[Federal Register Volume 71, Number 89 (Tuesday, May 9, 2006)]
[Notices]
[Pages 26995-27010]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 06-4243]
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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
[[Page 26996]]
determination by the Commission that such amendment involves no
significant hazards consideration, notwithstanding the pendency before
the Commission of a request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from April 14, 2006 to April 27, 2006. The last
biweekly notice was published on April 25, 2006 (71 FR 23952).
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, 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
[[Page 26997]]
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].
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2 New London County, Connecticut
Date of amendment request: January 26, 2006.
Description of amendment request: The proposed amendment would
update the list of Nuclear Regulatory Commission-approved documents
specified in the Technical Specifications that describe the analytical
methods used to determine the core operating limits.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment adds a new document (No. 16) to TS
6.9.1.8 b to complement the list of documents used to determine the
core operating limits. These documents have been previously reviewed
and approved by the NRC. It also changes the word ``minimum'' to
``maximum'' in TS 5.3.1 to correctly state the limit on nominal
average enrichment of reload fuel. This change restores TS 5.3.1
wording to the wording previously approved by the NRC in Amendment
274. The proposed changes do not modify any plant equipment and do
not impact any failure modes that could lead to an accident.
Additionally, the proposed changes have no effect on the consequence
of any analyzed accident since the changes do not affect the
function of any equipment credited for accident mitigation. Based on
this discussion, the proposed amendment does not increase the
probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes do not modify any plant equipment and there
is no impact on the capability of existing equipment to perform its
intended functions. No system setpoints are being modified and no
changes are being made to the method in which plant operations are
conducted. No new failure modes are introduced by the proposed
change. The proposed amendment does not introduce accident
initiators or malfunctions that would cause a new or different kind
of accident. Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment adds a new document (No. 16) to TS
6.9.1.8 b to complement the list of documents used to determine the
core operating limits. These documents have been previously reviewed
and approved by the NRC. It also changes the word ``minimum'' to
``maximum'' in TS 5.3.1 to correctly state the limit on nominal
average enrichment of reload fuel. This change restores TS 5.3.1
wording to the wording previously approved by the NRC in Amendment
274. The proposed changes have no impact on plant equipment
operation. The proposed changes do not revise any setpoints nor do
they change the acceptance criteria used in the accident analyses.
Therefore, the proposed changes will not result in a 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, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT
06385.
NRC Branch Chief: Darrell J. Roberts.
Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power
Station, Unit No. 3 New London County, Connecticut
Date of amendment request: March 28, 2006.
Description of amendment request: The proposed amendment would
delete the license condition, Section 2.F of Facility Operating License
No. NPF-49, which requires reporting of violations of the requirements
in Section 2.C of Facility Operating License No. NPF-49. The change is
consistent with the notice published in the Federal Register on
November 4, 2005, as part of the consolidated line item improvement
process.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
[[Page 26998]]
The proposed change involves the deletion of a reporting
requirement. The change does not affect plant equipment or operating
practices and therefore does not significantly increase 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?
Response: No.
The proposed change is administrative in that it deletes a
reporting requirement. The change does not add new plant equipment,
change existing plant equipment, or affect the operating practices
of the facility. Therefore, the change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change deletes a reporting requirement. The change
does not affect plant equipment or operating practices and therefore
does not involve a significant reduction in a margin of safety.
Based on the above, the NRC staff proposes that the change presents
no significant hazards consideration under the standards set forth in
10 CFR 50.92(c).
Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT
06385.
NRC Branch Chief: Darrell J. Roberts.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: June 15, 2005.
Description of amendment request: The proposed amendments would
revise the Technical Specifications to eliminate the out of date
requirements associated with the completion of the Keowee Refurbishment
modifications on both Keowee Hydro Units.
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. Involve a significant increase in the probability or
consequences of an accident previously evaluated:
The proposed change to the Oconee Technical Specification (TS)
3.8.1 removes out of date requirements associated with temporary
extensions to Required Action (RA) Completion Times (CTs) that are
no longer applicable because of the completion of the Keowee
Refurbishment modifications on both KHUs. The proposed change also
removes a Facility Operating License (FOL) License Condition that is
no longer needed since the associated TS change is no longer
applicable. As such, the proposed change is administrative. No
actual plant equipment, operating practices, or accident analyses
are affected by this change. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any kind of accident previously evaluated:
The proposed change to the Oconee TSs and FOLs removes
requirements associated with a temporary extension of TS 3.8.1 RA
CTs that are no longer applicable because of the completion of the
Keowee Refurbishment modifications on both KHUs. As such, the
proposed changes are administrative. No actual plant equipment,
operating practices, or accident analyses are affected by this
change. No new accident causal mechanisms are created as a result of
this change. The proposed change does not impact any plant systems
that are accident initiators; neither does it adversely impact any
accident mitigating systems. Therefore, this change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change does not adversely affect any plant safety
limits, set points, or design parameters. The change also does not
adversely affect the fuel, fuel cladding, Reactor Coolant System, or
containment integrity. The proposed change eliminates requirements
that are no longer applicable and is administrative in nature.
Therefore, the proposed change does not involve a 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: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Branch Chief: Evangelos C. Marinos.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: April 17, 2006.
Description of amendment request: The proposed change allows a
delay time for entering a supported system technical specification (TS)
when the inoperability is due solely to an inoperable snubber, if risk
is assessed and managed consistent with the program in place for
complying with the requirements of paragraph 50.65(a)(4) of Title 10 of
the Code of Federal Regulations (10 CFR). Limiting Condition for
Operation (LCO) 3.0.8 is added to the TS to provide this allowance and
define the requirements and limitations for its use.
This change was proposed by the industry's Technical Specification
Task Force (TSTF) and is designated TSTF-372, Revision 4. The NRC staff
issued a notice of opportunity for comment in the Federal Register on
November 24, 2004 (69 FR 68412), on possible amendments concerning
TSTF-372, 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 May 4, 2005 (70 FR 23252). The
licensee affirmed the applicability of the following NSHC determination
in its application dated April 17, 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 a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber if risk is assessed and managed. The postulated seismic
event requiring snubbers is a low-probability occurrence and the
overall TS system safety function would still be available for the
vast majority of anticipated challenges. Therefore, the probability
of an accident previously evaluated is not significantly increased,
if at all. The consequences of an accident while relying on
allowance provided by proposed LCO 3.0.8 are no different than the
consequences of an accident while relying on the TS required actions
in effect without the allowance provided by proposed LCO 3.0.8.
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 a physical alteration of
the plant (no new or different type of equipment will be installed).
Allowing delay times for entering supported system TS when
inoperability is due solely to inoperable snubbers, if risk is
assessed and
[[Page 26999]]
managed, 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
or 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 a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber, if risk is assessed and managed. The postulated seismic
event requiring snubbers is a low-probability occurrence and the
overall TS system safety function would still be available for the
vast majority of anticipated challenges. The risk impact of the
proposed TS changes was assessed following the three-tiered approach
recommended in RG [Regulatory Guide] 1.177. A bounding risk
assessment was performed to justify the proposed TS changes. [The
proposed LCO 3.0.8 defines limitations on the use of the provision
and includes a requirement for the licensee to assess and manage the
risk associated with operation with an inoperable snubber.] 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 proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: William A. Horin, Esq., Winston & Strawn,
1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: David Terao.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
2 (ANO-2), Pope County, Arkansas
Date of amendment request: March 20, 2006.
Description of amendment request: The proposed change removes
Arkansas Nuclear One, Unit 2 reactor coolant system (RCS) structural
integrity requirements contained in Technical Specification (TS)
3.4.10.1. The proposed change is consistent with NUREG-1432, ``Standard
Technical Specifications--Combustion Engineering Plants,'' Revision
3.1.
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 to remove the RCS structural integrity
controls from the TSs does not impact any mitigation equipment or
the ability of the RCS pressure boundary to fulfill any required
safety function. Since no accident mitigation or initiators are
impacted by this change, no design basis accidents are affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated?
Response: No.
The proposed change will not alter the plant configuration or
change the manner in which the plant is operated. No new failure
modes are being introduced by the proposed 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
the margin of safety?
Response: No.
Removal of TS 3.4.10.1 from the TSs does not reduce the controls
that are required to maintain the RCS pressure boundary for ASME
Code [American Society of Mechanical Engineers' Boiler and Pressure
Vessel Code] Class 1, 2, or 3 components. No equipment or RCS safety
margins are impacted due to the proposed change.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: David Terao.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of amendment request: January 27, 2006.
Description of amendment request: The proposed amendment involves
changes to Technical Specifications Section 3/4 9.1, ``Boron
Concentration,'' Section 3/4 9.14, ``Spent Fuel Storage,'' and Section
3/4 5.5.1, ``Fuel Storage Criticality.'' The proposed license amendment
removes reliance on Boraflex as a neutron absorber in Turkey Point
Units 3 and 4 spent fuel pool storage racks. To preclude continued loss
of reactivity margin due to the ongoing degradation of Boraflex, the
neutron absorbing function currently performed by Boraflex will be
replaced by some combination of rod cluster control assemblies, Metamic
rack inserts, and administrative controls that require mixing higher
reactivity fuel with lower-reactivity fuel.
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. Would operation of the facility in accordance with the
proposed amendment involve a significant increase in the probability
or consequences of an accident previously evaluated?
No. Operation in accordance with proposed amendment does not
involve a significant increase in the probability or consequences of
an accident previously evaluated. The proposed amendments do not
change or modify the fuel, fuel handling processes, spent fuel
storage racks, number of fuel assemblies that may be stored in the
spent fuel pool (SFP), decay heat generation rate, or the spent fuel
pool cooling and cleanup system. The proposed amendment was
evaluated for impact on the following previously evaluated events
and accidents:
a. A fuel handling accident (FHA),
b. A cask drop accident,
c. A fuel mispositioning event,
d. A spent fuel pool boron dilution event,
e. A seismic event, and
f. A loss of spent fuel pool cooling event.
The probability of a FHA is not significantly increased because
implementation of the proposed amendment will employ the same
equipment and process to handle fuel assemblies that is currently
used. Also, tests have confirmed that the Metamic inserts can be
installed and removed without damaging the host fuel assemblies. The
FHA radiological consequences are not increased because the
radiological source term of a single fuel assembly will remain
unchanged. Therefore, the proposed amendments do not significantly
increase the probability or consequences of a FHA.
The proposed amendments do not increase the probability of
dropping a fuel transfer cask because they do not introduce any new
heavy loads to the SFP and do not affect heavy load handling
processes. Also, the insertion of Metamic rack inserts does not
increase the consequences of the cask drop accident because the
radiological source term of that accident is developed from a non-
mechanistically derived quantity of damaged fuel stored in the spent
fuel pool. Therefore, the proposed amendments do not significantly
increase the probability or consequences of a cask drop accident.
Operation in accordance with the proposed amendment will not
change the probability of a fuel mispositioning event because fuel
movement will continue to be controlled by approved fuel handling
procedures. These procedures continue to require identification
[[Page 27000]]
of the initial and target locations for each fuel assembly that is
moved. The consequences of a fuel mispositioning event are not
changed because the reactivity analysis demonstrates that the same
subcriticality criteria and requirements continue to be met for the
worst-case fuel mispositioning event.
Operation in accordance with the proposed amendment will not
change the probability of a boron dilution event because the systems
and events that could affect spent fuel soluble boron are unchanged.
The consequences of a boron dilution event are unchanged because the
proposed amendment reduces the soluble boron requirement below the
currently required value and the maximum possible water volume
displaced by the inserts is an insignificant fraction of the total
spent fuel pool water volume.
Operation in accordance with the proposed amendment will not
change the probability of a seismic event, which is an Act of God.
The consequences of a seismic event are not significantly increased
because the forcing functions for seismic excitation are not
increased and because the mass of storage racks with Metamic inserts
is not appreciably increased. Seismic analyses demonstrate adequate
stress levels in the storage racks when inserts are installed.
Operation in accordance with the proposed amendment will not
change the probability of a loss of SFP cooling event because the
systems and events that could affect SFP cooling are unchanged. The
consequences are not significantly increased because there are no
changes in the SFP heat load or SFP cooling systems, structures or
components. Furthermore, conservative analyses indicate that the
current design requirements and criteria continue to be met with the
Metamic inserts installed.
Based on the above, it is concluded that the proposed amendments
do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Would operation of the facility in accordance with the
proposed amendment create the possibility of a new or different kind
of accident from any accident previously evaluated?
No. Operation in accordance with the proposed amendments do not
create the possibility of a new or different kind of accident from
any accident previously evaluated. The proposed amendments do not
change or modify the fuel, fuel handling processes, spent fuel
racks, number of fuel assemblies that may be stored in the pool,
decay heat generation rate, or the spent fuel pool cooling and
cleanup system. The effects of operating with the proposed amendment
are listed below. The proposed amendments were evaluated for the
potential of each effect to create the possibility of a new or
different kind of accident:
a. Addition of inserts to the spent fuel storage racks,
b. New storage patterns,
c. Additional weight from the inserts,
d. Insert movement above spent fuel, and
e. Displacement of fuel pool water by the inserts.
Each insert will be placed between a fuel assembly and the
storage cell wall, taking up some of the space available on two
sides of the fuel assembly. Tests confirm that the insert can be
installed and removed without damaging the fuel assembly. Analyses
demonstrate that the presence of the inserts does not adversely
affect spent fuel cooling, seismic capability, or subcriticality.
The aluminum (alloy 6061) and boron carbide materials of
construction have been shown to be compatible with nuclear fuel,
storage racks and spent fuel pool environments, and generate no
adverse material interactions. Therefore, placing the inserts into
the spent fuelpool storage racks can not cause a new or different
kind of accident.
Operation with the proposed fuel storage patterns will not
create a new or different kind of accident because fuel movement
will continue to be controlled by approved fuel handling procedures.
These procedures continue to require identification of the initial
and target locations for each fuel assembly that is moved. There are
no changes in the criteria or design requirements pertaining to
spent fuel safety, including subcriticality requirements, and
analyses demonstrate that the proposed storage patterns meet these
requirements and criteria with adequate margins. Therefore, the
proposed storage patterns can not cause a new or different kind of
accident.
Operation with the added weight of the Metamic inserts will not
create a new or different accident. The net effect of the adding the
maximum number of inserts is to add less than one percent to the
weight of the loaded racks. Furthermore, the analyses of the racks
with Metamic inserts installed demonstrate that the stress levels in
the rack modules continue to be considerably less than allowable
stress limits. Therefore, the added weight from the inserts can not
cause a new or different kind of accident.
Operation with the insert allowed to move above spent fuel will
not create a new or different kind of accident. The insert with its
handling tool weighs considerably less than the weight of a single
fuel assembly. Single fuel assemblies are routinely moved safely
over spent fuel assemblies and the same level of safety in design
and operation will be maintained when moving the inserts.
Furthermore, the effect of a dropped insert to block the top of a
storage cell has been evaluated in thermal-hydraulic analyses.
Therefore, the movement of inserts can not cause a new or different
kind of accident.
Whereas the installed rack inserts will displace a very small
fraction of the fuel pool water volume and impose a very small
reduction in operator response time to previously-evaluated SFP
accidents, the reduction will not promote a new or different kind of
accident. Also, displacement of water along two sides of a stored
fuel assembly may have some local reduction in the peripheral
cooling flow; however, this effect would be small compared to the
flow induced through the fuel assembly and would in no way promote a
new or different kind of accident.
Based on the above, it is concluded that operation with the
proposed amendment does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Would operation of the facility in accordance with the
proposed amendment involve a significant reduction in a margin of
safety?
No. Operation of the facility in accordance with the proposed
amendment does not significantly reduce the margin of safety. The
proposed change was evaluated for its effect on current margins of
safety related to criticality, structural integrity, and spent fuel
heat removal capability. The margin of safety for subcriticality
required by 10 CFR 50.68(b)(4) is unchanged. New criticality
analysis confirms that operation in accordance with the proposed
amendment continues to meet the required subcriticality margins.
Also, the margin of safety for SFP soluble boron concentration is
actually increased because new analyses require less soluble boron
than is currently required, and much less than the value required by
Technical Specifications. The structural evaluations for the racks
and spent fuel pool with Metamic inserts installed show that the
rack and spent fuel pool are unimpaired by loading combinations
during seismic motion, and there is no adverse seismic-induced
interaction between the rack and Metamic inserts.
The proposed change does not affect spent fuel heat generation
or the spent fuel cooling systems. A conservative analysis indicates
that the design basis requirements and criteria for spent fuel
cooling continue to be met with the Metamic inserts in place, and
displacing coolant. Thermal hydraulic analysis of the local effects
of an installed rack insert blocking peripheral flow show a small
increase in local water and fuel clad temperatures, but will remain
within acceptable limits including no departure from nucleate
boiling.
Based on these evaluations, operating the facility with the
proposed amendment does not involve a significant reduction in any
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Branch Chief: Michael L. Marshall, Jr.
Nuclear Management Company, LLC, Docket No. 50-306, Prairie Island
Nuclear Generating Plant, Unit 2, Goodhue County, Minnesota
Date of amendment request: March 13, 2006.
Description of amendment request: The proposed amendment would
involve revision of the surveillance test load in Technical
Specification (TS) 3.8.1, ``AC Sources--Operating,'' Surveillance
Requirement (SR) 3.8.1.3. This license amendment request proposes to
revise SR 3.8.1.3 to require
[[Page 27001]]
testing D5 and D6 monthly at or above 4000 kW to demonstrate TS
operability. In addition to the TS required testing, NMC will continue
monthly operation at or above 90 percent of the emergency diesel
generator (EDG) rated load to assist in early identification of
degraded EDG capabilities which could prevent performance of their
safety function.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
This license amendment request proposes to reduce the Prairie
Island Nuclear Generating Plant Unit 2 emergency diesel generator's
monthly test loading which demonstrates Technical Specification
operability. The proposed test load will continue to assure that
both Unit 2 emergency diesel generators have the capacity and the
capability to assume the maximum auto-connected loads for Unit 2.
The emergency diesel generators are required to be operable in
the event of a design basis accident coincident with a loss of
offsite power to mitigate the consequences of the accident. They are
also the alternate AC source for a station blackout on the other
Prairie Island Nuclear Generating Plant unit. The emergency diesel
generators are not accident initiators and therefore this change
does not involve a significant increase in the probability of an
accident previously evaluated.
The accident analyses assume that at least one safeguards bus is
provided with power either from the offsite sources or the emergency
diesel generators. The Technical Specification changes proposed in
this license amendment request will continue to assure that both
Unit 2 emergency diesel generators have the capacity and the
capability to assume the maximum auto-connected loads for Unit 2.
Thus, the changes proposed in this license amendment request do not
involve a significant increase in the consequences of an accident
previously evaluated.
The changes proposed in this license amendment do not involve a
significant increase 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.
This license amendment request proposes to reduce the Prairie
Island Nuclear Generating Plant Unit 2 emergency diesel generator's
monthly test loading which demonstrates Technical Specification
operability. The proposed test load will continue to assure that
both Unit 2 emergency diesel generators have the capacity and the
capability to assume the maximum auto-connected loads for Unit 2.
The proposed Technical Specification changes do not involve a
change in the plant design, system operation, or the use of the
emergency diesel generators. The proposed changes allow the
emergency diesel generator to be tested at a reduced load which
envelopes the required safety function loads and continues to
demonstrate the capability and capacity of the emergency diesel
generators to perform their required functions. There are no new
failure modes or mechanisms created due to testing the emergency
diesel generators at the proposed test loading. Testing of the
emergency diesel generators at the proposed test loading does not
involve any modification in the operational limits or physical
design of plant systems. There are no new accident precursors
generated due to the proposed test loading.
The Technical Specification changes proposed in this license
amendment 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.
This license amendment request proposes to reduce the Prairie
Island Nuclear Generating Plant Unit 2 emergency diesel generator's
monthly test loading which demonstrates Technical Specification
operability. The proposed test load will continue to assure that
both Unit 2 emergency diesel generators have the capacity and the
capability to assume the maximum auto-connected loads for Unit 2.
The proposed Technical Specification changes will continue to
demonstrate that the emergency diesel generators meet the Technical
Specification definition of operability, that is, the proposed
testing will demonstrate that the emergency diesel generators will
perform their safety function and the necessary emergency diesel
generator attendant instrumentation, controls, cooling, lubrication
and other auxiliary equipment required for the emergency diesel
generators to perform their safety function loads are also tested at
this loading. The proposed testing will also continue to demonstrate
the capability and capacity of the emergency diesel generators to
supply the required Unit 2 loss of offsite power coincident with
Unit 1 station blackout loads. Since the proposed surveillance
testing will continue to demonstrate operability, and the capability
and capacity to supply their required Unit 2 loss of offsite power
coincident with Unit 1 station blackout loads, the proposed
Technical Specification changes do not involve a significant
reduction in a margin of safety.
The Technical Specification changes proposed in this license
amendment do not involve a significant reduction in a margin of
safety.
Based on the above, the Nuclear Management Company concludes
that the proposed amendment presents no significant hazards
consideration under the standards set forth in 10 CFR 50.92(c) and,
accordingly, a finding of ``no significant hazards consideration''
is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Branch Chief: L. Raghavan.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of amendment request: February 1, 2006.
Description of amendment request: The proposed amendment would
clarify the Technical Specification (TS) testing frequency for the
Surveillance Requirements (SRs) in TS 3.1.4, ``Control Rod Scram
Times.''
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 control rod hydraulic scram insertion system is not an
initiator to any accident sequence analyzed in the Final Safety
Analysis Report (FSAR). The changes do not involve any physical
change to structures, systems, or components (SSCs) and do not alter
the method of operation or control of SSCs. The current assumptions
in the safety analysis regarding accident initiators and mitigation
of accidents (including assumed scram insertion times) are
unaffected by these changes. No additional failure modes or
mechanisms are being introduced and the likelihood of previously
analyzed failures remains unchanged.
Operation in accordance with the proposed Technical
Specification (TS) ensures that the control rods and associated
scram insertion function remain capable of performing the function
as described in the FSAR [Final Safety Analysis Report]. Therefore,
the mitigative scram functions will continue to provide the
protection assumed by the analysis.
Therefore, this 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
[[Page 27002]]
accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant. No new equipment is being introduced, and installed
equipment is not being operated in a new or different manner. There
are no setpoints affected by this change at which protective or
mitigative actions are initiated. This change will not alter the
manner in which equipment operation is initiated, nor will the
functional demands on credited equipment be changed. No alterations
in the procedures that ensure the plant remains within analyzed
limits are being proposed, and no changes are being made to the
procedures relied upon to respond to an off-normal event as
described in the FSAR. As such, no new failure modes are being
introduced. The change does not alter assumptions made in the safety
analysis and licensing basis.
[Therefore, this change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.]
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. Operation in accordance with the proposed TS ensures
that the control rod scram insertion system remains capable of
performing the function as described in the FSAR. Sufficiently rapid
insertion of control rods following certain accidents (scram time)
will prevent fuel damage, and thereby maintain a margin of safety to
fuel damage. No change is being made to the required insertion rate
specified in plant Technical Specifications. Clarifying when control
rod insertion times must be verified following movement of fuel
assemblies, without actually changing the requirement (verification
of insertion times will continue to be required whenever work that
might impact the rod insertion time is done), does not reduce the
margin of safety related to fuel damage.
Therefore, the change does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Branch Chief: Richard J. Laufer.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: October 7, 2005.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to clarify certain
requirements during fuel movement and core alterations. The amendment
would make the TSs consistent with the NRC-approved Revision 2 to
Technical Specification Task Force (TSTF) Standard Technical
Specification Change Traveler, TSTF-51, ``Revise Containment
Requirements During Handling Irradiated Fuel and Core Alterations,''
and NUREG-1433, ``Standard Technical Specifications General Electric
Plants, BWR [boiling water reactor]/4.''
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 analyzed?
Response: No.
The proposed changes would revise Technical Specifications (TS)
3.6.5.3.1, FRVS [filtration, recirculation and ventilation system]
Ventilation System, and 3.6.5.3.2, FRVS Recirculation System, ACTION
b from, ``* * * containment or operations * * * '' to read ``* * *
containment and operations * * * '' to be consistent with NUREG-
1433, ``Standard Technical Specifications General Electric Plants,
BWR/4'' (STS). Technical Specification 3.7.1.2, Service Water, and
3.8.3.2, Distribution--Shutdown, require the addition of
``recently'' to modify irradiated fuel consistent with NRC-approved
Revision 2 to Technical Specification Task Force (TSTF) Standard
Technical Specification Change Traveler, TSTF-51, ``Revise
Containment Requirements During Handling Irradiated Fuel and Core
Alterations.'' Technical Specifications 3.8.1.2, A.C. Sources--
Shutdown, 3.8.2.2, DC Sources--Shutdown, and 3.8.3.2, Distribution--
Shutdown, require that ``CORE ALTERATIONS'' be added to ACTION a.
The proposed changes associated with the fuel handling accident
(FHA) do not involve a change to structures, components, or systems
that would affect the probability of an accident previously
evaluated in the Hope Creek Updated Final Safety Analysis Report
(UFSAR). The FHA for Hope Creek is defined as a drop of a fuel
assembly over irradiated assemblies in the reactor core 24 hours
after reactor shutdown. 10 CFR 50.67, ``Accident Source Term''
(AST), was used to evaluate the dose consequences of a postulated
accident. The FHA has been analyzed without credit for Secondary
Containment; Filtration, Recirculation and Ventilation System
(FRVS); and CREF [control room emergency filtration] system. The
resultant radiological consequences are within the acceptance
criteria set forth in 10 CFR 50.67 and Regulatory Guide (RG) 1.183.
This amendment does not alter the methodology or equipment used in
fuel handling operations. The equipment hatch, personnel air locks,
other containment penetrations, or any component thereof is not an
accident initiator. Actual fuel handling operations are not affected
by the proposed changes.
Consequently the probability of a previously analyzed FHA is not
affected by the proposed amendment. No other accident initiator is
affected by the proposed changes.
Therefore, this proposed amendment does not involve a
significant increase in the probability of occurrence or
radiological 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 analyzed?
Response: No.
The proposed changes would revise TS 3.6.5.3.1, FRVS Ventilation
System and 3.6.5.3.2, FRVS Recirculation System, ACTION b from, ``*
* * containment or operations * * * '' to read ``* * * containment
and operations * * * '' to be consistent with NUREG-1433, Standard
Technical Specifications General Electric Plants, BWR/4'' (STS). TS
3.7.1.2, Service Water, and 3.8.3.2, Distribution--Shutdown, require
the addition of ``recently'' to modify irradiated fuel consistent
with NRC-approved Revision 2 to Technical Specification Task Force
(TSTF) Standard Technical Specification Change Traveler, TSTF-51,
``Revise Containment Requirements During Handling Irradiated Fuel
and Core Alterations.'' TS 3.8.1.2 A.C. Sources--Shutdown, 3.8.2.2,
D.C. Sources--Shutdown, and 3.8.3.2, Distribution--Shutdown, require
that ``CORE ALTERATIONS'' be added to ACTION a.
The proposed amendment will not create the possibility of a new
or different type of accident from any accident previously evaluated
because changes to the allowable activity in the primary and
secondary systems do not result in changes to the design or
operation of these systems. The evaluation of the proposed changes
indicates that all design standard and applicable safety criteria
limits are met. Equipment important to safety will continue to
operate as designed. Component integrity is not challenged. The
changes do not result in any event previously deemed incredible
being made credible. The changes do not result in more adverse
conditions or result in any increase in the challenges to safety
systems. The systems affected by the changes are used to mitigate
the consequences of a potential accident and would not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the change involve a significant reduction in the margin
of safety?
Response: No.
The proposed changes would revise TS 3.6.5.3.1, FRVS Ventilation
System and 3.6.5.3.2 FRVS Recirculation System, ACTION b from ``* *
* containment or operations * * * '' to read ``* * * containment and
operations * * * '' to be consistent with NUREG-1433, ``Standard
Technical Specifications General Electric Plants, BWR/4'' (STS). TS
3.7.1.2, Service Water, and 3.8.3.2, Distribution--Shutdown, require
the addition of ``recently'' to modify irradiated fuel consistent
with NRC approved Revision 2 to Technical Specification Task
[[Page 27003]]
Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
51, ``Revise Containment Requirements During Handling Irradiated
Fuel and Core Alterations.'' TS 3.8.1.2 A.C. Sources--Shutdown,
3.8.2.2 D.C. Sources--Shutdown, and 3.8.3.2 Distribution--Shutdown,
require that ``CORE ALTERATIONS'' be added to ACTION a.
The proposed changes revise the TS operational conditions where
specific activities represent situations during which significant
radioactive releases can be postulated. These operational conditions
are consistent with the design basis analysis and are established
such that the radiological consequences remain at or below the
regulatory guidelines. Safety margins and analytical conservatisms
are retained to ensure that the analysis adequately bounds all
postulated event scenarios. The proposed TS continue to ensure that
the total effective dose equivalent (TEDE) for the control room
(CR), the exclusion area boundary (EAB), and low population zone
(LPZ) boundaries are below the corresponding acceptance criteria
specified in 10 CFR 50.67 and RG 1.183.
Therefore, these 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: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Darrell J. Roberts.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: February 23, 2006.
Description of amendment request: The amendment would revise the
Operating License Condition 2.C.(6), ``Fuel Storage and Handling,'' to
clarify that the condition does not apply to Nuclear Regulator
Commission (NRC)-approved dry spent fuel storage systems. The current
condition states no more than a total of three fuel assemblies shall be
out of approved shipping containers, fuel assembly storage racks or the
reactor at any one time.
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 is a clarification to the Hope Creek
operating license to recognize that the dry spent fuel storage
system used at the ISFSI [independent spent fuel storage
installation] is licensed separately by the NRC under 10 CFR part
72. The change does not affect any SSCs [structure, systems and
components] used to operate the reactor or produce electrical power.
The change also does not affect SSCs used to shut down the reactor,
maintain it in a safe shutdown condition, or mitigate accidents.
The dry storage cask system design is supported by an NRC-
approved criticality analysis that demonstrates the system will
remain safely subcritical under all normal, off-normal, and credible
accident conditions applicable to the dry spent fuel storage system,
as defined in the cask CoC holder's 10 CFR part 72 licensing basis.
Dry spent fuel storage system loading operations are not addressed
in any Part 50 accident as described in Chapter 15 of the HCGS [Hope
Creek Generating Station] FSAR [final safety analysis report]. Dry
spent fuel storage system loading in the spent fuel pool is governed
by procedures that are consistent with the requirements in the HI-
STORM 100 System 10 CFR part 72 FSAR. Heavy load handling inside the
Part 50 facility associated with cask loading is conducted in
accordance with procedures that comply with the site's existing
heavy load control program. Because this change does not affect
PSEG's [PSEG Nuclear, LLC] heavy load handling procedures and all
structures, systems and components used for cask handling will meet
the existing commitments to NUREG-0612, a cask drop event remains
non-credible as currently described in HCGS FSAR Section 15.7.5.
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 is a clarification to the Hope Creek
operating license to recognize that the dry spent fuel storage
system is licensed separately by the NRC under 10 CFR part 72. The
change does not affect any SSCs used to operate the reactor or
produce electrical power. The change also does not affect SSCs used
to shut down the reactor, maintain it in a safe shutdown condition,
or mitigate accidents.
The dry spent fuel storage system design is supported by an NRC-
approved criticality analysis that demonstrates the system will
remain safely subcritical under all normal, off-normal, and credible
accident conditions, as defined in the cask CoC holder's 10 CFR part
72 licensing basis. Dry spent fuel storage system loading in the
spent fuel pool is governed by procedures that are consistent with
the requirements in the HI-STORM 100 System 10 CFR 72 FSAR. Heavy
load handling inside the Part 50 facility associated with cask
loading is conducted in accordance with procedures that comply with
the site's existing heavy load control program. Because this change
does not affect PSEG's heavy load handling procedures and all
structures, systems and components used for cask handling will meet
the existing commitments to NUREG-0612, a cask drop event remains
non-credible as currently described in HCGS FSAR Section 15.7.5.
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
the margin of safety?
Response: No.
The proposed change is a clarification to the Hope Creek
operating license to recognize that dry spent fuel storage systems
are licensed separately by the NRC under 10 CFR Part 72. The change
does not affect any SSCs used to operate the reactor or produce
electrical power. The change also does not affect SSCs used to shut
down the reactor, maintain it in a safe shutdown condition, or
mitigate accidents.
All safety analyses are consistent with the operations described
in the dry spent fuel storage system FSAR and have been previously
approved by the NRC as having sufficient safety margins. This change
does not affect the dry spent fuel storage system operation
procedures or change any normal, off-normal, or accident condition
for which the dry spent fuel storage system is designed.
Therefore, the proposed change will not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: 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: April 17, 2006.
Description of amendment requests: The proposed amendments would
delete Section 2.G of the Facility Operating Licenses, which require
reporting of violations of the requirements in Sections 2.C(1), 2.C(3),
and 2.F of the Facility Operating Licenses.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on August 29, 2005 (70 FR 51098), including a model
safety evaluation and model no significant hazards consideration (NSHC)
determination, using the consolidated
[[Page 27004]]
line item improvement process. The licensee affirmed the applicability
of the following NSHC determination in its application dated April 17,
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:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change involves the deletion of a reporting
requirement. The change does not affect plant equipment or operating
practices and therefore does not significantly increase 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?
Response: No.
The proposed change is administrative in that it deletes a
reporting requirement. The change does not add new plant equipment,
change existing plant equipment, or affect the operating practices
of the facility. Therefore, the change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change deletes a reporting requirement. The change
does not affect plant equipment or operating practices and therefore
does not involve a significant reduction in a margin of safety.
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.
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: March 29, 2006.
Description of amendment request: The proposed amendment would
revise Vogtle Electric Generating Plant (VEGP), Units 1 and 2,
Technical Specifications (TSs) 5.5, ``Programs and Manuals,'' TS 5.6,
``Reporting Requirements,'' and TS Bases for LCO [Limiting Condition
for Operation] 3.6.1, ``Containment,'' to reflect the latest
requirements for tendon surveillance.
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. The proposed license amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change replaces the current TS requirement to
implement a Containment Tendon Surveillance Program based on
Regulatory Guide 1.35, Rev. 2, with a Containment Inspection Program
Plan that complies with the current requirements of 10 CFR 50.55a.
This regulation requires licensees to implement a Containment
Inspection Program Plan in compliance with the 1992 Edition with the
1992 Addenda of Subsection IWE, ``Requirements for Class MC and
Metallic Liners of Class CC Components of Light-Water Cooled
Plants,'' and with Subsection IWL, ``Requirements for Class CC
Concrete Components of Light-Water Cooled Plants,'' of Section XI,
Division 1, of the American Society of Mechanical Engineers Boiler
and Pressure Vessel Code (ASME Code) with additional modifications
and limitations as stated in 10 CFR 50.55a(b)(2)(ix). [Southern
Nuclear Operating Company, Inc.] SNC has implemented a Containment
Inspection Program Plan that complies with the regulatory
requirements. This proposed TS amendment is requested to update the
TS to the latest 10 CFR 50.55a regulatory requirements.
In addition, reporting requirements that are redundant to
existing regulations are deleted, minor editorial changes are made,
and the applicability of SR 3.0.2 to the tendon surveillance program
is deleted since surveillance frequencies and associated extensions
are specified in ASME Section XI, Subsection IWL.
By complying with the regulatory requirements described in 10
CFR 50.55a, the probability of a loss of containment structural
integrity is maintained as low as reasonably achievable. Maintaining
containment structural integrity as described in the revised
Containment Inspection Program Plan does not impact the operation of
the reactor coolant system (RCS), containment spray (CS) system, or
emergency core cooling system (ECCS). The Containment Inspection
Program ensures that the containment will function as designed to
provide an acceptable barrier to release of radioactive materials to
the environment. The proposed change does not alter or prevent the
ability of structures, systems, and components (SSCs) from
performing their intended function to mitigate the consequences of
an initiating event within the assumed acceptance limits.
The proposed change does not impact any accident initiators or
analyzed events, nor does it impact the types or amounts of
radioactive effluent that may be released offsite. Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Maintaining containment structural integrity does not impact the
operation of the RCS, CS system, or ECCS. The proposed change does
not involve a modification to the physical configuration of the
plant or a change in the methods governing normal plant operation.
The proposed change does not introduce a new accident initiator,
accident precursor, or malfunction mechanism. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed license amendment does not involve a significant
reduction in a margin of safety.
By complying with the regulatory requirements described in 10
CFR 50.55a, the probability of a loss of containment structural
integrity is maintained as low as reasonably achievable. The
Containment Inspection Program Plan ensures that the containment
will function as designed to provide an acceptable barrier to
release of radioactive materials to the environment. The proposed
change does not adversely affect plant operation or existing safety
analyses. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Branch Chief: Evangelos C. Marinos.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: March 28, 2006.
Description of amendment request: The amendment would delete
references to specific isolation valves in the chemical and volume
control system (CVCS) and to modify notes to allow (1) an exception for
decontamination activities and (2) an exception for CVCS resin vessel
operation. These are changes to Technical Specifications (TSs) 3.3.9,
``Boron Dilution Mitigation System (BDMS),'' and 3.9.2, ``Unborated
Water Source Isolation Valves.''
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 do not involve a significant increase in
the probability or
[[Page 27005]]
consequences of an inadvertent boron dilution accident by isolating
the CVCS resin vessels in MODE 6 or by isolating the purge line for
detector SJRE001 during flushing activities in MODE 6. By
recognizing these potential [boron] dilution sources and by making
TS 3.3.9 and TS 3.9.2 more generic for consideration of all
potential [boron] dilution sources, plant administrative controls
are revised such that the plant is put in a safer condition than
before. Specific isolation valves are removed from TS 3.3.9 and TS
3.9.2. They are relocated from the [Technical] Specifications to the
appropriate TS Bases. This is an administrative only change and is
consistent with the [Improved] Standard Technical Specifications,
NUREG-1431. [The Wolf Creek Technical Specifications are based on
NUREG-1431.] Allowing a [boron] dilution source path to be
unisolated under administrative controls, described in TS Bases
3.9.1 during refueling decontamination activities, is acceptable as
allowed by Amendment [No.] 97 to the Callaway Operating License and
does not involve a significant increase in the probability or
consequences of an inadvertent boron dilution accident. Allowing an
exception for CVCS resin vessel operation is acceptable because
chemistry controls may require some CVCS resin vessels to be
configured with resin intended for boron dilution. Plant conditions
may warrant their use. As allowed by the LCO [limiting condition for
operation] Note, these vessels may be unisolated under
administrative controls. The administrative controls ensure that the
resin vessels are not [boron] dilution sources [for the reactor
coolant system (RCS)]. These changes do not involve a significant
increase in the probability or consequences of an inadvertent boron
dilution accident.
The proposed changes do not involve a significant increase in
the probability or consequences of an inadvertent boron dilution
accident by requiring the isolation of all unborated water source
isolation valves in higher plant modes when both trains of BDMS are
inoperable or when a condition of no reactor coolant loop in
operation exists. Proposed TS 3.3.9 Required Actions [B.3.1, B.3.2,
C.1 and C.2] are generic and remain consistent with the plant
accident analyses. Allowing exceptions for CVCS resin vessel
operation is acceptable because chemistry controls may require some
CVCS resin vessels to be configured with resin intended for boron
dilution. Plant conditions may warrant their use. As allowed by
exception Notes, these vessels may be unisolated under
administrative controls. The administrative controls ensure that the
resin vessels are not [boron] dilution sources.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not create the possibility of a new or
different kind of accident. Although other potential [boron]
dilution sources are identified for administrative control[s], the
evaluation of a MODE 6 [boron] dilution event remains unchanged.
Isolating the CVCS resin vessels or isolating the purge line for
detector SJRE001 during flushing activities in MODE 6 and making TS
3.3.9 and TS 3.9.2 more generic does not impact the operability of
any safety related equipment required for plant operation. No new
equipment will be added and no new limiting single failures are
created. The plant will continue to be operated within the envelope
of the existing safety analysis. In addition[,] specific isolation
valves are removed from TS 3.3.9 and TS 3.9.2. They are relocated
from the [Technical] Specifications to the appropriate TS Bases.
This is an administrative only change and is consistent with the
[Improved] Standard Technical Specifications, NUREG-1431. Allowing a
[boron] dilution source path to be unisolated under administrative
controls, described in TS Bases 3.9.1 during refueling
decontamination activities, is acceptable as allowed by Amendment
[No.] 97 to the Callaway Operating License and does not create the
possibility of a new or different kind of inadvertent boron dilution
accident. Allowing an exception for CVCS resin vessel operation is
acceptable because chemistry controls may require some CVCS resin
vessels to be reconfigured with resin intended for boron dilution.
Plant conditions may warrant their use. As allowed by the LCO Note
these vessels may be unisolated under administrative controls. The
administrative controls ensure that the resin vessels are not
[boron] dilution sources. These changes do not create the
possibility of a new or different kind of accident from an
inadvertent boron dilution accident previously evaluated.
Requiring the isolation of unborated water source isolation
valves in higher plant modes when both trains of BDMS are inoperable
or when a condition of no RCS loop in operation exists, does not
create the possibility of a new or different kind of inadvertent
boron dilution accident. Proposed TS 3.3.9 is generic and remains
consistent with the plant accident analyses. Allowing exceptions for
CVCS resin vessel operation is acceptable because chemistry controls
may require some CVCS resin vessels to be configured with resin
intended for boron dilution. Plant conditions may warrant their use.
As allowed by exception Notes, these vessels may be unisolated under
administrative controls. The administrative controls ensure that the
resin vessels are not [boron] dilution sources.
Therefore, the proposed changes do not create a new or different
kind of accident from any accident previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not reduce the margin of safety.
Although other potential [boron] dilution sources are identified for
administrative control[s] and TS 3.3.9 and TS 3.9.2 are made generic
for consideration of all potential [boron] dilution sources, the
evaluated margin of safety for a [boron] dilution event in MODE 6
remains the same. Recognition of other potential [boron] dilution
sources, isolation of the CVCS resin vessels and the purge line for
detector SJRE001 during flushing activities in MODE 6, places the
plant in a safer condition than before. In addition[,] specific
isolation valves are removed from TS 3.3.9 and TS 3.9.2. They are
relocated from the [Technical] Specifications to the appropriate TS
Bases. This is an administrative only change and is consistent with
the [Improved] Standard Technical Specifications, NUREG-1431.
Finally, allowing a [boron] dilution source path to be unisolated
under administrative controls, described in TS Bases 3.9.1 during
refueling decontamination activities, is acceptable under Amendment
[No.] 97 to the Callaway Operating License and does not involve a
significant reduction in a margin of safety [ * * * ]. Allowing an
exception for CVCS resin vessel operation is acceptable because
chemistry controls may require some CVCS resin vessels to be
configured with resin intended for boron dilution. Plant conditions
may warrant their use. As allowed by the LCO Note these vessels may
be unisolated under administrative controls. The administrative
controls ensure that the resin vessels are not [boron] dilution
sources. This change does not involve a significant reduction in a
margin of safety [ * * * ].
Requiring the isolation of all unborated water source isolation
valves in higher plant modes when both trains of BDMS are inoperable
or when no reactor coolant loop is in operation does not involve a
significant reduction in the margin of safety. The changes to the
[Technical] Specifications make it generic and [remain] consistent
with the plant accident analyses. Allowing exceptions for CVCS resin
vessel operation is acceptable because chemistry controls may
require some CVCS resin vessels to be configured with resin intended
for boron dilution. Plant conditions may warrant their use. As
allowed by these exception Notes, these vessels may be unisolated
under administrative controls. The administrative controls ensure
that the resin vessels are not [boron] dilution sources.
Therefore, the proposed changes do not involve a significant
reduction in [a] margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: David Terao.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri.
Date of amendment request: March 28, 2006.
Description of amendment request: The amendment would revise
Technical Specification 5.0, ``Administrative
[[Page 27006]]
Controls,'' by changing position titles and department names. The
amendment would not change any specific responsibilities, job
functions, organizational commitments, or qualification requirements of
plant personnel.
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 do not affect accident initiators or
assumptions. The radiological consequences of accidents previously
evaluated remain unchanged. These changes involve administrative
changes concerning designations for position titles and department
names. The changes do not affect responsibilities, functions,
organizational commitments, or the qualification requirements of
plant personnel.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes are administrative in nature. The overall
operating philosophy of [the] Callaway Plant is unchanged. As such,
there are no hardware changes nor are there any changes in the
method by which any safety-related plant system performs its safety
function. This amendment will not affect the normal method of plant
operation or change any operating parameters. No new accident
scenarios, transient precursors, failure mechanisms, or limiting
single failures are introduced as a result of this amendment. There
will be no adverse effects or challenges imposed on any safety-
related system as a result of this amendment.
Therefore, the proposed changes do not create a new or different
kind of accident from any accident previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
There will be no effect on the manner in which safety limits or
limiting safety system settings are determined nor will there be any
effect on those plant systems necessary to assure the accomplishment
of protection functions. The changes do not involve any change in
overall organizational commitments. The changes to personnel titles
and department designations are administrative and will not reduce
any margin of safety.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: David Terao.
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].
Duke Energy Corporation, et al., Docket No. 50-414, Catawba Nuclear
Station, Unit 2, York County, South Carolina
Date of application for amendments: December 19, 2005, as
supplemented on February 2 and 28, 2006.
Brief description of amendments: The amendment made a one-time
change to the Technical Specifications regarding the required steam
generator (SG) tube repair criteria for Catawba Unit 2 during refueling
outage 14 and operating cycle 15. In addition, the proposed amendment
added a license condition that requires a reduction in the allowable
normal operating primary-to-secondary leakage rate from 150 gallons-
per-day to 75 gallons-per-day through any one SG and from 600 gallons-
per-day to 300 gallons-per-day through all SGs. The proposed license
condition will be applicable only for the duration of Catawba Unit 2
cycle 15 operation.
Date of issuance: March 31, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance March 31, 2006.
Amendment No.: 224.
Renewed Facility Operating License No. NPF-52: Amendments revised
the Technical Specifications and the license.
Date of initial notice in Federal Register: February 22, 2006 (71
FR 9169).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 31, 2006.
No significant hazards consideration comments received: No.
[[Page 27007]]
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Duke Energy Corporation, Docket No. 72-004, Oconee Independent Spent
Fuel Storage Installation, Oconee County, South Carolina
Date of application for amendments: August 5, 2005, as supplemented
by letters dated November 28 and December 14, 2005, and February 6,
2006.
Brief description of amendments: The amendments revised the
operating licenses approving the indirect transfer of the Renewed
Facility Operating Licenses for Catawba Nuclear Station, Units 1 and 2,
McGuire Nuclear Station, Units 1 and 2, and Oconee Nuclear Station,
Units 1, 2, and 3, and the Materials License for Oconee Independent
Spent Fuel Storage Installation from Duke Energy Corporation to a new
holding company, to be named Duke Energy Corporation, in connection
with a proposed corporate restructuring and merger involving Cinergy
Corporation.
Date of issuance: April 1, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 229, 225, 232, 214, 349, 351, 349 and 8
respectively.
Renewed Facility Operating License Nos. NPF-35 , NPF-52, NPF-9,
NPF-17, DPR-38, DPR-47, DPR-55, and SNM-2503: Amendments revised the
Operating Licenses.
Date of initial notice in Federal Register: December 30, 2005 (70
FR 77428).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 7, 2006 (ML060250498).
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: May 19, 2004.
Brief description of amendment: The change revises Technical
Specification (TS) 3.8.1, ``AC Sources--Operating,'' to permit a longer
completion time for the Division 1 and Division 2 diesel generators
(DGs). This is a risk-informed TS change that would extend the DG
completion time from 72 hours (the current limit) to 14 days.
Date of issuance: April 14, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days of the date of issuance.
Amendment No.: 197.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 22, 2004 (69 FR
34699).
The September 1, 2005, January 9, February 23, and March 20, 2006,
supplemental letters and March 30, 2006, e-mail 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 no significant hazards considerations determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 14, 2006.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: September 2, 2004, as
supplemented by letters dated August 9, 2005, December 29, 2005 and
March 22, 2006.
Brief description of amendment: The amendment allows continued
plant operation with a single recirculation loop operation at Pilgrim.
Date of issuance: April 12, 2006.
Effective date: As of the date of issuance, and shall be
implemented within 120 days.
Amendment No.: 219.
Facility Operating License No. DPR-35: The amendment revised the
Facility Operating License, Technical Specifications and Surveillance
Requirements.
Date of initial notice in Federal Register: December 21, 2004 (69
FR 76490).
The supplements dated August 9, 2005, December 29, 2005 and March
22, 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 consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 12, 2006.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: May 24, 2005.
Brief description of amendment: The amendment deletes the main
steam isolation valve twice per week partial stroke testing
surveillance specified in Technical Specification 4.7.A.2.b.1.c.
Date of issuance: April 13, 2006.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 220.
Facility Operating License No. DPR-35: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 16, 2005 (70 FR
48205).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 13, 2006.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: May 24, 2005, as supplemented by
letter dated December 6, 2005.
Brief description of amendment: The amendment revises the Technical
Specifications allowances for bypassing the rod worth minimizer.
Date of issuance: April 13, 2006.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 221.
Facility Operating License No. DPR-35: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 30, 2005 (70 FR
51380).
The supplement dated December 6, 2005, 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. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 13, 2006.
No significant hazards consideration comments received: No.
[[Page 27008]]
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: April 20, 2005.
Brief description of amendment: The changes revised the Technical
Specifications (TSs) to replace plant-specific position titles with
generic position titles. Also, the changes deleted TS 6.7, ``Safety
Limit Violations or Protective Limit Violation,'' and included a change
to TS 2.1.2, ``Reactor Core,'' associated with the deletion of TS 6.7.
Additionally, the changes relocated to the Davis-Besse Nuclear Power
Station Updated Safety Analysis Report the Process Control Program
requirements from TS 6.8, ``Procedures and Programs,'' and from TS
6.14, ``Process Control Program (PCP).'' Associated with this change,
TS Definition 1.30, ``Process Control Program,'' was deleted. Also, TS
6.15, ``Offsite Dose Calculation Manual (ODCM),'' was modified to
eliminate the requirement that changes to the ODCM be reviewed and
accepted by the Plant Operations Review Committee (PORC). These changes
to administrative requirements also eliminated the need to propose
additional changes in the future to plant-specific position/
organizational titles. The changes are consistent with NUREG-1430,
``Standard Technical Specifications--Babcock and Wilcox Plants,''
Revision 3, dated June 2004. Lastly, the changes revised in the TSs the
title ``Industrial Security Plan'' to ``Physical Security Plan.''
Date of issuance: February 7, 2006.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 272.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 24, 2005 (70 FR
29795).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 7, 2006.
No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie
Plant, Unit No. 2, St. Lucie County, Florida
Date of application for amendment: January 6, 2005, as supplemented
October 14, 2005, and February 13, 2006.
Brief description of amendment: The amendment revises Technical
Specification (TS) Section 3/4.4.5, ``Steam Generators,'' to allow
repair of steam generator tubes by installing Westinghouse Alloy 800
leak limiting sleeves.
Date of Issuance: April 18, 2006.
Effective Date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 144.
Renewed Facility Operating License No. NPF-16: Amendment revised
the TS.
Date of initial notice in Federal Register: March 1, 2005 (70 FR
9993). The October 14, 2005, and February 13, 2006, supplements did not
affect the original proposed no significant hazards determination, or
expand the scope of the request as noticed in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 18, 2006.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendments request: June 1, 2005, as supplemented on
February 13, 2006.
Brief Description of amendments: The amendments revise Technical
Specification (TS) Section 5.5.6, ``Pre-Stressed Concrete Containment
Tendon Surveillance Program,'' for consistency with the requirements of
10 CFR 50.55a(g)(4) for components classified as Code Class CC. The
amendments also delete the provisions of Surveillance Requirement 3.0.2
from this TS and delete the reporting requirements in TS 5.6.9,
``Tendon Surveillance Report.''
Date of issuance: April 14, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 172 and 165.
Renewed Facility Operating License Nos. NPF-2 and NPF-8: Amendments
revise the Technical Specifications.
Date of initial notice in Federal Register: June 21, 2005 (70 FR
35739). The February 13, 2006, supplemental letter provided clarifying
information that did not change the June 1, 2005, application and the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 14, 2006.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry
Nuclear Plant, Units 2 and 3, Limestone County, Alabama
Date of application for amendments: July 29, 2005.
Brief description of amendments: The proposed amendments revised
the technical specification testing frequency for the surveillance
requirement 3.1.4.2, control rod scram time testing, from 120 days
cumulative operation in MODE 1 to 200 days cumulative operation in MODE
1.
Date of issuance: January 9, 2006.
Effective date: As of the date of issuance and to be implemented
within 60 days.
Amendment Nos.: 295 and 253.
Facility Operating License Nos. DPR-52 and DPR-68: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 27, 2005 (70
FR 56504).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 9, 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 amendment request: January 24, 2005.
Brief description of amendments: The requested amendments revise
Technical Specification (TS) 3.7.5, ``Auxiliary Feedwater (AFW)
System.'' The change would add a Note to surveillance requirements
(SRs) 3.7.5.1, 3.7.5.3, and 3.7.5.4 that states, ``AFW train(s) may be
considered OPERABLE during alignment and operation for steam generator
level control, if it is capable of being manually realigned to the AFW
mode of operation.''
Date of issuance: April 24, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 126 and 126.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 8, 2005 (70 FR
67753).
No significant hazards consideration comments received: No.
[[Page 27009]]
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, (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
[[Page 27010]]
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.
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\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.
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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).
Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook
Nuclear Plant, Unit 2 (DCCNP-2), Berrien County, Michigan
Date of amendment request: April 10, 2006, as supplemented on April
12, and 13 (two letters), 2006.
Description of amendment request: The amendment revised
Surveillance Requirement 3.8.1.11 of the DCCNP-2 Technical
Specifications, raising the diesel generator load rejection voltage
test limit from 5000 volts to 5350 volts.
Date of issuance: April 13, 2006.
Effective date: April 13, 2006.
Amendment No.: 276.
Facility Operating License No. DPR-74: Amendment revises the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): No. The Commission's related evaluation of the
amendment, finding of emergency circumstances, state consultation, and
final NSHC determination are contained in a safety evaluation dated
April 13, 2006.
Attorney for licensee: James M. Petro, Jr., Esquire, One Cook
Place, Bridgman, MI 49106.
NRC Branch Chief: L. Raghavan.
Dated at Rockville, Maryland, this 1st day of May 2006.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 06-4243 Filed 5-8-06; 8:45 am]
BILLING CODE 7590-01-P