[Federal Register Volume 71, Number 243 (Tuesday, December 19, 2006)]
[Notices]
[Pages 75987-76003]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E6-21346]


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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 November 22, 2006 to December 7, 2006. The 
last

[[Page 75988]]

biweekly notice was published on December 5, 2006 (71 FR 70553).

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, Rulemaking, 
Directives and Editing Branch, Division of Administrative Services, 
Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and should cite the publication date and 
page number of this Federal Register notice. Written comments may also 
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the Commission's 
Public Document Room (PDR), located at One White Flint North, Public 
File Area O1F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. The filing of requests for a hearing and petitions for leave 
to intervene is discussed below.
    Within 60 days after the date of publication of this notice, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR part 2. Interested persons should 
consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding

[[Page 75989]]

the request for a hearing. Any hearing held would take place after 
issuance of the amendment. If the final determination is that the 
amendment request involves a significant hazards consideration, any 
hearing held would take place before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, [email protected]; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

    Date of amendment request: September 15, 2006.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 6.8.5, ``Reactor Building 
Leakage Rate Testing Program,'' to allow a one-time deferral of the 
next Type A, containment integrated leak rate test (ILRT) from ``no 
later than September 2008'' to ``prior to startup from T1R18 refueling 
outage.'' The NRC has previously approved a one-time 5-year extension 
to the Type A ILRT schedule for TMI-1 by issuance of Amendment No. 244, 
dated August 14, 2003. Amendment No. 244 changed the TSs to state that 
the Type A ILRT shall be performed no later than September 2008. The 
proposed amendment would add approximately 15 months to the currently-
approved 15-year interval. This deferral would allow the Type A ILRT to 
be performed during a steam generator replacement in the fall of 2009.
    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:

    Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change will revise TS 6.8.5 to reflect a one-time 
extension to the Three Mile Island, Unit 1 Type A Integrated Leak 
Rate Test (ILRT) as currently specified in the Technical 
Specifications. This change will extend the requirement to perform 
the Type A ILRT from the current requirement of ``no later than 
September 2008'' to ``prior to startup from the T1R18 refueling 
outage,'' which is currently scheduled for Fall 2009. The current 
Type A ILRT interval of 15 years, based on past performance, would 
be extended on a one-time basis by approximately 15 months.
    The function of the containment is to isolate and contain 
fission products released from the reactor coolant system 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 TMI, Unit 1 containment will not exceed allowable leakage rate 
values specified in the TS and will continue to perform its design 
function following an accident. The risk assessment of the proposed 
change has concluded that there is an insignificant increase in 
postulated total population dose rate and an insignificant increase 
in the postulated conditional containment failure probability. 
Additionally, containment inspections have also been performed which 
demonstrate the continued structural integrity of the primary 
containment.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change for a one-time extension of the Type A ILRT 
for TMI, Unit 1 will not affect the control parameters governing 
unit operation or the response of plant equipment to transient and 
accident conditions. The proposed change does not introduce any new 
equipment, modes of system operation or failure mechanisms.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    Does the proposed amendment involve a significant reduction in a 
margin of safety?
    Response: No.
    The integrity of the 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 containment is 
verified by a Type A ILRT, as required by 10 CFR [Part] 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 containment at the 
design basis accident pressure. The proposed change for a one-time 
extension of the Type A ILRT does not affect the method for Type A, 
B or C testing or the test acceptance criteria.
    AmerGen has conducted a risk assessment to determine the impact 
of a change to the TMI, Unit 1 Type A ILRT schedule from a baseline 
ILRT frequency of three times in 10 years to once in 15 years plus 
15 months for the risk measures of Large Early Release Frequency 
(i.e., LERF), Population Dose, and Conditional Containment Failure 
Probability (i.e., CCFP). This assessment indicated that the 
proposed TMI, Unit 1 ILRT interval extension has a small change in 
risk to the public and is an acceptable plant change from a risk 
perspective.
    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.

[[Page 75990]]

    Attorney for licensee: Mr. Brad Fewell, Assistant General Counsel, 
Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA 
19348.
    NRC Branch Chief: Harold K. Chernoff.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of amendment request: May 30, 2006, as supplemented by letter 
dated November 20, 2006.
    Description of amendment request: The proposed amendment would 
modify the Technical Specification (TS) requirements related to steam 
generator tube integrity. The amendment would adopt Nuclear Regulatory 
Commission (NRC)-approved Revision 4 to Technical Specification Task 
Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
449, ``Steam Generator Tube Integrity.''
    The NRC staff published a notice of opportunity for comment in the 
Federal Register on March 2, 2005 (70 FR 10298), on possible amendments 
adopting TSTF-449, including a model safety evaluation and model no 
significant hazards consideration (NSHC) determination, using the 
consolidated line item improvement process. The NRC staff subsequently 
issued a notice of availability of the models for referencing in 
license amendment applications in the Federal Register on May 6, 2005 
(70 FR 24126). The licensee affirmed the applicability of the following 
NSHC determination in its application dated May 30, 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:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change requires a SG [Steam Generator] Program that 
includes performance criteria that will provide reasonable assurance 
that the SG tubing will retain integrity over the full range of 
operating conditions (including startup, operation in the power 
range, hot standby, cooldown and all anticipated transients included 
in the design specification). The SG performance criteria are based 
on tube structural integrity, accident induced leakage, and 
operational LEAKAGE.
    A Steam Generator Tube Rupture (SGTR) 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 Main Steam Line Break 
(MSLB), 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 TSs 
identifies 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 TSs. 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 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 change does not affect the 
consequences of an MSLB, rod ejection, or a reactor coolant pump 
locked rotor event, or other previously evaluated accident.

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

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

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

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

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.

[[Page 75991]]

    NRC Branch Chief (Acting): Douglas V. Pickett.

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

    Date of amendment request: May 23, 2006, as supplemented by letter 
dated October 3, 2006.
    Description of amendment request: The proposed amendment would 
modify the Technical Specification (TS) requirements related to steam 
generator tube integrity. The amendment would adopt Nuclear Regulatory 
Commission (NRC)-approved Revision 4 to Technical Specification Task 
Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
449, ``Steam Generator Tube Integrity.''
    The NRC staff published a notice of opportunity for comment in the 
Federal Register on March 2, 2005 (70 FR 10298), on possible amendments 
adopting TSTF-449, including a model safety evaluation and model no 
significant hazards consideration (NSHC) determination, using the 
consolidated line item improvement process. The NRC staff subsequently 
issued a notice of availability of the models for referencing in 
license amendment applications in the Federal Register on May 6, 2005 
(70 FR 24126). The licensee affirmed the applicability of the following 
NSHC determination in its application dated May 23, 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:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change requires a SG [Steam Generator] Program that 
includes performance criteria that will provide reasonable assurance 
that the SG tubing will retain integrity over the full range of 
operating conditions (including startup, operation in the power 
range, hot standby, cooldown and all anticipated transients included 
in the design specification). The SG performance criteria are based 
on tube structural integrity, accident induced leakage, and 
operational LEAKAGE.
    A Steam Generator Tube Rupture (SGTR) 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 Main Steam Line Break 
(MSLB), 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 TSs 
identifies 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 TSs. 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 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 change does not affect the 
consequences of an MSLB, rod ejection, or a reactor coolant pump 
locked rotor event, or other previously evaluated accident.

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

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

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

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

    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 (Acting): Douglas V. Pickett.

[[Page 75992]]

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, (HBRSEP) Unit No. 2, Darlington County, South Carolina

    Date of amendment request: June 1, 2006, as supplemented by letter 
dated November 20, 2006.
    Description of amendment request: The proposed amendment would 
revise the surveillance requirements (SR) for the emergency core 
cooling system suction inlet in the containment as specified in 
Technical Specification SR 3.5.2.6.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The proposed surveillance change will continue to ensure 
that the emergency core cooling system (ECCS) containment sump inlet 
is inspected in a manner that will verify operability. Performance 
of the required system surveillances, in conjunction with the 
applicable operational and design requirements for the ECCS, provide 
assurance that the system will be capable of performing the required 
design functions for accident mitigation and that the system will 
perform in accordance with the functional requirements for the 
system as described in the Updated Final Safety Analysis Report for 
HBRSEP, Unit No. 2. The proposed rewording of the surveillance 
requirement will continue to ensure that the ECCS containment sump 
suction inlet is not restricted by debris and suction inlet 
strainers show no evidence of structural distress or abnormal 
corrosion for HBRSEP, Unit No. 2. This ensures that the rate of 
occurrence and consequences of analyzed accidents will not change. 
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 previously evaluated?
    No. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated. HBRSEP, 
Unit No. 2, is replacing the existing ECCS containment sump inlet 
trash racks and screens with new strainers in accordance with the 
response to Generic Letter 2004-02. The strainer is a passive 
component in the ECCS, which is a standby safety system used for 
accident mitigation. As such, the strainer cannot be an accident 
initiator. A change to Technical Specifications Surveillance 
Requirement 3.5.2.6 is needed to accommodate the change to the ECCS 
containment sump inlet design. This change does not alter the nature 
of events postulated in the HBRSEP, Unit No. 2, Updated Final Safety 
Analysis Report, nor does it introduce any unique precursor 
mechanisms. Therefore, this change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    No. The proposed change does not involve a significant reduction 
in the margin of safety. The proposed change to the ECCS containment 
sump inlet surveillance requirement provides appropriate and 
applicable surveillance for this system. The proposed change to this 
surveillance requirement for the ECCS system will continue to ensure 
system operability. The proposed change does not adversely affect 
any plant safety limits, setpoints, or design parameters. The change 
also does not adversely affect the fuel, fuel cladding, Reactor 
Coolant System (RCS), or containment integrity. Therefore, this 
change does not affect any margin of safety for HBRSEP, Unit No. 2.

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

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power 
Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: May 31, 2006.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) requirements related to steam 
generator (SG) tube integrity. In particular, Dominion Nuclear 
Connecticut, Inc. (DNC) is proposing to replace the existing SG tube 
surveillance program with the NRC-approved Technical Specifications 
Task Force (TSTF) 449, Revision 4. The proposed changes are consistent 
with the Consolidated Line Item Improvement Process (CLIIP) provided in 
the May 6, 2005, Federal Register notice (70 FR 24126). In addition, 
the Millstone Power Station, Unit No. 2 (MPS2) TSs are revised beyond 
the scope of the CLIIP to provide consistent terminology and format. 
Basis for proposed no significant hazards consideration determination: 
DNC proposed minor variations and/or deviations from the TS changes 
described in the CLIIP beyond the scope of the no significant hazards 
consideration determination published on March 2, 2005. DNC has 
evaluated the proposed beyond-scope TS changes and determined it does 
not represent a significant hazards consideration. As required by 10 
CFR 50.91(a), DNC has provided its analysis of the issue of no 
significant hazards consideration. The NRC staff has reviewed the 
licensee's analysis against the standards of 10 CFR 50.92(c). The NRC 
staff's review is presented below.
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes do not affect initiators of previously 
analyzed events or assumed mitigation of accident or transient events.
    Therefore, these changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes involve adding a new definition and rewording 
the existing TS to be consistent with NUREG-1432, Revision 3. In 
addition, the requested change for MPS2 incorporates a more 
conservative leakage limit of 75 gallons per day per steam generator as 
opposed to the CLIIP specified limit of 150 gallons per day per steam 
generator. The changes will not impose any requirements or eliminate 
any existing requirements that will create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety. Since the 
proposed changes do not have an impact on any safety analysis 
assumptions and accidents previously evaluated, there are no margin of 
safety issues involved.
    Therefore, the changes do not involve a significant reduction in a 
margin of safety.
    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 Branch Chief: Harold K. Chernoff.

[[Page 75993]]

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

    Date of amendment request: May 31, 2006.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) requirements related to steam 
generator (SG) tube integrity. In particular, Dominion Nuclear 
Connecticut, Inc. (DNC) is proposing to replace the existing SG tube 
surveillance program with the NRC-approved Technical Specifications 
Task Force (TSTF) 449, Revision 4. The proposed changes are consistent 
with the Consolidated Line Item Improvement Process (CLIIP) provided in 
the May 6, 2005, Federal Register notice (70 FR 24126). In addition, 
the Millstone Power Station, Unit No. 3 (MPS3) TSs are revised beyond 
the scope of the CLIIP to provide consistent terminology and format.
    Basis for proposed no significant hazards consideration 
determination: DNC proposed minor variations and/or deviations from the 
TS changes described in the CLIIP beyond the scope of the no 
significant hazards consideration determination published on March 2, 
2005. DNC has evaluated the proposed beyond-scope TS changes and 
determined it does not represent a significant hazards consideration. 
As required by 10 CFR 50.91(a), DNC has provided its analysis of the 
issue of no significant hazards consideration to support this 
conclusion. The NRC staff has reviewed the licensee's analysis against 
the standards of 10 CFR 50.92(c). The NRC staff's review is presented 
below.
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes involve rewording the existing technical 
specifications to be consistent with NUREG-1431, Revision 3. These 
proposed changes do not affect initiators of previously analyzed events 
or assumed mitigation of accident or transient events.
    Therefore, these changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    These proposed changes do not involve physical alteration of the 
plant (no new or different type of equipment will be installed). The 
changes will not impose any requirements or eliminate any existing 
requirements that will create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    Since the proposed changes do not have an impact on any safety 
analysis assumptions and accidents previously evaluated, there are no 
margin of safety issues involved.
    Therefore, the changes do not involve a significant reduction in a 
margin of safety.
    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 Branch Chief: Harold K. Chernoff.

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

    Date of amendment request: November 13, 2006.
    Description of amendment request: The proposed amendment would 
revise Grand Gulf Nuclear Station, Unit 1, Technical Specification (TS) 
Limiting Condition for Operation (LCO) 3.10.1, and the associated TS 
Bases, to expand its scope to include provisions for temperature 
excursions greater than 200 [deg]F as a consequence of inservice leak 
and hydrostatic testing, and as a consequence of scram time testing 
initiated in conjunction with an inservice leak or hydrostatic test, 
while considering operational conditions to be in MODE 4.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 21, 2006 (71 FR 48561), on possible 
amendments to revise the plant-specific TS, to expand the scope of TS 
LCO 3.10.1, to include provisions for temperature excursions greater 
than 200 [deg]F as a consequence of inservice leak and hydrostatic 
testing, and as a consequence of scram time testing initiated in 
conjunction with an inservice leak or hydrostatic test, while 
considering operational conditions to be in MODE 4, including a model 
safety evaluation and model No Significant Hazards Determination 
(NSHC), 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 October 27, 2006 (71 FR 63050). The licensee affirmed the 
applicability of the model NSHC determination in its application dated 
November 13, 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:

Criterion 1: The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    Technical Specifications currently allow for operation at 
greater than [200] [deg]F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. Extending the activities that can apply this allowance will not 
adversely impact the probability or consequences of an accident 
previously evaluated. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.

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

    Technical Specifications currently allow for operation at 
greater than [200] [deg]F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. No new operational conditions beyond those currently allowed by 
LCO 3.10.1 are introduced. The changes do not involve a physical 
alteration of the plant (i.e., no new or different type of equipment 
will be installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements or eliminate any existing requirements. The 
changes do not alter assumptions made in the safety analysis. The 
proposed changes are consistent with the safety analysis assumptions 
and current plant operating practice. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.

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

    Technical Specifications currently allow for operation at 
greater than [200] [deg]F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. Extending the activities that can apply this allowance will not 
adversely impact any margin of safety. Allowing completion of 
inspections and testing and supporting completion of scram time 
testing initiated in conjunction with an inservice leak or 
hydrostatic test prior to power operation results in enhanced safe 
operations by eliminating unnecessary maneuvers to control reactor 
temperature and pressure. Therefore, the proposed change

[[Page 75994]]

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: Terence A. Burke, Associate General 
Council--Nuclear, Entergy Services, Inc., 1340 Echelon Parkway, 
Jackson, Mississippi 39213.
    NRC Branch Chief: David Terao.

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York 
and Lancaster Counties, Pennsylvania

    Date of application for amendments: July 14, 2006.
    Description of amendment request: The proposed changes would modify 
Technical Specification (TS) requirements related to required end 
states for TS action statements. The changes are generally consistent 
with the NRC-approved Revision 0 to Technical Specification Task Force 
(TSTF) Change Traveler, TSTF-423, ``Risk Informed Modification to 
Selected Required Action End States for BWR Plants.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on December 14, 2005 (70 FR 74037), on possible 
amendments adopting TSTF-423, 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 March 23, 2006 (71 FR 14726).
    The licensee affirmed the applicability of the following TSTF-423 
model NSHC determination in its application dated July 14, 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:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change allows a change to certain required end 
states when the TS Completion Times for remaining in power operation 
will be exceeded. Most of the requested technical specification (TS) 
changes are to permit an end state of hot shutdown (Mode 3) rather 
than an end state of cold shutdown (Mode 4) contained in the current 
TS. The request was limited to: (1) Those end states where entry 
into the shutdown mode is for a short interval, (2) entry is 
initiated by inoperability of a single train of equipment or a 
restriction on a plant operational parameter, unless otherwise 
stated in the applicable technical specification, and (3) the 
primary purpose is to correct the initiating condition and return to 
power operation as soon as is practical. Risk insights from both the 
qualitative and quantitative risk assessments were used in specific 
TS assessments. Such assessments are documented in Section 6 of GE 
[General Electric] NEDC-32988, Revision 2, ``Technical Justification 
to Support Risk Informed Modification to Selected Required Action 
End States for BWR [boiling-water reactor] Plants.'' They provide an 
integrated discussion of deterministic and probabilistic issues, 
focusing on specific technical specifications, which are used to 
support the proposed TS end state and associated restrictions. The 
staff finds that the risk insights support the conclusions of the 
specific TS assessments. Therefore, the probability of an accident 
previously evaluated is not significantly increased, if at all. The 
consequences of an accident after adopting proposed TSTF-423, are no 
different than the consequences of an accident prior to adopting 
TSTF-423. Therefore, the consequences of an accident previously 
evaluated are not significantly affected by this change. The 
addition of a requirement to assess and manage the risk introduced 
by this change will further minimize possible concerns.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.

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

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
If risk is assessed and managed, allowing a change to certain 
required end states when the TS Completion Times for remaining in 
power operation are exceeded, i.e., entry into hot shutdown rather 
than cold shutdown to repair equipment, will not introduce new 
failure modes or effects and will not, in the absence of other 
unrelated failures, lead to an accident whose consequences exceed 
the consequences of accidents previously evaluated. The addition of 
a requirement to assess and manage the risk introduced by this 
change and the commitment by the licensee to adhere to the guidance 
in TSTF-IG-05-02, Implementation Guidance for TSTF-423, Revision 0, 
``Technical Specifications End States, NEDC-32988-A,'' will further 
minimize possible concerns.
    Thus, this change does not create the possibility of a new or 
different kind of accident from an accident previously evaluated.

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

    The proposed change allows, for some systems, entry into hot 
shutdown rather than cold shutdown to repair equipment, if risk is 
assessed and managed. The BWROG's [Boiling Water Reactor Owner's 
Group's] risk assessment approach is comprehensive and follows staff 
guidance as documented in RGs [Regulatory Guides] 1.174 and 1.177. 
In addition, the analyses show that the criteria of the three-tiered 
approach for allowing TS changes are met. The risk impact of the 
proposed TS changes was assessed following the three-tiered approach 
recommended in RG 1.177. A risk assessment was performed to justify 
the proposed TS changes. The net change to the margin of safety is 
insignificant.

    Therefore, this change does not involve a significant reduction in 
a margin of safety.
    The NRC staff has reviewed the analysis and, based on this review, 
it appears that the three standards of 10 CFR 50.92(c) are satisfied. 
Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Attorney for Licensee: Mr. Brad Fewell, Assistant General Counsel, 
Exelon Generation Company LLC, 200 Exelon Way, Kennett Square, PA 
19348.
    NRC Branch Chief: Harold K. Chernoff.

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York 
and Lancaster Counties, Pennsylvania

    Date of application for amendments: September 15, 2006.
    Description of amendment request: The proposed amendment would 
revise the required frequency for control rod scram time testing, as 
described in Technical Specification (TS) Surveillance Requirement 
3.1.4.2, from ``120 days cumulative operation in MODE 1'' to ``200 days 
cumulative operation in MODE 1.'' The proposed TS change is based on 
the NRC-approved Revision 0 to Technical Specification Task Force 
(TSTF) Change Traveler, TSTF-460, ``Control Rod Scram Time Testing 
Frequency.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on May 27, 2004 (69 FR 30339), on possible amendments 
adopting TSTF-460, 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 August 23, 
2004 (69 FR 51864).
    The licensee affirmed the applicability of the following TSTF-460

[[Page 75995]]

model NSHC determination in its application dated September 15, 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 amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The frequency 
of surveillance testing is not an initiator of any accident 
previously evaluated. The frequency of surveillance testing does not 
affect the ability to mitigate any accident previously evaluated, as 
the tested component is still required to be operable. Therefore, 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The proposed 
change does not result in any new or different modes of plant 
operation. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The proposed 
change continues to test the control rod scram time to ensure the 
assumptions in the safety analysis are protected. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.

    The NRC staff has reviewed the 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 NSHC.
    Attorney for Licensee: Mr. Brad Fewell, Assistant General Counsel, 
Exelon Generation Company LLC, 200 Exelon Way, Kennett Square, PA 
19348.
    NRC Branch Chief: Harold K. Chernoff.

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of amendment request: November 14, 2006.
    Description of amendment request: The proposed amendment would 
revise Specification 3.3.5.1-1 of the Technical Specifications (TSs) to 
permit a one-time extension of the quarterly surveillance interval 
(i.e., from 92 days to 140 days) for three low pressure coolant 
injection (LPCI) loop select logic functions.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) Part 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration (NSHC), 
which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This amendment requests a one-time extension to the performance 
interval for a limited number of TS surveillance requirements. The 
performance of these surveillances, or the failure to perform, is 
not a precursor and does not affect the probability of an accident. 
Therefore, the delay in performance proposed in this amendment 
request for these surveillance requirements does not increase the 
probability of an accident previously evaluated.
    A delay in performing these surveillances does not result in a 
system being unable to perform its required function. In the case of 
this one-time extension, the relatively short period of additional 
time period for the systems and components to be in service prior to 
the next performance of the surveillance will not affect the ability 
of those systems to operate as designed. Therefore, the systems 
required to mitigate accidents will remain capable of performing 
their required function. No new failure modes have been introduced 
because of this action and the consequences remain consistent with 
previously evaluated accidents. Therefore, the proposed delay in 
performance of the surveillance requirements in this amendment 
request does not involve a significant increase in the consequences 
of an accident.
    Therefore, operation of the facility in accordance with the 
proposed license amendment would not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind or accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment does not involve a physical alteration of 
any system, structure, or component (SSC) or a change in the way any 
SSC is operated. The proposed amendment does not involve operation 
of any SSCs in a manner or configuration different from those 
previously recognized or evaluated. No new failure mechanisms will 
be introduced by the one-time surveillance requirement deferrals 
being requested.
    Thus, the proposed amendment does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment is a one-time extension of the 
performance interval of a limited number of TS surveillance 
requirements. Extending these surveillance requirements does not 
involve a modification of any TS Limiting Condition for Operation. 
Extending these surveillance requirements does not involve a change 
to any limit on accident consequences specified in the license or 
regulations. Extending these surveillance requirements does not 
involve a change to how accidents are mitigated or a significant 
increase in the consequences of an accident. Extending these 
surveillance requirements does not involve a change in a methodology 
used to evaluate consequences of an accident. Extending these 
surveillance requirements does not involve a change in any operating 
procedure or process.
    The instrumentation and components involved in this request have 
exhibited reliable operation based on the results of their 
performance during past periodic ECCS [emergency core cooling 
system] functional testing.
    Based on the limited additional period of time that the systems 
and components will be in service before the surveillances are next 
performed, as well as the operating experience that these 
surveillances are typically successful when performed, it is 
reasonable to conclude that the margins of safety associated with 
these surveillance requirements will not be affected by the 
requested extension.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The U.S. Nuclear Regulatory Commission (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: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Branch Chief: L. Raghavan.

Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant, 
Van Buren County, Michigan

    Date of amendment request: November 6, 2006.

[[Page 75996]]

    Description of amendment request: The proposed amendment would add 
the realistic large break loss-of-coolant accident (RLBLOCA) 
methodology to the analytical methods referenced in Technical 
Specification (TS) 5.6.5.b.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (CFR) Section 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed license amendment adds approved analytical methods 
used to determine the core operating limits per Technical 
Specification 5.6.5.b. Accidents previously evaluated will be 
unaffected because they will continue to be analyzed using 
applicable methodologies approved by the Nuclear Regulatory 
Commission to ensure all required safety limits are met. The 
proposed amendment does not affect the acceptance criteria for any 
Final Safety Analysis Report (FSAR) safety analysis analyzed 
accidents and anticipated operational occurrences. As such, the 
proposed amendment does not increase the probability or consequences 
of an accident. The proposed amendment does not involve operation of 
the required structures, systems or components (SSCs) in a manner or 
configuration different from those previously recognized or 
evaluated.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment does not involve a physical alteration of 
any SSC or a change in the way any SSC is operated. The proposed 
amendment does not involve operation of any required SSCs in a 
manner or configuration different from those previously recognized 
or evaluated. No new failure mechanisms will be introduced by the 
changes being requested.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment does not, by itself, introduce a failure 
mechanism. The proposed amendment does not involve any physical 
changes to the plant or manner in which the plant is operated. The 
proposed changes do not affect the acceptance criteria for any FSAR 
safety analysis analyzed accidents or anticipated operational 
occurrences. All required safety limits would continue to be 
analyzed using methodologies approved by the Nuclear Regulatory 
Commission.
    Therefore, the proposed amendment would 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: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Branch Chief: L. Raghavan.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: November 13, 2006.
    Description of amendment request: The proposed amendment would 
relocate the requirements of Technical Specification (TS) 2.22, ``Toxic 
Gas Monitors,'' and TS Table 3-3, Item 29 to the Fort Calhoun Station, 
Unit No. 1, Updated Safety Analysis Report (USAR).
    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 relocates requirements for toxic gas 
monitors that do not meet the criteria for inclusion in the TS set 
forth in 10 CFR 50.36(c)(2)(ii). The requirements for toxic gas 
monitors are being relocated from [the] TS to the USAR, which will 
be maintained pursuant to 10 CFR 50.59, thereby reducing the level 
of regulatory control. The level of regulatory control has no impact 
on the probability or consequences of an accident previously 
evaluated. Therefore, the change does not involve a significant 
increase in the probability or consequences 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 change relocates requirements for toxic gas 
monitors that do not meet the criteria for inclusion in [the] TS set 
forth in 10 CFR 50.36(c)(2)(ii). The change does not involve a 
physical alteration of the plant (no new or different type of 
equipment will be installed) or make changes in the methods 
governing normal plant operation. The change will not impose 
different requirements, and adequate control of information will be 
maintained. This change will not alter assumptions made in the 
safety analysis and licensing basis. Therefore, the change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Does th[e] [proposed] change involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change relocates requirements for toxic gas 
monitors that do not meet the criteria for inclusion in [the] TS set 
forth in 10 CFR 50.36(c)(2)(ii). The change will not reduce a margin 
of safety since the location of a requirement has no impact on any 
safety analysis assumptions. In addition, the relocated requirements 
for toxic gas monitors remain the same as the existing TS. Since any 
future changes to these requirements or the surveillance procedures 
will be evaluated per the requirements of 10 CFR 50.59, there will 
be no reduction in a margin of safety. [Therefore, the TS change 
does not involve a significant reduction in the margin of safety.]

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn, 
1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: David Terao.

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

    Date of amendment request: May 31, 2006.
    Description of amendment request: The proposed amendments would 
correct administrative errors in the SSES 1 and 2 Technical 
Specifications (TSs) by adding a logical connector in Condition B of 
Limiting Condition for Operation (LCO) 3.8.1 (SSES 1 TS only) and 
correct the routing of Interstate 80 (I-80) on Figure 4.1-2 in the SSES 
1 and 2 TSs Section 4.0.
    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?

[[Page 75997]]

    Response: No.

Change to Technical Specification 3.8.1

    The proposed change is administrative in nature and does not 
impact any accident initiators or analyzed events or assumed 
mitigation of accident or transient events. They do not involve the 
addition or removal of any equipment, or any design changes to the 
facility. Therefore, this proposed change does not represent a 
significant increase in the probability or consequences of an 
accident previously evaluated.

Change to Technical Specification Figure 4.1-2

    The proposed change is administrative in nature and does not 
impact any accident initiators or analyzed events or assumed 
mitigation of accident or transient events. It does not involve the 
addition or removal of any equipment or any design changes to the 
facility.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.

Change to Technical Specification 3.8.1

    The proposed change is an administrative change and does not 
involve a modification to the physical configuration of the plant 
(i.e., no new equipment will be installed) or change in the methods 
governing normal plant operation. The proposed change will not 
impose any new or different requirements or introduce a new accident 
initiator, accident precursor, or malfunction mechanism. 
Additionally, there is no change in the types or increases in the 
amounts of any effluent that may be released off-site, and there is 
no increase in individual or cumulative occupational exposure. 
Therefore, this proposed change does not create the possibility of 
an accident of a different kind than previously evaluated.

Change to Technical Specification Figure 4.1-2

    The proposed change is an administrative change and will not 
impose any new or different requirements or introduce a new accident 
initiator, accident precursor, or malfunction mechanism. 
Additionally, there is no change in the types or increases in the 
amounts of any effluent that may be released off-site, and there is 
no increase in individual or cumulative occupational exposure.
    Therefore, this 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.

Change to Technical Specification 3.8.1

    The proposed change revises Condition B in LCO 3.8.1 to be 
consistent with Technical Specification 1.2, ``Logical Connectors.'' 
This change is administrative in nature. Therefore, this proposed 
change does not involve a significant reduction in a margin of 
safety.

Change to Technical Specification Figure 4.1-2

    The proposed change is administrative in nature and does not 
affect any plant systems.
    Therefore, this 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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Branch Chief: Richard J. Laufer.

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

    Date of amendment request: September 7, 2006.
    Description of amendment request: The proposed amendments would 
revise the SSES 1 and 2 Technical Specification (TSs) Section 5.5.6, 
``Inservice Testing Program,'' and TS 5.5.12, ``Primary Containment 
Leakage Rate Testing Program,'' to be consistent with the requirements 
of Title 10 of the Code of Federal Regulations (10 CFR) Section 
50.55a(f)(4) and 10 CFR 50.55a(g)(4), respectively. The proposed 
amendments would implement TS Task Force (TSTF) 343, Revision 1 and 
TSTF 479, Revision 0.
    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.

Change to Technical Specification 5.5.6

    The proposed change revises the Inservice Testing Program for 
consistency with the requirements of 10 CFR 50.55a(f)(4) for pumps 
and valves which are classified as American Society of Mechanical 
Engineers (ASME) Code Class 1, Class 2 and Class 3.
    The proposed change does not impact any accident initiators or 
analyzed events or assumed mitigation of accident or transient 
events. It does not involve the addition or removal of any 
equipment, or any design changes to the facility. Therefore, this 
proposed change does not represent a significant increase in the 
probability or consequences of an accident previously evaluated.

Change to Technical Specification 5.5.12

    The proposed change revises the TS administrative controls 
programs for consistency with the requirements of 10 CFR [Part] 50, 
paragraph 55a (g)(4) for components classified as Code Class CC.
    The proposed change affects the frequency of visual examinations 
that will be performed for the concrete surfaces of the containment 
for the purpose of the Primary Containment Leakage Rate Testing 
Program. In addition, the proposed change allows those examinations 
to be performed during power operation as opposed to during a 
refueling outage. The frequency of visual examinations of the 
concrete surfaces of the containment and the mode of operation 
during which those examinations are performed has no relationship to 
or adverse impact on the probability of any of the initiating events 
assumed in the accident analyses. The proposed change would allow 
visual examinations that are performed pursuant to NRC approved ASME 
Section XI Code requirement (except where relief has been granted by 
the NRC) to meet the intent of visual examinations required by 
Regulatory Guide 1.163, without requiring additional visual 
examinations pursuant to the Regulatory Guide. The intent of early 
detection of deterioration will continue to be met by the more 
rigorous requirements of the Code required visual examinations. As 
such, the safety function of the containment as a fission product 
barrier is maintained.
    The proposed change does not impact any accident initiators or 
analyzed events or assumed mitigation of accident or transient 
events. It does not involve the addition or removal of any 
equipment, or any design changes to the facility.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.

Change to Technical Specification 5.5.6

    The proposed change revises the Inservice Testing Program for 
consistency with the requirements of 10 CFR 50.55a(f)(4) for pumps 
and valves which are classified as American Society of Mechanical 
Engineers (ASME) Code Class 1, Class 2 and Class 3.
    The proposed change does not involve a modification to the 
physical configuration of the plant (i.e., no new equipment will be 
installed) or change in the methods governing normal plant 
operation. The proposed change will not impose any new or different 
requirements or introduce a new accident initiator, accident 
precursor, or malfunction mechanism. Additionally, there is no 
change in the types or increases in the amounts of any effluent that 
may be released off-site and there is no increase in individual or 
cumulative occupational exposure. Therefore, this proposed change 
does not

[[Page 75998]]

create the possibility of an accident of a different kind than 
previously evaluated.

Change to Technical Specification 5.5.12

    The proposed change revises the TS administrative controls 
programs for consistency with the requirements of 10 CFR [Part] 50, 
paragraph 55a (g)(4) for components classified as Code Class CC.
    The change affects the frequency of visual examinations that 
will be performed for the concrete surfaces containments. In 
addition, the proposed change allows those examinations to be 
performed during power operation as opposed to during a refueling 
outage. The proposed change does not involve a modification to the 
physical configuration of the plant (i.e., no new equipment will be 
installed) or change in the methods governing normal plant 
operation. The proposed change will not impose any new or different 
requirements or introduce a new accident initiator, accident 
precursor, or malfunction mechanism. Additionally, there is no 
change in the types or increases in the amounts of any effluent that 
may be released off-site and there is no increase in individual or 
cumulative occupational exposure.
    Therefore, this 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.

Change to Technical Specification 5.5.6

    The proposed change revises the Inservice Testing Program for 
consistency with the requirements of 10 CFR 50.55a(f)(4) for pumps 
and valves which are classified as American Society of Mechanical 
Engineers (ASME) Code Class 1, Class 2 and Class 3. The safety 
function of the affected pumps and valves will be maintained. 
Therefore, this proposed change does not involve a significant 
reduction in a margin of safety.

Change to Technical Specification 5.5.12

    The proposed change revises the Improved Standard Technical 
Specification Administrative Controls program requirements for 
consistency with the requirements of 10 CFR [Part] 50, paragraph 55a 
(g)(4) for components classified as Code Class CC.
    The change affects the frequency of visual examinations that 
will be performed for the concrete surfaces of containments. In 
addition, the proposed change allows those examinations to be 
performed during power operation as opposed to during a refueling 
outage. The safety function of the containment as a fission product 
barrier will be maintained.
    [Therefore, this 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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Branch Chief: Richard J. Laufer.

PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric 
Station, Unit 2 (SSES 2), Luzerne County, Pennsylvania

    Date of amendment request: November 16, 2006.
    Description of amendment request: The proposed amendment would 
revise the SSES 2 Technical Specification (TS) Section 2.1.1.2 to 
reflect the Unit 2 Cycle 14 (U2C14) Minimum Critical Power Ratio (MCPR) 
Safety Limits for two-loop and single-loop operation. Additionally, TS 
Section 5.6.5.b would be revised to reflect the NRC-approved 
methodology used in the MCPR Safety Limit Analysis.
    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:

    2. Does the proposed change involve a significant increase in 
the probability [* * *] or consequences of an accident previously 
evaluated?
    Response: No.
    The proposed change to the two-loop and single-loop MCPR Safety 
Limits do not directly or indirectly affect any plant system, 
equipment, component, or change the processes used to operate the 
plant. Further, the proposed U2C14 MCPR Safety Limits were generated 
using NRC-approved methodology and meet the applicable acceptance 
criteria. Thus, this proposed amendment does not involve a 
significant increase in the probability of occurrence or 
consequences of an accident previously evaluated.
    Prior to the startup of U2C14, licensing analyses are performed 
(using NRC-approved methodology referenced in Technical 
Specification Section 5.6.5.b) to determine changes in the critical 
power ratio as a result of anticipated operational occurrences. 
These results are added to the MCPR Safety Limit values to generate 
the MCPR operating limits in the U2C14 COLR [Core Operating Limits 
Report]. These limits could be different from those specified for 
the previous Unit 2 COLR. The COLR operating limits thus assure that 
the MCPR Safety Limit will not be exceeded during normal operation 
or anticipated operational occurrences. Postulated accidents are 
also analyzed prior to the startup of U2C14 and the results shown to 
be within the NRC-approved criteria.
    The changes to the references in Section 5.6.5.b were made to 
properly reflect the NRC-approved methodology used to generate the 
U2C14 core operating limits. The use of this approved methodology 
does not increase the probability [* * *] or consequences of an 
accident previously evaluated.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability [* * *] or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The changes to the two-loop and single-loop MCPR Safety Limits 
do not directly or indirectly affect any plant system, equipment, or 
component and therefore does not affect the failure modes of any of 
these items. Thus, the proposed change does not create the 
possibility of a previously unevaluated operator error or a new 
single failure.
    The changes to the references in Section 5.6.5.b were made to 
properly reflect the NRC-approved methodology used to generate the 
U2C14 core operating limits. The use of this approved methodology 
does not create the possibility of a new or different kind of 
accident.
    Therefore, this proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Since the proposed changes do not alter any plant system, 
equipment, component, or the processes used to operate the plant, 
the proposed change will not jeopardize or degrade the function or 
operation of any plant system or component governed by Technical 
Specifications. The proposed two-loop and single-loop MCPR Safety 
Limits do not involve a significant reduction in the margin of 
safety as currently defined in the Bases of the applicable Technical 
Specification sections because the MCPR Safety Limits calculated for 
U2C14 preserve the required margin of safety.
    The changes to the references in Section 5.6.5.b were made to 
properly reflect the NRC-approved methodology used to generate the 
U2C14 core operating limits. This approved methodology is used to 
demonstrate that all applicable criteria are met, thus, 
demonstrating that there is no reduction in 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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Branch Chief: Richard J. Laufer.

[[Page 75999]]

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 4, 2006.
    Description of amendment request: The amendments would revise the 
Technical Specifications (TSs) to allow the movement of irradiated fuel 
inside containment to commence at 24 hours after shutdown or at the 
decay time calculated using the licensee's spent fuel pool integrated 
decay heat management program, whichever is later.
    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 of occurrence or consequences of an accident previously 
evaluated?
    Response: No.
    The proposed license amendment would allow fuel assemblies to be 
removed from the reactor core and be stored in the Spent Fuel Pool 
in less time after subcriticality (but more accurately calculated), 
than currently allowed by the TSs. Decreasing the decay time of the 
fuel affects the radionuclide make-up of the fuel to be offloaded as 
well as the amount of decay heat that is present from the fuel at 
the time of offload. The proposed changes do not involve a 
significant increase in the probability of occurrence of an accident 
previously evaluated. The accident previously evaluated that is 
associated with the proposed license amendment is the fuel handling 
accident. Allowing the fuel to be offloaded based on the IDHM 
[integrated decay heat management program] calculated time after 
subcriticality does not impact the manner in which the fuel is 
offloaded. The accident initiator is the dropping of the fuel 
assembly. Since earlier offload does not affect fuel handling, there 
is no increase in the probability of occurrence of a fuel handling 
accident. The time frame in which the fuel assemblies are moved has 
been evaluated against the 10 CFR 50.67 dose limits for members of 
the public, licensee personnel and control room. Additionally, the 
guidance provided in Reg. Guide 1.183 was used for the selective 
application of Alternative Source Term. All dose limits are met with 
the reduced core offload times; and significant margin is 
maintained, as the minimum decay time prior to movement of fuel for 
the FHA [fuel handling accident] analysis is 24 hours.
    Therefore, the proposed license amendment does not increase the 
probability of occurrence or the consequences of accidents 
previously evaluated are not increased.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Response: No.
    The proposed license amendment would allow core offload to occur 
in less time after subcriticality (but more accurately calculated), 
which affects the radionuclide make-up of the fuel to be offloaded 
as well as the amount of decay heat that is present from the fuel at 
the time of offload. The radionuclide makeup of the fuel assemblies 
and the amount of decay heat produced by the fuel assemblies do not 
currently initiate any accident. A change in the radionuclide makeup 
of the fuel at the time of core offload or an increase in the decay 
heat produced by the fuel being offloaded will not cause the 
initiation of any accident. The accident previously evaluated that 
is associated with fuel movement is the fuel handling accident. 
There is no change to the manner in which fuel is being handled or 
in the equipment used to offload or store the fuel. The effects of 
the additional decay heat load have been analyzed. The analysis 
demonstrated that the existing Spent Fuel Pool cooling system and 
associated systems under worst-case circumstances would maintain the 
integrity of the Spent Fuel Pool. The proposed method of offload 
does not create a new or different kind of accident from any 
accident previously evaluated.
    Therefore, the proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The margin of safety pertinent to the proposed changes is the 
dose consequences resulting from a fuel handling accident. The 
shorter decay time prior to fuel movement has been evaluated against 
10 CFR part 50.67 and all limits continue to be met. All dose limits 
are met with the reduced core offload times; and significant margin 
is maintained, as the minimum decay time prior to movement of fuel 
for the FHA analysis is 24 hours. Decay heatup calculations 
performed prior to each refueling outage as part of the IDHM program 
ensure that planned spent fuel transfer to the SFP [spent fuel pool] 
will not result in maximum SFP temperature exceeding the design 
basis limit of 149 [deg]F (with both heat exchangers available) or 
180 [deg]F (with one heat exchanger alternating between the two 
pools). As stated above, the changes in radionuclide makeup and 
additional heat load do not impact any safety settings and do not 
cause any safety limit to not be met. In addition, the integrity of 
the Spent Fuel Pool is maintained.
    The time frame in which the fuel assemblies are moved has been 
evaluated against the 10 CFR 50.67 dose limits for members of the 
public, licensee personnel and control room. Additionally, the 
guidance provided in Reg. Guide 1.183 was used. Calculations 
performed conclude that expected dose limits following a Fuel 
handling Accident are met with the proposed decay time prior to 
commencing fuel movement.
    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 Branch Chief: Harold K. Chernoff.

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: November 7, 2006.
    Description of amendment requests: The amendments request to revise 
Main Steam Safety Valve Requirements and Actions (Technical 
Specification 3.7.1).
    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. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Based on a detailed plant transient analysis, the Limiting 
Conditions for Operation (LCOs) and Action statements will continue 
to restrict operation to within the regions that provide acceptable 
results. The safety analysis was performed in accordance with the 
Nuclear Regulatory Commission (NRC) approved San Onofre Units 2 and 
3 reload analysis methodology, and considered the concerns 
identified in NRC Information Notice 94-60.
    The increase in Completion Time for Required Action 3.7.1.A.2 
from 12 hours to 36 hours is consistent with NUREG-1432 Revision 3, 
``Standard Technical Specifications for Combustion Engineering 
Plants.''
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not add any new equipment, modify any 
interfaces with any existing equipment, alter the equipment's 
function, or change the method of operating the equipment. The 
proposed change does not alter plant conditions in a manner that 
could affect other plant components. The proposed change does not

[[Page 76000]]

cause any existing equipment to become an accident initiator.
    The increase in Completion Time for Required Action 3.7.1.A.2 
from 12 hours to 36 hours is consistent with NUREG-1432 Revision 3, 
``Standard Technical Specifications for Combustion Engineering 
Plants.''
    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 Limiting Conditions for Operation (LCOs) and Action 
statements will continue to restrict operation such that the 
American Society of Mechanical Engineers (ASME) code requirements 
continue to be met. The analyses were performed using the NRC 
approved San Onofre Units 2 and 3 reload analysis methodology. 
Therefore, the proposed change will have no impact on the margins as 
defined in the Technical Specification bases.
    The increase in Completion Time for Required Action 3.7.1.A.2 
from 12 hours to 36 hours is consistent with NUREG-1432 Revision 3, 
``Standard Technical Specifications for Combustion Engineering 
Plants.''
    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 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.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, Limestone County, 
Alabama

    Date of amendment request: November 15, 2006 (TS-459).
    Description of amendment request: The proposed amendment requests 
revision to the Fire Protection License Condition for Units 1, 2, and 
3, condition number (13), (14), and (7), respectively, to accommodate 
operation of Units 1, 2, and 3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    No. The proposed change revises the license condition to reflect 
a combined Units 1, 2 and 3 Fire Protection Report. Compliance with 
the applicable Appendix R requirements is ensured through 
implementation of the Fire Protection Program and the Appendix R 
Safe Shutdown Program including Regulatory Issue Summary 2006-10, 
``Regulatory Expectations with Appendix R Paragraph III.G.2 Post-
Fire Manual Actions.'' The change does not affect any design bases 
accident or the ability of any safe shutdown equipment to perform 
its function. Also, although modifications were required to bring 
BFN in compliance with 10 CFR 50 Appendix R, there are no physical 
modifications required to implement this license amendment.
    Therefore, this proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    3. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The proposed change revises the license condition to reflect 
a combined Units 1, 2 and 3 Fire Protection Report. Compliance with 
the applicable Appendix R requirements is ensured through 
implementation of the Fire Protection Program and Appendix R Safe 
Shutdown Program including Regulatory Issue Summary 2006-10, 
``Regulatory Expectations with Appendix R Paragraph III.G.2 Post-
Fire Manual Actions.'' This change does not affect any design basis 
accident or the ability of any safe shutdown equipment to perform 
its function. Also, there are no physical modifications required to 
implement this license amendment. Therefore, this proposed change 
does not create the possibility of a new or different kind of 
accident from an accident previously evaluated.
    4. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The proposed change revises the license condition to reflect 
a combined Units 1, 2 and 3 Fire Protection Report. Compliance with 
the applicable Appendix R requirements is ensured through the 
implementation of the Fire Protection Program and Appendix R Safe 
Shutdown Program (Units 1, 2, and 3 Fire Protection Report) 
including Regulatory Issue Summary 2006-10, ``Regulatory 
Expectations with Appendix R Paragraph III.G.2 Post-Fire Manual 
Actions.'' The proposed change does not affect any design basis 
accident and does not reduce or adversely affect the capability to 
achieve and maintain safe shutdown in the event of a fire. 
Furthermore, no reductions to the requirements for equipment 
operability, surveillance requirements or setpoints are being made 
which could result in reduction in the margin of safety. Therefore, 
this proposed change will not result in a reduction in the margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: L. Raghavan.

Notice of Issuance of Amendments to Facility Operating Licenses

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

[[Page 76001]]

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

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

    Date of application for amendments: December 23, 2005, as 
supplemented by letters dated May 4 and August 3, 2006.
    Brief description of amendments: These amendments revise Technical 
Specification (TS) 3.8.1, ``AC Sources--Operating,'' to extend the 
allowed out of service time for one inoperable emergency diesel 
generator from 72 hours to 10 days. TS 3.8.3, ``Diesel Fuel Oil, Lube 
Oil, and Starting Air,'' is revised by the addition of a clarifying 
note to Condition F of this specification. Additionally, TS 3.4.9, 
``Pressurizer,'' is revised to delete the words contained in the 
limiting condition for operation which require that the two groups of 
pressurizer heaters are capable of being powered from an emergency 
power supply.
    Date of issuance: December 5, 2006.
    Effective date: As of the date of issuance to be implemented within 
90 days from the date of issuance.
    Amendment Nos.: Unit 1--164, Unit 2--164, Unit 3--164
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Operating Licenses for all three units.
    Date of initial notice in Federal Register: January 31, 2006 (71 FR 
5080). The May 4 and August 3, 2006, supplemental letters provided 
additional information that clarified the application as originally 
noticed, and did not change the staff's original no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 5, 2006.
    No significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of application for amendment: September 30, 2004, as 
supplemented by letters dated March 16, September 29, 2005, and March 
21, August 7, August 24, and September 11, 2006.
    Brief description of amendment: The license amendment request 
revised the technical specifications and the final safety analysis 
report to amend the Columbia Generating Station's licensing and design 
bases to reflect the application of the alternative source term 
methodology with an exception. That exception is the Technical 
Information Document (TID)-14844, ``Calculation of Distance Factors for 
Power and Test Reactor Sites,'' which will continue to be used as the 
radiation dose basis for equipment qualification, and radiation zone 
maps/shielding calculations.
    Date of issuance: November 27, 2006.
    Effective date: As of its date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment No.: 199.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications and Final Safety Analysis Report.
    Date of initial notice in Federal Register: October 26, 2004 (69 FR 
62472). The March16, September 29, 2005, and March 21, August 7, August 
24, and September 11, 2006, supplemental letters provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 27, 2006.
    No significant hazards consideration comments received: No.

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

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

    Date of application for amendment: October 3, 2005.
    Brief description of amendment: The amendment revised the Technical 
Specification (TS) Section 1.1, ``Definitions,'' description of the 
Pressure and Temperature Limits Report (PTLR), by deleting reference to 
specifications containing limits in the PTLR; (2) revised 
administrative controls TS 5.6.6, ``Reactor Coolant System (RCS) 
Pressure and Temperature Limits Report (PTLR),'' by requiring the NRC 
approval documents to be identified by date and topical reports to be 
identified by number and title; and (3) added Westinghouse Electric 
Company, LLC report, WCAP-16143, ``Reactor Vessel Closure Head/Vessel 
Flange Requirements Evaluation for Byron/Braidwood Units 1 and 2,'' to 
the list of analytical methods provided in TS 5.6.6. The amendment also 
revises the title of the NRC letter dated August 8, 2001 to clarify the 
regulation being referenced.
    Date of issuance: November 27, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 148, 148, 142, 142.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications and License.
    Date of initial notice in Federal Register: March 14, 2006 (71 FR 
13175).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 27, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 2, 2006, as supplemented by 
letters dated August 18 and October 5, 2006.
    Brief description of amendment: The amendments revised Technical 
Specification Surveillance Requirement 3.1.7.10, ``Standby Liquid 
Control System Sodium Pentaborate Isotopic Enrichment'' such that the 
required enrichment increases from >= 30.0 atom percent boron-10 to >= 
45.0 atom percent boron-10.
    Date of issuance: November 16, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 222/214.
    Renewed Facility Operating License Nos. DPR-19 and DPR-25: The 
amendments revised the Technical Specification Surveillance Requirement 
and Licenses.
    Date of initial notice in Federal Register: (71 FR 46931; August 
15, 2006).
    The August 18 and October 5, 2006, supplements provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the NRC 
staff's original proposed no significant hazards consideration 
determination as published in the Federal Register on August 15, 2006 
(71 FR 46931). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 16, 2006.
    No significant hazards consideration comments received: No.

[[Page 76002]]

FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of application for amendment: November 14, 2005, as 
supplemented by letter dated September 1, 2006.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) Table 3.3.6.1-1, ``Primary Containment Isolation 
Instrumentation,'' to eliminate the Main Steamline Radiation Monitor 
trip function.
    Date of issuance: November 15, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 261.
    Facility Operating License No. DPR-49: The amendment revises the 
TSs.
    Date of initial notice in Federal Register: (71 FR 43533) August 1, 
2006.
    The supplement provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the NRC staff's original proposed no 
significant hazards consideration determination, as published in the 
Federal Register on August 1, 2006 (71 FR 43533).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 15, 2006.
    No significant hazards consideration comments received: No.

FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of application for amendment: December 22, 2005.
    Brief description of amendment: The amendment revises the Duane 
Arnold Energy Center licensing basis, as described in the Updated Final 
Safety Analysis Report (UFSAR), to replace the current plant-specific 
reactor pressure vessel material surveillance program with the Boiling 
Water Reactor Vessel and Internals Project Integrated Surveillance 
Program as the basis for demonstrating compliance with the requirements 
of Appendix H to Part 50 of Title 10 of the Code of Federal 
Regulations, ``Reactor Vessel Material Surveillance Program 
Requirements.''
    Date of issuance: November 27, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 262.
    Facility Operating License No. DPR-49: The amendment authorizes 
changes to the UFSAR.
    Date of initial notice in Federal Register: (71 FR 43533) August 1, 
2006.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 27, 2006.
    No significant hazards consideration comments received: No.

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 application for amendments: November 9, 2005, supplemented 
by letter dated May 15, 2006.
    Brief description of amendments: The amendments modify the 
Technical Specifications (TS) for Prairie Island Nuclear Generating 
Plant Units 1 and 2, to clarify which TS Surveillance Requirements 
shall be met for the TS systems which include more components 
(installed spare components) than are required to satisfy the TS 
Limiting Conditions for Operation. These amendments revise TS 3.7.8, 
``Cooling Water (CL) System,'' TS 3.8.1, ``AC Sources-Operating,'' and 
TS 3.9.3, ``Nuclear Instrumentation.'' The amendments also make minor 
corrections for some of these TSs.
    Date of issuance: November 14, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 175 and 165.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 14, 2006 (71 
FR 7809).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 14, 2006. The supplemental 
information provided in letter May 15, 2006, did not impact the 
conclusions of the Determination of No Significant Hazards 
Consideration and Environmental Assessment presented in the November 9, 
2005 submittal.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket No. 50-275, Diablo Canyon 
Nuclear Power Plant, Unit No. 1, San Luis Obispo County, California

    Date of application for amendment: December 16, 2005, as 
supplemented by a letter dated September 27, 2006.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 5.6.5, ``Core Operating Limits Report (COLR).'' 
Specifically, the change added Westinghouse Topical Report WCAP-12945-
P-A, Addendum 1-A, Revision 0, ``Method for Satisfying 10 CFR 50.46 
Reanalysis Requirements for Best Estimate LOCA [Loss-of-Coolant 
Accident] Evaluation Models,'' dated December 2004, to the list of 
approved analytical methods in TS 5.6.5.b.
    Date of issuance: November 21, 2006.
    Effective date: As of its date of issuance, and shall be 
implemented within 90 days of issuance.
    Amendment No.: 191.
    Facility Operating License No. DPR-80: The amendment revised the 
Technical Specifications and Facility Operating License.
    Date of initial notice in Federal Register: February 14, 2006 (71 
FR 7810).
    The September 27, 2006, supplemental letter provided additional 
information that clarified the application, and did not expand the 
scope of the application as originally noticed.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 21, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 6, 2005, as 
supplemented by letters dated March 14 and November 30, 2006.
    Brief description of amendments: The amendment deleted Technical 
Specification (TS) Limiting Condition for Operation (LCO) 3.3.10, 
``Fuel Handling Isolation Signal (FHIS),'' and TS LCO 3.7.14, ``Fuel 
Handling Building Post-Accident Cleanup Filter System,'' and their 
associated surveillance requirements. The amendment also deleted the 
Fuel Handling Building Post-Accident Cleanup Filter Systems from the 
Ventilation Filter Testing Program in administrative TS 5.5.2.12.
    Date of issuance: December 4, 2006.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: Unit 2--208; Unit 3--200.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: January 3, 2006 (71 FR 
155). The March 14 and November 30, 2006, supplemental letters provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed,

[[Page 76003]]

and did not change the staff's original no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 4, 2006.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-259 Browns Ferry Nuclear 
Plant, Unit 1, Limestone County, Alabama

    Date of application for amendment: October 12, 2004, as 
supplemented by letters dated September 7 and November 1, 2006.
    Brief description of amendment: To remove License Condition 2.C(4).
    Date of issuance: November 28, 2006.
    Effective date: As of date of issuance, to be implemented within 30 
days.
    Amendment No.: 265
    Renewed Facility Operating License Nos. DPR-33: Amendment revised 
the Renewed Operating License.
    Date of initial notice in Federal Register: August 15, 2006 (71 FR 
46937). The supplements dated September 7 and November 1, 2006, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 28, 2006.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: May 25, 2006, as supplemented 
by letter dated September 1, 2006.
    Brief description of amendments: The requested changes provide a 
revision to the design and licensing basis for the containment sump 
debris transport analysis as described in the Sequoyah Nuclear Plant 
(SQN) Updated Final Safety Analysis Report (UFSAR). The current 
transport analysis for SQN is a two-dimensional physical transport 
model, and Tennessee Valley Authority (TVA) is requesting to update the 
analysis to a three-dimensional computational fluid dynamics transport 
model. The results of the reanalysis will be used to size the flow area 
of the advanced design containment sump strainers which will replace 
the original sump intake structure.
    Date of issuance: November 7, 2006.
    Effective date: Implementation of the amendment is the 
incorporation into the next UFSAR update made in accordance with 10 CFR 
50.71(e), of the changes to the description of the facility as 
described in TVA's application dated May 25, 2006, as supplemented by 
letter dated September 1, 2006, and evaluated in the staff's Safety 
Evaluation attached to this amendment.
    Amendment Nos. 313 and 302.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revised the Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: June 20, 2006 (71 FR 
35460). The supplemental letter dated September 1, 2006, provided 
clarifying information that was within the scope of the initial notice 
and did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 7, 2006.
    No significant hazards consideration comments received: No.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application for amendment: September 20, 2006, as 
supplemented by letter dated November 20, 2006.
    Brief description of amendment: The amendment revised (1) the 
definition of the Pressure and Temperature Limits Report (PTLR) in 
Technical Specification (TS) 1.1, ``Definitions,'' and (2) TS 5.6.6, 
``Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT 
(PTLR).''
    Date of issuance: December 5, 2006.
    Effective date: As of its date of issuance, and shall be 
implemented within 90 days of the date of issuance.
    Amendment No.: 177.
    Facility Operating License No. NPF-30: The amendment revised the 
Operating License and Technical Specifications.
    Date of initial notice in Federal Register: October 6, 2006 (71 FR 
59136).
    The supplemental letter dated November 20, 2006, provided 
additional clarifying information, did not expand the scope of the 
application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 5, 2006.
    No significant hazards consideration comments received: No.

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.

Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear 
Plant, Unit 1, Limestone County, Alabama

    Date of application for amendments: November 9, 2006 (TS-458).
    Description of amendments request: The proposed amendment would 
delete the Technical Specification Surveillance Requirement to verify 
the position of a low pressure coolant injection crosstie valve.
    Date of publication of individual notice in the Federal Register: 
November 20, 2006 (71 FR 67166).
    Expiration date of individual notice: December 20, 2006 (Public 
comments) and January 19, 2007 (Hearing requests).

    Dated at Rockville, Maryland, this 11th day of December 2006.

    For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. E6-21346 Filed 12-18-06; 8:45 am]
BILLING CODE 7590-01-P