Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 28717-28728 [E7-9523]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 72, Number 98 (Tuesday, May 22, 2007)]
[Notices]
[Pages 28717-28728]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E7-9523]


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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 April 27, 2007, to May 10, 2007. The last 
biweekly notice was published on May 8, 2007 (72 FR 26173).

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

[[Page 28718]]

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.

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit No. 1, DeWitt County, Illinois

    Date of amendment request: December 12, 2006.
    Description of amendment request: The proposed amendment would 
revise Surveillance Requirement (SR) 3.3.1.1.8 and SR 3.3.1.3.2 to 
increase the interval between local power range monitor (LPRM) 
calibrations from 1000 megawatt-days per ton (MWD/T) average core 
exposure to 2000 MWD/T average core exposure. The proposed increase in 
the interval between required LPRM calibrations is acceptable due to 
improvements in fuel analytical bases, core monitoring processes, and 
nuclear instrumentation.
    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 amendment revises the surveillance interval for the 
LPRM calibration from 1000 MWD/T average core exposure to 2000 MWD/T 
average core exposure. Increasing the frequency interval between 
required LPRM calibrations is acceptable due to improvements in fuel 
analytical bases, core monitoring processes, and nuclear 
instrumentation. Therefore, the revised surveillance interval 
continues to ensure that the LPRM detector signal will continue to 
be adequately calibrated.
    This change will not alter the operation of process variables, 
structures, systems, or components as described in the CPS [Clinton 
Power Station] Updated Safety Analysis Report (USAR). The proposed 
change does not alter the initiation conditions or operational 
parameters for the LPRM subsystem and there is no new equipment 
introduced by the extension of the LPRM calibration interval. The 
performance of the Average Power Range Monitor (APRM) system, 
Oscillation Power Range Monitor (OPRM) system, Rod Control and 
Information System (RC&IS) and 3D MONICORE core monitoring system is 
not significantly affected by the proposed surveillance interval 
increase. The proposed LPRM calibration interval extension will have 
no significant effect on the Reactor Protection System (RPS) 
instrumentation accuracy during power maneuvers or transients and 
will therefore not significantly affect the performance of the RPS. 
As such, the probability of occurrence for a previously evaluated 
accident is not increased.
    The radiological consequences of an accident can be affected by 
the thermal limits existing at the time of the postulated accident; 
however, LPRM chamber exposure has no significant affect on the 
calculated thermal limits since LPRM accuracy does not

[[Page 28719]]

significantly deviate with exposure. For the LPRM extended 
calibration interval, the total nodal power uncertainty remains less 
than the uncertainty assumed in the General Electric BWR [boiling 
water reactor] Thermal Analysis Basis (GETAB) safety limit, 
maintaining the accuracy of the thermal limit calculation. 
Therefore, the thermal limit calculation is not significantly 
affected by LPRM calibration frequency, and thus the radiological 
consequences of any accident previously evaluated are not increased.
    Based on the above information, 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 previously evaluated?
    Response: No.
    The performance of the APRM, OPRM, RC&IS and 3D MONICORE systems 
is not significantly affected by the proposed LPRM surveillance 
interval increase. The proposed change does not affect the control 
parameters governing unit operation or the response of plant 
equipment to transient conditions. The proposed amendment does not 
change or introduce any new equipment, modes of system operation or 
failure mechanisms.
    Therefore, based on the above information, 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 has no impact on equipment design or 
fundamental operation, and there are no changes being made to safety 
limits or safety system allowable values that would adversely affect 
plant safety as a result of the proposed LPRM surveillance interval 
increase. The performance of the APRM, OPRM, RC&IS and 3D MONICORE 
systems is not significantly affected by the proposed change. The 
proposed LPRM calibration interval extension will have no 
significant effect on RPS instrumentation accuracy during power 
maneuvers or transients and will therefore not significantly affect 
the performance of the RPS. The margin of safety can be affected by 
the thermal limits existing at the time of the postulated accident; 
however, uncertainties associated with LPRM chamber exposure have no 
significant effect on the calculated thermal limits. The thermal 
limit calculation is not significantly affected since LPRM 
sensitivity with exposure is well defined. LPRM accuracy remains 
within the total nodal power uncertainty assumed in the GETAB, 
therefore maintaining thermal limits and the safety margin. The 
proposed change does not affect safety analysis assumptions or 
initial conditions and therefore, the margin of safety in the 
original safety analyses is maintained.
    Based on the above information, 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 J. Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Russell Gibbs.

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit No. 1, DeWitt County, Illinois

    Date of amendment request: January 26, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.3.1.1, ``Reactor Protection 
System (RPS) Instrumentation,'' Table 3.3.1.1-1, ``Reactor Protection 
System Instrumentation,'' Function 8, ``Scram Discharge Volume Water 
Level--High,'' item b, ``Float Switches,'' by replacing Surveillance 
Requirement (SR) 3.3.1.1.9 with SR 3.3.1.1.12. This change will 
effectively revise the surveillance frequency for the scram discharge 
volume (SDV) level float switch from every 92 days to every 24 months.
    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 TS change involves a change in the surveillance 
frequency for the SDV water level float switch channel functional 
test. The proposed TS change does not physically impact the plant. 
The proposed change does not affect the design of the SDV water 
level instruments, the operational characteristics or function of 
the instruments, the interfaces between the instruments and the RPS, 
or the reliability of the SDV water level instruments. The proposed 
TS change does not degrade the performance of, or increase the 
challenges to, any safety systems assumed to function in the 
accident analysis. As noted in the Bases to TS 3.3.1.1, even though 
the two types of SDV Water Level--High Functions are an input to the 
RPS logic, no credit is taken for a scram initiated from these 
functions for any of the design basis accidents or transients 
evaluated in the CPS [Clinton Power Station] Updated Safety Analysis 
Report (USAR). An inoperable SDV water level instrument is not 
considered as an initiator of any analyzed event. The proposed TS 
change does not impact the usefulness of the SRs in evaluating the 
operability of required systems and components, or the way in which 
the surveillances are performed. In addition, the frequency of 
surveillance testing is not considered an initiator of any analyzed 
accident, nor does a revision to the frequency introduce any 
accident initiators. Therefore, the proposed change does not involve 
a significant increase in the probability of an accident previously 
evaluated.
    The consequences of a previously analyzed event are dependent on 
the initial conditions assumed in the analysis, the availability and 
successful functioning of equipment assumed to operate in response 
to the analyzed event, and the setpoints at which these actions are 
initiated. The consequences of a previously evaluated accident are 
not significantly increased by the proposed change. The proposed 
change does not affect the performance of any equipment credited to 
mitigate the radiological consequences of an accident. The risk 
assessment of the proposed changes has concluded that there is an 
insignificant increase in the core damage frequency as well as the 
total population dose rate. Historical review of surveillance test 
results and associated maintenance records did not find evidence of 
failures that would invalidate the above conclusions.
    Therefore, the proposed change does not alter the ability to 
detect and mitigate events and, as such, does not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any previously evaluated?
    Response: No.
    The proposed TS change does not introduce any failure mechanisms 
of a different type than those previously evaluated, since there are 
no physical changes being made to the facility. No new or different 
equipment is being installed. No installed equipment is being 
operated in a different manner. There is no change being made to the 
parameters within which CPS is operated. There are no setpoints at 
which protective or mitigative actions are initiated that are 
affected by this proposed action. The change does not alter 
assumptions made in the safety analysis. This proposed action will 
not alter the manner in which equipment operation is initiated, nor 
will the function demands on credited equipment be changed. No 
alteration in the procedures, which ensure the unit remains within 
analyzed limits, is proposed, and no change is being made to 
procedures relied upon to respond to an off-normal event. As a 
result, no new failure modes are being introduced. The way 
surveillance tests are performed remains unchanged. A historical 
review of surveillance test results and associated maintenance 
records indicated there was no evidence of any failures that would 
invalidate the above conclusions.
    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.
    Margins of safety are established in the design of components, 
the configuration of components to meet certain performance 
parameters, and in the establishment of setpoints to initiate alarms 
or actions. The

[[Page 28720]]

proposed TS change involves a change in the surveillance frequency 
for the SDV water level float switch channel functional test. There 
is no change in the design of the affected systems, no alteration of 
the setpoints at which alarms or actions are initiated, and no 
change in plant configuration from original design. The proposed 
change does not significantly impact the condition or performance of 
structures, systems, and components relied upon for accident 
mitigation. The proposed change does not result in any hardware 
changes or in any changes to the analytical limits assumed in 
accident analyses. Existing operating margin between plant 
conditions and actual plant setpoints is not significantly reduced 
due to these changes. The proposed change does not significantly 
impact any safety analysis assumptions or results.
    AmerGen has conducted a risk assessment to determine the impact 
of a change to the SDV water level instrument surveillance frequency 
from the current once every 92 days to once every 24 months for the 
risk measures of Core Damage Frequency (CDF) and Large Early Release 
Frequency (LERF). This assessment indicated that the proposed CPS 
surveillance frequency 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 J. Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Russell Gibbs.

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

    Date of amendment request: April 30, 2007.
    Description of amendment request: The amendment will revise the 
technical specifications to use other narrow range containment sump 
water level instrumentation rather than the existing redundant 
instruments to allow installation of new emergency core cooling system 
recirculation sumps strainers as specified in the Nuclear Regulatory 
Commission Generic Letter 2004-02, Potential Impact of Debris Blockage 
on Emergency Recirculation during Design Basis Accidents at Pressurized 
Water Reactors. 
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated, and it does not change an accident previously evaluated 
in the Final Safety Analysis Report (FSAR). The use of other narrow 
range containment sump water level instruments rather than the 
existing narrow range containment recirculation sump water level 
instruments, which have level elements located inside the emergency 
core cooling system (ECCS) recirculation sumps, will continue to 
ensure that acceptable narrow range containment sump water level 
monitoring is maintained during post-accident conditions. Operation 
of the containment spray and residual heat removal systems is 
unchanged as a result of the proposed amendment. The level elements 
associated with the existing narrow range containment recirculation 
sump water level instruments are not accident initiators, and the 
FSAR does not credit these level elements in the dose analyses for 
loss-of-coolant accidents. The proposed amendment does not adversely 
affect the ability of structures, systems, or components (SSCs) to 
perform their design function. SSCs required for post-accident 
recirculation remain capable of performing their design functions.
    Therefore, this amendment 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 amendment does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated, and it does not change an accident previously evaluated 
in the Final Safety Analysis Report (FSAR). The use of other narrow 
range containment sump water level instruments rather than the 
existing narrow range containment recirculation sump water level 
instruments supports the replacement of the existing containment 
recirculation sump screens with new strainers in accordance with the 
response to Generic Letter 2004-02, Potential Impact of Debris 
Blockage on Emergency Recirculation during Design Basis Accidents at 
Pressurized-Water Reactors. The proposed amendment does not change 
the design function or the operation of the containment spray and 
residual heat removal systems associated with the containment 
recirculation sumps. The proposed amendment does not create new 
failure mechanisms or malfunctions or accident initiators. The 
proposed amendment will continue to ensure that acceptable narrow 
range containment sump water level monitoring is maintained during 
post-accident conditions, and that SSCs required for post-accident 
recirculation remain capable of performing their design functions.
    Therefore, this amendment 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 involve a significant reduction 
in a margin of safety. The proposed amendment does not adversely 
affect a plant safety limit or a limiting safety system setting, and 
does not alter a design basis limit for a parameter evaluated in the 
FSAR. The use of other narrow range containment sump water level 
instruments, which meet the requirements of the FSAR, rather than 
the existing narrow range containment recirculation sump water level 
instruments, will continue to ensure that acceptable narrow range 
containment sump water level monitoring is maintained during post-
accident conditions. The proposed amendment does not adversely 
affect the ability of SSCs to perform their design functions or the 
reliability of equipment to mitigate accidents evaluated in the 
FSAR. The proposed amendment will continue to ensure that SSCs 
required for post-accident recirculation remain capable of 
performing their design functions.
    Therefore, this amendment 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: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Branch Chief: Thomas H. Boyce.

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

    Date of amendment request: April 24, 2007.
    Description of amendment request: The proposed change will add 
Optimized ZIRLOTM as an acceptable fuel rod cladding 
material in the Waterford Steam Electric Station, Unit 3 (Waterford 3), 
Technical Specification (TS) 5.3.1, ``Fuel Assemblies.'' TS 5.3.1 
currently identifies, in part, Zircaloy or ZIRLOTM fuel rod 
cladding as the allowable fuel rod cladding material.
    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

[[Page 28721]]

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 NRC-approved topical report WCAP-12610-P-A and CENPD-404-P-
A, Addendum 1-A, ``Optimized ZIRLOTM,'' prepared by 
Westinghouse Electric Company, LLC (Westinghouse), addresses 
Optimized ZIRLOTM and demonstrates that Optimized 
ZIRLOTM has essentially the same properties as currently 
licensed ZIRLOTM. The fuel cladding itself is not an 
accident initiator and does not affect accident probability. Use of 
Optimized ZIRLOTM fuel cladding has been shown to meet 
all 10 CFR 50.46 design criteria and, therefore, will not increase 
the consequences of an accident.
    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.
    Use of Optimized ZIRLOTM clad fuel will not result in 
changes in the operation or configuration of the facility. Topical 
report WCAP-12610-P-A and CENPD-404-P-A demonstrated that the 
material properties of Optimized ZIRLOTM are similar to 
those of standard ZIRLOTM. Therefore, Optimized 
ZIRLOTM fuel rod cladding will perform similarly to those 
fabricated from standard ZIRLOTM, thus precluding the 
possibility of the fuel becoming an accident initiator and causing a 
new or different type of accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change will not involve a significant reduction in 
the margin of safety because it has been demonstrated that the 
material properties of the Optimized ZIRLOTM are not 
significantly different from those of standard ZIRLOTM. 
Optimized ZIRLOTM is expected to perform similarly to 
standard ZIRLOTM for all normal operating and accident 
scenarios, including both loss-of-coolant accident (LOCA) and non-
LOCA scenarios. For LOCA scenarios, where the slight difference in 
Optimized ZIRLOTM material properties relative to 
standard ZIRLOTM could have some impact on the overall 
accident scenario, plant-specific LOCA analyses using Optimized 
ZIRLOTM properties will be performed prior to the use of 
fuel assemblies with fuel rods containing Optimized 
ZIRLOTM. These LOCA analyses will demonstrate that the 
acceptance criteria of 10 CFR 50.46 will be satisfied when Optimized 
ZIRLOTM fuel rod cladding is implemented.
    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.

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

    Date of amendment request: March 30, 2007.
    Description of amendment request: The proposed amendment would 
change the NMP2 Technical Specifications to reflect an expanded 
operating domain resulting from implementation of Average Power Range 
Monitor/Rod Block Monitor/Technical Specifications/Maximum Extended 
Load Line Analysis (ARTS/MELLLA). The Average Power Range Monitor 
(APRM) flow-biased simulated thermal power Allowable Value would be 
revised to permit operation in the MELLLA region. The current flow-
biased Rod Block Monitor (RBM) would be replaced by a power dependent 
RBM, which also would require new Allowable Values. The flow-biased 
APRM simulated thermal power setdown requirement would be replaced by 
more direct power and flow dependent thermal limits administration. The 
Surveillance Requirement for the standby liquid control (SLC) system 
would be revised to require each SLC pump to deliver required flow at a 
discharge pressure >=1325 psig in lieu of >=1320 psig; the SLC relief 
valve setpoint would be increased from 1394 psig to 1400 psig. Finally, 
the proposed amendment employs a new model for performing the 
anticipated transients without scram (ATWS) analysis for ARTS/MELLLA 
conditions.
    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 eliminates the APRM flow-biased simulated 
thermal power setdown requirement and substitutes power and flow 
dependent adjustments to the Minimum Critical Power Ratio (MCPR) and 
Linear Heat Generation Rate (LHGR) thermal limits. Thermal limits 
will be determined using NRC [Nuclear Regulatory Commission] 
approved analytical methods. The proposed change will have no effect 
upon any accident initiating mechanism. The power and flow dependent 
adjustments will ensure that the MCPR safety limit will not be 
violated as a result of any Anticipated Operational Occurrence 
(AOO), and that the fuel thermal and mechanical design bases will be 
maintained.
    The proposed change also expands the power and flow operating 
domain by relaxing the restrictions imposed by the formulation of 
the APRM flow-biased simulated thermal power Allowable Value and the 
replacement of the current flow-biased RBM with a new power 
dependent RBM. The APRM and RBM are not involved in the initiation 
of any accident, and the APRM flow-biased simulated thermal power 
function is not credited in any NMP2 safety analyses. The proposed 
change will not introduce any initial conditions that would result 
in NRC approved criteria being exceeded and the APRM and RBM will 
remain capable of performing their design functions.
    The Standby Liquid Control (SLC) System is provided to mitigate 
anticipated transients without scram (ATWS) events and, as such, is 
not considered an initiator of an ATWS event or any other analyzed 
accident. The revised SLC discharge pump test pressure neither 
reduces the ability of the SLC system to respond to or mitigate an 
ATWS event nor increases the likelihood of a system malfunction that 
could increase the consequences of an accident.
    Based on the above discussion, it is concluded that 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 eliminates the APRM flow-biased simulated 
thermal power setdown requirement and substitutes power and flow 
dependent adjustments to the MCPR and LHGR thermal limits. Because 
the thermal limits will continue to be met, no analyzed transient 
event will escalate into a new or different type of accident due to 
the initial starting conditions permitted by the adjusted thermal 
limits.
    The proposed change also expands the power and flow operating 
domain by relaxing the restrictions imposed by the formulation of 
the APRM flow-biased simulated thermal power Allowable Value and the 
replacement of the current flow-biased RBM with a new power 
dependent RBM. Changing the formulation for the APRM flow-biased 
simulated thermal power Allowable Value and changing from a flow-
biased RBM to a power dependent RBM does not change their respective 
functions and manner of

[[Page 28722]]

operation. The change does not introduce a sequence of events or 
introduce a new failure mode that would create a new or different 
[kind] of accident. While not credited, the APRM flow-biased 
simulated thermal power Allowable Value and associated scram trip 
setpoint will continue to initiate a scram to protect the MCPR 
safety limit. The power dependent RBM will prevent rod withdrawal 
when the power dependent RBM rod block setpoint is reached. No new 
failure mechanisms, malfunctions, or accident initiators are being 
introduced by the proposed change. In addition, operating within the 
expanded power flow map will not require any systems, structures or 
components to function differently than previously evaluated and 
will not create initial conditions that would result in a new or 
different kind of accident from any accident previously evaluated.
    The proposed change to the SLC pump test discharge pressure is 
consistent with the functional requirements of the ATWS rule (10 CFR 
50.62). This proposed change does not involve the installation of 
any new or different type of equipment, does not introduce any new 
modes of plant operation, and does not change any methods governing 
normal plant operation.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change eliminates the APRM flow-biased simulated 
thermal power setdown requirement and substitutes power and flow 
dependent adjustments to the MCPR and LHGR thermal limits. 
Replacement of the APRM setdown requirement with power and flow 
dependent adjustments to the MCPR and LHGR thermal limits will 
continue to ensure that margins to the fuel cladding Safety Limit 
are preserved during operation at other than rated conditions. 
Thermal limits will be determined using NRC approved analytical 
methods. The power and flow dependent adjustments will ensure that 
the MCPR safety limit will not be violated as a result of any AOO, 
and that the fuel thermal and mechanical design bases will be 
maintained.
    The proposed change also expands the power and flow operating 
domain by relaxing the restrictions imposed by the formulation of 
the APRM flow-biased simulated thermal power Allowable Value and the 
replacement of the current flow-biased RBM with a new power 
dependent RBM. The APRM flow-biased simulated thermal power 
Allowable Value and associated scram trip setpoint will continue to 
initiate a scram to protect the MCPR safety limit. The RBM will 
continue to prevent rod withdrawal when the power dependent RBM rod 
block setpoint is reached. The MCPR and LHGR thermal limits will be 
developed to ensure that fuel thermal mechanical design bases remain 
within the licensing limits during a control rod withdrawal error 
event and to ensure that the MCPR safety limit will not be violated 
as a result of a control rod withdrawal error event. Operation in 
the expanded operating domain will not alter the manner in which 
safety limits, limiting safety system settings, or limiting 
conditions for operation are determined. AOOs and postulated 
accidents within the expanded operating domain will continue to be 
evaluated using NRC approved methods. The 10 CFR 50.46 acceptance 
criteria for the performance of the ECCS [emergency core cooling 
system] following postulated LOCAs [loss-of-coolant accidents] will 
continue to be met.
    The proposed change to the SLC pump discharge test pressure does 
not alter the results of any accident analyses. The proposed change 
is consistent with the functional requirements of the ATWS rule (10 
CFR 50.62). The ability of the SLCS to respond to and mitigate an 
ATWS event is not affected.
    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: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006.
    NRC Branch Chief: Mark G. Kowal.

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

    Date of amendment requests: April 17, 2007.
    Description of amendment requests: A change is proposed to the 
standard technical specifications (STS) (NUREGs 1430 through 1434) and 
plant-specific technical specifications (TS), to strengthen TS 
requirements regarding control room envelope (CRE) habitability by 
changing the action and surveillance requirements associated with the 
limiting condition for operation operability requirements for the CRE 
emergency ventilation system, and by adding a new TS administrative 
controls program on CRE habitability. Accompanying the proposed TS 
change are appropriate conforming technical changes to the TS Bases. 
The proposed revision to the Bases also includes editorial and 
administrative changes to reflect applicable changes to the 
corresponding STS Bases, which were made to improve clarity, conform 
with the latest information and references, correct factual errors, and 
achieve more consistency among the STS NUREGs. The proposed revision to 
the TS and associated Bases is consistent with STS as revised by TS 
Task Force (TSTF)-448, Revision 3, ``Control Room Envelope 
Habilitability.''
    The proposed amendment would revise the TS Improvement To Modify 
Requirements Regarding CRE Habitability using the Consolidated Line 
Item Improvement Process, based on the NRC-approved to TSTF-448, 
Revision 3. The NRC staff issued a notice of opportunity for comment in 
the Federal Register on October 17, 2006 (71 FR 61075), on possible 
amendments adopting TSTF-448, 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 following NSHC determination in its application 
dated April 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

[[Page 28723]]

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 licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Branch Chief: Thomas G. Hiltz.

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: February 28, 2007.
    Brief description of amendments: The proposed amendment request 
would revise the language of Technical Specification (TS) 3.7.1.2, 
``Auxiliary Feedwater System,'' Action b from ``MODE 3 may be entered 
with an inoperable turbine-driven auxiliary feedwater pump for the 
purposes of performing Surveillance Requirement 4.7.1.2.1a.2'' to 
``MODE 3 may be entered with an inoperable turbine-driven auxiliary 
feedwater pump.''
    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 deletion of the existing words in TS 3.7.1.2 Action 
b is an administrative change that will clarify the Licensing Basis 
for the turbine-driven auxiliary feedwater pump. Since this change 
does not change the Licensing Basis for TS 3.7.1.2, this change 
cannot affect the probability or consequence of any accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed deletion of the existing words in TS 3.7.1.2 Action 
b is an administrative change that will clarify the Licensing Basis 
for the turbine-driven auxiliary feedwater pump. Since this change 
does not change the Licensing Basis for TS 3.7.1.2, this change 
cannot affect 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 deletion of the existing words in TS 3.7.1.2 Action 
b is an administrative change that will clarify the Licensing Basis 
for the turbine-driven auxiliary feedwater pump. Since this change 
does not change the Licensing Basis for TS 3.7.1.2, this change 
cannot 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: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Branch Chief: Thomas G. Hiltz.

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

    Date of amendment request: April 5, 2007.
    Description of amendment request: The proposed amendments would 
revise technical specifications (TSs) to change the surveillance 
frequency for the turbine trip functions of the reactor trip system 
instrumentation. The current frequency is prior to each reactor startup 
and the proposed change will revise this to be prior to exceeding the 
Permissive P-9 interlock whenever the unit has been in hot standby. The 
proposed change is consistent with NRC-approved Technical Specification 
Task Force Traveler TSTF-311, as incorporated into the latest revision 
of Standard TSs (NUREG-1431, Revision 3).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes revise the surveillance frequency for 
reactor trip functions from a turbine trip event. These changes do 
not alter these functions physically or how they are maintained. 
Delaying the performance of the surveillance up to the P-9 interlock 
will continue to ensure operability of the function before the plant 
is in a condition that would benefit from the associated actuation. 
The incorporation of a surveillance frequency that is consistent 
with the applicability for the function eliminates potential 
misapplication of the TS requirements. The frequency changes support 
turbine trip operability during plant startup and are consistent 
with their ability to perform the reactor trip functions. Since 
these changes will not affect the ability of these trips to perform 
the initiation of reactor trips when appropriate, the off-site dose 
consequences for an accident will not be impacted. Equally, the 
potential to cause an accident is not affected because no plant 
system or component has been altered by the proposed changes. 
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 changes only affect the surveillance frequency 
requirement for the turbine trip functions. This does not affect

[[Page 28724]]

any physical features of the plant or the manner in which these 
functions are utilized. The proposed surveillance frequency will 
require the functions to be verified operable before the turbine 
trip functions are applicable and able to perform their trip 
functions. Delaying the performance of the surveillance up to the P-
9 interlock will continue to ensure operability of the function 
before the plant is in a condition that would benefit from the 
associated actuation. Therefore, the proposed change does not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not alter any plant setpoints or 
functions that are assumed to actuate in the event of postulated 
accidents. In fact, the proposed changes do not alter any plant 
feature and only alter the requirements for when the function must 
be verified to be operable through surveillance testing. The 
proposed changes ensure the functionality of the turbine trips when 
assumed in the analysis for accident mitigation. 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Thomas H. Boyce.

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: December 19, 2006.
    Brief description of amendments: The proposed amendment request 
would revise the requirements in Technical Specification (TS) 5.5.8, 
``Inservice Testing Program,'' to update references to the American 
Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, 
Section XI, as the source of requirements for the inservice testing of 
ASME Code Class 1, 2, and 3 pumps and valves, and address the 
applicability of Surveillance Requirement 3.0.2 to other normal and 
accelerated frequencies specified as 2 years or less in the Inservice 
Testing Program.
    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 changes revise TS 5.5.8, ``Inservice Testing 
Program,'' for consistency with the requirements of 10 CFR 
50.55a(f)(4) regarding the inservice testing of pumps and valves. 
The proposed change incorporates revisions to the ASME Code that 
result in a net improvement in the measures for testing pumps and 
valves.
    The proposed changes do not impact any accident initiators or 
analyzed events or assumed mitigation of accident or transient 
events. They do not involve the addition or removal of any 
equipment, or any design changes to the facility. Therefore, the 
proposed changes do not represent 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 changes revise TS 5.5.8, ``Inservice Testing 
Program,'' for consistency with the requirements of 10 CFR 50.55a(f) 
regarding the inservice testing of pumps and valves. The proposed 
changes incorporate revisions to the ASME Code that result in a net 
improvement in the measures for testing pumps and valves.
    The proposed changes do not involve a modification to the 
physical configuration of the plant (i.e., no new equipment will be 
installed) or change in the methods governing normal plant 
operation. The proposed changes will not impose any new or different 
requirements or introduce a new accident initiator, accident 
precursor, or malfunction mechanism. Additionally, there is no 
change in the types or increases in the amounts of any effluent that 
may be released off-site and there is no increase in individual or 
cumulative occupational exposure. Therefore, these proposed changes 
do not create the possibility of an accident of a different kind 
than previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes revise TS 5.5.8, ``Inservice Testing 
Program,'' for consistency with the requirements of 10 CFR 
50.55a(f)(4) regarding the inservice testing of pumps and valves. 
The proposed changes incorporate revisions to the ASME Code that 
result in a net improvement in the measures for testing pumps and 
valves. The safety function of the affected pumps and valves will be 
maintained. Therefore, these proposed changes do not involve a 
significant reduction in a margin of safety.

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

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for a Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, 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,

[[Page 28725]]

(301) 415-4737 or by e-mail to pdr@nrc.gov.

Consumers Energy Company, Entergy Nuclear Palisades, LLC, and Entergy 
Nuclear Operations, Inc., Docket No. 50-155, Big Rock Point Facility, 
Charlevoix County, Michigan

    Date of application for amendment: October 31, 2006.
    Brief description of amendment: The license amendment reflects the 
changes in ownership and operating authority for the Big Rock Facility 
and its Independent Spent Fuel Storage Installation.
    Date of issuance: April 11, 2007.
    Effective date: As of the date of issuance.
    Amendment No.: 127.
    Facility Operating License No. DPR-06: The license amendment 
reflects the changes in ownership and operating authority for the Big 
Rock Facility and its Independent Spent Fuel Storage Installation.
    Date of ini