Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 45454-45466 [E7-15459]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 72, Number 156 (Tuesday, August 14, 2007)]
[Notices]
[Pages 45454-45466]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E7-15459]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from July 19, 2007, to August 1, 2007. The last 
biweekly notice was published on July 31, 2007 (72 FR 41780).

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.

[[Page 45455]]

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

[[Page 45456]]

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of amendments request: June 29, 2007.
    Description of amendments request: The amendment would modify 
Technical Specification (TS) requirements related to control room 
envelope (CRE) habitability in TS 3.7.8, ``Control Room Emergency 
Ventilation System (CREVS),'' and TS 5.5, ``Programs and Manuals.'' The 
changes are consistent with the Nuclear Regulatory Commission approved 
Technical Specification Task Force (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 item improvement process (CLIIP). In addition, 
the amendment would remove a footnote currently contained in the 
Completion Time of TS 3.7.8, Required Action D. The footnote was added 
in Amendment Nos. 250/227 and was only applicable during the Unit 1 
2002 refueling outage.
    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 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.
    The removal of a footnote [to TS 3.7.8] that is no longer 
applicable is an editorial change that does not affect accident 
initiators or precursors, nor alter the design assumptions, 
conditions or configuration of the facility. The proposed change 
also does not affect the ability of SSCs to perform their intended 
function to mitigate the consequences of an accident. Therefore, the 
proposed editorial change does not increase the probability or 
consequences of an accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not 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.
    The proposed change is the editorial removal of a footnote [to 
TS 3.7.8] that no longer applies. The removal of a footnote that no 
longer applies does not impact the accident analyses. Additionally, 
it does not add or modify any existing plant equipment and does not 
introduce any new operational methods. Therefore, the proposed 
editorial change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change 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.
    The proposed editorial change [removal of a footnote to TS 
3.7.8] does not affect safety analyses acceptance criteria or safety 
system operation. Removal of a footnote that is no longer applicable 
does not result in plant operation outside the design basis. 
Therefore, the proposed editorial change does not involve a 
reduction in the margin of safety.
    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: Carey Fleming, Sr. Counsel--Nuclear 
Generation, Constellation Generation Group, LLC, 750 East Pratt Street, 
17th floor, Baltimore, MD 21202.
    NRC Branch Chief: Mark G. Kowal.

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: February 20, 2007.
    Description of amendment request: The proposed change would revise 
Limerick Generating Station (LGS), Units 1 and 2, Technical 
Specifications (TSs), Section 6.8.4.g, ``Primary Containment Leakage 
Rate Testing Program,'' to allow a one-time extension of no more than 5 
years for the Type A, Integrated Leakage Rate Test (ILRT) interval. 
This revision is a one-time exception to the 10-year frequency of the 
performance-based leakage rate testing program for Type A tests as 
defined in Nuclear Energy Institute (NEI) document NEI 94-01, Revision 
0, ``Industry Guideline For Implementing Performance-Based Option of 10 
CFR Part 50, Appendix J,'' pursuant to Title 10 of the Code of Federal 
Regulations (10 CFR) Part 50, Appendix J, Option B. The requested 
exception is to allow the ILRT to be performed within 15 years from the 
last ILRT as opposed to the current 10-year frequency. The most recent 
containment Type A ILRTs for LGS Units 1 and 2 were performed on May 
15, 1998, and May 21, 1999, respectively.
    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

[[Page 45457]]

consequences of an accident previously evaluated?
    Response: No.
    The proposed change will revise TS 6.8.4.g (``Primary 
Containment Leakage Rate Testing Program'') of the LGS, Units 1 and 
2 TS to reflect a one-time extension to the Type A Integrated Leak 
Rate Test (ILRT) as currently specified in the Technical 
Specifications. This change will extend the requirement to perform 
the Type A ILRT from the current requirement of 10 to 15 years, 
which is ``no later than May 15, 2013'' for LGS, Unit 1 and is ``no 
later than May 21, 2014'' for Unit 2.
    The function of the containment is to isolate and contain 
fission products released from the reactor coolant system following 
a design basis Loss of Coolant Accident (LOCA) and to confine the 
postulated release of radioactive material to within limits. The 
test interval associated with Type A ILRTs is not a precursor of any 
accident previously evaluated. Type A ILRTs provide assurance that 
the LGS, Units 1 and 2 containments will not exceed allowable 
leakage rate values specified in the TS and will continue to perform 
their design function following an accident. The risk assessment of 
the proposed change has concluded that there is an insignificant 
increase in Large Early Release Frequency, Person-Rem, and 
Conditional Containment Failure Frequency. Additionally, containment 
inspections have also been performed which demonstrate the continued 
structural integrity of the primary containment and will be 
performed in the future as required by the ASME Code.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change for a one-time extension of the Type A ILRTs 
for LGS, Units 1 and 2 will not affect the control parameters 
governing unit operation or the response of plant equipment to 
transient and accident conditions. The proposed change does not 
introduce any new equipment, modes of system operation or failure 
mechanisms.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?

    Response: No.
    The integrity of the containment penetrations and isolation 
valves is verified through Type B and Type C local leak rate tests 
(LLRTs) and the overall leak tight integrity of the containment is 
verified by a Type A ILRT, as required by 10 CFR 50, Appendix J, 
``Primary Reactor Containment Leakage Testing for Water-Cooled Power 
Reactors.'' These tests are performed to verify the essentially leak 
tight characteristics of the containment at the design basis 
accident pressure. The proposed change for a one-time extension of 
the Type A ILRT does not affect the method for Type A, B or C 
testing or the test acceptance criteria.
    EGC has conducted a risk assessment to determine the impact of a 
change to the LGS, Units 1 and 2 Type A ILRT from 10 to 15 years. 
This risk assessment measured the impact to the Large Early Release 
Frequency, Person-Rem, and Conditional Containment Failure 
Frequency. This assessment indicated that the proposed LGS, Units 1 
and 2 Type A ILRT interval extension has a very small change in risk 
to the public and is an acceptable plant change from a risk 
perspective.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    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.

Florida Power and Light Company, Docket Nos. 50-335 and 50-389, St. 
Lucie Plant, Units 1 and 2, St. Lucie County, Florida

    Date of amendment request: June 4, 2007.
    Description of amendment request: The proposed amendment would 
remove the Technical Specification (TS) requirements that reference 
hydrogen recombiners and hydrogen monitors. The proposed amendment 
suggests changes support implementation of the revisions to 10 CFR 
50.44, ``Standards for Combustible Gas Control System in Light Water 
Cooled Power Reactors,'' that became effective on September 16, 2003. 
The changes would be consistent with Revision 1 of the NRC-approved 
Technical Specification Task Force (TSTF) Standard Technical 
Specification Change Traveler, TSTF-447, ``Elimination of Hydrogen 
Recombiners and Change to Hydrogen and Oxygen Monitors.'' The 
particular TS improvement in question was announced in the Federal 
Register Notice on September 25, 2003, as part of the consolidated line 
item improvement process.
    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 revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen [and oxygen] monitors are no longer required to mitigate 
design-basis accidents and, therefore, the hydrogen monitors do not 
meet the definition of a safety-related component as defined in 10 
CFR 50.2. RG [Regulatory Guide] 1.97 Category 1, is intended for key 
variables that most directly indicate the accomplishment of a safety 
function for design-basis accident events. The hydrogen [and oxygen] 
monitors no longer meet the definition of Category 1 in RG 1.97. As 
part of the rulemaking to revise 10 CFR 50.44 the Commission found 
that Category 3, as defined in RG 1.97, is an appropriate 
categorization for the hydrogen monitors because the monitors are 
required to diagnose the course of beyond design-basis accidents. 
[Also, as part of the rulemaking to revise 10 CFR 50.44, the 
Commission found that Category 2, as defined in RG 1.97, is an 
appropriate categorization for the oxygen monitors, because the 
monitors are required to verify the status of the inert 
containment.]
    The regulatory requirements for the hydrogen [and oxygen] 
monitors can be relaxed without degrading the plant emergency 
response. The emergency response, in this sense, refers to the 
methodologies used in ascertaining the condition of the reactor 
core, mitigating the consequences of an accident, assessing and 
projecting offsite releases of radioactivity, and establishing 
protective action recommendations to be communicated to offsite 
authorities. Classification of the hydrogen monitors as Category 3, 
[classification of the oxygen monitors as Category 2] and removal of 
the hydrogen [and oxygen] monitors from TS will not prevent an 
accident management strategy through the use of the SAMGs [severe 
accident management guidelines], the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen [and oxygen] monitor 
requirements, including removal of these

[[Page 45458]]

requirements from TS, does not involve a significant increase in the 
probability or the 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 elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen [and oxygen] monitor requirements, 
including removal of these requirements from TS, will not result in 
any failure mode not previously analyzed. The hydrogen recombiner 
and hydrogen [and oxygen] monitor equipment was intended to mitigate 
a design-basis hydrogen release. The hydrogen recombiner and 
hydrogen [and oxygen] monitor equipment are not considered accident 
precursors, nor does their existence or elimination have any adverse 
impact on the pre-accident state of the reactor core or post 
accident confinement of radionuclides within the containment 
building.
    Therefore, this change does 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 elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen [and oxygen] monitor requirements, 
including removal of these requirements from TS, in light of 
existing plant equipment, instrumentation, procedures, and programs 
that provide effective mitigation of and recovery from reactor 
accidents, results in a neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the TMI [Three-Mile Island], 
Unit 2 accident can be adequately met without reliance on safety-
related hydrogen monitors.
    [Category 2 oxygen monitors are adequate to verify the status of 
an inerted containment.]
    Therefore, this change does not involve a significant reduction 
in the margin of safety. The intent of the requirements established 
as a result of the TMI, Unit 2 accident can be adequately met 
without reliance on safety-related oxygen monitors.] Removal of 
hydrogen [and oxygen] monitoring from TS will not result in a 
significant reduction in their functionality, reliability, and 
availability.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Branch Chief: Thomas H. Boyce.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of amendment request: June 13, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TS) 5.5.9, ``Ventilation Filter 
Testing Program (VFTP),'' to impose lower (i.e., more restrictive) 
limits on the maximum pressure drop across the combined high efficiency 
particulate air filters and charcoal adsorbers in three safety-related 
ventilation systems. These ventilation systems are the Control Room 
Emergency Ventilation System, the Engineered Safety Features 
Ventilation System, and the Fuel-Handling Area Exhaust Ventilation 
System.
    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 of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    The proposed change consists of establishing more restrictive 
criteria in the Technical Specification (TS) for the maximum 
pressure drop across high efficiency particulate air filters (HEPA) 
and charcoal adsorbers in safety-related ventilation systems. These 
TS criteria are used to determine the acceptability of periodic test 
results. These criteria are not accident initiators. Therefore, 
there will be no effect on the probability of an accident. The 
safety-related ventilation systems involved in the proposed change 
function to mitigate the consequences of accidents. The proposed 
change will provide increased assurance that the HEPA filters and 
charcoal adsorbers in these systems will be capable of performing 
their safety function of reducing the release of radioactive 
material resulting from evaluated accidents. Therefore, there will 
be no increase in the consequences of those accidents.
    Therefore, the proposed change will 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 consists of establishing more restrictive 
acceptance criteria for existing TS[-]required tests. The proposed 
change does not affect the manner in which the tests are performed. 
The proposed change will not result in any new or different methods 
or modes of operation of existing structures, systems, or 
components. The proposed change will not introduce any new 
structures, system, or components.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety associated with the proposed change is the 
capability of the applicable safety-related ventilation systems to 
prevent radiation exposures from exceeding acceptable limits due to 
the release of radioactive material caused by an evaluated accident. 
The proposed change will provide increased assurance that the HEPA 
filters and charcoal adsorbers in these systems will be capable of 
performing this function.
    Therefore, the proposed change will not involve a significant 
reduction in the 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 requests involve no 
significant hazards consideration.
    Attorney for licensee: Kimberly Harshaw, Esquire, One Cook Place, 
Bridgman, MI 49106.
    NRC Acting Branch Chief: Travis Tate.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of amendment request: June 27, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TS) Surveillance Requirements 3.8.1.2, 
8, 12, 13, 16, and 19, changing the steady state frequency of all 
diesel generators (DGs) from the current allowed frequency range of 
59.4-61.2 Hz, to 59.4-60.5 Hz (i.e., a decrease of the upper limit, 
resulting in narrowing of

[[Page 45459]]

the current range). The licensee stated that the current frequency 
range is nonconservative and could result in undesirable effects such 
as centrifugal charging pump motor brake horsepower exceeding its 
nameplate maximum horsepower, and overloading the DGs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided 
its analysis of the issue of no significant hazards consideration. The 
NRC staff has performed its own analysis, which is presented below:
    (1) Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    The more restrictive steady state frequency range ensures that 
the diesel generators and equipment being powered by the diesel 
generators will function as designed to mitigate an accident as 
described in the Update Final Safety Analysis Report (UFSAR). The 
DGs and the equipment they power are part of the systems required to 
mitigate accidents; no accident analyzed in the UFSAR is initiated 
by mitigation equipment. Therefore, the proposed change to the 
allowed frequency range of the DGs will not have any impact on the 
probability of an accident previously evaluated. Furthermore, other 
than narrowing the allowed frequency range of the DGs, there is no 
other design or operational change. Therefore, the proposed change 
does not increase the probability of malfunction of the DGs or the 
equipment they power.
    Narrowing of the DG maximum steady state frequency limit will 
ensure that the DGs and equipment powered by the DGs will perform as 
originally designed and analyzed to mitigate the consequences of any 
accident described in the UFSAR. Therefore, the proposed change does 
not involve a significant increase in the consequences of an 
accident previously evaluated in the UFSAR.
    (2) Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    There is no design change associated with the proposed 
amendment. Making an existing DG requirement more restrictive alone 
will not alter plant configuration because no new or different type 
of equipment will be installed, and because no methods governing 
plant operation will be changed. The proposed change to allowed 
frequency range will not have any effect on the assumptions of 
accident scenarios previously made in the UFSAR. 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.
    Despite the proposed change to the DG maximum steady state 
frequency limit, the DGs and equipment powered by the DGs will 
continue to perform as originally designed, and originally analyzed 
in the UFSAR. There is no associated change to the methods and 
assumptions used to analyze DG performance. The proposed change will 
maintain the required function of the DGs and the equipment powered 
by the DGs to ensure that operation of structures, systems, or 
components is as currently set forth in the UFSAR. Therefore, the 
proposed change does not involve a significant reduction in the 
margin of safety.

    The Nuclear Regulatory Commission (NRC) staff has reviewed the 
licensee's analysis and, based on its own analysis, it appears that the 
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC 
staff proposes to determine that the proposed amendment involves no 
significant hazards consideration.
    Attorney for licensee: Kimberly Harshaw, Esquire, One Cook Place, 
Bridgman, MI 49106.
    NRC Acting Branch Chief: Travis L. Tate.

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

    Date of amendment request: July 9, 2007.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS), an NRC-controlled document, 
by moving the Table of Contents (TOC) out of the TS and making the TOC 
into a licensee-controlled document.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) Part 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration (NSHC) 
which is reproduced below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No.
    The proposed change is administrative and affects control of a 
document, the TOC, listing the specifications in the plant TS. 
Transferring control from the NRC to NMC (the licensee) does not 
affect the operation, physical configuration, or function of plant 
equipment or systems. It does not impact the initiators or 
assumptions of analyzed events, nor does it impact the mitigation of 
accidents or transient events. The change has no impact on, and 
hence cannot increase, the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No.
    The proposed change is administrative and does not alter the 
plant configuration, require installation of new equipment, alter 
assumptions about previously analyzed accidents, or impact the 
operation or function of plant equipment or systems. Therefore, this 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No.
    The proposed change is administrative. The TOC is not required 
by regulation to be in the TS. [Its] removal does not impact any 
safety assumptions or have the potential to reduce a margin of 
safety as described in the TS Bases. The change involves a transfer 
of control of the TOC from the NRC to NMC. No change in the 
technical content of the TS [ ] is involved. Consequently, transfer 
from the NRC to NMC has no impact on the margin of safety, and hence 
cannot involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this analysis, it appears that the three standards of 10 CFR 50.92(c) 
are satisfied. Therefore, the NRC staff proposes to determine that the 
proposed amendment 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.

PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric 
Station, Unit 2

    (SSES 2), Luzerne County, Pennsylvania.
    Date of amendment request: March 2, 2007.
    Description of amendment request: The proposed amendment would add 
an ACTIONS Note 3 to the SSES 2 Technical Specification 3.8.1, ``AC 
Sources--Operating,'' to allow a Unit 1 4160 volt subsystem to be de-
energized and removed from service to perform bus maintenance.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    This change does not involve any physical change to structures, 
systems, or components (SSCs) and does not alter the method of 
operation of any SSCs. The current assumptions in the safety 
analysis regarding accident initiators and mitigation of accidents 
are unaffected by these changes. No

[[Page 45460]]

SSC failure modes or mechanisms are being introduced, and the 
likelihood of previously analyzed failures remains unchanged.
    Operation in accordance with the proposed new ACTIONS Note 3 in 
Unit 2 Technical Specification 3.8.1 ensures that the AC 
[alternating current] distribution system and supported equipment 
remain capable of performing their functions as described in the 
Final Safety Analysis Report (FSAR). There are no changes to any 
accident initiators or to the mitigating capability of safety-
related equipment supported by the Class 1E Electrical AC system. 
The protection provided by these safety-related systems will 
continue to be provided as assumed by the safety analysis.
    Therefore, this proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
any plant equipment. No new equipment is being introduced, and 
installed equipment is not being operated in a new or different 
manner. There are no setpoints, at which protective or mitigative 
actions are initiated, affected by this change. This change does not 
alter the manner in which equipment operation is initiated, nor will 
the function demands on credited equipment be changed. No 
alterations in the procedures that ensure the plant remains within 
analyzed limits are being proposed, and no changes are being made to 
the procedures relied upon to respond to an off-normal event as 
described in the FSAR [final safety analysis report]. As such, no 
new failure modes are being introduced. The change does not alter 
assumptions made in the safety analysis and licensing basis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The proposed change is acceptable because the new 
ACTIONS Note 3 has been established to be consistent with the 
existing completion times for declaring required equipment 
inoperable that has no offsite power or DG [diesel generator] power 
available. Therefore, the plant response to analyzed events is not 
affected by this change and will continue to provide the margin of 
safety assumed by the safety analysis.
    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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Branch Chief: Mark G. Kowal.

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

    Date of amendment request: July 26, 2007
    Description of amendment request: The proposed amendment would 
remove values for turbine first stage pressure equivalent to 
Pbypass from the Technical Specifications. 
Pbypass is the reactor power level below which the turbine 
stop valve closure and the turbine control valve fast closure reactor 
protection system trip functions and the end-of-cycle recirculation 
pump trip are bypassed automatically.
    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 proposed removal of values for turbine first stage pressure 
associated with Pbypass from the Technical Specifications 
does not alter the requirements for component operability or 
surveillance currently in the Technical Specifications. The proposed 
change will have no impact on any safety related structures, systems 
or components.
    The probability of occurrence of a previously evaluated accident 
is not increased because this change does not introduce any new 
potential accident initiating conditions. The consequences of 
accidents previously evaluated in the UFSAR [Updated Final Safety 
Analysis Report] are not affected because the ability of the 
components to perform their required function is not affected.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change is administrative in nature, and does not 
result in physical alterations or changes in the method by which any 
safety related system performs its intended function. The proposed 
change does not affect any safety analysis assumptions. The proposed 
change does not create any new accident initiators or involve an 
activity that could be an initiator of an accident of a different 
type.
    All components will continue to be tested to the same 
requirements as defined in the Technical Specification Surveillance 
Requirements. The proposed revision does not make changes in any 
method of testing or how any safety related system performs its 
safety functions.
    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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change to remove values for turbine first stage 
pressure associated with Pbypass from the Technical 
Specifications does not alter the Technical Specification 
requirements for reactor protection system operability. The turbine 
first stage pressure setpoint will be controlled in accordance with 
plant procedures and will be verified during post-installation 
testing.
    The proposed change will not affect the current Technical 
Specification requirements or the components to which they apply.
    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: Mr. Jeffrie J. Keenan, Esquire, PSEG 
Nuclear--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Harold K. Chernoff.

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

    Date of amendment request: June 26, 2007.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TSs) to establish more effective 
and appropriate action, surveillance, and administrative requirements 
related to ensuring the habitability of the control room envelope (CRE) 
in accordance with Nuclear Regulatory Commission (NRC)-approved TS Task 
Force (TSTF) Standard Technical Specification change traveler TSTF-448, 
Revision 3, ``Control Room Habitability.'' Specifically, the proposed 
amendment would modify TS 3.7.7, Control Room Makeup and Cleanup 
Filtration System (CRMCFS) and TS Section 6.8, ``Administrative 
Controls-Procedures, Programs, and Manuals.'' The NRC staff issued a 
``Notice of Availability of Technical Specification Improvement to

[[Page 45461]]

Modify Requirements Regarding Control Room Envelope Habitability Using 
the Consolidated Line Item Improvement Process'' associated with TSTF-
448, Revision 3, in the Federal Register on January 17, 2007 (72 FR 
2022). The notice included a model safety evaluation, a model no 
significant hazards consideration (NSHC) determination, and a model 
license amendment request. In its application dated June 26, 2007, the 
licensee affirmed the applicability of the model NSHC determination 
which is presented below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC adopted by the licensee 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 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 has reviewed the analysis adopted by the licensee 
and, based on this review, it appears that the standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the request for amendments involves NSHC.
    Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Branch Chief: Thomas G. Hiltz.

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

    Date of amendment request: April 10, 2007.
    Brief description of amendments: The proposed amendments would 
revise Technical Specifications (TS) 3.1, ``Reactivity Control 
Systems,'' TS 3.2, ``Power Distribution Limits,'' TS 3.3, 
``Instrumentation,'' and TS 5.6.5b, ``Core Operating Limits Report 
(COLR).'' The requested change proposes to incorporate standard 
Westinghouse-developed and NRC-approved analytical methods into the 
lists of methodologies used to establish the core operating limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    No physical plant changes or changes in manner in which the 
plant will be operated as a result of the methodology changes. The 
proposed changes do not impact the condition or performance of any 
plant structure, system or component. The core operating limits are 
established to support Technical Specifications 3.1, 3.2, 3.3, and 
3.4. The core operating limits ensure that fuel design limits are 
not exceeded during any conditions of normal operation or in the 
event of any Anticipated Operational Occurrence (AOO). The methods 
used to establish the core operating limits for each operating cycle 
are based on methods previously found acceptable by the NRC and 
listed in Technical Specifications section 5.6.5.b. Application of 
these NRC-approved methods will continue to ensure that acceptable 
operating limits are established to protect the fuel cladding 
integrity during normal operation and AOOs. The requested Technical 
Specification changes, including those changes proposed to conform 
with the NRC-approved analysis methodologies, do not involve any 
plant modifications or operational changes that could affect system 
reliability, performance, or possibility of operator error. The 
requested changes do not affect any postulated accident precursors, 
does not affect any accident mitigation systems, and does not 
introduce any new accident initiation mechanisms.
    As a result, the proposed changes to the CPSES [Comanche Peak 
Steam Electric Station] Technical Specifications do not involve any 
increase in the probability or the consequences of any accident or 
malfunction of equipment important to safety previously evaluated 
since neither accident probabilities nor consequences are being 
affected by this proposed change.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    There are no physical changes being made to the plant. No new 
modes of plant operation are being introduced. The parameters 
assumed in the analyses are within the design limits of the existing 
plant equipment. All plant systems will perform as designed during 
the response to a potential accident.
    Therefore, the proposed change to the CPSES Technical 
Specifications does not create the possibility of a new or different 
kind of accident or malfunction of equipment important to safety 
from any accident previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The NRC-approved accident analysis methodologies include 
restrictions on the

[[Page 45462]]

choice of inputs, the degree of conservatism inherent in the 
calculations, and specified event acceptance criteria. Analyses 
performed in accordance with these methodologies will not result in 
adverse effects on the regulated margin of safety. Similarly, the 
use of axial power distribution controls based on the relaxed axial 
offset control strategy is a time-proven and NRC-approved method. 
The method is consistent with the accident analyses assumptions as 
described in the list of NRC-approved methodologies proposed to be 
used to establish the core operating limits. Finally, the proposed 
changes to allow operation with the BEACON [Best Estimate Analyzer 
for Core Operation Nuclear] power distribution monitoring tool 
provide additional information to the reactor operators on the state 
of the reactor core. Again, the use of the BEACON tool and the 
methodology used to develop the inputs to the tool are consistent 
with and controlled by the NRC-approved methodologies used to 
establish the core operating limits. As such, the margin of safety 
assumed in the plant safety analysis is not adversely affected by 
the proposed changes.
    Based on the above evaluations, TXU Power concludes that the 
proposed amendment(s) present 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.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Branch Chief: Thomas G. Hiltz.

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

    Date of amendment request: May 22, 2007.
    Brief description of amendments: The proposed amendment would 
revise the Technical Requirements Surveillance (TRS) 13.3.33.2, Cycling 
Frequency for the Turbine Stop and Control Valves. The proposed change 
would increase the frequency interval for the turbine stop and control 
valves testing from 12 to 26 weeks.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change will increase the frequency interval for 
testing the high pressure (HP) and low pressure (LP) turbine stop 
and control valves to 26 weeks. This test requires the movement of 
the HP and LP turbine valves through one complete cycle once every 
26 weeks. The test verifies freedom of movement of the valve 
components and is beneficial in early detection of problems with 
valve operation. [The test ensures that all turbine steam inlet 
valves are capable of closing to protect the turbine from excessive 
overspeed, which could generate potentially damaging missiles.]
    Siemens, the turbine manufacturer for Comanche Peak Steam 
Electric Station (CPSES), has evaluated the change in the 
probability of generating external/high-trajectory turbine missiles 
resulting from a hypothetical LP turbine disk failure which could 
adversely affect safety-related SSCs [structures, systems, and 
components] due to the change in the surveillance interval weeks 
using a previously approved missile probability analysis 
methodology. The results of the analysis show the new valve test 
interval of 26 weeks with a turbine inspection interval of 100,000 
hours is safe and acceptable as the probability of occurrence of a 
turbine missile per turbine year is less than the Nuclear Regulatory 
Commission (NRC) limit of 1E-4 per 8760 hours (turbine year) or 
11.42E-4 at 100,000 hours (Reference 7.4 [of the licensee's May 22, 
2007, application]). Therefore, the risk of the loss of an essential 
system from a single event