Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 51852-51869 [E7-17864]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 72, Number 175 (Tuesday, September 11, 2007)]
[Notices]
[Pages 51852-51869]
From the Federal Register Online via the Government Printing 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, https://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final

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

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