[Federal Register Volume 71, Number 186 (Tuesday, September 26, 2006)]
[Notices]
[Pages 56189-56199]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 06-8014]
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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 September 1, 2006, to September 14, 2006.
The last biweekly notice was published on September 12, 2006 (71 FR
53715).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
[[Page 56190]]
Written comments may be submitted by mail to the Chief, Rulemaking,
Directives and Editing Branch, Division of Administrative Services,
Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the Commission's
Public Document Room (PDR), located at One White Flint North, Public
File Area O1F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. The filing of requests for a hearing and petitions for leave
to intervene is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
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
[email protected]. 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, http://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 [email protected].
[[Page 56191]]
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendments request: July 20, 2006.
Description of amendments request: The proposed amendments would
revise Technical Specification 3.1.6, ``Shutdown Control Element
Assembly (CEA) Insertion Limits.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Safety analyses require that the shutdown CEAs insert into the
core at least 90% within 4 seconds of the safety signal initiating
the shutdown sequence with the assumption that the shutdown CEAs'
starting positions are at 150 inches withdrawn. This assumption will
not be altered with the new proposed withdrawal limit.
The positioning of control rods (shutdown CEAs) to a new limit
of >=147.75 inches withdrawn is not a precursor to any accident
analyzed at Palo Verde nor do these conditions affect any accident
precursor; thus, initial control rod position does not change the
probability of an accident previously evaluated.
To assess the effect control rod position would have on the
safety analyses with the rods positioned at the new limit, several
events and specific parameters were analyzed. The events were chosen
because of their sensitivity to rod position. The specific
parameters were analyzed to determine if, with the rods positioned
at the new limit, the power distribution in the core was still
within the assumptions made in the safety analyses.
Since none of the related safety analyses resulted in a
significant change in the previously calculated values and the
limiting parameters associated for those analyses were not exceeded,
the consequences of these accidents remain unchanged. Therefore, the
new insertion limit for the shutdown CEAs will not increase the
consequences of any accident analyzed in our licensing bases
documents.
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 accident from any accident previously evaluated?
Response: No.
PVNGS [Palo Verde Nuclear Generating Station] licensing bases
documents describe the design function of the control rods as
components that include a positive means (gravity) for inserting the
control rods and are capable of reliably controlling the nuclear
reactor to assure that under conditions of normal operation,
including anticipated accidents, fuel design limits are not
exceeded. The proposed amendment, new control rod (shutdown CEA)
insertion limit, does not create the possibility of a new or
different kind of accident from any accident previously evaluated
nor does it affect the control rods ability to perform its design
function.
Control rods placed at the new insertion limit will not cause
fuel design limits to be exceeded during normal operations or
accidents. Placing the control rods at the new insertion limit in no
way impedes their insertion due to gravity. These CEAs are tested to
ensure that they will insert greater than 90% into the core in less
than 4 seconds from a completely withdrawn position (150 inches) and
this requirement will continue to be met.
Establishing a new insertion limit for the control rods does not
modify any of the existing components or systems used to position
the control rods. The new insertion limit will also satisfy the
assumptions made in the safety analyses.
In conclusion, the new insertion limit stills [sic] allows the
control rods to fulfill their design function and does not create a
new or different accident than is already described in the licensing
bases documents. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed amendment, new shutdown CEA insertion limit, does
not involve a reduction in the margin of safety. The new shutdown
CEA insertion limit does not affect any of the limits used to
determine the acceptability of newly designed cores. The safety
analyses in the licensing bases documents remain acceptable when
this new (more restrictive) shutdown CEA insertion limit is applied.
Additionally, the design basis of the control rods is unaffected by
the new insertion limit. The design function of the control rods is
to provide a positive means (gravity) for inserting the control rods
and is capable of reliably controlling the nuclear reactor to assure
that under conditions of normal operation, including anticipated
accidents, fuel design limits are not exceeded. Since the bounding
safety analyses limits used remain the same and the control rod
design basis is unaffected, the fuel design limits associated with
the clad material; which houses the fuel; and the design limits of
the coolant system; which houses the fuel assemblies; remain
unchanged. Therefore, the margin of safety is not reduced.
In conclusion, since the bounding limits used for safety
analyses are unaffected by the new shutdown CEA insertion limit, the
safety limits associated with the fuel and the coolant system remain
unchanged. The design basis on the control rods is to ensure the
fuel safety limits are not exceeded and since they remain unchanged,
the design basis is still achieved. Therefore, there is no reduction
in the margin of safety.
Therefore, APS [Arizona Public Service] has concluded that the
proposed license amendment request does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: Janet S. Mueller, Director, Law Department,
Arizona Public Service Company, P.O. Box 52034, Mail Station 8695,
Phoenix, Arizona 85072-2034.
NRC Branch Chief: David Terao.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: August 2, 2006.
Description of amendment request: The proposed change would delete
Waterford 3 Technical Specification Surveillance Requirement (SR)
4.6.1.7.2. This SR is the augmented testing requirement for containment
purge supply and exhaust isolation valves with resilient seal materials
and allows the surveillance intervals to be set in accordance with the
Containment Leakage Rate Testing Program.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This change deletes the augmented testing requirement for these
containment isolation valves and allows the surveillance intervals
to be set in accordance with the Containment Leakage Rate Testing
Program. This change does not affect the system function or design.
The purge valves are not an initiator of any previously analyzed
accident. Leakage rates do not affect the probability of the
occurrence of any accident. Operating history has demonstrated that
the valves do not degrade and cause leakage as previously
anticipated. Because these valves have been demonstrated to be
reliable, these valves can be expected to perform the containment
isolation function as assumed in the accident analyses. Therefore,
there is no significant increase in the consequences of any
previously evaluated accident.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
[[Page 56192]]
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Extending the test intervals has no influence on, nor does it
contribute in any way to, the possibility of a new or different kind
of accident or malfunction from those previously analyzed. No change
has been made to the design, function or method of performing
leakage testing. Leakage acceptance criteria have not changed. No
new accident modes are created by extending the testing intervals.
No safety-related equipment or safety functions are altered as a
result of this change.
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 only margin of safety that has the potential of being
impacted by the proposed change involves the offsite dose
consequences of postulated accidents which are directly related to
the containment leakage rate. The proposed change does not alter the
method of performing the tests nor does it change the leakage
acceptance criteria. Sufficient data has been collected to
demonstrate these resilient seals do not degrade at an accelerated
rate.
Because of this demonstrated reliability, this change will
provide sufficient surveillance to determine an increase in the
unfiltered leakage prior to the leakage exceeding that assumed in
the accident analysis.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn, 1700
K Street NW., Washington, DC 20006-3817.
NRC Branch Chief: David Terao.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2 (NMP2), Oswego County, New York
Date of amendment request: May 11, 2006.
Description of amendment request: The proposed change would revise
Technical Specification (TS) 3.1.7, ``Standby Liquid Control (SLC)
System,'' to change the minimum required SLC pump discharge pressure
specified in surveillance requirement (SR) SR 3.1.7.7 from 1235 psig to
1320 psig. This change is in response to Nuclear Regulatory Commission
Information Notice 2001-13, ``Inadequate Standby Liquid Control System
Relief Valve Margin.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the surveillance requirements for
the SLC system to correspond to the maximum expected pressure in the
reactor pressure vessel for an ATWS [anticipated transient without
scram] event. This proposed increase in the specified SLC pump
discharge pressure involves only the SLC system. No other NMP2
structures, systems, or components are affected. The SLC system is
provided to mitigate ATWS events and, as such, is not considered to
be an initiator of an ATWS event or any other analyzed accident. The
revised TS surveillance requirement, and the associated change to
the SLC pump discharge relief valve set pressure (not described in
the TS), neither reduce the ability of the SLC system to respond to
and mitigate an ATWS event nor increase the likelihood of a system
malfunction that could increase the consequences of an accident.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change to the SLC pump TS surveillance requirement,
and the associated change to the SLC pump discharge relief valve set
pressure (not described in the TS), are consistent with the
functional requirements of the ATWS rule (10 CFR 50.62). The
proposed change does not involve the installation of any new or
different type of equipment, does not introduce any new modes of
plant operation, and does not change any methods governing normal
plant operation. The proposed change does not introduce any new
accident initiators, and therefore does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not alter any assumptions, initial
conditions or results from any accident analyses. The proposed
change to the SLC pump TS surveillance requirement, and the
associated changes to the SLC pump discharge relief valve set
pressure (not described in the TS), are consistent with the
functional requirements of the ATWS rule (10 CFR 50.62). The ability
of the SLC system to respond to and mitigate an ATWS event is not
affected. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mark J. Wetterhahn, Esq., Winston & Strawn,
1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Richard J. Laufer.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2 (NMP2), Oswego County, New York
Date of amendment request: August 11, 2006.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.3.2.1, ``Control Rod Block
Instrumentation,'' to revise the number of startups allowed with the
rod worth minimizer (RWM) inoperable from one per calendar year to two
per operating cycle (approximately 2 years).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change redefines the frequency at which plant
startup is permitted without using the RWM. The relevant design
basis accident is the control rod drop accident (CRDA), which
involves multiple failures to initiate the event. This
administrative change does not increase the probability of
occurrence of any of the failures that are necessary for a CRDA to
occur. Use of the RWM or the alternate use of a qualified human
checker to ensure the correct control rod withdrawal sequence is not
in itself an accident initiator, and redefining the startup
allowance frequency does not involve any plant hardware changes or
new operator actions that could serve to initiate a CRDA. The
proposed change will have no adverse effect on plant operation, or
the availability or operation of any accident mitigation equipment.
Also, since the banked position withdrawal sequence (BPWS) will
continue to be enforced by either the RWM or verification by a
second qualified individual, the initial conditions of the CRDA
radiological consequence analysis presented in the U[F]SAR [Updated
Final Safety Analysis Report] are not affected.
[[Page 56193]]
Therefore, there will be no increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not introduce any new modes of plant
operation and will not result in a change to the design function or
operation of any structure, system, or component that is used for
accident mitigation. The proposed redefinition of the frequency at
which plant startup is permitted without using the RWM does not
result in any credible new failure mechanisms, malfunctions, or
accident initiators not considered in the design and licensing
basis. This administrative change does not affect the ability of
safety-related systems and components to perform their intended
safety functions. Therefore, the proposed change will not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change redefines the frequency at which plant
startup is permitted without using the RWM. This administrative
change does not affect the overall frequency of use of the
allowance. The proposed change will have no adverse affect on plant
operation or equipment important to safety. The relevant design
basis accident is the control rod drop accident (CRDA), which
involves multiple failures to initiate the event. The CRDA analysis
consequences and related initial conditions remain unchanged when
invoking the proposed change. The plant response to the CRDA will
not be affected and the accident mitigation equipment will continue
to function as assumed in the accident analysis. Therefore, there
will be 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: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Richard J. Laufer.
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: July 5, 2006.
Description of amendment request: The proposed change would revise
the main control room (MCR) and emergency switchgear room (ESGR) air
conditioning system (ACS) Technical Specifications (TSs) to reflect the
completion of permanent modifications to the equipment and associated
power supply configuration. The revisions include the addition of
requirements and/or action statements addressing the inoperability of
two or more air handling units (AHUs) on a unit, as well as AHUs
powered from an H emergency bus. The proposed change, paralleling
requirements in the Improved Technical Specifications (ITS), also adds
MCR and ESGR ACS requirements during refueling operations and
irradiated fuel movement in the fuel building. In addition, the
proposed change clarifies the service water (SW) requirements for the
ACS chillers that serve the MCR and ESGRs.
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 license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
The proposed change does not impact the condition or performance
of any plant structure, system, or component. The proposed change
does not affect the initiators of analyzed events or the assumed
mitigation of accident or transient events. No physical changes to
the ACS or SW System are involved, and accident operation of the ACS
will not change. As a result, the proposed change to the Surry
Technical Specifications does not involve any significant increase
in the probability or the consequences of any accident or
malfunction of equipment important to safety previously evaluated
since neither accident probabilities nor consequences are being
affected by this proposed change.
2. Does the proposed license amendment 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 or a change in the methods used to respond to plant
transients. No new or different equipment is being installed, and no
installed equipment is being removed. There is no alteration to the
parameters with which the plant is normally operated or in the
setpoints, which initiate protective or mitigative actions. The ACS
will continue to perform its required function. Consequently, no new
failure modes are introduced by the proposed change. Therefore, the
proposed change to the Surry Technical Specifications does not
create the possibility of a new or different kind of accident or
malfunction of equipment important to safety from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
The proposed TS change does not impact any plant structure,
system, or component that is relied upon for accident mitigation.
Margin of safety is established through the design of the plant
structures, systems, and components, the parameters within which the
plant is operated, and the establishment of the setpoints for the
actuation of equipment relied upon to respond to an event. Since ACS
performance is not affected by the proposed change, the ACS will
continue to be available to perform its required function.
Furthermore, the change does not affect the condition or performance
of structures, systems, or components relied upon the accident
mitigation or any safety analysis assumptions. Therefore, the
proposed change to the Surry Technical Specifications 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: Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Branch Chief: Evangelos C. Marinos.
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.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of amendment request: July 20, 2006.
Brief description of amendment request: The proposed amendment
would revise the Vogtle Electric Generating Plant (VEGP), Units 1 and
2, Technical Specifications (TS) 5.5.9,
[[Page 56194]]
``Steam Generator (SG) Tube Surveillance Program,'' to incorporate
changes in the SG inspection scope for VEGP, Unit 1 during Refueling
Outage 13 and the subsequent operating cycle, and VEGP Unit 2 during
Refueling Outage 12 and the subsequent operating cycle. The proposed
changes modify the inspection requirements for portions of SG tubes
within the tubesheet region of the SGs.
Date of publication of individual notice in Federal Register: July
31, 2006 (71 FR 43225).
Expiration date of individual notice: 30-day August 30, 2006; 60-
day, September 29, 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 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, http://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 [email protected].
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: October 18, 2005, as
supplemented by letter dated May 26, 2006.
Brief description of amendment: The amendment revised the Oyster
Creek Nuclear Generating Station Technical Specifications (TSs)
Surveillance Requirement (SR) 4.4.B.1 to provide an alternative means
for testing the electromatic relief valves located on the main steam
system. The revised SR allows demonstration of the capability of the
valves to perform their function without requiring that the valves be
cycled with steam pressure while installed.
Date of Issuance: September 1, 2006.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 260.
Facility Operating License No. DPR-16: The amendment revised the
TSs.
Date of initial notice in Federal Register: December 20, 2005 (70
FR 75490). The May 26, 2006, letter provided clarifying information
within the scope of the original application and did not change the
staff's initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated September 1, 2006.
No significant hazards consideration comments received: No.
Duke Power Company LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: October 27, 2004.
Brief description of amendments: The amendments revised the
facility operating licenses by removal of Section 2.E, that lists
reporting requirements with regard to Maximum Power Level, Updated,
Fire Protection, Protection of the Environment (Unit 2 only) and
Physical Protection.
Date of issuance: September 7, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 233 and 215.
Renewed Facility Operating License Nos. NPF 9 and NPF-17:
Amendments revised the licenses.
Date of initial notice in Federal Register: July 5, 2005 (70 FR
38717).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 7, 2006.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: April 27, 2005, as supplemented
by letters dated November 22, 2005, and August 1, 2006. The August 1,
2006, submittal reduced the scope of the changes to only revise
Technical Specification Limiting Condition for Operation 3.8.4, ``DC
Sources-Operating.''
Brief description of amendment: The amendment revises the Technical
Specifications to allow a battery charger to be out of service for up
to 7 days.
Date of issuance: September 14, 2006.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 286.
Facility Operating License No. DPR-59: The amendment revised the
License and the Technical Specifications.
Date of initial notice in Federal Register: July 19, 2005 (70 FR
41444). The November 22, 2005, and August 1, 2006, supplements provided
additional information that clarified the application, 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 as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 14, 2006.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2
Will County, Illinois
Date of application for amendment: February 15, 2005, as
supplemented by
[[Page 56195]]
letters dated November 28 and December 9, 2005 (two letters), and
January 27, February 13, March 17 and July 14, 2006.
Brief description of amendment: The amendments fully implement an
alternative source term.
Date of issuance: September 8, 2006.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment Nos.: 147, 147, 140 and 140.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications and License.
Date of initial notice in Federal Register: May 10, 2005 (70 FR
24650). The November 28 and December 9, 2005 (two letters), and January
27, February 13, March 17 and July 14, 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 September 8, 2006.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Exelon Generation Company, LLC and MidAmerican Energy Company, Docket
Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and
2, Rock Island County, Illinois
Date of application for amendments: October 10, 2002, as
supplemented by letters dated March 21, March 28, August 4, September
15 and October 31, 2003, and June 30, August 6, September 3, September
10, September 22, November 2 and November 5, 2004, and March 3, August
22, September 3 and September 27, 2005, and February 17 and May 25,
2006.
Brief description of amendments: The amendments adopt the
alternative source term methodology as prescribed in Title 10 to the
Code of Federal Regulations Section 50.67.
Date of issuance: September 11, 2006.
Effective date: As of the date of issuance and shall be implemented
within 180 days.
Amendment Nos.: 221/212, 233/229.
Renewed Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and
DPR-30. The amendments revised the Technical Specifications,
Surveillance Requirements and Licenses.
Date of initial notice in Federal Register: August 19, 2003 (68 FR
49816). The supplements dated March 21, March 28, August 4, September
15 and October 31, 2003, and June 30, August 6, September 3, September
10, September 22, November 2, and November 5, 2004, and March 3, August
22, September 3 and September 27, 2005, and February 17 and May 25,
2006, 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 September 11, 2006.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook
Nuclear Plant, Unit 1, Berrien County, Michigan
Date of application for amendment: April 10, 2006, as supplemented
by letters dated April 12, 13 (2 letters), and June 27, 2006.
Brief description of amendment: The amendment revised Surveillance
Requirement 3.8.1.11 of the DCCNP-1 Technical Specifications, raising
the diesel generator load rejection voltage test limit from 5000 volts
to 5350 volts.
Date of issuance: September 1, 2006.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment No.: 295.
Facility Operating License No. DPR-58: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 1, 2006 (71 FR
43534). The supplemental letters contained clarifying information and
did not change the initial no significant hazards consideration
determination, and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 1, 2006.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: September 29, 2005, as supplemented by
letters dated January 16, and April 7, 2006.
Brief description of amendment: The amendment eliminated
operability requirements for secondary containment, secondary
containment isolation valves, the standby gas treatment system, and
secondary containment isolation instrumentation when handling
irradiated fuel that has decayed for 24 hours since critical reactor
operations, and when performing core alterations. Similar technical
specification relaxations are granted for the Control Room Emergency
Filter System and its initiation instrumentation after a decay period
of 7 days.
Date of issuance: September 5, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 222.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 3, 2006 (71 FR
149). The supplements dated January 16 and April 17, 2006, 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 amendment is contained
in a Safety Evaluation dated September 5, 2006.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: March 7, 2006, as supplemented by letter
dated May 10, 2006.
Brief description of amendment: The amendment revised Technical
Specification (TS) Section 5.5.6, ``Inservice Testing Program,'' by
replacing references to Section Xl 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 (CFR)
requires that the Inservice Testing (IST) Program be updated to the
latest Edition and Addenda of the Code incorporated by reference in 10
CFR 50.55a(b) 12 months before the start of the applicable 10-year
interval. Section Xl 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 10-year interval that began March 1, 2006. In addition,
the amendment expanded the scope of frequencies specified to be within
the applicability
[[Page 56196]]
of Surveillance Requirement (SR) 3.0.2 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.
Date of issuance: September 6, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 223.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 5, 2006 (71 FR
38184).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 6, 2006.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: July 1, 2005, as supplemented on
September 16, 2005, November 15, 2005, December 14, 2005, February 16,
2006, and July 6, 2006.
Brief description of amendment: The amendment revises the Updated
Safety Analysis Report, Section 14.10, ``Malfunctions of the Feedwater
System,'' to describe an existing Emergency Operating Procedure
operator action to isolate the steam generator blowdown within 15
minutes of a reactor trip during a loss-of-main feedwater event.
Date of issuance: September 11, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 242.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Updated Safety Analysis Report.
Date of initial notice in Federal Register: August 2, 2005 (70 FR
44403). The September 16, 2005, November 15, 2005, December 14, 2005,
February 16, 2006, and July 6, 2006, supplemental letters provided
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.
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated September 11, 2006.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of application for amendments: September 26, 2005, as
supplemented by letter dated June 28, 2006.
Brief description of amendments: The proposed amendments revised
the Salem Technical Specifications (TSs) to eliminate certain
Surveillance Requirements (SRs) for containment isolation valves. The
changes deleted SR 4.6.3.1.1 and SR 4.6.3.1 for Salem Unit Nos. 1 and
2, respectively. These SRs require a complete valve stroke and stroke
time measurement when a valve is returned to service after maintenance,
repair, or replacement work. The changes are intended to minimize
unnecessary testing and plant transients. Other Salem TS containment
isolation valve SRs ensure that the valves remain operable.
Date of issuance: August 31, 2006.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment Nos.: 274 and 255.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the License and Technical Specifications.
Date of initial notice in Federal Register: July 18, 2006 (71 FR
40739).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 31, 2006.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of application for amendments: January 27, 2005, as
supplemented by letters dated September 30, 2005, and January 25 and
May 5, 2006.
Brief description of amendments: The amendments revised the
Technical Specifications by extending the surveillance test interval
for components of the reactor protection system.
Date of issuance: September 1, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 145 and 125
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the licenses and the technical specifications.
Date of initial notice in Federal Register: November 8, 2005 (70 FR
67751). The supplements dated September 30, 2005, and January 25 and
May 5, 2006, provided clarifying information that did not change the
scope of the January 27, 2005, application nor the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 1, 2006.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of application for amendments: July 20, 2006, as supplemented
by letter dated August 4, 2006.
Brief description of amendments: The amendments revised Technical
Specification 5.5.9, ``Steam Generator (SG) Tube Surveillance
Program,'' regarding the required SG inspection scope for Vogtle, Unit
1, during Refueling Outage 13 and the subsequent operating cycle and
Vogtle, Unit 2, during Refueling Outage 12, and the subsequent
operating cycle. The proposed changes modify the inspection
requirements for portions of the SG tubes within the hot leg tubesheet
region of the SGs.
Date of issuance: September 12, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 146 and 126.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the licenses and the technical specifications.
Date of initial notice in Federal Register: July 31, 2006 (71 FR
43225). The supplement dated August 4, 2006, provided clarifying
information that did not expand the scope of the July 20, 2006,
application nor the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 12, 2006.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama
Date of application for amendments: January 10, 2006 as
supplemented by letters April 14, August 1, September 5 and 14, 2006.
[[Page 56197]]
Description of amendment request: The amendments revised Technical
Specifications 3.3.1.1 and 3.3.5.1 to specify the methodology used for
determining, setting, and evaluating as-found setpoints for drift-
susceptible instruments that are necessary to ensure compliance with a
Safety Limit or are critical in ensuring the fuel peak cladding
temperature acceptance criterion are met.
Date of issuance: September 14, 2006.
Effective date: Date of issuance, to be implemented within 90 days.
Amendment Nos.: 257, 296 and 254.
Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: March 28, 2006 (71 FR
15487). The supplements dated April 14, August 1, September 5 and 14,
2006, provided additional information that clarified the application,
did not expand the scope of the application as originally noticed, and
did not change the NRC staff's original proposed no significant hazards
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 14, 2006.
No significant hazards consideration comments received: No.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendments request: December 16, 2005, as supplemented by
letter dated June 7, 2006.
Brief description of amendments: The amendments revised the steam
generator tube surveillance program technical specifications (TSs) to
be consistent with TS Task Force (TSTF) traveler TSTF-449, Revision 4,
``Steam Generator Tube Integrity.''
Date of issuance: September 12, 2006.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: 128/128.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 14, 2006 (71 FR
13181). The supplement dated June 7, 2006, 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 September 12, 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 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, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems
in accessing the documents located in ADAMS, contact the PDR Reference
staff at 1 (800) 397-4209,
[[Page 56198]]
(301) 415-4737 or by e-mail to [email protected].
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,
http://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 [email protected]. 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.
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 have the opportunity to participate fully in the conduct of the
hearing. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
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, [email protected];
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
[email protected]. 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 or 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).
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1 (FCS), Washington County, Nebraska
Date of amendment request: June 2, 2006.
Description of amendment request: The amendment deleted Technical
[[Page 56199]]
Specifications (TSs) 4.3.1.2b and TS 4.3.1.2c of the FCS TSs. The
amendment also made an administrative change to TS 4.3.1.2 to correct
the current wording of TS 4.3.1.2 and TS 4.3.1.2d. TS 4.3.1.2 implied
that more than one new fuel storage rack at FCS is installed when there
is actually only one new fuel storage rack. In addition, Omaha Public
Power District (OPPD) will complete additional procedural enhancements
of administrative controls for compliance with 10 CFR 50.68(b)(2) and
(b)(3) prior to receipt of new fuel for the 2006 Refueling.
Date of issuance: June 27, 2006.
Effective date: The license amendment is effective as of its date
of issuance and shall be implemented within 7 days of issuance. OPPD
will complete additional enhancements of administrative controls for
compliance with 10 CFR 50.68(b)(2) and (b)(3) prior to receipt of new
fuel for the 2006 Refueling.
Amendment No.: 240.
Renewed Facility Operating License No. DPR-40: Amendment revised
the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC):
Yes. Omaha World-Herald on June 11, 2006. The notice provided an
opportunity to submit comments on the Commission's proposed NSHC
determination. No comments have been received.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, state consultation, and final NSHC determination
are contained in a safety evaluation dated August 31, 2006.
Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn,
1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: David Terao.
Dated at Rockville, Maryland, this 18th Day of September 2006.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 06-8014 Filed 9-25-06; 8:45 am]
BILLING CODE 7590-01-P