Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 38180-38189 [06-5899]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 71, Number 128 (Wednesday, July 5, 2006)]
[Notices]
[Pages 38180-38189]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 06-5899]


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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

[[Page 38181]]

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 June 9, 2006 to June 22, 2006. The last 
biweekly notice was published on June 20, 2006 (71 FR 35456).

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, 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

[[Page 38182]]

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.

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

    Date of amendment request: April 28, 2006.
    Description of amendment request: The proposed amendment would 
revise the Seabrook Station Unit No. 1 (Seabrook) Technical 
Specifications (TSs) consistent with the NRC-approved Revision 9 to 
Technical Specification Task Force (TSTF) Standard Technical 
Specification Change Traveler, TSTF-359, ``Increased Flexibility in 
MODE Restraints.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 2, 2002 (67 FR 50475), on possible 
amendments adopting TSTF-359, 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 April 4, 2003 (68 FR 16579). The licensee affirmed the applicability 
of the following NSHC determination in its application dated April 28, 
2006.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    Criterion 1-- The proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions [are] not an initiator 
of any accident previously evaluated. Therefore, the probability of 
an accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on required actions as 
allowed by [the] proposed LCO [limiting condition of operation] 
3.0.4 are no different than the consequences of an accident while 
entering and relying on the required actions while starting in a 
condition of applicability of the TS. Therefore, the consequences of 
an accident previously evaluated are not significantly affected by 
this change. The addition of a requirement to assess and manage the 
risk introduced by this change will further minimize possible 
concerns. Therefore, this change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Criterion 2-- The proposed change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    The proposed change does not involve the physical alteration of 
the plant (no new or different type of equipment will be installed). 
Entering into a mode or other specified condition in the 
applicability of a TS, while in a TS condition statement and the 
associated required actions of the TS, will not introduce new 
failure modes or effects and will not, in the absence of other 
unrelated failures, lead to an accident whose consequences exceed 
the consequences of accidents previously evaluated. The addition of 
a requirement to assess and manage the risk introduced by this 
change will further minimize possible concerns. Thus, this change 
does not create the possibility of a new of different kind of 
accident from an accident previously evaluated.
    Criterion 3-- The proposed change does not involve a significant 
reduction in the margin of safety.
    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. The TS 
allow operation of the plant without the full compliment of 
equipment through the conditions for not meeting the TS Limiting 
Conditions for Operation (LCO). The risk associated with this 
allowance is managed by the imposition of required actions that must 
be performed within the prescribed times. The net effect of being in 
a TS condition on the margin of safety is not considered 
significant. The proposed change does not alter the required actions 
or completion times of the TS. The proposed change allows TS 
conditions to be entered, and the associated required actions and 
completion times to be used in new circumstances. This use is 
predicated upon the licensee's performance of a risk assessment and 
the management of plant risk. The change also eliminates current 
allowances for utilizing required actions and completion times in 
similar circumstances, without assessing and managing risk. The new 
change to the margin of safety is insignificant. Therefore, this 
change does not involve a significant reduction in a margin of 
safety.

    Based upon the reasoning presented above 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: M. S. Ross, Florida Power & Light Company, 
P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Darrell J. Roberts.

Indiana Michigan Power Company, Docket No. 50-315, D. C. Cook Nuclear 
Plant, Unit 1, Berrien County, Michigan

    Date of amendment request: May 30, 2006.

[[Page 38183]]

    Description of amendment request: The proposed amendment would 
revise the Technical Specifications, deleting from Surveillance 
Requirement (SR) 3.3.1.15 a note which specifies that the surveillance 
includes ``verification of Reactor Coolant System [RCS] resistance 
temperature detector [RTD] bypass loop flow rate.'' Approval of this 
proposed amendment would permit the licensee to effect a plant design 
change, removing the RTD bypass piping and install a replacement system 
using fast response thermowell-mounted RTDs located in the RCS loop 
piping.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided a 
no significant hazards determination analysis.
    The NRC staff has reviewed the licensee's analysis and performed 
its own as follows:

    (1) Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    No. The RTD bypass system is the passive hardware associated 
with RCS instrumentation with control and indication functions. The 
RTD bypass system was not considered a precursor to any previously 
analyzed accident, and was not considered a factor in the scenario 
leading to accident consequences. The new system replacing the RTD 
bypass system will perform the same control and indication 
functions, and similarly will not be considered a precursor to any 
accident, or a factor affecting accident consequences in previously 
analyzed accident scenarios. Therefore, replacement of the existing 
RTD bypass system with the new system will not increase the 
probability of occurrence of an accident, and will not increase 
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?
    No. The replacement of the existing RTD bypass with the 
replacement system would not create new failure modes, and the 
replacement system is not an initiator of any new or different kind 
of accident. The proposed deletion of the note in SR 3.3.1.15 does 
not affect the interaction of the replacement system with any system 
whose failure or malfunction can initiate an 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?
    No. Margins of safety are established in the design of 
components, the configuration of components to meet certain 
performance parameters, and in the models and associated assumptions 
used to analysis the system's performance. The replacement system 
will continue to perform the same temperature detection function to 
the same level of reliability as defined in the D.C. Cook Updated 
Safety Analysis Report. Therefore, the proposed amendment does not 
involve a significant reduction in a margin of safety.

    The NRC staff evaluated the licensee's analysis, and based on this 
evaluation, the NRC staff proposes to determine that the requested 
amendment does not involve a significant hazards consideration.
    Attorney for licensee: James M. Petro, Jr., Esquire, One Cook 
Place, Bridgman, MI 49106.
    NRC Branch Chief: L. Raghavan.

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

    Date of amendment request: May 26, 2006.
    Description of amendment request: The licensee proposed to amend 
each unit's Technical Specifications in accordance with Revision 4 to 
Technical Specifications Task Force (TSTF) Standard TS Change 
Traveller, TSTF-449, ``Steam Generator Tube Integrity'' (see 70 FR 
24126). Specifically, the following Sections will be revised per TSTF-
449: Section 1.1, Definitions; Section 3.4.13, Reactor Coolant System 
Operational LEAKAGE; Section 5.5.7, Steam Generator (SG) Program; and 
Section 5.6.7, Steam Generator Tube Inspection Report. Also, a new 
Section 3.4.17, SG Tube Integrity, will be added. The proposed changes 
are necessary in order to implement the guidance for the industry 
initiative in Nuclear Energy Institute (NEI) 97-06, Steam Generator 
Program Guidelines.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, by referencing the NRC staff's model analysis published 
in 70 FR 10298 (March 2, 2005). The NRC staff's model analysis is 
reproduced 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 [main steam line 
break], 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.

[[Page 38184]]

    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 requests involve no significant hazards consideration.
    Attorney for licensee: James M. Petro, Jr., Esquire, One Cook 
Place, Bridgman, MI 49106.
    NRC Branch Chief: L. Raghavan.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: March 7, 2006, and as supplemented by 
letter dated May 10, 2006.
    Description of amendment request: The proposed changes would revise 
Technical Specification (TS) Section 5.5.6, ``Inservice Testing 
Program,'' by replacing references to Section XI of the American 
Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code 
with ASME Code for Operation and Maintenance of Nuclear Power Plants 
(OM Code). Section 50.55a of Title 10 of the Code of Federal 
Regulations (10 CFR) requires that the Inservice Testing (IST) Program 
be updated to the latest Edition and Addenda of the ASME OM Code 
incorporated by reference in 10 CFR 50.55a(b) 12 months before the 
start of the 10-year interval. Section XI of the ASME Boiler and 
Pressure Vessel Code has been replaced with the ASME OM Code as the 
code of reference for IST programs. Thus, the ASME OM Code is the code 
of reference for the IST Program for the next 10-year interval that 
began March 1, 2006. In addition, the scope of frequencies specified to 
be within the applicability of Surveillance Requirement (SR) 3.0.2 is 
expanded by adding mention of other normal and accelerated frequencies 
specified in the IST Program. This will eliminate any confusion 
regarding the applicability of SR 3.0.2 to IST Program Frequencies.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes revise the CNS [Cooper Nuclear Station] TS 
for the IST Program to be consistent with the requirements of 10 CFR 
50.55a(f)(4) for pumps and valves which are classified as ASME Code 
Class 1, Class 2, and Class 3. 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 impact any accident initiators, 
analyzed events, or assumed mitigation of accident or transient 
events. They do not involve addition or removal of any equipment, 
nor any design changes to the facility.
    Based on the above, NPPD [Nebraska Public Power District] 
concludes that the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes revise the CNS TS for the IST Program to be 
consistent with the requirements of 10 CFR 50.55a(f)(4) for pumps 
and valves which are classified as ASME Code Class 1, Class 2, and 
Class 3. 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 a change in the methods governing normal plant 
operation. The proposed changes will not introduce a new accident 
initiator, accident precursor, or malfunction mechanism. 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.
    Based on the above NPPD concludes that these proposed changes do 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes revise the CNS TS for the IST Program to be 
consistent with the requirements of 10 CFR 50.55a(f)(4) for pumps 
and valves which are classified as ASME Code Class 1, Class 2, and 
Class 3. 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. Based on the above, NPPD concludes that 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: Mr. John C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Branch Chief: David Terao.

[[Page 38185]]

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

    Date of amendment request: April 25, 2006.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) requirements for mode change 
limitations in TSs 3.0.4 and 4.0.4, using the CLIIP described in the 
Nuclear Regulatory Commission (NRC) approved Technical Specification 
Task Force (TSTF) change, TSTF-359, Revision 9.
    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 change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change allows entry into a MODE while relying on 
ACTIONS. Being in an ACTION is not an initiator of any accident 
previously evaluated. Consequently, the probability of an accident 
previously evaluated is not significantly increased. The 
consequences of an accident while relying on ACTIONS as allowed by 
the proposed LCO [limiting condition of operation] 3.0.4 are no 
different than the consequences of an accident while relying on 
ACTIONS for other reasons, such as equipment inoperability. 
Therefore, the consequences of an accident previously evaluated are 
not significantly increased by this change. Therefore, this change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Thus, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated; 
there is no change to the design basis.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change allows entry into a MODE or other specified 
conditions in the Applicability while relying on ACTIONS. The 
Technical Specifications allow operation of the plant without a full 
complement of equipment. The risk associated with this allowance is 
managed by the imposition of ACTIONS and Completion Times. The net 
effect of ACTIONS and Completion Times on the margin of safety is 
not considered significant. The proposed change does not change the 
ACTIONS or Completion Times of the Technical Specifications. The 
proposed change allows the ACTIONS and Completion Times to be used 
in new circumstances. However, this use is predicated on an 
assessment that focuses on managing plant risk. In addition, most 
current allowances to utilize the ACTIONS and Completion Times that 
do not require risk assessment are eliminated. As a result, the net 
change to the margin of safety is insignificant. Therefore, this 
change does not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit-N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Darrell J. Roberts.

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: February 28, 2006.
    This revised amendment request completely supercedes the licensee's 
submittal of December 17, 2004. Likewise, the biweekly Federal Register 
(FR) notice--notice of consideration of issuance of amendments to 
facility operating licenses, proposed no significant hazards 
consideration determination, and opportunity for a hearing, which was 
published in the FR on January 18, 2005 (70 FR 2897) is being 
superceded by the publication of this biweekly FR notice.
    Description of amendment requests: The proposed amendment revises 
Technical Specifications (TSs) 3.8.1, ``AC [alternating current] 
Sources--Operating,'' 3.8.4, ``DC [direct current] Sources--
Operating,'' 3.8.5, ``DC Sources--Shutdown,'' 3.8.6, ``Battery Cell 
Parameters,'' 3.8.7, ``Inverters--Operating,'' and 3.8.9, 
``Distribution Systems--Operating.'' This change will also add a new 
Battery Monitoring and Maintenance Program, Section 5.5.2.16.
    The proposed TS changes will provide operational flexibility 
supported by DC electrical subsystem design upgrades that are in 
progress. These upgrades will provide increased capacity batteries, 
additional battery chargers, and the means to cross-connect DC 
subsystems while meeting all design battery loading requirements. With 
these modifications in place, it will be feasible to perform routine 
surveillances as well as battery replacements online.
    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. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to Technical Specifications (TS) 3.8.4 and 
3.8.6 would allow extension of the Completion Time (CT) for 
inoperable Direct Current (DC) distribution subsystems to manually 
cross-connect DC distribution buses of the same safety train of the 
operating unit for a period of 30 days. Currently the CT only allows 
for 2 hours to ascertain the source of the problem before a 
controlled shutdown is initiated. Loss of a DC subsystem is not an 
initiator of an event. However, complete loss of a Train A 
(subsystems A and C) or Train B (subsystems B and D) DC system would 
initiate a plant transient/plant trip.
    Operation of a DC Train in cross-connected configuration does 
not affect the quality of DC control and motive power to any system. 
Therefore, allowing the cross-connect of DC distribution systems 
does not significantly increase the probability of an accident 
previously evaluated in Chapter 15 of the Updated Final Safety 
Analysis Report (UFSAR).
    The above conclusion is supported by Probabilistic Risk 
Assessment (PRA) evaluation which encompasses all accidents, 
including UFSAR Chapter 15.
    Modification to the Frequency for Surveillance Requirements in 
TS 3.8.4, 3.8.5, and 3.8.6 are consistent with previously described 
recommendations. Enhancements from TSTF-360, Rev. 1 and IEEE 450-
2002 have been incorporated into Limiting Conditions for Operation 
(LCOs) 3.8.4, 3.8.5, and 3.8.6. These changes do not impact the 
probability or consequences of an accident previously evaluated.
    Further changes are made of an editorial nature or provide 
clarification only. For example, discussions regarding electrical 
`Trains' and `Subsystems' will be in more conventional terminology. 
LCOs affected by editorial changes include 3.8.1, 3.8.4, 3.8.5, 
3.8.6, 3.8.7, and 3.8.9.
    The changes being proposed in the TS do not affect assumptions 
contained in other safety analyses or the physical design of the 
plant, nor do they affect other Technical Specifications that 
preserve safety analysis assumptions.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant increase in the 
probability or consequences of an accident previously analyzed.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of new or different kind of 
accident from any accident previously evaluated?
    Response: No.
    The proposed change modifies surveillances and LCOs for 
batteries and chargers to meet the requirements of IEEE

[[Page 38186]]

450-2002 whose intent is to maintain the same equipment capability 
as previously assumed in our commitment to IEEE 450-1980.
    The proposed change will allow the cross-tie of DC subsystems 
and allow extension of the CT for an inoperable subsystem to 30 
days. Failure of the crosstied DC buses and/or associated 
battery(ies) is bounded by existing evaluations for the failure of 
an entire electrical train.
    Swing battery chargers are added to increase the overall DC 
system reliability. Administrative and mechanical controls will be 
in place to ensure the design and operation of the DC systems 
continue to meet the UFSAR design basis.
    LCOs 3.8.1, 3.8.4, 3.8.5, 3.8.6, 3.8.7, and 3.8.9 revisions are 
editorial clarifications and do not affect plant design.
    Therefore, operation of the facility in accordance with this 
proposed change will not create the possibility of new or different 
kind of accident from any accident previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    Changes in accordance with IEEE 450-2002 and TSTF-360, Rev. 1 
maintain the same level of equipment performance stated in the UFSAR 
and the current Technical Specifications.
    Swing battery chargers are added to increase the overall DC 
system reliability. Administrative and mechanical controls will be 
in place to ensure the design and operation of the DC systems 
continue to meet the UFSAR design basis.
    The addition of the DC cross-tie capability proposed for LCO 
3.8.4 has been evaluated, as described previously, using PRA and 
determined to be of acceptable risk as long as the duration while 
cross-tied is limited to 30 days. An LCO has been included as part 
of this proposed change to ensure that plant operation, with DC 
buses cross-tied, will not exceed 30 days.
    All remaining changes are editorial.
    Therefore, operation of the facility in accordance with the 
proposed amendment would 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: David Terao.

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 51.12(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.

Duke Power Company LLC, 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: March 1, 2006, supplemented 
April 26, 2006.
    Brief description of amendments: The amendments revised the 
Technical Specifications to reconcile the criticality requirements of 
Title 10 of the Code of Federal Regulations (10 CFR), Part 50, and 10 
CFR part 72 for loading and unloading dry spent fuel pool canisters in 
the spent fuel pool.
    Date of Issuance: June 15, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 351/353/352.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Licenses and Technical Specifications.
    Date of initial notice in Federal Register: April 11, 2006 (71 FR 
18373).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 15, 2006.
    The supplement dated April 26, 2006, provided clarifying 
information that did not change the scope of the original application 
and the initial proposed no significant hazards consideration 
determination.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 22, 2005.
    Brief description of amendment: The amendments revise the 
surveillance requirements (SRs) for Technical Specification 3.3.5, 
``Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation.'' 
Specifically, a note was added to IP2 SR 3.3.5.2 to indicate that the 
verification of the setpoint is not required for the 480 volt (V) bus 
degraded voltage function when performing the trip actuating device 
operational test. A similar note was added to IP3 SR 3.3.5.1 for the 
480 V degraded voltage and undervoltage functions.
    Date of issuance: June 7, 2006.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 247 and 231.
    Facility Operating License Nos. DPR-26 and DPR-64: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 7, 2005 (70 FR 
33213).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 7, 2006.
    No significant hazards consideration comments received: No.

[[Page 38187]]

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

    Date of application for amendments: April 30, 2004, as supplemented 
by letters dated December 17, 2004; June 30, 2004; July 5, 2005; 
September 30, 2005; and June 1, 2006.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 3.3.1.3, ``Oscillation Power Range Monitor (OPRM) 
Instrumentation''; TS 3.4.1, ``Recirculation Loops Operating''; and TS 
5.6.5, ``Core Operating Limits Report (COLR)''; to insert a new TS 
section for the ORPM instrumentation, delete the current thermal-
hydraulic instability administrative requirements, and add the 
appropriate references for the OPRM trip setpoints and methodology.
    Date of issuance: June 13, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 150 days.
    Amendment Nos.: 177/163.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications and License.
    Date of initial notice in Federal Register: June 8, 2004 (69 FR 
32073).
    The December 17, 2004; June 30, 2004; July 5, 2005; September 30, 
2005; and June 1, 2006, supplements contained clarifying information 
and did not change the NRC staff's initial proposed finding of no 
significant hazards consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 13, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: February 25, 2005.
    Brief description of amendments: The amendments deleted the 
sections of the Facility Operating Licenses that require reporting of 
violations of the requirements in Sections 2.C and 2.E of the Facility 
Operating Licenses.
    Date of issuance: June 14, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 178/164.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the License.
    Date of initial notice in Federal Register: April 26, 2005 (70 FR 
21456).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 14, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: April 21, 2005.
    Brief description of amendments: The amendments revised Technical 
Specifications (TSs) to adopt seven TS Task Force (TSTF) generic 
changes (TSTF nos. 5, 65, 101, 258, 299, 308, and 361) that delete 
redundant safety limit violation notification requirements; adopt use 
of generic titles for utility positions; change the auxiliary feedwater 
pump test requirements to be consistent with the inservice test 
program; remove redundant requirements and add other requirements to 
Section 5.0, Administrative Controls; clarify the meaning of 
``refueling cycle'' for system integrated leak test intervals in the 
Primary Coolant Sources Outside Containment program; 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 (RHR) requirements during Mode 6 low water 
level operations that allows one required RHR loop to be inoperable for 
up to 2 hours for surveillance testing provided the other RHR loop is 
operable and in operation.
    Date of issuance: June 19, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos: 199 and 146.
    Renewed Facility Operating License Nos. DPR-67 and NPF-16: 
Amendments revised the TSs.
    Date of initial notice in Federal Register: July 5, 2005 (70 FR 
38720).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 19, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 12, 2004, as 
supplemented by letters dated January 30 and March 6, 2006.
    Brief description of amendments: The amendments revise Technical 
Specification 5.5.7, ``Inservice Testing Program'' to update the 
references to the American Society of Mechanical Engineers Code and 
certain associated periodicities for inservice testing activities, 
consistent with the requirements of 10 CFR 50.55a.
    Date of issuance: June 8, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 222 and 228.
    Renewed Facility Operating License Nos. DPR-24 and DPR-27: 
Amendments revise the Technical Specifications.
    Date of initial notice in Federal Register: January 17, 2006 (71 FR 
2592).
    The January 30, 2006, supplement withdrew a portion of the original 
request and the March 6, 2006, supplement contained clarifying 
information.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 8, 2006.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement Or Emergency 
Circumstances)

    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 for the 
amendment 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.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination

[[Page 38188]]

of no significant hazards consideration. The Commission has provided a 
reasonable opportunity for the public to comment, using its best 
efforts to make available to the public means of communication for the 
public to respond quickly, and in the case of telephone comments, the 
comments have been recorded or transcribed as appropriate and the 
licensee has been informed of the public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective 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 
application for amendment, (2) the amendment to Facility Operating 
License, 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 System's (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.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. 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, and electronically on the Internet at the NRC Web site, 
https://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are 
problems in accessing the document, contact the PDR Reference staff at 
1 (800) 397-4209, (301) 415-4737, or by e-mail to pdr@nrc.gov. If a 
request for a hearing or petition for leave to intervene is filed by 
the above date, 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 identify 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 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to those specific sources and 
documents of which the petitioner is aware and on which the petitioner 
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.\1\ 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 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.
---------------------------------------------------------------------------

    \1\ To the extent that the applications contain attachments and 
supporting documents that are not publicly available because they 
are asserted to contain safeguards or proprietary information, 
petitioners desiring access to this information should contact the 
applicant or applicant's counsel and discuss the need for a 
protective order.
---------------------------------------------------------------------------

    Each contention shall be given a separate numeric or alpha 
designation within one of the following groups:
    1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the 
applications.
    2. Environmental--primarily concerns/issues relating to matters 
discussed or referenced in the environmental analysis for the 
applications.

[[Page 38189]]

    3. Miscellaneous--does not fall into one of the categories outlined 
above.
    As specified in 10 CFR 2.309, if two or more petitioners/requestors 
seek to co-sponsor a contention, the petitioners/requestors shall 
jointly designate a representative who shall have the authority to act 
for the petitioners/requestors with respect to that contention. If a 
petitioner/requestor seeks to adopt the contention of another 
sponsoring petitioner/requestor, the petitioner/requestor who seeks to 
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that 
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the 
petitioners/requestors with respect to that contention.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and