG-1493. The NUREG notes that very few 
potential containment leakage paths are not identified by Type B and 
C tests. It concludes that even reducing the Type A (ILRT) testing 
frequency to once per twenty years leads to an imperceptible 
increase in risk.
    DNC provides a high degree of assurance through indirect testing 
and inspection that the containment will not degrade in a manner 
detectable only by Type A testing. The last two Type A tests 
identified containment leakage within acceptance criteria, 
indicating a very leak-tight containment. Inspections required by 
the ASME Code [American Society of Mechanical Engineers Boiler and 
Pressure Vessel Code] are also performed in order to identify 
indications of containment degradation that could affect leak-
tightness. Separately, Type B and C testing required by Technical 
Specifications, identifies any containment opening from design 
penetrations, such as valves, that would otherwise be detected by a 
Type A test. These factors establish that a one-time, five-year 
extension to the Millstone Power Station Unit 3 Type A test interval 
will not represent a significant increase in the consequences of an 
accident.
    Criterion 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 revision to Technical Specifications adds a one-
time extension to the current interval for Type A testing for 
Millstone Power Station Unit 3. The current test interval of ten 
years, based on past performance, would be extended on a one-time 
basis to fifteen years from the last Type A test. The proposed 
extension to Type A testing does not create the possibility of a new 
or different type of accident since there are no physical changes 
being made to the plant and there are no changes to the operation of 
the plant that could introduce a new failure.
    Criterion 3:
    Does the proposed amendment involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed revision to Millstone Power Station Unit 3 
Technical Specifications adds a one-time extension to the current 
interval for Type A testing. The current test interval of ten years, 
based on past performance, would be extended on a one-time basis to 
fifteen years from the last Type A test for Millstone Power Station 
Unit 3. RG [Regulatory Guide] 1.174 provides guidance for 
determining the risk impact of plant-specific changes to the 
licensing basis. RG 1.174 defines very small changes in risk as 
resulting in increases of CDF [core damage frequency] below 
10-6/yr and increases in LERF [large early release 
frequency] below 10-7/yr. Since the ILRT [integrated leak 
rate testing] does not impact CDF, the relevant criterion is LERF. 
The increase in LERF resulting from a change in the Type A ILRT test 
interval from a once-per-ten-years to a once-per-fifteen-years is 
3.1 x 10-7/yr, based on internal events. RG 1.174 states 
that when the calculated increase in LERF is in the range of 
10-7/yr to 10-6/yr, applications will be 
considered if it can be shown that the total [LERF] is less than 
10-5/yr. Since the total LERF for the 15-year metric is 
6.3 x 10-7/yr, then the change is considered acceptable. 
Increasing the ILRT interval from ten to fifteen years is, 
therefore, considered non-risk significant and will not 
significantly reduce the margin of safety. The NUREG-1493 generic 
study of the effects of extending containment leakage testing found 
that a 20-year interval in Type A leakage testing resulted in an 
imperceptible increase in risk to the public. NUREG-1493 generically 
concludes that the design containment leakage rate contributes about 
0.1 percent of the overall risk. Decreasing the Type A testing 
frequency would have a minimal [e]ffect on this risk since 95% of 
the Type A detectable leakage paths would already be detected by 
Type B and C testing.

    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, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Acting Branch Chief: Brooke D. Poole.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Miami-Dade County, Florida

    Date of amendment request: June 21, 2006.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3.7.3, Action a, to extend the 
allowed outage time (AOT) for one inoperable intake cooling water (ICW) 
pump from 7 days to 14 days. The proposed amendments were prepared in 
accordance with the guidance provided by the NRC in Regulatory Guide 
1.174, ``An Approach for Using Probabilistic Risk Assessment in Risk 
Informed Decisions on Plant-Specific Changes to the Licensing Basis'' 
and Regulatory Guide 1.177, ``An Approach for Plant-Specific, Risk-
Informed Decisionmaking; Technical Specifications.''
    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

[[Page 53718]]

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 affects the AOT for TS 3.7.3, 
Action a. The proposed change allows an extension of the current AOT 
for an inoperable ICW pump from 7 days to 14 days. The proposed 
change does not affect the design of the ICW System, the operational 
characteristics or function of the ICW System, the interfaces 
between the ICW System and other plant systems, or significantly 
affect the reliability of the ICW System. Limiting conditions for 
operation and their associated allowed outage times are not 
considered initiating conditions for any accident previously 
evaluated, nor is the ICW System considered an initiator for any 
accident previously evaluated. The ICW System provides the cooling 
water to the safety related CCW [component cooling water] heat 
exchangers. The ICW System also provides cooling water to the TPCW 
[turbine plant cooling water] heat exchangers and supplies water to 
the Lube Water System. During accident conditions, the ICW System 
performs the accident mitigation function of removing the heat load 
from the CCW System to support both reactor heat removal and 
containment heat removal requirements. The consequences of accidents 
previously evaluated are not affected by the proposed change in AOT. 
To fully evaluate the effect of the proposed ICW AOT extension, PRA 
[probabilistic risk assessment] methods and a deterministic analysis 
were utilized. The results of the analysis show no significant 
increase in Core Damage Frequency or Large Early Release Frequency 
based upon the guidance provided in Regulatory Guides 1.174 and 
1.177.
    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 probability of a new or 
different accident from any accident previously evaluated?
    Response: No. The proposed change does not involve a change in 
the design, configuration, or method of operation of the plant. The 
proposed change will not alter the manner in which equipment 
operation is initiated, nor will the functional demands on credited 
equipment be changed. The proposed change allows operation of a 
Turkey Point unit to continue while an ICW pump is repaired and 
tested. The proposed extension does not affect the interaction of an 
ICW pump with any system whose failure or malfunction can initiate 
an accident. As such, no new failure modes are being introduced.
    Therefore, the proposed action 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 does not alter the plant 
design, nor does it affect the assumptions contained in the safety 
analyses. Specifically, there are no changes being made to the ICW 
design, including instrument setpoints. The proposed change has been 
evaluated both deterministically, and using risk-informed methods. 
Based upon these evaluations, margins of safety ascribed to ICW 
availability and to plant risk have been determined to not be 
significantly reduced. The evaluation has concluded the following 
with respect to the proposed change:
    Applicable regulatory requirements will continue to be met, 
adequate defense-in-depth will be maintained, sufficient safety 
margins will be maintained, and any increases in CDF [core damage 
frequency] and LERF [large early release frequency] are small and 
consistent with the NRC Safety Goal Policy Statement (Federal 
Register, Vol. 5.1, P. 30028 (51 FR 30028), August 4, 1986) as 
interpreted by NRC Regulatory Guides 1.174 and 1.177. Furthermore, 
increases in risk posed by potential combinations of equipment out 
of service during the proposed extended ICW pump AOT will be managed 
under a configuration risk management program consistent with 10 CFR 
50.65, ``Requirements for monitoring the effectiveness of 
maintenance at nuclear power plants,'' paragraph (a)(4).
    The availability of the other ICW pumps and the use of on-line 
risk assessment tools provide adequate compensation for the 
potential small incremental increase in plant risk associated with 
the extended ICW pump AOT.
    Therefore, the proposed change does not involve a significant 
reduction in margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Branch Chief: L. Raghavan.

Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of amendment request: July 6, 2006.
    Description of amendment request: The proposed amendments would 
incorporate new large-break loss-of-coolant accident (LBLOCA) analyses 
using the realistic LBLOCA methodology in the NRC-approved WCAP-16009-
P-A, ``Realistic Large Break LOCA [loss-of-coolant-accident] Evaluation 
Methodology Using Automated Statistical Treatment of Uncertainty Method 
(ASTRUM),'' and would revise Technical Specification (TS) 5.6.5.b to 
include reference to WCAP-16009-P-A.
    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.
    This license amendment request proposes to incorporate large 
break loss of coolant accident analyses using the ASTRUM 
methodology, documented in WCAP-16009-P-A, ``Realistic Large Break 
LOCA Evaluation Methodology Using the Automated Statistical 
Treatment of Uncertainty Method (ASTRUM)'', in the Prairie Island 
Nuclear Generating Plant licensing basis and add reference to WCAP-
16009-P-A in the Technical Specification's list of approved 
methodologies for establishing core operating limits.
    Accident analyses are not accident initiators, therefore, this 
proposed licensing basis change does not involve a significant 
increase in the probability of an accident. The analyses using 
ASTRUM demonstrated that the acceptance criteria in 10 CFR 50.46, 
``Acceptance criteria for emergency core cooling systems for light-
water nuclear power reactors'', were met. The NRC has approved WCAP-
16009-P-A for application to two-loop Westinghouse plants with upper 
plenum injection. Since the Prairie Island Nuclear Generating Plant 
is a two-loop Westinghouse plants with upper head injection and the 
analysis results meet the 10 CFR 50.46 acceptance criteria, this 
change does not involve a significant increase in the consequences 
of an accident.
    Addition of the reference to WCAP-16009-P-A in the Technical 
Specifications is an administrative change that does not affect the 
probability or consequences of an accident previously evaluated.
    The changes proposed in this license amendment 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.
    This license amendment request proposes to incorporate large 
break loss of coolant accident analyses using the ASTRUM 
methodology, documented in WCAP-16009-P-A, ``Realistic Large Break 
LOCA Evaluation Methodology Using the Automated Statistical 
Treatment of Uncertainty Method (ASTRUM)'', in the Prairie Island 
Nuclear Generating Plant licensing basis and add reference to WCAP-
16009-P-A in the Technical Specification's list of approved 
methodologies for establishing core operating limits.
    There are no physical changes being made to the plant as a 
result of using the Westinghouse ASTRUM analysis methodology in 
WCAP-16009-P-A for performance of the large break loss of coolant

[[Page 53719]]

accident analyses. No new modes of plant operation are being 
introduced. The configuration, operation and accident response of 
the structures or components are unchanged by utilization of the new 
analysis methodology. Analyses of transient events have confirmed 
that no transient event results in a new sequence of events that 
could lead to a new accident scenario. The parameters assumed in the 
analysis are within the design limits of existing plant equipment.
    In addition, employing the Westinghouse ASTRUM large break loss 
of coolant accident analysis methodology does not create any new 
failure modes that could lead to a different kind of accident. The 
design of all systems remains unchanged and no new equipment or 
systems have been installed which could potentially introduce new 
failure modes or accident sequences. No changes have been made to 
any reactor protection system or emergency safeguards features 
instrumentation actuation setpoints.
    Based on this review, it is concluded that no new accident 
scenarios, failure mechanisms or limiting single failures are 
introduced as a result of the proposed methodology changes.
    Addition of the reference to WCAP-16009-P-A in the Technical 
Specifications is an administrative change that does not create the 
possibility of a new or different kind of accident.
    The licensing basis and Technical Specification changes proposed 
in this license amendment do 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.
    This license amendment request proposes to incorporate large 
break loss of coolant accident analyses using the ASTRUM 
methodology, documented in WCAP-16009-P-A, ``Realistic Large Break 
LOCA Evaluation Methodology Using the Automated Statistical 
Treatment of Uncertainty Method (ASTRUM)'', in the Prairie Island 
Nuclear Generating Plant licensing basis and add reference to WCAP-
16009-P-A in the Technical Specification's list of approved 
methodologies for establishing core operating limits.
    The analyses using ASTRUM demonstrated that the applicable 
acceptance criteria in 10 CFR 50.46, ``Acceptance criteria for 
emergency core cooling systems for light-water nuclear power 
reactors'' are met. Margins of safety for large break loss of 
coolant accidents include quantitative limits for fuel performance 
established in 10 CFR 50.46. These acceptance criteria and the 
associated margins of safety are not being changed by this proposed 
new methodology. The NRC has approved WCAP-16009-P-A for application 
to two-loop Westinghouse plants with upper head injection. Since the 
Prairie Island Nuclear Generating Plant is a two-loop Westinghouse 
plants with upper plenum injection and the analysis results meet the 
10 CFR 50.46 acceptance criteria, this change does not involve a 
significant reduction in a margin of safety.
    Addition of the reference to WCAP-16009-P-A in the Technical 
Specifications is an administrative change that does not involve a 
significant reduction in a margin of safety.
    Addition of the reference to WCAP-16009-P-A in the Technical 
Specifications is an administrative change that does not involve a 
significant reduction in a margin of safety.
    The licensing basis and Technical Specification changes proposed 
in this license amendment do not involve a significant reduction in 
a margin of safety.

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

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 amendment request: August 11, 2006.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to relocate response time 
limit tables for the reactor trip system and engineered safety features 
actuation system to the Updated Final Safety Analysis Report.
    The August 11, 2006, application supersedes the previous 
application related to relocation of response time limits, dated August 
19, 2005, which was noticed in the Federal Register on December 20, 
2005 (70 FR 75496). Instead of changing the response time definitions 
in TSs 1.12 and 1.26, as proposed in the August 19, 2005, application, 
the licensee proposes to revise certain TS Bases to clarify that 
Nuclear Regulatory Commission approval would be required to use a means 
other than testing to verify that response times are within limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment relocates the instrument response time 
limits for the reactor trip system (RTS) and engineered safety 
feature actuation system (ESFAS) from the technical specifications 
to the Updated Final Safety Analysis Report (UFSAR). The proposed 
amendment conforms to the guidance given in Enclosures 1 and 2 of 
Generic Letter 93-08. Neither the response time limits nor the 
surveillance requirements for performing response time testing will 
be altered by this submittal. The overall RTS and ESFAS functional 
capabilities will not be changed and assurance that action 
requirements of the reactor trip and engineered safety features 
systems are completed within the time limits assumed in the accident 
analyses is unaffected by the proposed amendment.
    Therefore, operation of the facility in accordance with the 
proposed amendment will 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 amendment will not change the physical plant or the 
modes of plant operation defined in the operating license. The 
change does not involve the addition or modification of equipment 
nor does it alter the design or operation of plant systems.
    Therefore, operation of the facility in accordance with the 
proposed amendment will 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 measurement of instrumentation response times at the 
frequencies specified in the technical specification provides 
assurance that actions associated with the reactor trip and 
engineered safety features are accomplished within the time limits 
assumed in the accident analyses. The response time limits and the 
measurement frequencies remain unchanged by the proposed amendment.
    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.
    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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Acting Branch Chief: Brooke D. Poole.

[[Page 53720]]

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 amendment requests: July 14, 2006.
    Description of amendment requests: This amendment application 
proposes to delete duplicative notifications, reporting, and restart 
requirements if a safety limit is violated; replace plant-specific 
position titles with generic position titles; and make several 
additional administrative changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to remove the duplicative safety limit 
reporting requirements from the TSs [Technical Specifications] does 
not affect the plant or operation of the plant. The change simply 
removes duplicative information from the TSs that is covered in the 
NRC regulations. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed changes to make plant-specific position/
organizational titles more generic do not affect any plant 
structures, systems, and components, and have no effect on plant 
operations. The proposed changes are administrative and do not 
affect any existing limits. Accident initial conditions, 
probability, and assumptions remain as previously analyzed. The 
proposed changes will have no effect on accident initiation 
frequency. The proposed changes do not invalidate the assumptions 
used in evaluating the radiological consequences of any accident. 
Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The remaining changes are administrative and do not modify the 
qualifications, responsibilities, or requirements for the positions. 
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?
    Response: No.
    The proposed change to remove the duplicative safety limit 
reporting requirements from the TSs does not introduce any new 
accident scenarios, failure mechanisms, or limiting single failures. 
All systems, structures, and components previously required for the 
mitigation of a transient remain capable of fulfilling their 
intended design functions. The proposed change has no adverse effect 
on any safety-related system or component and does not challenge the 
performance or integrity of any safety related system. This change 
is considered an administrative action to remove duplicative 
reporting requirements. Therefore, this proposed change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    The proposed changes to make plant-specific position/
organizational titles more generic are administrative and do not 
introduce any new or different accident initiators. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any previously evaluated.
    The remaining proposed changes are administrative and do not 
modify the qualifications, responsibilities, or requirements for the 
positions. 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?
    Response: No.
    The proposed changes are administrative and do not involve any 
reduction in a margin of safety. Removal of duplicative information, 
replacing plant-specific position titles with generic position 
titles, and the other proposed administrative changes do not affect 
compliance with the regulations. 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: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Branch Chief: David Terao.

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 amendment requests: July 14, 2006.
    Description of amendment requests: The proposed change incorporates 
a description of the parent tube inspection limitation adjacent to the 
nickel band portion of the lower sleeve joint and provides the basis 
for the structural and leakage integrity of the joint being ensured 
with the existing inspection of the parent tube adjacent to the nickel 
band region.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This proposed change revises the San Onofre Units 2 and 3 
Technical Specifications (TS) Section 5.5.2.11.f.1.j to provide a 
description of the parent tube inspection limitation adjacent to the 
nickel band and to provide the basis for the structural and leakage 
integrity. This is supported by Westinghouse Topical Report SG-SGDA-
05-48-P Revision 1, ``WOG [Westinghouse Owners Group] PA-MSC-0190, 
Revision 1: Test Results Related to TIG [tungsten inert gas] and 
Alloy 800 Sleeve Installation in \3/4\ Inch and \7/8\ Inch OD SG 
[steam generator] Tubing In-Service Inspection Requirements.''
    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.
    Steam generator tube leakage and structural integrity will be 
maintained during all plant conditions upon implementation of the 
proposed inspection scope and repair limit changes to the San Onofre 
Units 2 and 3 Technical Specifications. This change does not 
introduce any new mechanisms that might result in a different kind 
of accident from those previously evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Structural and leakage integrity of the steam generator sleeve 
joint is ensured with the existing inspection of the parent tube 
adjacent to the nickel band region.
    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 requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Branch Chief: David Terao.

[[Page 53721]]

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: February 28, 2006.
    Brief description of amendments: This request proposes changes to 
Technical Specification (TS) 3/4.8.2.1, ``DC Sources--Operating,'' and 
3/4.8.2.2, ``DC Sources--Shutdown,'' and the addition of a new TS 3/
4.8.2.3, ``Battery Parameters.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    [1.] The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change rearranges the Technical Specifications for 
the direct current electrical power system, and adds new Conditions 
and required actions with revised completion times to allow for 
battery charger inoperability. Neither the direct current electrical 
power subsystem nor associated battery chargers are initiators of an 
accident sequence previously evaluated. Performance of plant 
operational activities in accordance with the proposed Technical 
Specification changes ensures that the direct current electrical 
power subsystem is capable of performing its function as previously 
described. Therefore, the mitigating functions supported by the 
subject subsystem will continue to provide the protection assumed by 
the safety analysis.
    Relocation of preventive maintenance surveillances and certain 
operating limits and actions to a ``Battery Monitoring and 
Maintenance Program'' will not challenge the ability of the subject 
subsystem to perform its design function. Maintenance and monitoring 
currently required will continue to be performed. In addition, the 
direct current electrical power subsystem is within the scope of 10 
CFR 50.65, ``Requirements for monitoring the effectiveness of 
maintenance at nuclear power plants,'' which will ensure continued 
control of maintenance activities associated with the subject 
subsystem.
    Revision of battery performance test interval to 12 months from 
18 months in 4.8.2.1.f (now 4.8.2.3.f.1) is a conservative change 
that is intended to ensure continued battery operability. In 
addition, a surveillance requirement will be added as 4.8.2.3.f.2 to 
require performance discharge tests at least once per 24 months for 
any battery reaching 85% of the service life expected for the 
application and capacity is equal to or greater than 100% of the 
manufacturer's rating. Surveillance requirement 4.8.2.3.f.2 is an 
additional criterion that supplements 4.8.2.3.f.1. Modified 
performance tests of batteries that have reached 85% of their 
service life are to be performed at 12-month intervals with capacity 
less [than] 100% of the manufacturer's rating, and at 24-month 
intervals if the capacity is 100% or greater. These surveillance 
requirements are consistent with the requirements of IEEE-450.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    [2.] The proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    The proposed change does not involve any physical alteration of 
the units. No new equipment is introduced, and installed equipment 
is not operated in a new or different manner. The proposed changes 
do not affect setpoints for initiation of protective or mitigating 
actions.
    Revision of battery performance test interval to 12 months from 
18 months in 4.8.2.1.f (now 4.8.2.3.f.1) is a conservative change 
that is intended to ensure continued battery operability. In 
addition, a surveillance requirement will be added as 4.8.2.3.f.2 to 
require performance discharge tests at least once per 24 months for 
any battery reaching 85% of the service life expected for the 
application and capacity is equal to or greater than 100% of the 
manufacturer's rating. Surveillance requirement 4.8.2.3.f.2 is an 
additional criterion that supplements 4.8.2.3.f.1. Modified 
performance tests of batteries that have reached 85% of their 
service life are to be performed at 12-month intervals with capacity 
less [than] 100% of the manufacturer's rating, and at 24-month 
intervals if the capacity is 100% or greater. These surveillance 
requirements are consistent with the requirements of IEEE-450.
    Operability of the DC [direct currrent] electrical power 
subsystems in accordance with the proposed technical specifications 
is consistent with the initial assumptions of the accident analyses 
and is based upon meeting the design basis of the plant.
    The proposed changes will not alter the manner in which 
equipment operation is initiated, nor will the functional demands on 
credited equipment be changed. No alteration in the operating 
procedures is proposed, and no change is being made to procedures 
relied upon in response to an off-normal event. No new failure modes 
are being introduced, and the proposed change does not alter 
assumptions made in the safety analyses.
    Therefore, the proposed change does not create the possibility 
of a new or different accident from any accident previously 
evaluated.
    [3.] The proposed change does not involve a significant 
reduction in the margin of safety.
    The proposed change will not adversely affect operation of plant 
equipment and will not result in a change to the setpoints at which 
protective actions are initiated. Sufficient DC capacity to support 
operation of mitigation equipment is ensured. The provisions of the 
Battery Monitoring and Maintenance Program will ensure that the 
station batteries are maintained in a highly reliable manner.
    Revision of battery performance test interval to 12 months from 
18 months in 4.8.2.1.f (now 4.8.2.3.f.1) is a conservative change 
that is intended to ensure continued battery operability. In 
addition, a surveillance requirement will be added as 4.8.2.3.f.2 to 
require performance discharge tests at least once per 24 months for 
any battery reaching 85% of the service life expected for the 
application and capacity is equal to or greater than 100% of the 
manufacturer's rating. Surveillance requirement 4.8.2.3.f.2 is an 
additional criterion that supplements 4.8.2.3.f.1. Modified 
performance tests of batteries that have reached 85% of their 
service life are to be performed at 12-month intervals with capacity 
less [than] 100% of the manufacturer's rating, and at 24-month 
intervals if the capacity is 100% or greater. These surveillance 
requirements are consistent with the requirements of IEEE-450.
    The equipment fed by the DC electrical system will continue to 
provide adequate power to safety-related loads in accordance with 
analysis assumptions.
    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: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Branch Chief: David Terao.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: June 7, 2006.
    Brief description of amendments: The proposed change would revise 
the Spent Fuel Pool (SFP) and In-Containment Storage Area Criticality 
Analysis as described in Section 5.6 of the Technical Specifications 
(TSs).
    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.
    There is no increase in the probability of an accident. The 
proposed change does allow a greater number of fuel storage 
configurations in SFP. While this could increase the probability of 
a fuel misloading, the presence of sufficient soluble boron in

[[Page 53722]]

the SFP precludes criticality as a result of the misloading. Fuel 
assembly placement will continue to be controlled pursuant to 
approved fuel handling procedures and will be in accordance with the 
TS and the spent fuel rack storage configuration limitations of 
UFSAR [updated final safety analysis report] Chapter 9.1.2.
    Reactivity changes due to SFP temperature changes have been 
evaluated. The base case criticality analysis covers a ``normal'' 
SFP temperature range of 50 [deg]F to 160 [deg]F. Spent fuel pool 
temperature accidents are considered outside the normal temperature 
range extending from 50 [deg]F to 240 [deg]F. In all SFP temperature 
accident cases, sufficient reactivity margin is available to the 
0.95 keff limit without requiring additional soluble 
boron above the base case level. Because adequate soluble boron will 
be maintained in the SFP water to maintain keff < 0.95, 
the consequences of a loss of normal cooling to the SFP will not be 
increased.
    There is no increase in the consequences of the accidental 
misloading of spent fuel assemblies into the SFP racks. The 
criticality analysis demonstrates that the pool keff will 
remain <= 0.95 following an accidental misloading due to the boron 
concentration of the pool. The current TS limitation will ensure 
that an adequate SFP boron concentration is maintained.
    The criticality analysis shows the consequences of a fuel 
assembly drop accident in the SFP are not affected when considering 
the presence of soluble boron. The rack keff remains <= 
0.95.
    The editorial changes proposed in this license amendment request 
do not impact the probability or consequences of an accident.
    Therefore, based on the conclusions of the above evaluation, 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?
    Response: No.
    Spent fuel handling accidents are not new or different types of 
accidents; they have been analyzed in Section 15.7.4 of the UFSAR.
    Criticality accidents in the SFP are not new or different types 
of accidents. They have been analyzed in the UFSAR and in 
Criticality Analysis Reports associated with specific licensing 
amendments for fuel enrichments that are assumed for the proposed 
change. Because the proposed SFP storage configuration limitations 
will be similar to the current ones, the new limitations will not 
have any significant effect on normal SFP operations and 
maintenance, and will not create any possibility of a new or 
different kind of accident. Verifications will continue to be 
performed to ensure that the SFP loading configuration meets 
specified requirements.
    The misloading of a fuel assembly in the required storage 
configuration has been evaluated. In all cases, the rack 
keff remains <= 0.95. Removal of an RCCA [rod cluster 
control assembly] from a checkerboard storage configuration has been 
analyzed and found to be bounded by the misloading of a fuel 
assembly.
    As discussed above, the proposed changes will not create the 
possibility of a new or different kind of accident. There is no 
significant change in plant configuration, equipment design, or 
equipment.
    The editorial changes proposed in this license amendment request 
do not impact the design basis accidents of STP [South Texas 
Project].
    Under the proposed amendment, no changes are being made to the 
racks themselves, to any other systems, or to the physical 
structures of the Fuel Handling Building.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed TS changes and the resulting spent fuel storage 
operation limits will provide [an] adequate safety margin to ensure 
that the stored fuel assembly array always remains subcritical. 
Those limits are based on a plant-specific criticality analysis 
performed in accordance with Westinghouse spent fuel rack 
criticality analysis methodology.
    While the criticality analysis utilized credit for soluble 
boron, storage configurations have been defined using 95/95 
keff calculations to ensure that the spent fuel rack 
keff is < 1.0 with no soluble boron. Soluble boron credit 
is used to offset uncertainties, tolerances, and off-normal 
conditions, and to provide subcritical margin such that the SFP 
keff is maintained <= 0.95.
    The loss of substantial amounts of soluble boron from the SFP 
that could lead to keff exceeding 0.95 has been 
previously evaluated and approved (Ref. 4 and 5) and shown to be not 
credible. A safety evaluation has been performed which shows that 
dilution of the SFP boron concentration from 2500 ppm [part per 
million] to 700 ppm is not credible. Also, the spent fuel rack 
keff will remain < 1.0 (with a 95/95 confidence level) 
with the SFP flooded with unborated water. These safety analyses 
demonstrate a level of safety comparable to the conservative 
criticality analysis methodology required by Westinghouse WCAP-
14416-P-A.
    The editorial changes proposed in this license amendment request 
do not affect the margin of safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Branch Chief: David Terao.

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

    Date of amendment request: March 31, 2006.
    Brief description of amendments: The amendments requested would 
revise Technical Specifications (TS) requirement 5.0, ``ADMINISTRATIVE 
CONTROLS.''
    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 change involves organizational changes at the 
executive level and does not impact nor effect accident analysis 
assumptions. The method and tools used to maintain, and produce 
proposed changes to, the Technical Specifications has no bearing on 
any accident analysis assumptions. Therefore, these assumptions are 
preserved and there is no change in the probability or consequences 
of any previously evaluated accident.
    Therefore, the proposed change does 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 change involves an organizational change due to a 
change in title. There are no changes in existing reporting 
relationships or assigned responsibilities for safe operation of 
CPSES. The proposed re-issuance of the entire Technical 
Specifications stems from a change in the software utilized by TXU 
Power to produce and maintain the Technical Specifications. This 
software is not used to operate the plant nor is it used to 
establish any operational limits.
    There are no hardware changes nor are there any changes in the 
method by which any safety-related plant system performs its safety 
function. The proposed change will not effect the normal method of 
plant operation. No performance requirements will be affected or 
eliminated. The proposed change will not result in physical 
alteration to any plant system nor will there be any change in the 
method by which any safety-related plant system performs its safety 
function.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of this change. There will be no adverse effect or challenges 
imposed as a result of this

[[Page 53723]]

change. There will be no adverse effect or challenges imposed on any 
safety-related system as a result of these changes.
    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 do not affect the acceptance criteria for 
any analyzed event nor is there a change to any Safety Analysis 
Limit (SAL). There will be no effect on the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined nor will there be any effect on those plant 
systems necessary to assure the accomplishment of protection 
functions.
    Therefore, the proposed change does not involve a reduction in a 
margin of safety. q

    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.

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

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

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: July 20, 2006.
    Brief description of amendment request: The proposed amendment 
would revise the Vogtle Electric Generating Plant (VEGP), Units 1 and 
2, Technical Specifications (TS) 5.5.9, ``Steam Generator (SG) Tube 
Surveillance Program,'' to incorporate changes in the SG inspection 
scope for VEGP, Unit 1 during Refueling Outage 13 and the subsequent 
operating cycle, and VEGP Unit 2 during Refueling Outage 12 and the 
subsequent operating cycle. The proposed changes modify the inspection 
requirements for portions of SG tubes within the tubesheet region of 
the SGs.
    Date of publication of individual notice in Federal Register: July 
31, 2006 (71 FR 43225).
    Expiration date of individual notice: 30-day, August 30, 2006; 60-
day, September 29, 2006.

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 (