[Federal Register Volume 72, Number 175 (Tuesday, September 11, 2007)]
[Notices]
[Pages 51852-51869]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-17864]


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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

[[Page 51853]]

determination by the Commission that such amendment involves no 
significant hazards consideration, notwithstanding the pendency before 
the Commission of a request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from August 16, 2007 to August 29, 2007. The last 
biweekly notice was published on August 28, 2007 (72 FR 49568).

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

[[Page 51854]]

determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held. If 
the final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards consideration, any hearing held 
would take place before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, [email protected]; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment, which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by email to [email protected].

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

    Date of amendment request: June 12, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3.7.4 to add an Action Statement for two 
inoperable control center air conditioning (AC) subsystems. The 
proposed new Action Statement would allow a finite time to restore one 
control center AC subsystem to operable status and require verification 
that control room temperature remains < 90 [deg]F every 4 hours.
    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 by a reference to a generic analysis published in the 
Federal Register on December 18, 2006 (71 FR 75774), which is presented 
below:

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

    The proposed change is described in Technical Specification Task 
Force (TSTF) Standard TS Change Traveler TSTF-477 adds an action 
statement for two inoperable control room subsystems.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
The proposed changes add an action statement for two inoperable 
control room subsystems.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
The proposed changes add an action statement for two inoperable 
control room subsystems. The equipment qualification temperature of 
the control room equipment is not affected. Future changes to the 
Bases or licensee-controlled document will be evaluated pursuant to 
the requirements of 10 CFR 50.59, ``Changes, test and experiments'', 
to ensure that such changes do not result in more than a minimal 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, and 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not adversely 
affect the ability of structures, systems and components (SSCs) to 
perform their intended safety function to mitigate the consequences 
of an initiating event within the assumed acceptance limits. The 
proposed changes do not affect the source term, containment 
isolation, or radiological consequences of any accident previously 
evaluated. Further, the proposed changes do not increase the types 
and the amounts of radioactive effluent that may be released, nor 
significantly increase individual or cumulative occupation/public 
radiation exposures.
    Therefore, the changes do not involve a significant increase in 
the probability or consequences of any 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 changes add an action statement for two inoperable 
control room subsystems. The changes do not involve a physical 
altering of the plant (i.e., no new or different type of equipment 
will be installed) or a change in methods governing normal plant 
operation. The requirements in the TS continue to require 
maintaining the control room temperature within the design limits.
    Therefore, the changes do not create the possibility of a new or 
different kind of accident from any previously evaluated.

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

    The proposed changes add an action statement for two inoperable 
control room subsystems. Instituting the proposed changes will 
continue to maintain the control room temperature within design 
limits. Changes to the Bases or license[e-] controlled document are 
performed in accordance with 10 CFR 50.59. This approach provides an 
effective level of regulatory control and ensures that the control 
room temperature will be maintained within design limits.
    The proposed changes maintain sufficient controls to preserve 
the current margins of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David G. Pettinari, Legal Department, 688 
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279.
    NRC Acting Branch Chief: Travis L. Tate.

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

    Date of amendment request: July 17, 2007.

[[Page 51855]]

    Description of amendment request: The proposed changes would modify 
Technical Specification (TS) requirements related to control room 
envelope (CRE) habitability in TS 3.7.3, ``Control Room Emergency 
Ventilation Air Supply (CREVAS) System'' and adds new TS 5.5.14, 
``Control Room Envelope Habitability Program.''
    These changes were proposed by the industry's TS Task Force (TSTF) 
and is designated TSTF-448. The NRC staff issued a notice of 
opportunity for comment in the Federal Register on October 17, 2006 (71 
FR 61075), on possible amendments concerning TSTF-448, including a 
model safety evaluation and model no significant hazards (NSHC) 
determination, using the consolidated line item improvement process 
(CLIIP). The NRC staff subsequently issued a notice of availability of 
the models for referencing in license amendment applications in the 
Federal Register on January 17, 2007 (72 FR 2022). The licensee 
affirmed the applicability of the following NSHC determination in its 
application dated July 17, 2007.
    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 does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change revises the TS for the CRE emergency 
ventilation system, which is a mitigation system designed to 
minimize unfiltered air leakage into the CRE and to filter the CRE 
atmosphere to protect the CRE occupants in the event of accidents 
previously analyzed. An important part of the CRE emergency 
ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated. Therefore, the probability of any accident 
previously evaluated is not increased. Performing tests to verify 
the operability of the CRE boundary and implementing a program to 
assess and maintain CRE habitability ensure that the CRE emergency 
ventilation system is capable of adequately mitigating radiological 
consequences to CRE occupants during accident conditions, and that 
the CRE emergency ventilation system will perform as assumed in the 
consequence analyses of design basis accidents. Thus, the 
consequences of any accident previously evaluated are not increased. 
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

    The proposed change does not impact the accident analysis. The 
proposed change does not alter the required mitigation capability of 
the CRE emergency ventilation system, or its functioning during 
accident conditions as assumed in the licensing basis analyses of 
design basis accident radiological consequences to CRE occupants. No 
new or different accidents result from performing the new 
surveillance or following the new program. The proposed change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a significant 
change in the methods governing normal plant operation. The proposed 
change does not alter any safety analysis assumptions and is 
consistent with current plant operating practice. Therefore, this 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.

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

    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition. Therefore, the proposed change does not involve 
a significant reduction in a margin of safety.

    Based on the above, the NRC staff concludes that the proposed 
change presents no significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of 
no significant hazards consideration is justified.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Mark G. Kowal.

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

    Date of amendment request: July 25, 2007.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TS) by adding an Action statement 
to the Limiting Condition for Operation (LCO) for TS 3.7.4, Control 
Room Air Conditioning (AC) System. The new Action statement allows a 
finite time to restore one control room AC subsystem to operable status 
(72 hours) and requires verification that control room temperature 
remains less than 104 [deg]F every 4 hours. The licensing basis control 
room air temperature for the James A. FitzPatrick Nuclear Power Plant 
(JAFNPP) is 104 [deg]F.
    This change was proposed by the industry's TS Task Force (TSTF) and 
is designated TSTF-477. The NRC staff issued a notice of opportunity 
for comment in the Federal Register on December 18, 2006 (71 FR 75774), 
on possible amendments concerning TSTF-477, including a model safety 
evaluation and model no significant hazards (NSHC) determination, using 
the consolidated line item improvement process (CLIIP). The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on March 26, 2007 (72 FR 14143). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated July 25, 2007.
    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 Changes Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change as described in Technical Specification Task 
Force (TSTF) Standard TS Change Traveler TSTF-477 adds an action 
statement for two inoperable control room subsystems.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
The proposed changes add an action statement for two inoperable 
control room subsystems. The equipment qualification temperature of 
the control room equipment is not affected. Future changes to the 
Bases or licensee controlled document will be evaluated pursuant to 
the requirements of 10 CFR 50.59, ``Changes, test and experiments,'' 
to ensure that such changes do not result in more than a minimal 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, and 
configuration of the facility or the manner in

[[Page 51856]]

which the plant is operated and maintained. The proposed changes do 
not adversely affect the ability of structures, systems and 
components (SSCs) to perform their intended safety function to 
mitigate the consequences of an initiating event within the assumed 
acceptance limits. The proposed changes do not affect the source 
term, containment isolation, or radiological consequences of any 
accident previously evaluated. Further, the proposed changes do not 
increase the types and the amounts of radioactive effluent that may 
be released, nor significantly increase individual or cumulative 
occupation/public radiation exposures.
    Therefore, the changes do not involve a significant increase in 
the probability or consequences of any 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

    The proposed changes add an action statement for two inoperable 
control room subsystems. The changes do not involve a physical 
altering of the plant (i.e., no new or different type of equipment 
will be installed) or a change in methods governing normal plant 
operation. The requirements in the TS continue to require 
maintaining the control room temperature within the design limits.
    Therefore, the changes do not create the possibility of a new or 
different kind of accident from any previously evaluated.

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

    The proposed changes add an action statement for two inoperable 
control room subsystems. Instituting the proposed changes will 
continue to maintain the control room temperature within design 
limits. Changes to the Bases or license controlled document are 
performed in accordance with 10 CFR 50.59. This approach provides an 
effective level of regulatory control and ensures that the control 
room temperature will be maintained within design limits.

    The proposed changes maintain sufficient controls to preserve 
the current margins of safety.

    Based on the above, the NRC staff concludes that the proposed 
change presents no significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of 
no significant hazards consideration is justified.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Mark G. Kowal.

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

    Date of amendment request: July 2, 2007.
    Description of amendment request: The proposed amendment would 
modify RBS technical specification (TS) requirements for MODE change 
limitations in limiting condition for operation (LCO) 3.0.4 and 
surveillance requirement (SR) 3.0.4. The proposed TS changes are 
consistent with Revision 9 of Nuclear Regulatory Commission (NRC) 
approved Industry TS Task Force (TSTF) Standard TS Change Traveler, 
TSTF-359, ``Increase Flexibility in MODE Restraints.'' In addition, the 
proposed amendment would also change TS section 1.4, Frequency, Example 
1.4-1, ``Surveillance Requirements,'' to accurately reflect the changes 
made by TSTF-359, which is consistent with NRC-approved TSTF-485, 
Revision 0, ``Correct Example 1.4-1.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 2, 2002 (67 FR 50475), as part of the 
Consolidated Line Item Improvement Process (CLIIP), on possible 
amendments to revise the plant-specific TS to modify requirements for 
MODE change limitations in LCO 3.0.4 and SR 3.0.4.
    The NRC staff subsequently issued a notice of availability of the 
models for Safety Evaluation and No Significant Hazards Consideration 
Determination for referencing in license amendment applications in the 
Federal Register on April 4, 2003 (68 FR 16579). The licensee affirmed 
the applicability of the CLIIP, including the model No Significant 
Hazards Consideration Determination, in its application dated February 
8, 2007.
    The proposed TS changes are consistent with NRC-approved Industry 
TSTF Standard TS change, TSTF-359, Revision 8, as modified by 68 FR 
16579. TSTF-359, Revision 8, was subsequently revised to incorporate 
the modifications discussed in the April 4, 2003, Federal Register 
notice and other minor changes. TSTF-359, Revision 9, was subsequently 
submitted to the NRC on April 28, 2003, and was approved by the NRC on 
May 9, 2003.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the NRC staff's analysis 
of the issue of no significant hazards consideration is presented 
below:

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

    The proposed changes in TS Section 1.4, Frequency, Example 1.4-
1, would accurately reflect the changes made by TSTF-359 in LCO 
3.0.4 and SR 3.0.4, which are consistent with NRC-approved TSTF-485, 
Revision 0. These changes are considered administrative in that they 
modify the example to demonstrate the proper application of LCO 
3.0.4 and SR 3.0.4. The requirements of LCO 3.0.4 and SR 3.0.4 are 
clear and are clearly explained in the associated Bases. As a 
result, modifying the example will not result in a change in usage 
of the TS.
    The proposed changes in LCO 3.0.4 and SR 3.0.4 allow entry into 
a mode or other specified condition in the applicability of a TS, 
while in a TS condition statement and the associated required 
actions of the TS. The proposed changes do not adversely affect 
accident initiators or precursors, the ability of structures, 
systems, and components to perform their intended function to 
mitigate the consequences of an initiating event within the assumed 
acceptance limits, or radiological release assumptions used in 
evaluating the radiological consequences of an accident previously 
evaluated. Being in a TS condition and the associated required 
actions are not an initiator of any accident previously evaluated. 
Therefore, the probability of an accident previously evaluated is 
not significantly increased. The consequences of an accident while 
relying on required actions as allowed by proposed LCO 3.0.4, are no 
different than the consequences of an accident while entering and 
relying on the required actions while starting in a condition of 
applicability of the TS. Therefore, the consequences of an accident 
previously evaluated are not significantly affected by these 
changes. The addition of a requirement to assess and manage the risk 
introduced by these changes will further minimize possible concerns. 
Therefore, these changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

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

    No new or different accidents result from utilizing the proposed 
changes. The proposed changes do not involve a physical alteration 
of the plant (no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements or eliminate any existing requirements. The 
proposed changes do not alter assumptions made in the safety 
analysis and are consistent with the safety analysis assumptions and 
current plant operating practice. Entering into a mode or other 
specified condition in the applicability of a TS, while in a TS 
condition statement and the associated required actions of the TS, 
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 these changes will further minimize possible 
concerns. Thus, these changes do not create the possibility of a new 
or different kind of accident from an accident previously evaluated.

[[Page 51857]]

Criterion 3--The Proposed Changes Do Not Involve a Significant 
Reduction in the Margin of Safety

    The proposed changes in TS section 1.4, Example 1.4-1, are 
considered administrative and will have no effect on the application 
of the TS requirements. Therefore, the margin of safety provided by 
the TS requirements is unchanged.
    The proposed changes in TS LCO 3.0.4 and SR 3.0.4 allow entry 
into a mode or other specified condition in the applicability of a 
TS, while in a TS condition statement and the associated required 
actions of the TS. The RBS TS allows operation of the plant without 
the full complement of equipment through the TS conditions for not 
meeting the TS LCO. The risk associated with this allowance is 
managed by the imposition of required actions that must be performed 
within the prescribed completion times. The net effect of being in a 
TS LCO condition on the margin of safety is not considered 
significant. The proposed changes do not alter the required actions 
or completion times of the TS. The proposed changes allow TS 
conditions to be entered, and the associated required actions and 
completion times to be used in new circumstances. This use is 
predicated upon the licensee's performance of a risk assessment and 
the management of plant risk. The changes also eliminate current 
allowances for utilizing required actions and completion times in 
similar circumstances, without assessing and managing risk. The net 
change to the margin of safety is insignificant. Therefore, these 
changes do 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: Thomas G. Hiltz.

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

    Date of amendment request: July 16, 2007, as supplemented by letter 
dated August 7, 2007.
    Description of amendment request: The proposed amendment would 
revise the facility operating license (FOL), Paragraph 2.C, and 
technical specifications (TS) 3.7.2 and TS 5.5.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on October 17, 2006 (71 FR 61075), on possible 
amendments to revise the plant-specific TS, to strengthen requirements 
regarding control room envelope (CRE) habitability by changing the 
action and surveillance requirements associated with the limiting 
condition for operability requirements for the CRE emergency 
ventilation system. A new TS administrative controls program on CRE 
habitability is being added, 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 January 17, 2007 (72 FR 2022). The licensee affirmed the 
applicability of the model NSHC determination in its application dated 
July 16, 2007, as supplemented by letter dated August 7, 2007.
    Basis for proposed NSHC 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 does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change revises the TS for the CRE emergency 
ventilation system, which is a mitigation system designed to 
minimize unfiltered air leakage into the CRE and to filter the CRE 
atmosphere to protect the CRE occupants in the event of accidents 
previously analyzed. An important part of the CRE emergency 
ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated. Therefore, the probability of any accident 
previously evaluated is not increased. Performing tests to verify 
the operability of the CRE boundary and implementing a program to 
assess and maintain CRE habitability ensure that the CRE emergency 
ventilation system is capable of adequately mitigating radiological 
consequences to CRE occupants during accident conditions, and that 
the CRE emergency ventilation system will perform as assumed in the 
consequence analyses of design basis accidents. Thus, the 
consequences of any accident previously evaluated are not increased. 
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

    The proposed change does not impact the accident analysis. The 
proposed change does not alter the required mitigation capability of 
the CRE emergency ventilation system, or its functioning during 
accident conditions as assumed in the licensing basis analyses of 
design-basis accident radiological consequences to CRE occupants. No 
new or different accidents result from performing the new 
surveillance or following the new program. The proposed change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a significant 
change in the methods governing normal plant operation. The proposed 
change does not alter any safety analysis assumptions and is 
consistent with current plant operating practice. Therefore, this 
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 the Margin of Safety

    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition. Therefore, the proposed change does not involve 
a significant reduction in a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Thomas G. Hiltz.

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

    Date of amendment request: August 17, 2007.
    Description of amendment request: The proposed amendment would 
revise the date for performing the ``Type A test'' in the RBS technical 
specification (TS) 5.5.13, ``Primary Containment Leak Rate Testing 
Program,'' from ``prior to December 14, 2007'' to ``April 14, 2008.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or

[[Page 51858]]

consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment to TS 5.5.13 allows a one-time extension 
to the current interval for the ILRT [integrated leak rate test]. 
The current interval of 15 years 4 months, based on past 
performance, would be extended on a one-time basis to 15 years and 8 
months from the date of the last test. The proposed extension to the 
ILRT cannot increase the probability of an accident since there are 
no design or operating changes involved and the test is not an 
accident initiator. The proposed extension of the test interval does 
not involve a significant increase in the consequences since 
analysis has shown that, the proposed extension of the ILRT and DWBT 
[drywell bypass test] frequency has a minimal impact on plant risk. 
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 extension to the interval for the ILRT does not 
involve any design or operational changes that could lead to a new 
or different kind of accident from any accidents previously 
evaluated. The tests are not being modified, but are only being 
performed after a longer interval. The proposed change does not 
involve a physical alteration of the plant (no new or different type 
of equipment will be installed) or a change in the methods governing 
normal plant operation. Therefore, 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.

    An evaluation of extending the ILRT DWBT surveillance frequency 
from once in 10 years to once in 15 years and 8 months has been 
performed using methodologies based on the approved ILRT methodologies. 
This evaluation assumed that the DWBT frequency was being adjusted in 
conjunction with the ILRT frequency. This analysis used realistic, but 
still conservative, assumptions with regard to developing the frequency 
of leakage classes associated with the ILRT and DWBT. The results from 
this conservative analysis indicates that the proposed extension of the 
ILRT frequency has a minimal impact on plant risk and therefore, the 
proposed change does not involve a significant reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Thomas G. Hiltz.

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

    Date of amendment request: August 2, 2007.
    Description of amendment request: The proposed changes to the 
technical specifications (TSs) will add new analytical methods and 
modify the containment average air temperature and safety injection 
tank level to support the implementation of Combustion Engineering 16 x 
16 Next Generation Fuel (NGF) as defined in Westinghouse Topical Report 
WCAP-16500-P beginning in Cycle 16 commencing after the spring 2008 
refueling outage. The fuel design is intended to provide improved fuel 
reliability by reducing grid-to-rod fretting issues, improved fuel 
performance for high duty operation, and enhanced operating margin.
    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.

Core Operating Limits Report (COLR)

    The proposed changes to the COLR TS are administrative in nature 
and have no impact on any plant configuration or system performance 
relied upon to mitigate the consequences of an accident. Changes to 
the calculated core operating limits may only be made using NRC 
approved methodologies, must be consistent with all applicable 
safety analysis limits, and are controlled by the 10 CFR 50.59 
process.
    The proposed change will add the following topical reports to 
the list of referenced core operating analytical methods.

WCAP-16500-P and Final Safety Evaluation (SE)

    Westinghouse topical report WCAP-16500-P describes the methods 
and models that will be used to evaluate the acceptability of CE 16 
x 16 NGF at CE plants. Entergy has demonstrated that the Limitations 
and Conditions associated with the NRC SE will be met. Prior to 
implementation of NGF the new core design will be analyzed with 
applicable NRC staff approved codes and methods.

WCAP-12610-P-A and CENPD-404-P-A Addendum 1-A

    The proposed change allows the use of methods required for the 
implementation of Optimized ZIRLOTM clad fuel rods. Entergy has 
demonstrated that the Limitations and Conditions associated with the 
NRC SE will be met.

WCAP-16523-P and Final Safety Evaluation

    This topical report describes the departure from nucleate 
boiling correlations that will be used to account for the impact of 
the CE 16 x 16 NGF fuel assembly design. Entergy has demonstrated 
that the Limitations and Conditions associated with the NRC SE will 
be met. Prior to implementation of NGF the new core design will be 
analyzed with applicable NRC staff approved codes and methods.

CENPD-387-P-A

    The proposed addition of this topical report provides the 
departure from nucleate boiling (DNB) correlation that will be used 
to evaluate the DNB impact of non-mixing vane grid spans for CE 16 x 
16 standard and NGF assemblies Entergy has demonstrated that the 
Limitations and Conditions associated with the NRC SE will be met.

CENPD-132, Supplement 4-P-A, Addendum 1-P and Final Safety 
Evaluation

    The addendum provides an optional steam cooling model that can 
be used for Emergency Core Cooling System (ECCS) Performance 
analyses to support the implementation of the CE 16 x 16 NGF fuel 
assembly design. Entergy has demonstrated that the Limitations and 
Conditions associated with the NRC SE will be met.
    Assumptions used for accident initiators and/or safety analysis 
acceptance criteria are not altered by the addition of these topical 
reports.

Safety Injection Tank Water Level and Containment Average Air 
Temperature

    These values are used as inputs to the LBLOCA and SBLOCA 
analyses. The new limits ensure that the analyzed LBLOCA remain 
acceptable. The limits have no impact to the SBLOCA analysis 
results. The changes do not cause an increase in the probability of 
an accident or an increase in the dose consequences associated with 
a LBLOCA.
    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.

Core Operating Limits Report (COLR)

    The proposed change identifies changes in the codes used to 
confirm the values of selected cycle-specific reactor physics 
parameter limits. The proposed change allows the use of methods 
required for the implementation of CE 16 x 16 NGF. The proposed 
addition of the referenced topical reports has no impact on any 
plant configurations or on system performance that

[[Page 51859]]

is relied upon to mitigate the consequences of an accident. The 
change to the COLR is administrative in nature and does not result 
in a change to the physical plant or to the modes of operation 
defined in the facility license.

WCAP-16500-P and Final Safety Evaluation

    The proposed change adds Westinghouse topical report WCAP-16500-
P, which describes the methods and models that will be used to 
evaluate the acceptability of CE 16 x 16 NGF at CE plants. Entergy 
has demonstrated that the Limitations and Conditions associated with 
the NRC SE will be met. Prior to implementation of NGF, the new core 
design will be analyzed with applicable NRC staff approved codes and 
methods.

WCAP-12610-P-A and CENPD-404-P-A Addendum 1-A

    The proposed change allows the use of methods required for the 
implementation of Optimized ZIRLOTM clad fuel rods. Entergy has 
demonstrated that the Limitations and Conditions associated with the 
NRC SE will be met.

WCAP-16523-P and Final Safety Evaluation

    This topical report describes the departure from nucleate 
boiling correlations that will be used to account for the impact of 
the CE 16 x 16 NGF fuel assembly design. Entergy has demonstrated 
that the Limitations and Conditions associated with the SE will be 
met.

CENPD-387-P-A

    The proposed addition of this topical report provides the 
departure from nucleate boiling (DNB) correlation that will be used 
to evaluate the DNB impact of non-mixing vane grid spans for CE 16 x 
16 standard and NGF assemblies. Entergy has demonstrated that the 
Limitations and Conditions associated with the NRC SE will be met.

CENPD-132, Supplement 4-P-A, Addendum 1-P and Final Safety 
Evaluation

    The addendum provides an optional steam cooling model that can 
be used for ECCS Performance analyses to support the implementation 
of the CE 16 x 16 NGF fuel assembly design. Entergy has demonstrated 
that the Limitations and Conditions associated with the NRC SE will 
be met.

Safety Injection Tank Water Level and Containment Average Air 
Temperature

    The safety injection tank (SIT) system provides a passive means 
of adding a large quantity of borated water to the reactor core in 
the event of a LBLOCA. The SIT system serves no other purpose. 
Reducing the maximum volume will not create any new or different 
accidents.
    The containment average air temperature ensures that the peak 
cladding temperature and cladding oxidation remain within limits 
during a LBLOCA. The change in the minimum allowable containment 
average temperature does not create any new or different accidents.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.

Core Operating Limits Report (COLR)

    The addition of the following topical reports to the list of 
analytical methods referenced in the COLR is administrative in 
nature:

 WCAP-16500-P and Final Safety Evaluation for Westinghouse 
Electric Company (Westinghouse) Topical Report (TR) WCAP-16500-P, 
Revision 0, ``CE [Combustion Engineering] 16x16 Next Generation Fuel 
[(NGF)] Core Reference Report''

 WCAP-12610-P-A and CENPD-404-P-A Addendum 1-A

 WCAP-16523-P and Final Safety Evaluation for Westinghouse 
Electric Company (Westinghouse) Topical Report (TR), WCAP-16523-P, 
``Westinghouse Correlations WSSV and WSSV-T for Predicting Critical 
Heat Flux in Rod Bundles with Side-Supported Mixing Vanes''

 CENPD-387-P-A

 CENPD-132, Supplement 4-P-A, Addendum 1-P and Final Safety 
Evaluation for Westinghouse Electric Company (Westinghouse) Topical 
Report (TR) CENPD-132 Supplement 4-P-A, Addendum 1-P, ``Calculative 
Methods for the CE [Combustion Engineering] Nuclear Power Large Break 
LOCA Evaluation Model--Improvement to 1999 Large Break LOCA EM Steam 
Cooling Model for Less Than 1 in/sec Core Reflood''

Safety Injection Tank Water Level and Containment Average Air 
Temperature

    The change to the allowable range for these two parameters does 
not reduce a margin of safety. The changes add to the margin of 
safety and provide assurance that the peak cladding temperature and 
cladding oxidation remain within limits.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Thomas G. Hiltz.

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, Units 1 and 2, Will 
County, Illinois

    Date of amendment request: July 31, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 5.5.2, ``Primary Coolant Sources Outside 
Containment,'' to clarify the intent of refueling cycle intervals 
(i.e., 18 month intervals) with respect to system integrated leak test 
requirements and to add a statement that the provisions of Surveillance 
Requirement 3.0.2 are applicable.
    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 amendment affects only the interval at which 
integrated system leak tests are performed, not the effectiveness of 
the integrated system leak test requirements. Revising the 
integrated system leak test requirements from ``at refueling cycle 
interval or less'' to ``at least once per 18 months'' is considered 
to be an administrative change because Braidwood Station, Units 1 
and 2, and Byron Station, Units 1 and 2, operate on 18-month fuel 
cycles. Incorporation of the allowance to extend the 18-month 
interval by 25%, as allowed by Surveillance Requirement (SR) 3.0.2, 
does not significantly degrade the reliability that results from 
performing the Surveillance at its specified Frequency.
    Test intervals are not considered as initiators of any accident 
previously evaluated. As a result, the probability of any accident 
previously evaluated is not significantly increased by the proposed 
amendment. Technical Specification (TS) 5.5.2 continues to require 
the performance of periodic integrated system leak tests. Therefore, 
accident analysis assumptions will still be verified. As a result, 
the consequences of any accident previously evaluated are not 
significantly increased.
    Based on the above discussion, the proposed changes do not 
involve an increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment affects only the interval at which 
integrated system leak tests are performed; they do not alter the 
design

[[Page 51860]]

or physical configuration of the plant. No changes are being made to 
Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, 
that would introduce any new accident causal mechanisms.
    Based on this evaluation, the proposed change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed amendment does not change the design or function of 
plant equipment. The proposed amendment does not significantly 
reduce the level of assurance that any plant equipment will be 
available to perform its function.
    The proposed amendment provides operating flexibility without 
significantly affecting plant operation.
    Based on this evaluation, the proposed change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Russell Gibbs.

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

    Date of amendment request: June 18, 2007.
    Description of amendment request: The proposed amendments would 
revise Technical Specification 3.7.5, ``Control Room Area Ventilation 
Air Conditioning (AC) System,'' to add an Action Statement for two 
inoperable control room area ventilation AC subsystems.
    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 is described in Technical Specification Task 
Force (TSTF) Standard TS Change Traveler TSTF-477 adds an action 
statement for two inoperable control room subsystems. The proposed 
change does not involve a physical alteration of the plant (no new 
or different type of equipment will be installed). The proposed 
changes add an action statement for two inoperable control room 
subsystems. The equipment qualification temperature of the control 
room equipment is not affected. Future changes to the Bases or 
licensee-controlled document will be evaluated pursuant to the 
requirements of 10 CFR 50.59, ``Changes, Test and Experiments,'' to 
ensure that such changes do not result in more than a minimal 
increase in the probability or consequences of an accident 
previously evaluated. The proposed changes do not adversely affect 
accident initiators or precursors nor alter the design assumptions, 
conditions, and configuration of the facility or the manner in which 
the plant is operated and maintained. The proposed changes do not 
adversely affect the ability of structures, systems and components 
to perform their intended safety function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed changes do not affect the source term, 
containment isolation, or radiological consequences of any accident 
previously evaluated. Further, the proposed changes do not increase 
the types and the amounts of radioactive effluent that may be 
released, nor significantly increase individual or cumulative 
occupation/public radiation exposures. Therefore, the changes do not 
involve a significant increase in the probability or consequences of 
any 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 changes add an action statement for two inoperable 
control room subsystems. The changes do not involve a physical 
altering of the plant (i.e., no new or different type of equipment 
will be installed) or a change in methods governing normal plant 
operation. The requirements in the TS continue to require 
maintaining the control room temperature within the design limits. 
Therefore, the changes do not create the possibility of a new or 
different kind of accident from any previously evaluated.

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

    The proposed changes add an action statement for two inoperable 
control room subsystems. Instituting the proposed changes will 
continue to maintain the control room temperature within design 
limits. Changes to the Bases or license controlled document are 
performed in accordance with 10 CFR 50.59. This approach provides an 
effective level of regulatory control and ensures that the control 
room temperature will be maintained within design limits. The 
proposed changes maintain sufficient controls to preserve the 
current margins of safety.

    Based upon the reasoning above, it appears that the three standards 
of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the requested amendments involve no significant hazards 
consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Russell Gibbs.

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

    Date of amendment request: June 27, 2007.
    Description of amendment request: The proposed amendment would 
remove the operability and surveillance requirements for the drywell 
air temperature and suppression chamber air temperature instrumentation 
from the Limerick Generating Station (LGS) technical specifications. 
This will allow a relocation of these requirements to the LGS technical 
requirements manual, a licensee controlled document.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The failure of the drywell air temperature or suppression 
chamber air temperature instrumentation is not assumed to be an 
initiator of any analyzed event in the UFSAR [Updated Final Safety 
Analysis Report]. The proposed changes do not alter the physical 
design of this instrumentation or any other plant structure, system, 
or component. The proposed changes relocate the drywell air 
temperature and suppression chamber air temperature instrumentation 
operability and surveillance requirements from the Limerick 
Generating Station (LGS) Technical Specifications (TS) to a 
licensee-controlled document under the control of 10 CFR 50.59 
[Title 10 of the Code of Federal Regulations (10 CFR) Part 50, 
Section 50.59].
    The proposed changes conform to NRC regulatory requirements 
regarding the content of plant TS as identified in 10 CFR 50.36, and 
also the guidance as approved by the NRC in NUREG-1433, ``Standard 
Technical Specifications-General Electric Plants, BWR/4.''
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes relocate the drywell air temperature and 
suppression chamber air temperature instrumentation operability and 
surveillance requirements from the LGS TS to a licensee-controlled 
document under the control of 10 CFR 50.59. The proposed

[[Page 51861]]

changes do not alter the physical design, safety limits, or safety 
analysis assumptions associated with the operation of the plant. 
Accordingly, the proposed changes do not introduce any new accident 
initiators, nor do they reduce or adversely affect the capabilities 
of any plant structure, system, or component in the performance of 
their safety function.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The subject instrumentation does not provide primary information 
required to permit operators to take specific manually controlled 
actions for which no automatic control is provided, and that are 
required for safety systems to accomplish their safety functions for 
design basis accident events. The instrumentation provides only 
drywell air temperature indication and suppression chamber air 
temperature indication, and does not provide an input to any 
automatic safety function. Operability and surveillance requirements 
will be established in a licensee-controlled document to ensure the 
reliability of drywell air temperature and suppression chamber air 
temperature instrumentation capability. Changes to these 
requirements will be subject to the controls of 10 CFR 50.59, 
providing the appropriate level of regulatory control.
    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: Mr. Bradley Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Harold K. Chernoff.

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

    Date of amendment request: April 12, 2007.
    Description of amendment request: The proposed amendment request 
would make the operating license and technical specification changes 
necessary to allow an increase in the rated thermal power from 2772 
megawatts thermal (MWt) to 2817 MWt (approximately 1.63 percent), based 
on the use of Caldon, Inc. Leading Edge Flow Meter 
CheckPlusTM System instrumentation to improve the accuracy 
of the plant power calorimetric measurement.
    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.
    Under contract to the FirstEnergy Nuclear Operating Company, 
AREVA NP Inc. performed evaluations of the Davis-Besse Nuclear Power 
Station (DBNPS) Nuclear Steam Supply System (NSSS) and balance of 
plant systems, components, and analyses that could be affected by 
the proposed change to the licensed power level. A power uncertainty 
calculation was performed and the effect of increasing core thermal 
power by 1.63 percent to 2817 MWt on the DBNPS design and licensing 
basis was evaluated. The evaluations determined that all structures, 
systems and components will continue to be capable of performing 
their design function at the proposed uprated power level of 2817 
MWt. An evaluation of the accident analyses demonstrates that the 
applicable analysis acceptance criteria continue to be met with the 
proposed changes. No accident initiators are affected by the power 
uprate and no challenges to any plant safety barriers are created by 
any of the proposed changes.
    The proposed change to the licensed power level does not affect 
the release paths, the frequency of release, or the analyzed source 
term for any accidents previously evaluated in the DBNPS Updated 
Final Safety Analysis Report (UFSAR). Systems, structures, and 
components required to mitigate transients will continue to be 
capable of performing their design functions with the proposed 
changes, and thus were found acceptable. The reduced uncertainty in 
the power calorimetric measurement ensures that applicable accident 
analyses acceptance criteria will continue to be met with operation 
at the proposed power level of 2817 MWt. Analyses performed to 
assess the effects of mass and energy remain valid. The source term 
used to assess radiological consequences has been reviewed and 
determined to bound operation at the proposed power level.
    The proposed change to the RPS high flux setpoint Allowable 
Value does not alter the typical manner in which systems or 
components are operated, and, therefore, will not result in an 
increase in the probability of an accident. The proposed High Flux 
Trip Allowable Values preserve assumptions of current accident 
analyses at the higher thermal power allowed by the proposed 
amendment, irrespective of the source of Heat Balance calculation 
input data. This proposed change does not alter any assumption 
previously made in the radiological consequence evaluations, nor 
does it affect mitigation of the radiological consequences of an 
accident previously evaluated. Therefore, this proposed change does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The addition of references to Note 10 to Functional Unit 2, High 
Flux, in Table 4.3-1 is administrative and does not impact the 
probability or consequences of an accident previously evaluated 
because its inclusion does not involve an accident initiator or 
impact any radiological analyses. This change is made to incorporate 
NRC guidance in a manner previously determined to be acceptable in 
DBNPS License Amendment No. 274.
    The proposed change to the volume of the condensate storage 
tanks does not alter the typical manner in which the system or 
component is operated, and, therefore, will not result in a 
significant increase in the probability of an accident. The 
condensate storage tanks are not accident initiators. The proposed 
change preserves the assumptions previously made in the radiological 
consequence evaluations and the radiological consequences of 
accidents previously evaluated. Therefore, this proposed change does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes to the Core Operating Limits Report (COLR) 
portion of the Administrative Controls Section of the TS are 
administrative and do not impact the probability or consequences of 
an accident previously evaluated because their inclusion do not 
involve accident initiators or impact any radiological analyses. 
These changes are made to include the NRC-approved documents 
pertaining to the Caldon Leading Edge Flow Meter.
    In summary, none of the proposed changes involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    No new accident scenarios, failure mechanisms, or single 
failures are introduced as a result of any of the proposed changes. 
Use of the Caldon CheckPlus\TM\ System has been analyzed, and 
failures of the system will have no adverse effect on any safety-
related system or any systems, structures, and components required 
for transient mitigation. Systems, structures, and components 
previously required for the mitigation of a transient continue to be 
capable of fulfilling their intended design functions. The proposed 
changes have no significant adverse affect on any safety-related 
structures, systems or components and do not significantly change 
the performance or integrity of any safety-related system.
    The proposed changes do not adversely affect any current system 
interfaces or create any new interfaces that could result in an 
accident or malfunction of a different kind than previously 
evaluated. Operating at a core power level of 2817 MWt does not 
create any new accident initiators or precursors. The reduced 
uncertainty in the power calorimetric measurement ensures that 
applicable accident analyses acceptance criteria continue to be met, 
to support

[[Page 51862]]

operation at the proposed core power level of 2817 MWt. Credible 
malfunctions continue to be bounded by the current accident analyses 
of record or recent evaluations that demonstrate that applicable 
criteria will continue to be met with the proposed changes.
    The proposed change to the RPS high flux setpoint Allowable 
Value does not introduce new accident scenarios, failure mechanisms 
or single failures. The change does not alter the manner in which 
plant systems or components are operated. The proposed High Flux 
Trip Allowable Values preserve assumptions of current accident 
analyses at the higher thermal power allowed by the proposed 
amendment, irrespective of the source of Heat Balance calculation 
input data. Therefore, this proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The addition of a reference to Note 10 to Functional Unit 2, 
High Flux, in Table 4.3-1 is administrative and will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated because its inclusion will not change the 
manner in which any equipment is operated. The proposed change to 
the volume of the condensate storage tanks does not introduce new 
accident scenarios, failure mechanisms or single failures. 
Therefore, this proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes to the COLR portion of the Administrative 
Controls Section of the TS are administrative and will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated because their inclusion will not 
change the manner in which any equipment is operated.
    In summary, none of the proposed changes will 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 margins of safety associated with the power uprate are those 
pertaining to core thermal power. These include those associated 
with the fuel cladding, Reactor Coolant System pressure boundary, 
and containment barriers. An engineering evaluation of the proposed 
1.63 percent increase in core thermal power was performed. The power 
uprate required revised NSSS design thermal and hydraulic parameters 
to be established to serve as the basis for all of the NSSS analyses 
and evaluations. This engineering review identified the design 
modifications necessary to accommodate the revised NSSS design 
conditions. Evaluations determined that the NSSS systems and 
components will continue to operate satisfactorily at the uprated 
power level with these modifications and the proposed changes. The 
NSSS accident analyses were evaluated at the uprated power level. In 
all cases, the evaluations demonstrate that the applicable analyses 
acceptance criteria will continue to be met with approval of the 
proposed changes. As such, the margins of safety will continue to be 
bounded by the analyses for all the changes being proposed.
    Therefore, none of the proposed changes will involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy 
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Russell Gibbs.

Florida Power Corporation, et. al., Docket No. 50-302, Crystal River 
Unit 3 Nuclear Generating Plant (CR-3), Citrus County, Florida

    Date of amendment request: April 25, 2007, as supplemented by 
letter dated June 28, 2007.
    Description of amendment request: The proposed amendment would 
change the operating license and technical specifications to increase 
the maximum power level from 2568 megawatts thermal (MWt) to 2609 MWt. 
The approximately 1.6 percent increase in power level would be achieved 
by use of the Caldon Leading Edge Flowmeter CheckPlus system to 
accurately measure power level.
    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 not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed change will increase the maximum core power level 
from 2568 MWt to 2609 MWt. This increase will only require 
adjustments and calibrations of existing plant instrumentation and 
control systems. The only equipment upgrades necessary for this 
uprate are spool pieces containing multiple ultrasonic flow 
instruments, which will be installed in each feedwater line, as well 
as more accurate instrumentation for feedwater pressure and steam 
pressure and temperature. Indication and control functions will 
continue to be performed by the currently installed feedwater 
instrumentation.
    Nuclear steam supply systems (NSSS) and balance-of-plant (BOP) 
systems and components that could be affected by the proposed change 
have been evaluated using revised NSSS design parameters based on a 
core power level of 2609 MWt. The results of these evaluations, 
which used well-defined analysis input assumptions/parameter values 
and currently approved analytical techniques, indicate that CR-3 
systems and components will continue to function within their design 
parameters and remain capable of performing their required safety 
functions at 2609 MWt. Since the revised NSSS parameters remain 
within the design conditions of the Reactor Coolant System (RCS) 
functional specification, the proposed change will not result in any 
new design transients or adversely affect the current CR-3 design 
transient analyses.
    The accidents analyzed in Chapter 14 of the CR-3 Final Safety 
Analysis Report (FSAR) have been reviewed for the impact of the 
uprate. Based on the power levels assumed in the current safety 
analyses, it has been determined that all FSAR and supporting 
analyses bound the uprate. This includes the dose calculations for 
the design basis radiological accidents, which assume a power level 
of 2619 MWt (2568 MWt plus an assumed 2 percent measurement 
uncertainty). Since the proposed change relies on less than 0.4% 
uncertainty, the assumed power level of 100.4% of 2609 MWt remains 
2619 MWt. Therefore, analyses performed at this power remain 
bounding.
    (2) Does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    As discussed above, the only equipment upgrades necessary for 
this uprate are spool pieces containing multiple ultrasonic flow 
instruments, which will be installed in each feedwater line, as well 
as more accurate instrumentation for feedwater pressure and steam 
pressure and temperature. All CR-3 systems and components will 
continue to function within their design parameters and remain 
capable of performing their required safety functions. The proposed 
change does not impact current CR-3 design transients or introduce 
any new transients. Equipment failure modes are expected to be the 
same as for existing instruments. Protective and control functions 
will continue to be performed by the currently installed feedwater 
instrumentation. Therefore, the proposed change will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) Does not involve a significant reduction in a margin of 
safety
    Challenges to the fuel, RCS pressure boundary and containment 
were evaluated for uprate conditions. Core analyses show that the 
implementation of the power uprate will continue to meet the current 
nuclear design basis. Impacts to components associated with RCS 
pressure boundary structural integrity, and factors such as 
pressure/temperature limits, vessel fluence, and pressurized thermal 
shock (PTS) were determined to be bounded by current analyses.
    As discussed above, all systems will continue to operate within 
their design parameters and remain capable of performing their 
intended safety functions following implementation of the proposed 
change. Finally, the current CR-3 safety analyses, including the 
design basis radiological accident dose calculations, bound the 
uprate.

[[Page 51863]]

Therefore, this change does not involve a significant reduction in 
the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Thomas H. Boyce.

Nine Mile Point Nuclear Station (NMPNS), LLC, Docket No. 50-220, Nine 
Mile Point Nuclear Station Unit No. 1 (NMP1), Oswego County, New York

    Date of amendment request: July 12, 2007.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) requirements related to control 
room envelope (CRE) habitability in TS 3.4.5, ``Control Room Air 
Treatment System,'' and TS 6.5, ``Programs and Manuals.'' The proposed 
changes are consistent with TS Task Force (TSTF) change TSTF-448, 
Revision 3, ``Control Room Habitability.'' The availability of the TS 
improvement was published in the Federal Register on January 17, 2007 
(72 FR 2022) as part of the consolidated line item improvement process. 
The licensee affirmed the applicability of the model no significant 
hazards consideration determination in its application.
    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 does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change revises the TS for the CRE emergency 
ventilation system, which is a mitigation system designed to 
minimize unfiltered air leakage into the CRE and to filter the CRE 
atmosphere to protect the CRE occupants in the event of accidents 
previously analyzed. An important part of the CRE emergency 
ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated. Therefore, the probability of any accident 
previously evaluated is not increased. Performing tests to verify 
the operability of the CRE boundary and implementing a program to 
assess and maintain CRE habitability ensure that the CRE emergency 
ventilation system is capable of adequately mitigating radiological 
consequences to CRE occupants during accident conditions, and that 
the CRE emergency ventilation system will perform as assumed in the 
consequence analyses of design basis accidents. Thus, the 
consequences of any accident previously evaluated are not increased. 
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

    The proposed change does not impact the accident analysis. The 
proposed change does not alter the required mitigation capability of 
the CRE emergency ventilation system, or its functioning during 
accident conditions as assumed in the licensing basis analyses of 
design basis accident radiological consequences to CRE occupants. No 
new or different accidents result from performing the new 
surveillance or following the new program. The proposed change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a significant 
change in the methods governing normal plant operation. The proposed 
change does not alter any safety analysis assumptions and is 
consistent with current plant operating practice. Therefore, this 
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 the [a] Margin of Safety

    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition. Therefore, the proposed change does not involve 
a significant reduction in a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006.
    NRC Branch Chief: Mark G. Kowal.

Nine Mile Point Nuclear Station (NMPNS), LLC, Docket No. 50-220, Nine 
Mile Point Nuclear Station Unit No. 1 (NMP1), Oswego County, New York

    Date of amendment request: July 23, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) section 3.1.1, ``Control Rod 
System,'' to incorporate a provision that should the rod worth 
minimizer (RWM) become inoperable before a reactor startup is commenced 
or before the first 12 control rods have been withdrawn, startup would 
be allowed to continue. This provision would rely on the RWM function 
being performed manually and would require a double check of compliance 
with the control rod program by a second licensed operator or other 
qualified member of the technical staff. The use of this allowance 
would be limited to one startup in the last calendar year.
    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 allows plant startup to proceed if the RWM 
becomes inoperable prior to withdrawing the first 12 control rods. 
The relevant design basis accident is the control rod drop accident 
(CRDA), which involves multiple failures to initiate the event. This 
change does not increase the probability of occurrence of any of the 
failures that are necessary for a CRDA to occur. Use of the RWM or 
the alternate use of a second qualified individual to ensure the 
correct control rod withdrawal sequence is not in itself an accident 
initiator, and adding the new startup allowance does not involve any 
plant hardware changes or new operator actions that could serve to 
initiate a CRDA. The proposed change will have no adverse effect on 
plant operation, or the availability or operation of any accident 
mitigation equipment. Also, since the control rod program will 
continue to be enforced by either the RWM or verification by a 
second qualified individual, the initial conditions of the CRDA 
radiological consequence analysis presented in the Updated Final 
Safety Analysis Report are not affected. Therefore, there will be no 
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.

[[Page 51864]]

    The proposed change does not introduce any new modes of plant 
operation and will not result in a change to the design function or 
operation of any structure, system, or component that is used for 
accident mitigation. The proposed change allows plant startup to 
proceed if the RWM becomes inoperable prior to withdrawing the first 
12 control rods, with verification of control rod movement in the 
correct sequence performed by a second qualified individual. This 
change does not result in any credible new failure mechanisms, 
malfunctions, or accident initiators not considered in the design 
and licensing basis. This change does not affect the ability of 
safety-related systems and components to perform their intended 
safety functions. Therefore, the proposed change will not create the 
possibility of a new or different kind of accident from any 
[accident] previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change allows plant startup to proceed if the RWM 
becomes inoperable prior to withdrawing the first 12 control rods. 
The proposed change will have no adverse effect on plant operation 
or equipment important to safety. The relevant design basis accident 
is the [CRDA], which involves multiple failures to initiate the 
event. The CRDA analysis consequences and related initial conditions 
remain unchanged when invoking the proposed change. The plant 
response to the CRDA will not be affected and the accident 
mitigation equipment will continue to function as assumed in the 
accident analysis. Therefore, there will be no significant reduction 
in a margin of safety.

    The 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: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006.
    NRC Branch Chief: Mark G. Kowal.

Nine Mile Point Nuclear Station (NMPNS), LLC, Docket No. 50-410, Nine 
Mile Point Nuclear Station Unit No. 2 (NMP2), Oswego County, New York

    Date of amendment request: July 12, 2007.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) requirements related to control 
room envelope (CRE) habitability in TS 3.7.2, ``Control Room Envelope 
Filtration (CREF) System,'' and TS 5.5, ``Programs and Manuals.'' The 
proposed changes are consistent with TS Task Force (TSTF) change TSTF-
448, Revision 3, ``Control Room Habitability.'' The availability of the 
TS improvement was published in the Federal Register on January 17, 
2007 (72 FR 2022) as part of the consolidated line item improvement 
process. The licensee affirmed the applicability of the model no 
significant hazards consideration determination in its application.
    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 does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change revises the TS for the CRE emergency 
ventilation system, which is a mitigation system designed to 
minimize unfiltered air leakage into the CRE and to filter the CRE 
atmosphere to protect the CRE occupants in the event of accidents 
previously analyzed. An important part of the CRE emergency 
ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated. Therefore, the probability of any accident 
previously evaluated is not increased. Performing tests to verify 
the operability of the CRE boundary and implementing a program to 
assess and maintain CRE habitability ensure that the CRE emergency 
ventilation system is capable of adequately mitigating radiological 
consequences to CRE occupants during accident conditions, and that 
the CRE emergency ventilation system will perform as assumed in the 
consequence analyses of design basis accidents. Thus, the 
consequences of any accident previously evaluated are not increased. 
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

    The proposed change does not impact the accident analysis. The 
proposed change does not alter the required mitigation capability of 
the CRE emergency ventilation system, or its functioning during 
accident conditions as assumed in the licensing basis analyses of 
design basis accident radiological consequences to CRE occupants. No 
new or different accidents result from performing the new 
surveillance or following the new program. The proposed change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a significant 
change in the methods governing normal plant operation. The proposed 
change does not alter any safety analysis assumptions and is 
consistent with current plant operating practice. Therefore, this 
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 the [a] Margin of Safety

    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition. Therefore, the proposed change does not involve 
a significant reduction in a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006.
    NRC Branch Chief: Mark G. Kowal.

Nine Mile Point Nuclear Station (NMPNS), LLC, Docket No. 50-410, Nine 
Mile Point Nuclear Station Unit No. 2 (NMP2), Oswego County, New York

    Date of amendment request: July 30, 2007.
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TSs) by changing the testing 
frequency for drywell spray nozzles specified in TS Surveillance 
Requirement (SR) 3.6.1.6.3 from ``10 years'' to ``following maintenance 
that could result in nozzle blockage.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change modifies the surveillance requirement (SR) 
to verify that the drywell spray nozzles are unobstructed after 
maintenance that could introduce material that could result in 
nozzle blockage. The spray nozzles are not assumed to be initiators 
of any previously analyzed

[[Page 51865]]

accident. Therefore, the proposed change does not increase the 
probability of any accident previously evaluated. The spray nozzles 
are used in the accident analyses to mitigate design basis 
accidents. The revised SR to verify system operability following 
maintenance is considered adequate to ensure operability of the 
Residual Heat Removal (RHR) Drywell Spray System.
    Since the system will still be able to perform its accident 
mitigation function, the consequences of accidents previously 
evaluated are not increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises the SR to verify that the RHR 
Drywell Spray System nozzles are unobstructed after maintenance that 
could result in nozzle blockage. The change does not introduce a new 
mode of plant operation and does not involve physical modification 
to the plant. The change will not introduce new accident initiators 
or impact the assumptions made in the safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises the frequency for performance of the 
SR to verify that the RHR Drywell Spray System nozzles are 
unobstructed. The frequency is changed from every 10 years to 
following maintenance that could result in nozzle blockage. This 
requirement, along with the foreign material exclusion program, the 
normal environmental conditions for the system, and the remote 
physical location of the spray nozzles, provide assurance that the 
spray nozzles will remain unobstructed. As the spray nozzles are 
expected to remain unobstructed and able to perform their post-
accident mitigation function, plant safety is not significantly 
affected.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The 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: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006.
    NRC Branch Chief: Mark G. Kowal.

Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant (PBNP), Units 1 and 2, Town of Two Rivers, 
Manitowoc County, Wisconsin

    Date of amendment request: June 29, 2007.
    Description of amendment request: The proposed amendments would 
modify the Technical Specifications (TSs) 3.7.2, by removing the 
specific isolation time for the main steam isolation valves from the 
associated TS Surveillance Requirements (SRs) and by replacing it with 
the requirement to verify the valve isolation time is within limits. 
The changes are consistent with Nuclear Regulatory Commission (NRC) 
approved Industry/Technical Specification Task Force (TSTF)-491, 
``Removal of the Main Steam and Main Feedwater Valve Isolation Time 
from Technical Specifications,'' Revision 2. The proposed amendments 
deviate from TSTF-491 in that the current PBNP TS 3.7.3, and associated 
SRs do not include the main feedwater valve closure times, and thus 
TSTF-491 changes to TS 3.7.3 are not applicable to the PBNP TSs.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on October 5, 2006 (71 FR 58884), on possible 
amendments concerning the Consolidated Line Item Improvement Process 
(CLIIP), including a model safety evaluation and a model no significant 
hazards consideration determination. The NRC staff subsequently issued 
a notice of availability of the models for referencing in license 
amendment applications in the Federal Register on December 29, 2006 (71 
FR 78472) as part of the CLIIP. In its application dated June 29, 2007, 
the licensee affirmed the applicability of the following determination.
    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 relocating main steam [ ] valve 
isolation times to the Licensee Controlled Document that is 
referenced in the Bases. The proposed change is described in 
Technical Specification Task Force (TSTF) Standard TS Change 
Traveler TSTF-491 related to relocating the main steam [ ] valve 
isolation times to the Licensee Controlled Document that is 
referenced in the Bases and replacing the isolation time with the 
ph[r]ase, ``within limits.''
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
The proposed changes relocate the main steam [ ] isolation valve 
times to the Licensee Controlled Document that is referenced in the 
Bases. The requirements to perform the testing of these isolation 
valves are retained in the TS. Future changes to the Bases or 
licensee-controlled document will be evaluated pursuant to the 
requirements of 10 CFR 50.59, ``Changes, test and experiments,'' to 
ensure that such changes do not result in more than minimal increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, and 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not adversely 
affect the ability of structures, systems and components (SSCs) to 
perform their intended safety function to mitigate the consequences 
of an initiating event within the assumed acceptance limits. The 
proposed changes do not affect the source term, containment 
isolation, or radiological consequences of any accident previously 
evaluated. Further, the proposed changes do not increase the types 
and the amounts of radioactive effluent that may be released, nor 
significantly increase individual or cumulative occupation/public 
radiation exposures.
    Therefore, the changes do not involve a significant increase in 
the probability or consequences of any 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 changes relocate the main steam [ ] valve isolation 
times to the Licensee Controlled Document that is referenced in the 
Bases. In addition, the valve isolation times are replaced in the TS 
with the ph[r]ase ``within limits.'' The changes do not involve a 
physical altering of the plant (i.e., no new or different type of 
equipment will be installed) or a change in methods governing normal 
p[l]ant operation. The requirements in the TS continue to require 
testing of the main steam [ ] isolation valves to ensure the proper 
functioning of these isolation valves.
    Therefore, the changes do not create the possibility of a new or 
different kind of accident from any previously evaluated.

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

    The proposed changes relocate the main steam [ ] valve isolation 
times to the Licensee Controlled Document that is referenced in the 
Bases. In addition, the valve isolation times are replaced in the TS 
with the ph[r]ase ``within limits.'' Instituting the proposed 
changes will continue to ensure the testing of main steam [ ] 
isolation valves. Changes to the Bases or license controlled 
document are performed in accordance with 10 CFR 50.59. This 
approach provides an effective level of regulatory control and 
ensures that main steam [ ] isolation valve testing is

[[Page 51866]]

conducted such that there is no significant reduction in the margin 
of safety.
    The margin of safety provided by the isolation valves is 
unaffected by the proposed changes since there continue to be TS 
requirements to ensure the testing of main steam [ ] isolation 
valves. The proposed changes maintain sufficient controls to 
preserve the current margins 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 Acting Branch Chief: Travis L. Tate.

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 15, 2007.
    Description of amendment request: The proposed amendment would 
revise the licensing basis, as described in Appendix 3A of the Salem 
Updated Final Safety Analysis Report (UFSAR), regarding the method of 
calculating the net positive suction head available (NPSHa) for the 
emergency core cooling system (ECCS) and containment heat removal 
system pumps. The proposed change relates to issues associated with 
Generic Letter 2004-02, ``Potential Impact of Debris Blockage on 
Emergency Recirculation During Design Basis Accidents at Pressurized-
Water Reactors.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The change in NPSH methodology for ECCS pumps allows the use of 
initial containment air pressure in calculating NPSHa. Although this 
change is a non-conservative change in the Salem methodology for 
calculation of RHR [residual heat removal] pump NPSHa during post 
LOCA [loss-of-coolant accident] recirculation (per 10 CFR 
50.59(c)(1)(viii) [Title 10 of the Code of Federal Regulations, Part 
50, Section 50.59(c)(1)(viii)]), the proposed new methodology is in 
accordance with NPSHa calculation methodologies provided in Safety 
Guide 1, Regulatory Guides [RG] 1.1, and 1.82, and the guidance of 
NEI [Nuclear Energy Institute] 04-07, [``]Pressurized Water Reactor 
Sump Performance Evaluation Methodology[,''] (GSI [generic safety 
issue]--191) and accompanying SER [safety evaluation report]. The 
containment air pressure value used in the NPSHa calculation is 
based on the containment conditions prior to the accident only and 
does not include any credit for accident pressure conditions, is 
conservatively determined based on minimum containment initial 
pressure, and maximum temperature and relative humidity conditions. 
In addition, the vapor pressure term for the sump water being pumped 
is also included in the NPSHa equation, and the value chosen for the 
NPSHa calculation is based on the highest temperature of the sump 
fluid for the condition being evaluated. This, in conjunction with 
the more rigorous GSI-191 analyses, provides assurance that the ECCS 
pumps can perform their design function. Consequently, the ECCS 
pumps will continue to perform their design function and there is no 
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 ECCS pumps take suction from the containment sump during the 
recirculation phase of the LOCA to provide long term core cooling. 
This system is not utilized during normal operation of the plant. 
Therefore, it does not cause initiation of any accident.
    However, the ECCS pumps will continue to perform their design 
function during the recirculation phase. Crediting initial 
containment air pressure in the NPSH methodology does not create any 
new or different kind of accident from any accident previously 
evaluated. This change removes an additional conservatism built into 
the original methodology. By changing the UFSAR described 
methodology to credit the containment initial air pressure in the 
RHR pump NPSHa calculation, a more realistic methodology is 
established. The sole purpose of the additional conservatism was to 
ensure credit was not taken for post-LOCA pressure. The revised 
methodology continues to meet this requirement.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change removes conservatism from the existing UFSAR 
methodology. However, the purpose of the conservatism (equating 
containment pressure to sump vapor pressure) was solely to ensure 
that no credit was taken for transient (post-LOCA) pressure in the 
NPSHa calculation. The purpose was not to deny credit for initial 
containment air pressure. Consequently, removing the conservatism 
does not alter the basic intent of the NPSH methodology per RG 1.1 
requirements, and is consistent with the requirements of RG 1.82, 
Revision 1 and NEI 04-07. This change to include a containment air 
pressure value establishes a more realistic methodology that still 
encompasses adequate conservatisms; no credit is given for the 
higher accident pressure conditions, and the value is conservatively 
determined based on minimum initial containment air pressure and 
maximum temperature and relative humidity conditions. In addition, 
the vapor pressure term for the sump water being pumped is also 
added to the NPSHa equation, and the value chosen for the NPSHa 
calculation is based on the highest temperature of the sump fluid 
for the condition being evaluated. Consequently, 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 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.

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

    Date of amendment request: August 20, 2007.
    Description of amendment request: The amendment would increase the 
minimum volume of fuel required for the emergency diesel generators 
(EDGs) in Technical Specification (TS) 3.8.3, ``Diesel Fuel Oil, Lube 
Oil, and Starting Air.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to the minimum required fuel oil volume 
required in the EDG storage tanks have no impact on the frequency of 
occurrence of any of the accidents evaluated in the FSAR [Final 
Safety Analysis Report for Callaway]. Changing the minimum required 
fuel oil volume in the EDG fuel oil storage tank has no impact on 
the likelihood of occurrence of a loss of coolant accident (LOCA), 
line break, plant transient, loss of offsite power, or any such 
accident because the precursors for such accidents do not involve 
the fuel oil storage tanks.
    The EDGs are designed to provide [alternating current] 
electrical power to systems required for mitigating the effects of 
accidents in the event of a loss of the

[[Page 51867]]

preferred (offsite) power source (i.e., from the grid). However, the 
failure or malfunction of an EDG (due, for example, to a loss or 
interruption of [the] fuel oil supply) is not itself an initiator of 
any accident previously evaluated.
    Based on these considerations, the proposed changes have no 
impact on the probability of occurrence of any accident evaluated in 
the FSAR, and therefore the proposed changes do not involve a 
significant increase in the probability of an accident previously 
evaluated.
    With respect to the consequences of postulated accidents 
addressed in [the] FSAR, the support function provided by the EDGs 
for accident mitigation is not affected by the proposed TS changes. 
[The proposed changes are to provide additional margin for 
precluding adverse effects that could result from air entrapment 
caused by a vortex condition during fuel oil transfer pump operation 
and, thus, to ensure that the EDG has sufficient fuel oil to provide 
its support function when needed.] Each of the diesel fuel oil 
storage tanks has adequate excess capacity to more than accommodate 
a slight increase in the usable volume of fuel oil contained 
therein. Thus, even with this increase, the tanks will still be 
fully capable of storing the required fuel oil volume needed to 
ensure EDG operation throughout the assumed duration of an accident. 
At the same time, the proposed changes to TS 3.8.3 will serve to 
ensure that the unusable volume in the tanks provides adequate 
margin against potentially adverse vortex effects (by precluding the 
potential for air ingestion into the fuel oil transfer pumps). On 
this basis, the proposed changes have no impact on the capability of 
the EDGs to perform their required mitigation/support function for 
accidents involving a loss of offsite power. Since the proposed 
changes have no impact on accident mitigation capability, they 
involve no increase in the consequences of any accident evaluated in 
the FSAR.
    Based on the above, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes involve a slight change to the minimum fuel 
oil volume required for the EDGs, but they do not involve hardware 
changes or changes to EDG operation or testing that would create any 
new failure modes for the EDGs or any other [safety-related] system 
or component, or that would adversely affect plant operation. The 
changes do not involve the addition of any new equipment. No changes 
to accident assumptions, including any new limiting single failures, 
are involved. With respect to the proposed changes, the plant will 
continue to be operated within the envelope of the existing safety 
analyses.
    Therefore, based on the above, the proposed changes do not 
create [the possibility of] a new or different kind of accident 
[from any accident] previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is related to the confidence in the ability 
of the fission product barriers to perform their design functions 
during and following an accident situation. These barriers include 
the fuel cladding, the reactor coolant system, and the containment 
system. The proposed changes do not directly affect these barriers, 
nor do they involve or cause any adverse impact on the EDGs which 
serve to support these barriers in the event of an accident 
concurrent with a loss of offsite power.
    [The margin of safety is also related to the ability of the 
safety-related systems to perform their safety function as described 
in the safety analyses in the FSAR. The proposed changes are to 
provide additional margin for precluding adverse effects that could 
result from air entrapment caused by a vortex condition during fuel 
oil transfer pump operation and, thus, to ensure that the EDG has 
sufficient fuel oil to provide its support function when needed. 
Therefore, the proposed changes are to increase margin for the 
EDGs.]
    The proposed changes do not alter the manner in which safety 
limits or limiting safety system settings are determined, nor is 
[the] basis of any limiting condition for operation changed or 
affected. The safety analysis acceptance criteria are not impacted 
by these changes. The proposed changes will not result in plant 
operation in a configuration outside the design basis.
    Therefore, the proposed changes do not involve a significant 
reduction in 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(copyright) are satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: John O'Neill, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: Thomas G. Hiltz.

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 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) The 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].

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 application for amendment: March 1, 2007.
    Brief description of amendment: The amendment revised the Grand 
Gulf Nuclear Station, Unit 1 (GGNS) Technical Specification (TS) to add 
a note to the Required Actions of TS 3.6.1.3, ``Primary Containment 
Isolation Valves (PCIVs)''. GGNS TS 3.6.1.3 requires specific actions 
to be taken for inoperable PCIVs. The TS Required Actions include 
isolating the affected penetration by use of a closed and deactivated 
automatic valve, closed manual valve, blind flange, or check valve with 
flow through the valve secured. The new note would allow a relief valve 
to be used to comply with

[[Page 51868]]

TS 3.6.1.3, Actions A.1 and B.1 without being deactivated provided it 
has a relief setpoint of at least 1.5 times containment design pressure 
(i.e., at least 23 pounds per square inch gauge) and meets one of the 
following criteria:
    1. The relief valve is one-inch nominal size or less, or
    2. The flow path is into a closed system whose piping pressure 
rating exceeds the containment design pressure rating.
    Date of issuance: August 24, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No: 176.
    Facility Operating License No. NPF-29: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: April 24, 2007 (72 FR 
20382).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 24, 2007.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian 
Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New 
York

    Date of application for amendment: September 25, 2006, as 
supplemented March 12, 2007.
    Brief description of amendment: Entergy Nuclear Operations, Inc., 
requested an amendment to make editorial changes to the Technical 
Specifications of Indian Point Nuclear Generating Unit Nos. 2 and 3. 
The editorial changes consist of typographical corrections, update of 
references, and deletion of obsolete notes.
    Date of issuance: August 16, 2007.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 252 and 234.
    Facility Operating License Nos. DPR-26 and DPR-64: The amendment 
revised the License and the Technical Specifications.
    Date of initial notice in Federal Register: November 7, 2006 (71 FR 
65142).
    The March 12, 2007, 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.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 16, 2007.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 7, 2006, as supplemented by 
letters dated January 22, May 14, and August 7, 2007.
    Description of amendment request: The amendment revises the 
Seabrook Technical Specifications (TSs) to correct a joint-owner name 
in the operating license, remove a license condition from Appendix C to 
the FOL, and remove the list of Bases sections from the TS Index. 
Additionally, the amendment removes two manual valves from TS table 
3.3-9, ``Remote Shutdown System,'' adds the requirement that only one 
charging pump is permitted to be aligned for injection into the reactor 
coolant system in Modes 4, 5, and 6, removes a 1-hour reporting 
requirement for portable makeup pump system storage from TS 3.7.4, 
``Service Water System/Ultimate Heat Sink,'' deletes a footnote from TS 
3.7.6.2, ``Air Conditioning,'' and modifies TS 6.7.6, ``Radioactive 
Effluent Controls Program.''
    Date of issuance: August 23, 2007.
    Effective date: As of its date of issuance, and shall be 
implemented within 90 days.
    Amendment No.: 116.
    Facility Operating License No. NPF 86: The amendment revised the 
License and Technical Specification.
    Date of initial notice in Federal Register: June 5, 2007 (72 FR 
31101).
    The licensee's January 22, May 14, and August 7, 2007, supplements 
provided clarifying information that did not change the scope of the 
proposed amendment as described in the original notice of proposed 
action published in the Federal Register, and did not change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 23, 2007.
    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: August 14, 2006, as 
supplemented by letter dated July 16, 2007.
    Brief description of amendments: The amendments make miscellaneous 
improvements to the Technical Specifications (TSs) for Prairie Island 
Nuclear Generating Plant, Units 1 and 2. The amendments revise the 
wording in the section headers in TS 1.3, ``Completion Times''; remove 
an unnecessary Note in TS 3.1.4, ``Rod Group Alignment Limits''; remove 
applicable modes in TS 3.3.7, ``Spent Fuel Pool Special Ventilation 
System (SFPSVS) Actuation Instrumentation''; add reference to a TS 
Condition to clarify the requirements of TS 3.7.10, ``Control Room 
Special Ventilation System (CRSVS)''; and update a reference in TS 4.0, 
``Design Features.''
    Date of issuance: August 10, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 180 & 170.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the TSs.
    Date of initial notice in Federal Register: November 21, 2006 (71 
FR 67397).
    The supplemental letter contained clarifying information and did 
not change the initial no significant hazards consideration 
determination, and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 10, 2007.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 26, 2006, as 
supplemented on May 14, 2007.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 6.9.1.9, ``Core Operating Limits Report (COLR),'' to 
remove the revision numbers and dates from the list of topical reports 
that contain the analytical methods used in the COLR. The Salem Unit 2 
amendment also adds a new topical report to the list of COLR methods 
referenced in TS 6.9.1.9.
    Date of issuance: August 23, 2007.
    Effective date: The license amendments are effective as of the date 
of issuance. The Salem Unit 1 amendment shall be implemented prior to 
restart from the 19th refueling outage in fall 2008. The Salem Unit 2 
amendment shall be implemented prior to restart from the 16th refueling 
outage in spring 2008.
    Amendment Nos.: 284 and 267.

[[Page 51869]]

    Facility Operating License Nos. DPR 70 and DPR-75: The amendments 
revised the TSs and the License.
    Date of initial notice in Federal Register: November 7, 2006 (71 FR 
65143).
    The supplement dated May 14, 2007, 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 November 7, 2006 (71 FR 65143).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 23, 2007.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of application for amendment: November 15, 2006, as 
supplemented by letters dated June 21, and August 23, 2007.
    Brief description of amendment: The amendment deletes Technical 
Specification (TS) Table 3.6.3-1, ``Primary Containment Isolation 
Valves,'' and relocates the information to the Hope Creek Generating 
Station Technical Requirements Manual (TRM). The amendment also revises 
other TS sections that reference TS Table 3.6.3-1.
    Date of issuance: August 27, 2007.
    Effective date: As of the date of issuance, to be implemented 
within 90 days. Implementation shall include the relocation of 
information from the TSs to the TRM as described in the licensee's 
application dated November 15, 2006, and letters dated June 21, and 
August 23, 2007.
    Amendment No.: 171.
    Facility Operating License No. NPF-57: The amendment revised the 
TSs and the License.
    Date of initial notice in Federal Register: February 13, 2007 (72 
FR 6789).
    The supplements dated June 21, and August 23, 2007, provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination or expand the 
application beyond the scope of the original Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 27, 2007.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 5th day of September, 2007.

    For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing Office of Nuclear 
Reactor Regulation.
[FR Doc. E7-17864 Filed 9-10-07; 8:45 am]
BILLING CODE 7590-01-P