Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 32602-32614 [E6-8450]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 71, Number 108 (Tuesday, June 6, 2006)]
[Notices]
[Pages 32602-32614]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E6-8450]


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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 May 12, 2006 to May 24, 2006. The last 
biweekly notice was published on May 23, 2006 (71 FR 29671).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination. Within 60 days after the date of publication of this 
notice, the licensee may file a request for a hearing with respect to 
issuance of the amendment to the subject facility operating license and 
any person whose interest may be affected by this proceeding and who 
wishes to participate as a party in the proceeding must file a written 
request for a hearing and a petition for leave to intervene.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
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.

[[Page 32603]]

    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, HearingDocket@nrc.gov; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of amendments request: April 26, 2006.
    Description of amendment request: The proposed amendment would 
modify technical specification (TS) requirements for inoperable 
snubbers by adding Limiting Condition for Operation 3.0.8. The changes 
are consistent with Nuclear Regulatory Commission approved Industry/
Technical Specification Task Force (TSTF) standard TS change TSTF-372, 
Revision 4.
    The NRC staff issued a notice of availability of a model safety 
evaluation and model no significant hazards consideration (NSHC) 
determination for referencing in license amendment applications in the 
Federal Register on May 4, 2005 (70 FR 23252). The licensee affirmed 
the applicability of the model NSHC determination in its application 
dated April 26, 2006.
    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:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change allows a delay time before declaring supported 
TS systems inoperable when the associated snubber(s) cannot perform its 
required safety function. Entrance into Actions or delaying entrance 
into Actions is not an initiator of any accident previously evaluated.
    Consequently, the probability of an accident previously evaluated 
is not significantly increased. The consequences of an accident while 
relying on the delay time allowed before declaring a TS supported 
system inoperable and taking its Conditions and Required Actions are no 
different than the consequences of an accident under the same plant 
conditions while relying on the existing TS supported system Conditions 
and Required Actions.

[[Page 32604]]

    Therefore, the consequences of an accident previously evaluated are 
not significantly increased by this change. Therefore, this change does 
not involve a significant increase in the probability or consequences 
of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change allows a delay time before declaring supported 
TS systems inoperable when the associated snubber(s) cannot perform its 
required safety function. The proposed change does not involve a 
physical alteration of the plant (no new or different type of equipment 
will be installed) or a change in the methods governing normal plant 
operations. Thus, this 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 allows a delay time before declaring supported 
TS systems inoperable when the associated snubber(s) cannot perform its 
required safety function. The proposed change restores an allowance in 
the pre-ISTS conversion TS that was unintentionally eliminated by the 
conversion. The pre-ISTS TS were considered to provide an adequate 
margin of safety for plant operation, as does the post-ISTS conversion 
TS. Therefore, this change does not involve a significant reduction in 
a margin of safety.
    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Branch Chief: Michael L. Marshall, Jr.

Entergy Nuclear Operations, Inc., Docket No. 50-271, Vermont Yankee 
Nuclear Power Station (VYNPS), Vernon, Vermont

    Date of amendment request: April 22, 2006.
    Description of amendment request: The proposed amendment would 
relocate the Technical Specification (TS) requirements for shock 
suppressors (snubbers) to the Technical Requirements Manual (TRM) and 
add a new Limiting Condition for Operation (LCO) 3.0.8.
    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 relocate TS 3/4.6.1 to the TRM is 
administrative in nature and does not involve the modification of any 
plant equipment or affect basic plant operation. Snubber operability 
and surveillance requirements will be contained in the TRM to ensure 
design assumptions for accident mitigation are maintained.
    The proposed change to add LCO 3.0.8 allows a delay time before 
declaring supported TS systems inoperable when the associated 
snubber(s) cannot perform the required safety function. Entrance into 
actions or delaying entrance into actions is not an initiator of any 
accident previously evaluated. Consequently, the probability of an 
accident previously evaluated is not significantly increased. The 
station design and safety analysis assumptions included provisions for 
redundancy to provide for periods when redundant systems are out-of-
service per the TS. The proposed snubber LCO ensures that out-of-
service time is minimized and risk is managed per 10 CFR 50.65(a)(4).
    Therefore, the consequences of an accident previously evaluated are 
not significantly increased by this change.
    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 relocate TS 3/4.6.1 to the TRM is 
administrative and does not involve any physical alteration of plant 
equipment. The proposed change does not change the method by which any 
safety-related system performs its function. As such, no new or 
different types of equipment will be installed, and the basic operation 
of installed equipment is unchanged. The methods governing plant 
operation and testing remain consistent with current safety analysis 
assumptions.
    [* * *]
    The proposed change to add LCO 3.0.8 allows a delay time before 
declaring supported TS systems inoperable when the associated 
snubber(s) cannot perform the required safety function. The proposed 
change does not involve a physical alteration of the plant (no new or 
different type of equipment will be installed) or a change in the 
methods governing normal plant operation.
    Therefore, this 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 relocate TS 3/4.6.1 to the TRM is 
administrative in nature, does not negate any existing requirement, and 
does not adversely affect existing plant safety margins or the 
reliability of the equipment assumed to operate in the safety analysis. 
As such, there are no changes being made to safety analysis 
assumptions, safety limits or safety system settings that would 
adversely affect plant safety as a result of the proposed change. 
Margins of safety are unaffected by requirements that are retained, but 
relocated from the TS to the TRM.
    [* * *]
    The proposed change to add LCO 3.0.8 to TS allows a delay time 
before declaring supported TS systems inoperable when the associated 
snubber(s) cannot perform the required safety function. The proposed 
change retains an allowance in the current VYNPS TS while upgrading it 
to be more conservative for snubbers supporting multiple trains or sub-
systems of an associated system. The updated TS will continue to 
provide an adequate margin of safety for plant operation upon 
incorporation of LCO 3.0.8. The station design and safety analysis 
assumptions provide margin in the form of redundancy to account for 
periods of time when system capability is reduced. This proposed change 
does not reduce that margin.
    Therefore, this change does not involve a significant reduction in 
a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Travis C. McCullough, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White 
Plains, NY 10601.
    Branch Chief: Richard Laufer.

[[Page 32605]]

Exelon Generation Company, LLC (EGC), Docket No. 50-374, LaSalle County 
Station, Unit 2, LaSalle County, Illinois

    Date of amendment request: April 21, 2006.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 5.5.13, ``Primary 
Containment Leakage Rate Testing Program,'' to reflect a one-time 
extension of the LaSalle County Station (LSCS), Unit 2 primary 
containment Type A integrated leak rate test (ILRT) date from the 
current requirement of no later than December 7, 2008, to prior to 
startup following the twelfth LSCS, Unit 2 refueling outage (L2R12).
    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 changes will revise LSCS, Unit 2, TS 5.5.13, ``Primary 
Containment Leakage Rate Testing Program,'' to reflect a one-time 
extension of the primary containment Type A Integrated Leak Rate Test 
(ILRT) date to ``prior to startup following L2R12.'' The current Type A 
ILRT interval of 15 years, based on past performance, would be extended 
on a one-time basis by approximately 2% of the current interval.
    The function of the primary containment is to isolate and contain 
fission products released from the reactor Primary Coolant System (PCS) 
following a design basis Loss of Coolant Accident (LOCA) and to confine 
the postulated release of radioactive material to within limits. The 
test interval associated with Type A ILRTs is not a precursor of any 
accident previously evaluated. Type A ILRTs provide assurance that the 
LSCS Unit 2 primary containment will not exceed allowable leakage rate 
values specified in the TS and will continue to perform their design 
function following an accident. The risk assessment of the proposed 
changes has concluded that there is an insignificant increase in total 
population dose rate and an insignificant increase in the conditional 
containment failure probability.
    Therefore, the proposed changes do 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 changes for a one-time extension of the Type A ILRT 
for LSCS Unit 2 will not affect the control parameters governing unit 
operation or the response of plant equipment to transient and accident 
conditions. The proposed changes do not introduce any new equipment, 
modes of system operation or failure mechanisms.
    Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Does the change involve a significant reduction in a margin of 
safety?
    Response: No.
    LSCS Unit 2 is a General Electric BWR/5 plant with a Mark II 
primary containment. The Mark II primary containment consists of two 
compartments, the drywell and the suppression chamber. The drywell has 
the shape of a truncated cone, and is located above the cylindrically 
shaped suppression chamber. The primary containment is penetrated by 
access, piping and electrical penetrations.
    The integrity of the primary containment penetrations and isolation 
valves is verified through Type B and Type C local leak rate tests 
(LLRTs) and the overall leak tight integrity of the primary containment 
is verified by a Type A ILRT, as required by 10 CFR 50, Appendix J, 
``Primary Reactor Containment Leakage Testing for Water-Cooled Power 
Reactors.'' These tests are performed to verify the essentially leak 
tight characteristics of the primary containment at the design basis 
accident pressure. The proposed changes for a one-time extension of the 
Type A ILRTs do not affect the method for Type A, B or C testing or the 
test acceptance criteria.
    EGC has conducted a risk assessment to determine the impact of a 
change to the LSCS Unit 2 Type A ILRT schedule from a baseline ILRT 
frequency of three times in ten years to once in 16.25 years (i.e., 15 
years plus 15 months) for the risk measures of Large Early Release 
Frequency (i.e., LERF), Total Population Dose, and Conditional 
Containment Failure Probability (i.e., CCFP). This assessment indicated 
that the proposed LSCS ILRT interval extension has a minimal impact on 
public risk.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Assistant General 
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett 
Square, PA 19348.
    NRC Branch Chief: Daniel S. Collins

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

    Date of amendment request: May 1, 2006.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 1.1, ``Definitions,'' TS 3.4.13, 
``RCS [reactor coolant system] Operational Leakage,'' TS 5.5.8, ``Steam 
Generator Program,'' and add new specifications (TS 3.4.17) for ``Steam 
Generator (SG) Tube Integrity'' and (TS 5.6.7) for ``Steam Generator 
Tube Inspection Report.'' The proposed changes are necessary in order 
to implement the guidance for the industry initiative on Nuclear Energy 
Institute (NEI) 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 Technical Specification Task Force Change Traveller 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 May 1, 2006.
    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--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated
    The proposed change requires an 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,

[[Page 32606]]

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.
    An 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 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 typical 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 
[500 gallons per day or 720 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 [kind] of accident from any accident previously 
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in [a] Margin of Safety
    The 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, 
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: Daniel F. Stenger, Ballard Spahr Andrews & 
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC 
20005.
    NRC Branch Chief: Richard J. Laufer.

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: April 28, 2006.
    Description of amendment requests: The proposed change will 
increase the minimum allowed boron concentration of the spent fuel pool 
and allow credit for soluble boron, guide tube inserts (GT-Inserts) 
made from borated stainless steel, and fuel storage patterns in place 
of Boraflex.
    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.

[[Page 32607]]

Dropped Fuel Assembly

    There is no significant increase in the probability of a fuel 
assembly drop accident in the spent fuel pool when assuming a complete 
loss of the Boraflex panels in the spent fuel pool racks and 
considering the presence of soluble boron in the spent fuel pool water 
for criticality control.
    Neither the presence of soluble boron in the spent fuel pool water, 
nor the placement of borated stainless steel guide tube inserts (GT-
Inserts) in the fuel assemblies for criticality control, will increase 
the probability of a fuel assembly drop accident. The handling of the 
fuel assemblies in the spent fuel pool has always been performed in 
borated water, and the quantity of Boraflex remaining in the racks or 
GT-Inserts placed in the fuel assemblies, has no affect on the 
probability of such a drop accident.
    Southern California Edison (SCE) has performed a criticality 
analysis which shows that the consequences of a fuel assembly drop 
accident in the spent fuel pool are not affected when considering a 
complete loss of the Boraflex in the spent fuel racks and the presence 
of soluble boron. The rack Keff remains less than or equal 
to 0.95.
    The fuel, the fuel rack, and the fuel pool qualifications have been 
evaluated and determined to be unaffected by the installation of the 
GT-Inserts. The mechanical design configuration of the GT-Inserts is 
similar to the shape, size, and weight of a control element assembly 
(CEA) finger. Each of the GT-Inserts are approximately 0.78 inch 
outside diameter (OD) solid stainless steel, with a boron content of 
approximately 2 weight percent (w/o). A small counterbore is machined 
at the top for handling and a rounded bottom is machined. The OD of 
these GT-Inserts is less than that of a CEA finger. The material 
(borated stainless steel) is American Society for Testing and Materials 
(ASTM) approved and has been licensed by the United States Nuclear 
Regulatory Commission (NRC) for use in spent fuel storage technologies 
and spent fuel pools. The structural effect of the weight of the GT-
Inserts on the fuel, the fuel rack, and the fuel pool structural 
interfaces and drop qualifications are unaffected. This is because the 
addition of five GT-Inserts (which increases the dry weight of a fuel 
assembly by 110 lbs.) brings the total weight to 1551 lbs. which is 
enveloped by the 2904 lbs. assumed in the calculation for fuel rack 
design.

Fuel Misloading

    There is no significant increase in the probability of the 
accidental misloading of spent fuel assemblies into the spent fuel 
racks when assuming a complete loss of the Boraflex panels and 
considering the presence of soluble boron in the pool water for 
criticality control. Fuel assembly placement will continue to be 
controlled pursuant to approved fuel handling procedures and will be in 
accordance with Technical Specification (TS) 3.7.18[,] ``Spent Fuel 
Assembly Storage[,]'' and Licensee Controlled Specification (LCS) 
4.0.100, ``Fuel Storage Patterns,'' which will specify spent fuel rack 
storage configuration limitations.
    There is no increase in the consequences of the accidental 
misloading of a spent fuel assembly into the spent fuel racks. The 
criticality analysis, performed by SCE, demonstrates that the pool 
Keff will be maintained less than or equal to 0.95 following 
an accidental misloading by the boron concentration of the pool. The 
proposed TS 3.7.17[,] ``Fuel Storage Pool Boron Concentration[,]'' will 
ensure that an adequate spent fuel pool boron concentration is 
maintained.

Change in Spent Fuel Temperature

    There is no significant increase in the probability of either the 
loss of normal cooling to the spent fuel pool water or a decrease in 
pool water temperature from a large emergency makeup when assuming a 
complete loss of the Boraflex panels and considering the presence of 
soluble boron in the spent fuel pool water. A high proposed 
concentration (>2000 parts per million (ppm)) of soluble boron is 
consistent with current operating practices maintained in the spent 
fuel pool water. The proposed minimum boron concentration of 2000 ppm 
in TS 3.7.17 will ensure that an adequate concentration is maintained 
in the spent fuel pools.
    A loss of normal cooling to the spent fuel pool water causes an 
increase in the temperature of the water passing through the stored 
fuel assemblies. This causes a decrease in the water density, and when 
coupled with the assumption of a complete loss of Boraflex, may result 
in a positive reactivity addition. However, the additional negative 
reactivity provided by the boron concentration limit in the proposed TS 
3.7.17 will compensate for the increased reactivity which could result 
from a loss of spent fuel pool cooling. Because adequate soluble boron 
will be maintained in the spent fuel pool water to maintain 
Keff less than or equal to 0.95, the consequences of a loss 
of normal cooling to the spent fuel pool will not be increased.
    The thermal considerations of the fuel are unaffected by the 
presence of the GT-Inserts because the guide tube is designed for the 
presence of a CEA; therefore, it is not a primary coolant flow area. 
The fuel rack normal thermal cooling and malfunctioned blocked cooling 
scenarios are unaffected by the presence of the GT-Inserts in the fuel 
assemblies.
    The proposed change does not involve an increase in the probability 
or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The consideration of criticality accidents in the spent fuel pool 
are not new or different. They have been analyzed in the Updated Final 
Safety Analysis Report (UFSAR) and in previous submittals to the NRC. 
Specific accidents considered and evaluated include fuel assembly drop, 
fuel assembly misloading in the racks, and spent fuel pool water 
temperature changes.
    The possibility for creating a new or different kind of accident is 
not credible. Neither Boraflex [n]or soluble boron are accident 
initiators. The proposed change takes credit for soluble boron in the 
spent fuel pool while maintaining the necessary margin of safety. 
Because soluble boron has always been present in the spent fuel pool, a 
dilution of the spent fuel pool soluble boron has always been a 
possibility. However, a criticality accident resulting from a dilution 
accident was not considered credible. For this proposed amendment, SCE 
performed a spent fuel pool dilution analysis, which demonstrated that 
a dilution of the boron concentration in the spent fuel pool water 
which could increase the rack Keff to greater than 0.95 
(constituting a reduction of the required margin to criticality) is not 
a credible event. The requirement to maintain boron concentration in 
the spent fuel pool water for reactivity control will have no effect on 
normal pool operations and maintenance. There are no changes in 
equipment design or plant configuration.
    The possibility of accidentally withdrawing a GT-Insert is 
minimized because special tooling is required to remove it, and it is 
completely contained within the guide tubes of the designated 
assemblies. Potential misloading of the GT-Inserts is minimized due to 
the design of the

[[Page 32608]]

installation equipment, procedural controls, and double verification 
that will be in place to ensure the GT-Inserts are installed properly.
    The possibility of accidentally withdrawing a CEA is minimized 
because specialized tooling is required for withdrawing a CEA from a 
fuel assembly. It is physically possible for the spent fuel handling 
tool to bind on a CEA after ungrappling from a fuel assembly and 
raising the tool. However, existing SONGS [San Onofre Nuclear 
Generating Station] procedures require that the operator validate 
``tool weight only'' on the spent fuel handling machine's load cell 
read out after ungrappling from a fuel assembly and raising the hoist 
slightly, and to report this information to the engineer directing the 
fuel movement.
    Therefore, the proposed change will not result in 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 TS changes proposed by this license amendment request and the 
resulting spent fuel storage operation limits will provide adequate 
safety margin to ensure that the stored fuel assembly array will always 
remain subcritical. Those limits are based on a San Onofre Nuclear 
Generating Station (SONGS) Units 2 and 3 plant specific analysis that 
was performed in accordance with a methodology previously approved by 
the NRC.
    The proposed change takes partial credit for soluble boron in the 
spent fuel pool. SCE's analyses show that spent fuel storage 
requirements meet the following NRC acceptance criteria for preventing 
criticality outside the reactor.
    (1) The neutron multiplication factor, Keff, including 
all uncertainties, shall be less than 1.0 when flooded with unborated 
water, and
    (2) The neutron multiplication factor, Keff, including 
all uncertainties, shall be less than or equal to 0.95 when flooded 
with borated water.
    The criticality analysis utilized credit for soluble boron to 
ensure Keff will be less than or equal to 0.95 under normal 
circumstances, and storage configurations have been defined using a 95/
95 Keff calculation to ensure that the spent fuel rack will 
be less than 1.0 with no soluble boron. Soluble boron credit is used to 
provide safety margin by maintaining Keff less than or equal 
to 0.95 including uncertainties, tolerances[,] and accident conditions 
in the presence of spent fuel pool soluble boron. SCE evaluated the 
loss of a substantial amount of soluble boron from the spent fuel pool 
water which could lead to Keff exceeding 0.95 and showed 
that it was not credible.
    Also, the spent fuel rack Keff will remain less than 1.0 
with the spent fuel pool flooded with unborated water.
    Decay heat, radiological effects, and seismic loads are unchanged 
by the absence of Boraflex.
    The mechanical properties and the weight of the fuel assemblies 
remain essentially unchanged with the inclusion of the weight of five 
GT-Inserts per assembly. The original mechanical and thermal analysis 
of the fuel assembly/fuel rack and fuel pool building interfaces 
currently approved remain valid and conservative.
    Therefore, the proposed change does not involve a significant 
reduction in the plant's 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.

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: March 30, 2006.
    Description of amendment request: The proposed amendments revise 
Technical Specification 3.3.3.6, ``Accident Monitoring 
Instrumentation,'' with respect to the required action for inoperable 
Wide Range Reactor Coolant Temperature, Wide Range Steam Generator 
Water Level, and Auxiliary Feedwater (AFW) Flow.
    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 increase in the allowed outage times for the Reactor 
Coolant Outlet Temperature--Wide Range, Reactor Coolant Inlet 
Temperature--Wide Range, Steam Generator [Water] Level--Wide Range, and 
the AFW Flow does not involve a significant increase in the probability 
of an accident previously evaluated because these are accident 
monitoring functions that have no effect on the potential for accident 
initiation. The proposed deletion of the existing requirements in 
ACTION 38 is an administrative change. Since these requirements are not 
currently applied to any plant equipment, this change cannot affect the 
probability of any accident previously evaluated.
    The proposed increase in the allowed outage times for the Reactor 
Coolant Outlet Temperature--Wide Range, Reactor Coolant Inlet 
Temperature--Wide Range, Steam Generator [Water] Level--Wide Range, and 
AFW Flow does not involve a significant increase in the consequences of 
an accident previously evaluated because the availability of redundant 
and diverse indications provides adequate assurance that the operator 
will be able to determine the post-accident status of the secondary 
heat sink.
    The proposed deletion of the existing requirements in ACTION 38 is 
an administrative change. Since these requirements are not currently 
applied to any plant equipment, this change cannot affect the 
consequence of any accident previously evaluated.
    (2) Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed increase in the allowed outage times for the Reactor 
Coolant Outlet Temperature--Wide Range, Reactor Coolant Inlet 
Temperature--Wide Range, Steam Generator [Water] Level--Wide Range, and 
the AFW Flow does not create the possibility of a new or different kind 
accident from any accident previously evaluated because the proposed 
change affects only the allowed outage time for accident monitoring 
instrumentation and involves no changes to plant design, plant 
configuration or operating procedures.
    The proposed deletion of the existing requirements in ACTION 38 is 
an administrative change. Since these requirements are not currently 
applied to any plant equipment, this change cannot create the 
possibility of any kind of accident.
    (3) Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed increase in the allowed outage times for the Reactor 
Coolant Outlet Temperature--Wide Range, Reactor Coolant Inlet 
Temperature--

[[Page 32609]]

Wide Range, Steam Generator [Water] Level--Wide Range, and AFW Flow 
does not involve a significant reduction in the margin of safety 
because the availability of redundant and diverse indications provides 
adequate assurance that the operator will be able to determine the 
post-accident status of the secondary heat sink.
    The proposed deletion of the existing requirements in ACTION 38 is 
an administrative change. Since these requirements are not currently 
applied to any plant equipment, this change cannot affect the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves 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, Units 1 and 2, Somervell County, Texas

    Date of amendment request: February 21, 2006.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 5.6.5 entitled, ``Core Operating Limits Report 
(COLR),'' to revise the listed Loss-of-Coolant Accident (LOCA) and non-
LOCA analysis methodologies used at Comanche Peak Steam Electric 
Station, Units 1 and 2.
    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 an administrative change only. 
Designation of the accident analysis methodologies, described in ERX-
04-004 and ERX-04-005, as approved analytical methods is required to 
maintain the accuracy of the Technical Specification 5.6.5 (Core 
Operating Limits Report) and to maintain consistency with the 
resolution of issues as prescribed in 10 CFR 50.46. 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 change involves an administrative change only. 
Technical Specification 5.6.5 is being changed to reference the revised 
accident analysis methodologies currently under NRC review. No actual 
plant equipment will be affected by the proposed change. 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.
    Margin of safety is associated with the confidence in the ability 
of the fission product barriers (i.e., fuel and fuel cladding, Reactor 
Coolant System pressure boundary, and containment structure) to limit 
the level of radiation dose to the public. This request involves an 
administrative change (subject to NRC approval) only to incorporate the 
NRC-approved methodologies into the allowable analysis methodologies 
specified in Technical Specification 5.6.5. No actual plant equipment 
will be affected by the proposed change. The compliance of the revised 
methodology with the requirements of 10 CFR 50.46 and Appendix K will 
be addressed through the NRC staff's review of the topical reports. 
Therefore, it is concluded that the use of the proposed methodology 
will not degrade the confidence in the ability of the fission product 
barriers to limit the level of radiation dose to the public. 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 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: February 21, 2006.
    Brief description of amendments: The amendments would revise 
Technical Specifications (TS) 3.3.1, 3.3.2, 3.4.5, 3.4.6, and 3.4.7, 
``Reactor Trip System (RTS) Instrumentation,'' ``Engineered Safety 
Feature System Actuation (ESFAS) Instrumentation,'' ``RCS [Reactor 
Coolant System] Mode 3,'' ``RCS Loops-Mode 4,'' and ``RCS Loops-Mode 5, 
Loops Filled,'' respectively. The revisions reflect the different steam 
generator water level trip setpoints and steam generator inventory 
requirements associated with the planned replacement of the steam 
generators in Comanche Peak Steam Electric Station, Unit 1.
    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 TS changes affect the protective and mitigative 
capabilities of the plant; 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 in which the steam generator water level trip functions are 
credited for initiating a protective or mitigative function. These 
transients and accidents have been analyzed and all relevant event 
acceptance criteria were shown to be satisfied. The radiological dose 
consequences are unaffected. Therefore, there is no increase in the 
consequences of an accident previously evaluated.
    The actual proposed setpoint values were determined using an 
uncertainty methodology previously approved by the NRC for this 
application. These values provide adequate assurance that required 
protective and mitigative functions will be initiated as assumed in the 
transient and accident analyses. Therefore, there is no increase in the 
consequences of an accident previously evaluated.
    The proposed revisions to the [Delta]76 steam generator inventory, 
required to ensure that the steam generators can provide an effective 
heat sink, are consistent with the current design requirements. 
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

[[Page 32610]]

accident from any accident previously evaluated?
    Response: No.
    No new accident scenarios, transient precursors, failure 
mechanisms, 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. 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. The possibility of a new or different malfunction 
of safety-related equipment 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 proposed changes to the Steam Generator Water Level-Low-Low and 
Steam Generator Water Level-High-High trip function setpoints protect 
the assumed safety analysis limits established in the transient and 
accident analyses. When used in the transient and accident analyses, 
all relevant event acceptance criteria are satisfied. Therefore, these 
proposed changes do not result in the reduction in a margin of safety.
    The proposed changes to the [Delta]76 steam generator inventory 
requirements, which ensure the steam generators can function as an 
effective heat sink during required shutdown operating modes, are 
consistent with the existing design and licensing bases. Therefore, 
these proposed changes do not result in the 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: 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.

Georgia Power Company, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia

    Date of amendment request: March 17, 2006.
    Brief description of amendment request: The proposed amendment 
would add a license condition to Section 2.C of the Edwin I. Hatch 
Nuclear Plant, Unit Nos. 1 and 2, Operating Licenses. This license 
condition will authorize the licensee to credit administering potassium 
iodide (KI) to reduce the 30-day post-accident thyroid radiological 
dose to the operators in the main control room for an interim period of 
approximately 4 years. In addition, the design-basis accident analysis 
section of the Updated Final Safety Analysis Reports will be updated to 
reflect crediting of KI.
    Date of publication of individual notice in Federal Register: March 
27, 2006 (71 FR 15223).
    Expiration date of individual notice: 30-day date April 26, 2006; 
60-day date May 26, 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 (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, 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.

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.
    Brief description of amendment: The amendment revised the Technical 
Specification (TS) to make permanent the temporary changes to TS Table 
3.3.8.1-1 previously approved by Amendment No. 147. TS Table 3.3.8.1-1 
is 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 
3.3.8.1.2, the monthly Channel Functional Test, on certain LOP time 
delay functions.
    Date of issuance: May 17, 2006.
    Effective date: As of the date of issuance and shall be implemented 
prior to expiration of the temporary change on June 1, 2006, provided 
by Amendment No. 147.
    Amendment No.: 151.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specfications.

[[Page 32611]]

    Date of initial notice in  Federal Register: March 14, 2006 (71 FR 
13173).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 17, 2006.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket No. 50-
278, Peach Bottom Atomic Power Station, Unit 3, York and Lancaster 
Counties, Pennsylvania

    Date of application for amendment: July 6, 2005, as supplemented 
March 15 and April 7, 2006.
    Brief description of amendments: The proposed changes extend the 
use of the Peach Bottom Atomic Power Station, Unit 3, pressure-
temperature (P-T) limits specified in the Technical Specifications 
(TSs) from 22 to 32 effective full-power years.
    Date of issuance: May 12, 2