Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 35456-35466 [E6-9434]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 71, Number 118 (Tuesday, June 20, 2006)]
[Notices]
[Pages 35456-35466]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E6-9434]


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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 May 25, 2006 to June 8, 2006. The last 
biweekly notice was published on June 6, 2006 (71 FR 32603).

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

[[Page 35457]]

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, Rules and 
Directives 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,

[[Page 35458]]

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.

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

    Date of amendment request: May 8, 2006.
    Description of amendment request: The proposed change will add an 
NRC-approved topical report to the analytical methods referenced in 
Technical Specification (TS) Section 5.6.5, ``Core Operating Limits 
Report (COLR).''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Core operating limits are established each operating cycle in 
accordance with TS 3.2, ``Power Distribution'' and TS 5.6.5, ``Core 
Operating Limits Report (COLR)''. These core operating limits ensure 
that the fuel design limits are not exceeded during any conditions 
of normal operation or in the event of any Anticipated Operational 
Occurrence (AOO). In addition, the Average Planar Linear Heat 
Generation Rate (APLHGR) operating limits imposed by Technical 
Specification 3.2.1 also ensure that the Peak Cladding Temperature 
(PCT) during the postulated design[-]basis LOCA [loss-of-coolant 
accident] does not exceed the 2200 [deg]F limit specified in 10 CFR 
50.46. The APLHGR is a measure of the average linear heat generation 
rate of all the fuel rods in a fuel assembly at any axial location.
    The methods used to determine the operating limits are those 
previously found acceptable by the NRC and listed in TS Section 
5.6.5.b. A change to TS Section 5.6.5.b is requested to include an 
updated LOCA analysis method, EXEM BWR-2000. The updated method will 
be used to determine the APLHGR operating limits imposed by 
Technical Specification 3.2.1. EXEM BWR-2000 has been reviewed and 
approved by the NRC and is applicable to the GGNS [Grand Gulf 
Nuclear Station, Unit 1] plant design and the FRA-ANP [Framatome-
Advance Nuclear Power] fuel being used at GGNS. The application of 
the LOCA analytical model will continue to ensure that the APLHGR 
operating limits are established to protect the fuel cladding 
integrity during normal operation, AOOs, and the design-basis LOCA. 
The requested TS changes concern the use of analytical methods and 
do not involve any plant modifications or operational changes that 
could affect any postulated accident precursors or accident 
mitigation systems and do not introduce any new accident initiation 
mechanisms.
    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 TS amendment will not change the design function, 
reliability, performance, or operation of any plant systems, 
components, or structures. It does not create the possibility of a 
new failure mechanism, malfunction, or accident initiators not 
considered in the design and licensing bases. Plant operation will 
continue to be within the core operating limits that are established 
using NRC[-]approved methods that are applicable to the GGNS design 
and the GGNS fuel.
    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 ECCS [emergency core cooling system] performance analysis 
methods are used to establish the APLHGR limits required by 
Technical Specification 3.2.1. The APLHGR limits are specified in 
the COLR and are the result of fuel design, design[-]basis accident 
(DBA), and transient analyses. Limits on the APLHGR are specified to 
ensure that the fuel design limits are not exceeded during 
anticipated operational occurrences (AOOs) and that the peak 
cladding temperature (PCT) during the postulated design[-]basis LOCA 
does not exceed the 2200 [deg]F limit specified in 10 CFR 50.46.
    The EXEM BWR-2000 evaluation model is an updated LOCA analytical 
method that has been approved by the NRC and is applicable to the 
GGNS plant design and the fuel being used at GGNS. A GGNS plant[-
]specific ECCS performance analysis has been performed with the EXEM 
BWR-2000 evaluation model. This evaluation concluded that the 
resulting PCT still afforded adequate margin to the 2200 [deg]F 
limit of 10 CFR 50.46.
    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: Nicholas S. Reynolds, Esquire, Winston and 
Strawn LLP, 1700 K Street, NW., Washington, DC 20006
    NRC Branch Chief: David Terao.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 
No. 1, Fairfield County, South Carolina

    Date of amendment request: May 24, 2006.
    Description of amendment request: This amendment revises TS 1.0, 
Definitions, TS 3/4.4.5, Steam Generator Tube Integrity, TS 3/4.4.6.2, 
Reactor Coolant System (RCS) Operational LEAKAGE, adds a new 
specification TS 6.8.4.k for Steam Generator Program and adds a new TS 
6.9.1.12, Steam Generator Tube Inspection Report. The proposed changes 
are necessary in order to implement the guidance for the industry 
initiative on NEI 97-06, ``Steam Generator Program Guidelines.'' The 
NRC staff issued a notice of availability of a model safety evaluation 
and model no significant hazards consideration (NSHC) determination for 
referencing in license amendment applications in the Federal Register 
on March 2, 2005, (70 FR 10298). The licensee affirmed the 
applicability of the model NSHC determination in its application dated 
May 24, 2006.

[[Page 35459]]

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The proposed change requires a SG Program that includes 
performance criteria that will provide reasonable assurance that the 
SG tubing will retain integrity over the full range of operating 
conditions (including startup, operation in the power range, hot 
standby, cooldown and all anticipated transients included in the 
design specification). The SG performance criteria are based on tube 
structural integrity, accident induced leakage, and operational 
LEAKAGE.
    A SGTR event is one of the design basis accidents that are 
analyzed as part of a plant's licensing basis. In the analysis of a 
SGTR event, a bounding primary to secondary LEAKAGE rate equal to 
the operational LEAKAGE rate limits in the licensing basis plus the 
LEAKAGE rate associated with a double-ended rupture of a single tube 
is assumed.
    For other design basis accidents such as MSLB, rod ejection, and 
reactor coolant pump locked rotor the tubes are assumed to retain 
their structural integrity (i.e., they are assumed not to rupture). 
These analyses typically assume that primary to secondary LEAKAGE 
for all SGs is 1 gallon per minute or increases to 1 gallon per 
minute as a result of accident induced stresses. The accident 
induced leakage criterion introduced by the proposed changes 
accounts for tubes that may leak during design basis accidents. The 
accident induced leakage criterion limits this leakage to no more 
than the value assumed in the accident analysis.
    The SG performance criteria proposed change to the TS identify 
the standards against which tube integrity is to be measured. 
Meeting the performance criteria provides reasonable assurance that 
the SG tubing will remain capable of fulfilling its specific safety 
function of maintaining reactor coolant pressure boundary integrity 
throughout each operating cycle and in the unlikely event of a 
design basis accident. The performance criteria are only a part of 
the SG Program required by the proposed change to the TS. The 
program, defined by NEI 97-06, Steam Generator Program Guidelines, 
includes a framework that incorporates a balance of prevention, 
inspection, evaluation, repair, and leakage monitoring. The proposed 
changes do not, therefore, significantly increase the probability of 
an accident previously evaluated.
    The consequences of design basis accidents are, in part, 
functions of the DOSE EQUIVALENT 1-131 in the primary coolant and 
the primary to secondary LEAKAGE rates resulting from an accident. 
Therefore, limits are included in the plant technical specifications 
for operational leakage and for DOSE EQUIVALENT 1-131 in primary 
coolant to ensure the plant is operated within its analyzed 
condition. The typical analysis of the limiting design basis 
accident assumes that primary to secondary leak rate after the 
accident is 1 gallon per minute with no more than 150 gallons per 
day in any one SG, and that the reactor coolant activity levels of 
DOSE EQUIVALENT 1-131 are at the TS values before the accident. The 
proposed change does not affect the design of the SGs, their method 
of operation, or primary coolant chemistry controls. The proposed 
approach updates the current TSs and enhances the requirements for 
SG inspections. The proposed change does not adversely impact any 
other previously evaluated design basis accident and is an 
improvement over the current TSs.
    Therefore, the proposed change does not affect the consequences 
of a SGTR accident and the probability of such an accident is 
reduced. In addition, the proposed changes do not affect the 
consequences of an MSLB, rod ejection, or a reactor coolant pump 
locked rotor event, or other previously evaluated accident.

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

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

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

    The SG tubes in pressurized water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
isolate the radioactive fission products in the primary coolant from 
the secondary system. In summary, the safety function of an SG is 
maintained by ensuring the integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical condition of the tube. The proposed 
change does not affect tube design or operating environment. The 
proposed change is expected to result in an improvement in the tube 
integrity by implementing the SG Program to manage SG tube 
inspection, assessment, repair, and plugging. The requirements 
established by the SG Program are consistent with those in the 
applicable design codes and standards and are an improvement over 
the requirements in the current TSs.
    For the above reasons, the margin of safety is not changed and 
overall plant safety will be enhanced by the proposed change to the 
TS.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    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: J. Hagood Hamilton, South Carolina Electric 
& Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: Evangelos C. Marinos.

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

    Date of amendment request: May 1, 2006 (TS-05-10).
    Description of amendment request: The proposed amendment would 
extend the burnup limit of the Mark-BW fuel design with advanced alloy 
material referred to as M5 alloy. This proposed change affects Section 
6.9.1.14.a of the Sequoyah Nuclear Plant Technical Specifications 
(TSs). The impact to Section 6.9.1.14.a includes adding an NRC-approved 
topical report (TR) associated with M5 alloy fuel assemblies. This TR 
will be utilized, among others, in the determination of core operating 
limits for each fuel cycle. In addition, the proposed amendment 
includes the adoption of Industry/Technical Specification Task Force 
(TSTF) Traveler, TSTF-363, Revision 0, ``Revised Topical Report 
References in Improved Technical Specification (ITS) 5.6.5, Core 
Operating Limits Report (COLR),'' which removes any references to 
dates, revision numbers, and supplements in the TS listing of TRs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the

[[Page 35460]]

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.
    In general, fuel assemblies and more specifically fuel rod 
cladding, of any burnup level, is not a precursor to accidents 
previously evaluated. An evaluation has been performed of the Mark-
BW design fuel assembly for all loss-of-coolant accidents (LOCA) and 
non-LOCA transient events. This evaluation confirmed and justified 
the use of Mark-BW fuel for operation in Sequoyah Nuclear Plant 
(SQN) Units 1 and 2.
    The ability of the M5 fuel rod cladding material to provide a 
barrier against the release of radioactive fuel material has not 
been reduced with respect to the Zircaloy-4 material. The approved 
TR evaluated postulated accidents that involved adverse core 
conditions and the release of radionuclides, and found that higher 
burnup limits have very little impact on the overall radiological 
consequences. Radiological consequences, as well as other safety 
limits, are evaluated on a cycle-to-cycle basis to confirm that the 
analyses of record remain bounding. If a proposed extended burnup 
core design exceeds bounding safety analysis values, then either the 
core design would be changed, or the safety values would be changed.
    Rod cladding failures are assumed to occur in the fuel handling 
accident; however, the consequences of this event are independent of 
the properties of the fuel rod cladding. This is based on the fuel 
handling event assuming the rupture of all fuel rods regardless of 
the rod cladding material.
    No change is proposed to the established safety analysis fuel 
assembly inputs, specifically fuel assemblies are still limited to a 
maximum 1500 effective full power day (EFPD) burnup and the reactor 
core average maximum burnup will remain at 1000 EFPD burnup ensuring 
the present accident analyses remain bounding. Based on above 
discussion, the proposed revision to extend the burnup limit of M5 
fuel rod cladding material will not significantly increase the 
consequences of an accident and the potential for the release of 
radioactive material to the environment.
    Removing revision numbers, dates, and parenthetical information 
from the listed TRs has no impact on the actual analytical methods 
used to determine the core operating limits, nor does the change 
have impact on the calculations performed for the current or future 
reloads. This change is administrative in nature. This change has no 
impact on plant equipment operation nor does it affect the 
likelihood 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.
    Mark-BW fuel design with M5 alloy has been demonstrated to have 
similar characteristics to that of the Mark-B fuel design. Extended 
burnup of the M5 material has not been shown to alter the functions 
of the rod cladding, which is to provide a barrier against the 
release of radioactive material. Initial plant conditions, which are 
considered in the accident analysis, will also be maintained such 
that no new plant conditions will exist that could affect the 
analysis results. Since plant functions and conditions are not 
impacted by the proposed revision and the higher burnup limit of the 
Mark-BW fuel design with M5 alloy material is not postulated to 
become an accident initiator based on the similarity with Mark-B 
fuel design and Zircaloy-4 material, the possibility of a new or 
different kind of accident is not created.
    The proposed changes will not alter the plant configuration or 
require any new or unusual operator actions. They do not alter the 
way any structure, system, or component functions and do not alter 
the manner in which the plant is operated. These changes do not 
introduce any new failure modes.
    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 margin of safety is established by the acceptance criteria 
used by NRC. Meeting the acceptance criteria assures that the 
consequences of accidents are within known and acceptable limits. 
The emergency core cooling system (ECCS) acceptance criteria are not 
exceeded. Testing has been performed on M5 alloy with respect to 
criteria for peak cladding temperature (PCT) and maximum cladding 
oxidation. These tests demonstrate that M5 alloy rod cladding 
remains within PCT of 2200 degrees Fahrenheit and conservatively 
bounded by the 17 percent limit for maximum cladding oxidation. M5 
alloy oxidation rates are lower than that of Zircaloy at 
temperatures less than 2200 degrees Fahrenheit and have similar 
rates for temperatures up to about 2300 degrees Fahrenheit. High-
temperature oxidation rates of M5 alloy remain equivalent to 
Zircaloy and, as such, respond as hydrogen generators to the same 
extent. Core geometry for amenable cooling is not directly related 
to rod cladding material; however, it applies equally well to all 
materials. The consequences of both thermal and mechanical 
deformation of fuel assemblies have been assessed, and the resultant 
deformations have been shown to maintain coolable core 
configurations. The ECCS is evaluated against the thermal power 
immediately after shutdown. The thermal power is largely a function 
of short-lived fission products which tend to saturate at relatively 
low burnup limits and are not appreciably affected by extended 
burnup. Therefore, with no system changes being proposed; long-term 
cooling is maintained. Additionally, the fuel storage cooling system 
is capable of supporting the long-term storage of the extended 
burnup fuel assemblies' decay heat.
    The changes to burnup limit have been evaluated against 
Departure from Nucleate Boiling (DNB) events and all applicable 
acceptance criteria are met. In addition, the proposed revision to 
allow an increase in the burnup limit of the Mark-BW fuel design 
with M5 alloy will not impact plant setpoints that maintain the 
margin of safety. Based on these results, it is concluded that the 
margin of safety is not significantly reduced. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.
    Removing revision numbers, dates, and parenthetical information 
from the listed TRs will not reduce a margin of safety because this 
information has no effect on any safety analysis assumption nor does 
it revise any setpoints assumed in the analysis of record. The 
proposed change is consistent with NUREG-1431, issued by the NRC 
staff, revising the TSs to reflect the approved level of detail, 
which indicates that there is no 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: Michael L. Marshall, Jr.

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

    Date of amendment request: May 25, 2006 (TSC 06-02).
    Description of amendment request: The proposed amendment would 
revise Section 6.2.1.6 of the Sequoyah Nuclear Plant (SQN) Updated 
Final Safety Analysis Report (UFSAR). This change would revise the 
methodology used for containment sump debris transport analysis and 
affects SQN's current design and licensing basis described in Section 
6.2.1.6 of the SQN UFSAR.
    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 design function of the sump during accident conditions is to 
support emergency core cooling systems (ECCS) and containment spray 
system operation for

[[Page 35461]]

recirculation. The sump is a passive feature that does not act as an 
accident initiator, (i.e., failure of the sump would not initiate a 
design basis accident).
    The proposed change to the UFSAR regarding debris transport 
analysis provides an overall improvement in the analysis for 
recirculation operation and does not change the consequences of 
accidents previously evaluated. The change in methodology is neutral 
with regard to probability. Consequently, the changes associated 
with the enclosed license amendment do not affect the frequency of 
occurrence for accidents previously evaluated in the UFSAR.
    Accident dose as previously evaluated in the UFSAR is unaffected 
by the proposed license amendment.
    Based on the above discussion, 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 sump is a passive component and is not an accident 
initiator; i.e., failure of the sump will not initiate a design 
basis accident. The sump transport methodology is used to confirm 
the ability of the sump to perform all safety functions during 
normal and accident conditions. Consequently, this activity does not 
create a possibility of a new or different type of accident than any 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The changes addressed in TVA's proposed amendment are associated 
with methodology for debris transport to the containment sump.
    The change does not affect specific safety limits, design 
limits, set points, or other critical parameters. The transport 
methodology is used to confirm that the ECCS and containment spray 
systems will perform their safety functions for all accident 
conditions within existing equipment performance capability margins.
    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: Michael L. Marshall, Jr.

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

    Date of amendment request: May 9, 2006.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) 1.1, ``Definitions,'' and 3.4.16, 
``RCS [reactor coolant system] Specific Activity.'' The revisions would 
replace the current Limiting Condition for Operation (LCO) 3.4.16 limit 
on RCS gross specific activity with limits on RCS Dose Equivalent I-131 
and Dose Equivalent Xe-133 (DEX). The conditions and required actions 
for LCO 3.4.16 not being met, and surveillance requirements for LCO 
3.4.16, are being revised. The modes of applicability for LCO 3.4.16 
would be extended. The current definition of [Emacr]--Average 
Disintegration Energy in TS 1.1 would be replaced by the definition of 
DEX. In addition, the current definition of Dose Equivalent I-131 in TS 
1.1 would be revised to allow alternate, NRC-approved thyroid dose 
conversion factors.
    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 change[s] involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes would add new thyroid dose conversion 
factor reference[s] to the definition of DOSE EQUIVALENT I-131, 
eliminate the definition of [Emacr]--AVERAGE DISINTEGRATION ENERGY, 
add a new definition of DOSE EQUIVALENT XE-133, replace the 
Technical Specification (TS) 3.4.16 limit on reactor coolant system 
(RCS) gross specific activity with a limit on noble gas specific 
activity in the form of a Limiting Condition for Operation (LCO) on 
DOSE EQUIVALENT XE-133, increase the Completion Time for Required 
Action B.1, replace TS Figure 3.4.16-1 with a maximum limit on DOSE 
EQUIVALENT I-131, extend the Applicability of LCO 3.4.16, and make 
corresponding changes to TS 3.4.16 to reflect all of the above. The 
proposed changes are not accident initiators and have no impact on 
the probability of occurrence of any design basis accidents.
    The proposed changes will have no impact on the consequences of 
a design basis accident because they will limit the RCS noble gas 
specific activity to be consistent with the values assumed in the 
radiological consequence analyses. The changes will also limit the 
potential RCS [radio]iodine concentration excursion to the value 
currently associated with full power operation, which is more 
restrictive on plant operation than the existing allowable RCS 
[radio]iodine specific activity at lower power levels.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. [Do] the proposed change[s] create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not alter any physical part of the plant 
nor do they affect any plant operating parameters besides the 
allowable specific activity in the RCS. The changes which impact the 
allowable specific activity in the RCS are consistent with the 
assumptions assumed in the current radiological consequence 
analyses. [The proposed changes are also not accident initiators.]
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. [Do] the proposed change[s] involve a significant reduction 
in a margin of safety?
    Response: No.
    The acceptance criteria related to the proposed changes involve 
the allowable control room and offsite radiological consequences 
following a design basis accident. The proposed changes will have no 
impact on the radiological consequences of a design basis accident 
because they will limit the RCS noble gas specific activity to be 
consistent with the values assumed in the radiological consequence 
analyses. The changes will also limit the potential RCS 
[radio]iodine specific activity excursion to the value currently 
associated with full power operation, which is more restrictive on 
plant operation than the existing allowable RCS [radio]iodine 
specific activity at lower power levels.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: David Terao.

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

    Date of amendment request: May 11, 2006.
    Description of amendment request: The proposed amendment would 
revise Surveillance Requirements 3.7.2.1, 3.7.3.1, and 3.7.3.3 on 
verifying the

[[Page 35462]]

closure time of the main steam isolation valves (MSIVs), main feedwater 
regulating valves (MFRVs), main feedwater regulating valve bypass 
valves (MFRVBVs), and main feedwater isolation valves (MFIVs) in the 
Technical Specifications (TSs). These valves are the Main Steam and 
Main Feedwater System isolation valves. The revisions would replace (1) 
the specified maximum acceptable valve closure time for the MSIVs, 
MFRVs, and MFRVBVs, and (2) TS Figure 3.7.3-1, which shows acceptable 
valve closure times for the MFIVs, by the reference to the valve 
closure time, is verified to be ``within limits.'' The maximum 
acceptable valve closure times for the MFRVs and MFRVBVs, and TS Figure 
3.7.3-1 will be relocated to the TS Bases. The maximum acceptable valve 
closure time for the MSIV is already in the TS Bases.
    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.
    Because the proposed change[s remove] specific isolation times 
from the TS and [relocate] the specific values to the TS Bases, 
there are no design or physical changes to the facility or to the 
Main Steam and Main Feedwater System isolation valves themselves. 
The design and functional performance requirements, operational 
characteristics, and reliability of these components remain 
unchanged. There is[,] therefore[,] no impact on the design safety 
function of the valves to close (as an accident mitigator), nor is 
there any change with respect to inadvertent closure (as a potential 
transient initiator). Since no failure mode or initiating condition 
that could cause an accident (including any plant transient) 
evaluated per the FSAR [Final Safety Analysis Report]-described 
safety analyses is created or affected, the change cannot involve a 
significant increase in the probability of an accident previously 
evaluated. The probability of an accident is not affected. The Main 
Steam and Main Feedwater System isolation valves are assumed to 
function to mitigate some accidents (for example, SLB [steam line 
break] and FWLB [main feedwater line break]). The proposed change[s] 
only [affect] the level of detail included in the TS. The TS 
requirements continue to provide the same level of assurance as 
before that the Main Steam and Main Feedwater System isolation 
valves are capable of performing their intended safety function. 
These isolation valves will continue to be verified operable in the 
same manner as before. As such, the proposed change[s do] not affect 
the ability of the isolation valves to perform their assumed 
mitigation function.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change[s] only [affect] the level of detail 
included in the TS. The TS requirements [are not being changed and 
they will] continue to provide the same level of assurance as before 
that the Main Steam and Main Feedwater System isolation valves are 
capable of performing their intended safety function. The Main Steam 
and Main Feedwater System isolation valves will continue to be 
verified operable in the same manner. As such, the proposed change[s 
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 change[s] 
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, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. [Do] the proposed change[s] involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change[s do] not reduce the margin of safety. The 
proposed change[s] only [affect] the level of detail included in the 
TS. The TS requirements [are not being changed and will] continue to 
provide the same level of assurance as before that the Main Steam 
and Main Feedwater System isolation valves will continue to be 
verified operable in the same manner as before. As such, the 
proposed change[s do] not affect the assumptions of any accident 
analysis or the availability or operability of any plant equipment.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: David Terao.

Previously Published Notices of Consideration of Issuance of Amendments 
to Facility Operating Licenses, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: February 14, 2006.
    Brief description of amendment request: The proposed amendments 
would add a requirement to the Title 10 of the Code of Federal 
Regulations, (10 CFR) part 50 license to restrict the minimum cooling 
time and burnup of spent fuel assemblies that will be placed into 
storage in the NUHOMS HD spent fuel dry storage system at Surry 
starting in the summer of 2006.
    Date of publication of individual notice in Federal Register: May 
16, 2006 (71 FR 28390).
    Expiration date of individual notice: 30 day expiration date, June 
15, 2006, and 60 day expiration date, July 17, 2006.

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

[[Page 35463]]

Hazards Consideration Determination, and Opportunity for A Hearing in 
connection with these actions was published in the Federal Register as 
indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) The 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, 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 1, DeWitt County, Illinois

    Date of application for amendment: April 26, 2004, as supplemented 
April 18 and October 11, 2005, and May 19, 2006.
    Brief description of amendment: The amendment revised Technical 
Specification 3.8.7, ``Inverters--Operating'' to change the completion 
time for restoration of an inoperable Division 1 or 2 inverter from the 
current 24 hours to 7 days.
    Date of issuance: May 26, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of the date of issuance.
    Amendment No.: 174.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: June 8, 2004 (69 FR 
32072). The supplements dated April 18 and October 11, 2005, and May 
19, 2006, provided additional information that clarified the 
application, but did not expand the scope of the application as 
originally noticed, and did not change the NRC staff's original 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 26, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 3, 2005, as supplemented 
by letter dated March 7, 2006.
    Brief description of amendments: The amendments revise the Updated 
Final Safety Analysis Report (UFSAR) to incorporate the description of 
the approved changes associated with the plant modifications made to 
the diesel generator cooling water system for each emergency diesel 
generator as described in the amendment application of June 3, 2005, as 
supplemented by letter dated March 7, 2006.
    Date of issuance: May 25, 2006.
    Effective date: As of the date of issuance to be implemented within 
90 days from the date of issuance.
    Amendment Nos.: Unit 1-160, Unit 2--160, Unit 3 -160.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revise the Operating Licenses and the UFSAR for all three 
units.
    Date of initial notice in Federal Register: July 5, 2005 (70 FR 
38715). The March 7, 2006, supplemental letter provided additional 
clarifying information, did not expand the scope of the application as 
originally noticed, and did not change the NRC staff's original 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 25, 2006.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of application of amendments: August 20, 2004, supplemented 
January 31, 2006.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 3.3.8, ``Post Accident Monitoring (PAM) 
Instrumentation,'' to eliminate TS requirements associated with the 
reactor building spray flow instruments commensurate with the 
importance of their post-accident function.
    Date of Issuance: June 1, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 350/352/351.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Licenses and the Technical Specifications.
    Date of initial notice in Federal Register: September 28, 2004 (69 
FR 57983).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 1, 2006.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
2, Pope County, Arkansas

    Date of amendment request: September 19, 2005.
    Brief description of amendment: The proposed changes would revise 
Technical Specification (TS) 3.1.1.5, ``Minimum Temperature for 
Criticality.'' The request proposes to change the current Limiting 
Condition for Operation (LCO) for TS 3.1.1.5 by raising the minimum 
temperature for criticality from the current value of >= 525 [deg]F to 
>= 540 [deg]F; to change the current Action statement for LCO 3.1.1.5 
to reflect this change; and to delete the current statement in 
Surveillance Requirement 4.1.1.5 and replace the statement with wording 
consistent with NUREG-1432, ``Standard Technical Specifications 
Combustion Engineering Plants.'' Also, changes will be made to the ANO-
2 TS Bases in accordance with the Technical Specifications (TS) Bases 
Control Program (ANO-2 TS 6.5.14).
    Date of issuance: May 30, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 264.
    Renewed Facility Operating License No. NPF-6: The amendment revised 
the Technical Specifications and Surveillance Requirements.
    Date of initial notice in Federal Register: December 6, 2005, (70 
FR 72672).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 30, 2006.
    No significant hazards consideration comments received: No.

[[Page 35464]]

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Miami-Dade County, Florida

    Date of application for amendments: January 20, 2005, as 
supplemented July 5, 2005.
    Brief description of amendments: The amendments revised several 
Technical Specifications (TSs) using six TS Task Force (TSTF) generic 
changes. The six TSTFs (nos. 5, 93, 258, 299, 308, and 361) delete 
redundant safety limit violation notification requirements; extend the 
pressurizer heater surveillance frequency from 92 days to 18 months; 
remove redundant requirements and add other requirements to the 
Administrative Controls section of the TSs; clarify the requirements 
regarding the frequency of testing for cumulative and projected dose 
contributions from radioactive effluents; and add a note to the 
residual heat removal requirements during Mode 6 low water level 
operations that allows one required residual heat removal (RHR) loop to 
be inoperable for up to 2 hours for surveillance testing provided the 
other RHR loop is operable and in operation.
    The amendments represent partial approval of the January 20, 2005, 
application for the proposed amendments. The Commission has granted the 
request of Florida Power and Light Company (the licensee) to withdraw 
portions of its January 20, 2005, application for the proposed 
amendment. The application also included TSTF-95, which would extend 
the completion time for reducing the Power Range High trip setpoint 
from 8 to 72 hours and TSTF-101, which would change the auxiliary 
feedwater pump test frequency to be consistent with the inservice test 
program frequency. However, by letter dated March 22, 2005, the 
licensee withdrew the request to adopt TSTF-95 and by letter dated 
October 13, 2005, the licensee withdrew the request to adopt TSTF-101.
    Date of issuance: May 26, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos: 229 and 225.
    Renewed Facility Operating License Nos. DPR-31 and DPR-41: 
Amendments revised the TSs.
    Date of initial notice in Federal Register: March 15, 2005 (70 FR 
12747). The supplement dated July 5, 2005, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 26, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 29, 2005, as supplemented 
by letters dated April 25 (two letters), May 4, and May 12, 2006.
    Brief description of amendment: The amendment converts the current 
Technical Specifications (CTSs) to the Improved Technical 
Specifications (ITSs) format and relocates certain requirements to 
other licensee-controlled documents. The ITSs are based on NUREG-1433, 
``Standard Technical Specifications General Electric Plants BWR/4,'' 
Revision 3, dated June 2004; the Commission's Final Policy Statement, 
``NRC Final Policy Statement on Technical Specification Improvements 
for Nuclear Power Reactors,'' dated July 22, 1993 (58 FR 39132); and 10 
CFR 50.36, ``Technical specifications.'' The purpose of the conversion 
is to provide clearer and more readily understandable requirements in 
the TSs for MNGP to ensure safer operation of the unit. In addition, 
the amendment includes a number of issues that are considered beyond 
the scope of NUREG-1433.
    Date of issuance: June 5, 2006.
    Effective date: As of the date of issuance and shall be implemented 
by September 30, 2006.
    Amendment No: 146.
    Facility Operating License No. DPR-22: Amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: November 16, 2005 (70 
FR 70889).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 5, 2006.
    No significant hazards consideration comments received: No.
    Amendment No: 146.
    Facility Operating License No. DPR-22: Amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: November 16, 2005 (70 
FR 70889).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 5, 2006.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California

    Date of application for amendments: October 19, 2005, as 
supplemented by letter dated December 23, 2005.
    Brief description of amendments: The amendments updated the 
Technical Specification (TS)5.3, ``Unit Staff Qualifications,'' 
operator minimum qualification requirements contained in the March 28, 
1980, NRC letter to all licensees with the more recent NRC-approved 
operator qualification requirements contained in American National 
Standards Institute/American Nuclear Society (ANSI/ANS) 3.1-1993. In 
addition, the changes removed the TS 5.3.1 plant staff retraining and 
replacement training program requirements, which have been superseded 
by requirements contained in 10 CFR 50.120.
    Date of issuance: May 26, 2006.
    Effective date: As of its date of issuance, and shall be 
implemented within 90 days of issuance.
    Amendment Nos.: Unit 1--187 ; Unit 2--189.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 20, 2005 (70 
FR 75495). The December 23, 2005, supplemental letter provided 
additional information that clarified the application, and did not 
expand the scope of the application as originally noticed.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 26, 2006.
    No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket No. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, 
Pennsylvania

    Date of application for amendments: February 28, 2006, as 
supplemented on April 7, 2006.
    Brief description of amendments: The amendments revise the SSES 1 
and 2 Technical Specification (TS) Surveillance Requirements 3.8.4.7 
and 3.8.4.8 to clarify that Diesel Generator ``E'' (DG E) electrical 
power subsystem testing does not require a mode restriction when the DG 
E diesel is not aligned to the Class 1E distribution system.
    Date of issuance: May 30, 2006.

[[Page 35465]]

    Effective date: As of the date of issuance and to be implemented 
within 30 days.
    Amendment Nos.: 235 and 212.
    Facil