[Federal Register Volume 72, Number 185 (Tuesday, September 25, 2007)]
[Notices]
[Pages 54471-54485]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-18634]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a determination by the Commission that such
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from August 30, 2007 to September 12, 2007. The
last biweekly notice was published on September 11, 2007 (72 FR 51852).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rulemaking,
Directives and Editing Branch, Division of Administrative Services,
Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the Commission's
Public Document Room (PDR), located at One White Flint North, Public
File Area O1F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. The filing of requests for a hearing and petitions for leave
to intervene is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, 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
[[Page 54472]]
at the hearing. The petitioner/requestor must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner/requestor intends to rely to
establish those facts or expert opinion. The petition must include
sufficient information to show that a genuine dispute exists with the
applicant on a material issue of law or fact. Contentions shall be
limited to matters within the scope of the amendment under
consideration. The contention must be one which, if proven, would
entitle the petitioner/requestor to relief. A petitioner/requestor who
fails to satisfy these requirements with respect to at least one
contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) e-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Detroit Edison Company, Docket No. 50-331, Duane Arnold Energy Center,
Linn County, Iowa
Date of amendment request: July 20, 2007.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.7.5 to add an Action Statement
for two inoperable control building chiller (CBC) subsystems. The
proposed new Action Statement would allow 72 hours to restore one CBC
subsystem to operable status and require verification once every 4
hours that control room temperature remains less than 90 [deg]F. The
proposed changes are consistent, with certain variations, with TS Task
Force (TSTF) Change Traveler TSTF-477, Revision 3, ``Adding an Action
Statement for Two Inoperable Control Room Air Conditioning
Subsystems.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration by a reference to a generic analysis published in the
Federal Register on December 18, 2006 (71 FR 75774), which 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 is described in Technical Specification Task
Force (TSTF) Standard TS Change Traveler TSTF-477 adds an action
statement for two inoperable control room subsystems.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
The proposed changes add an action statement for two inoperable
control room subsystems. The equipment qualification temperature of
the control room equipment is not affected. Future changes to the
Bases or licensee-controlled document will be evaluated pursuant to
the requirements of 10 CFR 50.59, ``Changes, test and experiments,''
to ensure that such changes do not result in more than a minimal
increase in the probability or consequences of an accident
previously evaluated.
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, and
configuration of the facility or the manner in which the plant is
operated and maintained. The proposed changes do not adversely
affect the ability of structures, systems and components (SSCs) to
perform their intended safety function to mitigate the consequences
of an initiating event within the assumed acceptance limits. The
proposed changes do not affect the source term, containment
isolation, or radiological consequences of any accident previously
evaluated. Further, the proposed changes do not increase the types
and the amounts of radioactive effluent that may be released, nor
significantly increase individual or cumulative occupation/public
radiation exposures.
Therefore, the changes do not involve a significant increase in
the probability or consequences of any 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 changes add an action statement for two inoperable
control room subsystems. The changes do not involve a physical
altering of the plant (i.e., no new or different type of equipment
will be installed) or a change in methods governing normal p[l]ant
operation. The requirements in the TS continue to require
maintaining the control room temperature within the design limits.
Therefore, the changes do not create the possibility of a new or
different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed changes add an action statement for two inoperable
control room subsystems. Instituting the proposed changes
[[Page 54473]]
will continue to maintain the control room temperature within design
limits. Changes to the Bases or license[e-] controlled document are
performed in accordance with 10 CFR 50.59. This approach provides an
effective level of regulatory control and ensures that the control
room temperature will be maintained within design limits.
The proposed changes maintain sufficient controls to preserve
the current margins 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: Marjan Mashhadi, Florida Power & Light
Company, 801 Pennsylvania Avenue, Suite 220, Washington, DC 20004.
NRC Acting Branch Chief: Travis L. Tate.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of amendment request: July 17, 2007, as supplemented by letter
dated August 7, 2007.
Description of amendment request: The proposed amendment would
revise the facility operating license (FOL), paragraph 2.C, and
technical specifications (TS) 3.7.2 and TS 5.5 for Grand Gulf Nuclear
Station, Unit 1.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on October 17, 2006 (71 FR 61075), on possible
amendments to revise the plant specific TS, to strengthen TS
requirements regarding control room envelope (CRE) habitability by
changing the action and surveillance requirements associated with the
limiting condition for operation operability requirements for the CRE
emergency ventilation system, and by adding a new TS administrative
controls program on CRE habitability, including a model safety
evaluation and model no significant hazards consideration (NSHC)
determination, using the consolidated line item improvement process.
The NRC staff subsequently issued a notice of availability of the
models for referencing in license amendment applications in the Federal
Register on January 17, 2007 (72 FR 2022). The licensee affirmed the
applicability of the model NSHC determination in its application dated
July 17, 2007, as supplemented by letter dated August 7, 2007.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility. The proposed change
does not alter or prevent the ability of structures, systems, and
components to perform their intended function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change revises the TS for the CRE emergency
ventilation system, which is a mitigation system designed to
minimize unfiltered air leakage into the CRE and to filter the CRE
atmosphere to protect the CRE occupants in the event of accidents
previously analyzed. An important part of the CRE emergency
ventilation system is the CRE boundary. The CRE emergency
ventilation system is not an initiator or precursor to any accident
previously evaluated. Therefore, the probability of any accident
previously evaluated is not increased. Performing tests to verify
the operability of the CRE boundary and implementing a program to
assess and maintain CRE habitability ensure that the CRE emergency
ventilation system is capable of adequately mitigating radiological
consequences to CRE occupants during accident conditions, and that
the CRE emergency ventilation system will perform as assumed in the
consequence analyses of design basis accidents. Thus, the
consequences of any accident previously evaluated are not increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Accident Previously
Evaluated
The proposed change does not impact the accident analysis. The
proposed change does not alter the required mitigation capability of
the CRE emergency ventilation system, or its functioning during
accident conditions as assumed in the licensing basis analyses of
design basis accident radiological consequences to CRE occupants. No
new or different accidents result from performing the new
surveillance or following the new program. The proposed change does
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a significant
change in the methods governing normal plant operation. The proposed
change does not alter any safety analysis assumptions and is
consistent with current plant operating practice. Therefore, this
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The proposed change does not affect safety
analysis acceptance criteria. The proposed change will not result in
plant operation in a configuration outside the design basis for an
unacceptable period of time without compensatory measures. The
proposed change does not adversely affect systems that respond to
safely shut down the plant and to maintain the plant in a safe
shutdown condition. Therefore, the proposed change does not involve
a significant reduction in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves NSHC.
Attorney for licensee: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Thomas G. Hiltz.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3 (Waterford 3), St. Charles Parish, Louisiana
Date of amendment request: August 16, 2007.
Description of amendment requests: A change is proposed to the
Waterford 3 Control
Room Emergency Air Filtration System technical specifications (TSs)
using the Nuclear Regulatory Commission (NRC) notice of availability
regarding Control Room Envelope (CRE) Habitability using the
Consolidated Line Item Improvement Process. The proposed amendment is
consistent with the NRC approved Industry/Technical Specification Task
Force (TSTF) change to the Standard Technical Specifications (STS),
TSTF-448, Revision 3, ``Control Room Habitability.''
The NRC staff issued a notice of opportunity for comment in the
Federal Register on October 17, 2006 (71 FR 61075), on possible
amendments adopting TSTF-448, including a model safety evaluation and
model no significant hazards consideration (NSHC) determination, using
the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on January 17, 2007 (72 FR 2022). The licensee affirmed the
applicability of the following NSHC determination in its application
dated August 16, 2007.
[[Page 54474]]
TSTF-448, Revision 3 is formatted to the Improved Technical
Specification (ITS) plants while the Waterford 3 TSs are based on the
CE standard technical specifications. Therefore, the information
contained in TSTF-448, Revision 3 has been modified to the Waterford 3
TS format.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility. The proposed change
does not alter or prevent the ability of structures, systems, and
components (SSCs) to perform their intended function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change revises the TS for the CRE emergency
ventilation system, which is a mitigation system designed to
minimize unfiltered air leakage into the CRE and to filter the CRE
atmosphere to protect the CRE occupants in the event of accidents
previously analyzed. An important part of the CRE emergency
ventilation system is the CRE boundary. The CRE emergency
ventilation system is not an initiator or precursor to any accident
previously evaluated. Therefore, the probability of any accident
previously evaluated is not increased. Performing tests to verify
the operability of the CRE boundary and implementing a program to
assess and maintain CRE habitability ensure that the CRE emergency
ventilation system is capable of adequately mitigating radiological
consequences to CRE occupants during accident conditions, and that
the CRE emergency ventilation system will perform as assumed in the
consequence analyses of design basis accidents. Thus, the
consequences of any accident previously evaluated are not increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Accident Previously
Evaluated
The proposed change does not impact the accident analysis. The
proposed change does not alter the required mitigation capability of
the CRE emergency ventilation system, or its functioning during
accident conditions as assumed in the licensing basis analyses of
design basis accident radiological consequences to CRE occupants. No
new or different accidents result from performing the new
surveillance or following the new program. The proposed change does
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a significant
change in the methods governing normal plant operation. The proposed
change does not alter any safety analysis assumptions and is
consistent with current plant operating practice. Therefore, this
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The proposed change does not affect safety
analysis acceptance criteria. The proposed change will not result in
plant operation in a configuration outside the design basis for an
unacceptable period of time without compensatory measures. The
proposed change does not adversely affect systems that respond to
safely shut down the plant and to maintain the plant in a safe
shutdown condition. Therefore, the proposed change does not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company,2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Branch Chief: Thomas G. Hiltz.
FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of amendment request: June 29, 2007.
Description of amendment request: The proposed amendment would
change Technical Specifications (TS) sections 3.7.4 and 5.5.13 to
strengthen TS requirements regarding control building envelope (CBE)
habitability. The proposed amendment would change the action and
surveillance requirements associated with the limiting condition for
operation operability requirements for the CBE standby filter unit and
add a new TS administrative controls program on CBE habitability. The
proposed changes to the TS and associated Bases are consistent with
certain exceptions with standard technical specifications (STS) as
revised by TS Task Force (TSTF) change traveler TSTF-448, Revision 3,
``Control Room Envelope Habitability'' to the extent that the amendment
request adopts by reference certain model TSTF-448 content, where
applicable.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on October 17, 2006 (71 FR 61075), on possible
amendments adopting TSTF-448, including a model safety evaluation and
model no significant hazards consideration (NSHC) determination, using
the Consolidated Line Item Improvement Process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on January 17, 2007 (72 FR 2022).
The licensee affirmed the applicability of the following NSHC
determination in its application dated June 29, 2007.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility. The proposed change
does not alter or prevent the ability of structures, systems, and
components (SSCs) to perform their intended function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change revises the TS for the CRE emergency
ventilation system, which is a mitigation system designed to
minimize unfiltered air leakage into the CRE and to filter the CRE
atmosphere to protect the CRE occupants in the event of accidents
previously analyzed. An important part of the CRE emergency
ventilation system is the CRE boundary. The CRE emergency
ventilation system is not an initiator or precursor to any accident
previously evaluated. Therefore, the probability of any accident
previously evaluated is not increased. Performing tests to verify
the operability of the CRE boundary and implementing a program to
assess and maintain CRE habitability ensure that the CRE emergency
ventilation system is capable of adequately mitigating radiological
consequences to CRE occupants during accident conditions, and that
the CRE emergency ventilation system will perform as assumed in the
consequence analyses of design basis accidents. Thus, the
consequences of any accident previously evaluated are not increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
[[Page 54475]]
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Accident Previously
Evaluated
The proposed change does not impact the accident analysis. The
proposed change does not alter the required mitigation capability of
the CRE emergency ventilation system, or its functioning during
accident conditions as assumed in the licensing basis analyses of
design basis accident radiological consequences to CRE occupants. No
new or different accidents result from performing the new
surveillance or following the new program. The proposed change does
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a significant
change in the methods governing normal plant operation. The proposed
change does not alter any safety analysis assumptions and is
consistent with current plant operating practice. Therefore, this
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The proposed change does not affect safety
analysis acceptance criteria. The proposed change will not result in
plant operation in a configuration outside the design basis for an
unacceptable period of time without compensatory measures. The
proposed change does not adversely affect systems that respond to
safely shut down the plant and to maintain the plant in a safe
shutdown condition. Therefore, the proposed change does not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Marjan Mashhadi, Florida Power & Light
Company, 801 Pennsylvania Avenue, Suite 220, Washington, DC 20004.
NRC Acting Branch Chief: Travis L. Tate.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment request: July 30, 2007.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) by adding a new Surveillance
Requirement (SR) 3.8.2.2 that would be applicable when onsite
electrical power is supplied to a unit via backfeed through the main
transformer, and the unit is in either Mode 5 or Mode 6, or during
movement of irradiated fuel. The proposed SR would correct a non-
conservatism in the TS and will assure the capability to transfer the
required safety-related loads from the backfeed source to the qualified
offsite circuit.
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 of occurrence or consequences of an accident
previously evaluated?
Response: No.
The proposed change will add a new Technical Specification
Surveillance Requirement applicable during shutdown conditions when
a backfeed configuration is used to provide power from the offsite
transmission network to required safety equipment via the main
transformer. The new Surveillance Requirement will require that
portions of an existing Surveillance Requirement be met. If not met,
the existing Surveillance Requirement must be performed before
establishing a backfeed configuration. It is highly unlikely that
the proposed change will necessitate performance of the existing
Surveillance Requirement more frequently than is currently required.
Even if more frequent performance of the existing Surveillance
Requirement were required, its performance would not significantly
increase the probability of a loss of offsite power. Consequently,
there is no significant change in the likelihood of any accident
associated with verifying the existing Surveillance Requirement has
been met. Therefore, the probability of occurrence of a previously
evaluated accident will not be significantly increased.
The verifications required by the new Surveillance Requirement
will assure that a unit's required safety-related equipment can be
transferred to a qualified offsite circuit while the equipment is
being provided power from the offsite transmission network using a
backfeed configuration while the unit is shutdown or while
irradiated fuel is [being] moved. This will provide assurance that
the systems needed to mitigate the consequences of the accidents in
these conditions will be provided with electrical power if the
systems are needed to perform their specified safety function.
Therefore, the consequences of a previously evaluated accident will
not be significantly increased.
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 addition of a new Technical Specification Surveillance
Requirement to verify that an existing Surveillance Requirement has
been met, or to perform that Surveillance Requirement if not met,
would not create the possibility of a new or different kind of
accident because the Surveillance Requirement has previously existed
and previously been performed. Therefore, the proposed change does
not involve any new systems, structures, or components, or any
different mode of operation of any existing systems, structures, or
components.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety associated with the proposed change
involves the availability of offsite electrical power to support
required safety equipment when a unit is shut down or during the
movement of irradiated fuel. The proposed change provides assurance
that the single required qualified offsite circuit from the
transmission network remains available while the required safety
equipment is powered by a different circuit from that network.
Consequently, the proposed change does not reduce the margin of
safety provided by the required qualified offsite circuit, and
enhances the margin of safety by acknowledging use of an additional
offsite circuit.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The Nuclear Regulatory Commission (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: Kimberly Harshaw, Esquire, One Cook Place,
Bridgman, MI 49106
NRC Acting Branch Chief: Travis Tate.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: August 10, 2007.
Description of amendment request: The proposed change to Technical
Specification 2.1.1.2 will revise two recirculation loop and single
recirculation loop safety limit minimum critical power ratio (SLMCPR)
values to reflect results of a cycle-specific calculation.
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:
[[Page 54476]]
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
Four accidents have been evaluated previously as reflected in
the CNS [Cooper Nuclear Station] Updated Safety Analysis Report
(USAR). These four accidents are (1) loss-of-coolant, (2) control
rod drop, (3) main steamline break, and (4) fuel handling. The
probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident.
Changing the SLMCPR does not increase the probability of an
evaluated accident. The change does not require any physical plant
modifications to the plant or any components, nor does it require a
change in plant operation. Therefore, no individual precursors of an
accident are affected.
The consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. This proposed change makes no modification to the
design or operation of the systems that are used in mitigation of
accidents. Limits have been established, consistent with NRC
approved methods, to ensure that fuel performance during normal,
transient, and accident conditions is acceptable. The proposed
change to the value of the SLMCPR continues to conservatively
establish this safety limit such that the fuel is protected during
normal operation and during any plant transients or anticipated
operational occurrences.
Based on the above NPPD [Nebraska Public Power District]
concludes that the proposed change does 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.
Creation of the possibility of a new or different kind of
accident from an accident previously evaluated would require
creation of precursors of that accident. New accident precursors may
be created by modification of the plant configuration or changes in
how the plant is operated. The proposed change does not involve a
modification of the plant configuration or in how the plant is
operated. The proposed change to the SLMCPR assures that safety
criteria are maintained.
Based on the above, NPPD concludes that the proposed change does
not create the possibility of a new or different kind of accident
from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The value of the proposed SLMCPR provides a margin of safety by
ensuring that no more than 0.1% of the rods are expected to be in
boiling transition if the Minimum Critical Power Ratio limit is not
violated. The proposed change will ensure the appropriate level of
fuel protection is maintained. Additionally, operational limits are
established based on the proposed SLMCPR to ensure that the SLMCPR
is not violated during all modes of operation. This will ensure that
the fuel design safety criteria (i.e., that at least 99.9% of the
fuel rods do not experience transition boiling during normal
operation as well as anticipated operational occurrences) are met.
Based on the above, NPPD concludes that 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: Mr. John C. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Branch Chief: Thomas G. Hiltz.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: August 16, 2007.
Description of amendment request: The proposed amendment revises
Technical Specification 5.5.6, ``Inservice Testing Program,'' to allow
a one-time extension of the five-year frequency requirement for
setpoint testing of safety valve MS-RV-70ARV.
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 function of SRVs [safety relief valves] and SVs [safety
valves] is to prevent overpressurization of the reactor coolant
system (RCS) during transients and abnormal operation that could
cause increases in RCS pressure. They are also used to depressurize
the RCS when needed to allow injection of water from the high-
volume, low-pressure Emergency Core Cooling System (ECCS) Low
Pressure Coolant Injection mode of the Residual Heat Removal System
into the reactor pressure vessel (RPV) as part of mitigation of an
accident. Actuation or failure to actuate of a SRV or SV is not an
initiator of any accident previously evaluated. Thus, this proposed
amendment would not result in a significant increase in the
probability of an accident previously evaluated.
A range or tolerance of plus-or-minus three percent of the
setpoint pressure is acceptable for the results of setpoint testing.
A 90-day extension of the interval for setpoint testing of one SV is
not expected to result in actuation of the SV outside of its
acceptable setpoint range. However, even if the single SV whose test
interval is being extended did actuate outside of its acceptable
range, it is not expected that this would result in a significant
degradation in the ability of the Nuclear System Pressure Relief
System to perform its safety function, since the remaining eight
SRVs and two other SVs would be unaffected by the proposed extension
of the testing interval for the single SV. The proposed change does
not modify the design of or alter the operation of systems or
components used in mitigating design basis accidents. Thus, this
proposed amendment would not result in a significant increase in the
consequences of any accident previously evaluated.
Based on the above, it is concluded that the proposed change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
A new or different kind of accident from any previously
evaluated might result from a modification of the plant design by
either addition of a new system or removal of an existing system, or
a change in how any of the plant systems function during the
operation of the plant. The proposed change does not modify the
plant design, nor does it alter the operation of the plant or
equipment involved in either routine plant operation or in the
mitigation of the design basis accidents.
Based on the above, it is concluded that the proposed change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The margin of safety applicable to this issue would be the
margin between the pressure at which the SRVs and SVs would actuate
and the allowable ASME [American Society of Mechanical Engineers]
Code overpressure limit of 1,375 psig [pounds per square inch gauge]
(110 percent of vessel design pressure, 1250 psig). This margin
would be impacted if the setpoint at which the applicable SV
actuated experienced drift greater than the allowable plus-or-minus
three percent of the setpoint pressure. This is not expected to
occur based on the results demonstrated by the setpoint testing
conducted over the last ten years. Those results were two actuations
of the SV at a pressure below the nameplate rating with less than
two percent deviation, and one actuation at a pressure above the
nameplate rating with less than one percent deviation. However, even
if this one SV did experience setpoint drift greater than the
allowable plus-or-minus three percent, there would not be a
significant reduction in the margin since it is expected that the
remaining eight SRVs and the two other SVs would actuate within the
allowable setpoint tolerance and begin to reduce RCS pressure as
needed. Furthermore, the proposed extension will not result in a
change to the steam discharge capacity and characteristics of the
applicable SV.
[[Page 54477]]
Based on the above, it is concluded that 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. John C. McClure, Nebraska Public Power
District, Post OfficeBox 499, Columbus, NE 68602-0499.
NRC Branch Chief: Thomas G. Hiltz.
Nine Mile Point Nuclear Station (NMPNS), LLC, Docket No. 50-410, Nine
Mile Point Nuclear Station Unit No. 2 (NMP2), Oswego County, New York
Date of amendment request: July 23, 2007.
Description of amendment request: The proposed amendment would
modify a footnote in NMP2 Technical Specification (TS) Table 3.3.2.1-1,
``Control Rod Block Instrumentation,'' such that a new banked position
withdrawal sequence (BPWS) shutdown sequence could be utilized. The
proposed change is consistent with TS Task Force (TSTF) change TSTF-
476, Revision 1, ``Improved BPWS Control Rod Insertion Process (NEDO-
33091).'' The availability of the TS change was published in the
Federal Register on May 23, 2007 (72 FR 29004) as part of the
consolidated line item improvement process. The licensee affirmed the
applicability of the model no significant hazards consideration
determination in its application.
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 changes modify the TS to allow the use of the
improved banked position withdrawal sequence (BPWS) during shutdowns
if the conditions of NEDO-33091-A, Revision 2, ''Improved BPWS
Control Rod Insertion Process,'' July 2004, have been satisfied. The
[NRC] staff finds that the licensee's justifications to support the
specific TS changes are consistent with the approved topical report
and TSTF-476, Revision 1. Since the change only involves changes in
control rod sequencing, the probability of an accident previously
evaluated is not significantly increased, if at all. The
consequences of an accident after adopting TSTF-476 are no different
than the consequences of an accident prior to adopting TSTF-476.
Therefore, the consequences of an accident previously evaluated are
not significantly affected by this change. 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 [Accident] Previously
Evaluated
The proposed change 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 control rod drop accident (CRDA)
is the design basis accident for the subject TS changes. This change
does not create the possibility of a new or different kind of
accident from [any] accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the [a] Margin of Safety
The proposed change, TSTF-476, Revision 1, incorporates the
improved BPWS, previously approved in NEDO-33091-A, into the
improved TS. The control rod drop accident (CRDA) is the design
basis accident for the subject TS changes. In order to minimize the
impact of a CRDA, the BPWS process was developed to minimize control
rod reactivity worth for BWR plants. The proposed improved BPWS
further simplifies the control rod insertion process, and in order
to evaluate it, the [NRC] staff followed the guidelines of Standard
Review Plan Section 15.4.9, and referred to General Design Criterion
28 of Appendix A to 10 CFR Part 50 as its regulatory requirement.
The TSTF stated the improved BPWS provides the following benefits:
(1) Allows the plant to reach the all-rods-in condition prior to
significant reactor cool down, which reduces the potential for re-
criticality as the reactor cools down; (2) reduces the potential for
an operator reactivity control error by reducing the total number of
control rod manipulations; (3) minimizes the need for manual scrams
during plant shutdowns, resulting in less wear on control rod drive
(CRD) system components and CRD mechanisms; and (4) eliminates
unnecessary control rod manipulations at low power, resulting in
less wear on reactor manual control and CRD system components. The
addition of procedural requirements and verifications specified in
NEDO-33091-A, along with the proper use of the BPWS will prevent a
control rod drop accident (CRDA) from occurring while power is below
the low power setpoint (LPSP). 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: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1700 K Street, NW., Washington, DC 20006.
NRC Branch Chief: Mark G. Kowal.
Nine Mile Point Nuclear Station (NMPNS), LLC, Docket No. 50-410, Nine
Mile Point Nuclear Station Unit No. 2 (NMP2), Oswego County, New York
Date of amendment request: July 30, 2007.
Description of amendment request: The proposed amendment would
modify Technical Specifications (TS) 3.7.3, ``Control Room Envelope Air
Conditioning (AC) System,'' by adding an Action Statement to the
Limiting Conditions for Operation. The new Action Statement allows a
finite time to restore one control room envelope AC subsystem to
operable status and requires verification that the control room
temperature remains < 90 [deg]F every 4 hours. The proposed changes are
consistent with Nuclear Regulatory Commission (NRC)-approved TS Task
Force (TSTF) TSTF-477, Revision 3, ``Adding an Action Statement for Two
Inoperable Control Room Air Conditioning Subsystems.'' The availability
of this TS improvement was published in the Federal Register on March
26, 2007 (72 FR 14143) as part of the consolidated line item
improvement process. The licensee affirmed the applicability of the
model no significant hazards consideration determination in its
application.
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 is described in Technical Specification Task
Force (TSTF) Standard TS Change Traveler TSTF-477 [and] adds an
action statement for two inoperable control room subsystems. The
proposed change does not involve a physical alteration of the plant
(no new or different type of equipment will be installed). The
proposed changes add an action statement for two inoperable control
room subsystems. The equipment qualification temperature of the
control room equipment is not affected. Future changes to the Bases
or licensee controlled documents will be evaluated pursuant to the
requirements of 10 CFR 50.59, ``Changes, test and experiments'', to
ensure that such changes do not result in more than a minimal
increase in the probability or consequences of an accident
previously evaluated. The proposed changes do not adversely affect
accident initiators or precursors nor alter the design assumptions,
conditions, and configuration of the facility or the manner in which
the plant is operated and maintained. The proposed changes do not
adversely affect the ability of structures, systems and components
(SSCs) to perform their intended safety function to mitigate the
[[Page 54478]]
consequences of an initiating event within the assumed acceptance
limits. The proposed changes do not affect the source term,
containment isolation, or radiological consequences of any accident
previously evaluated. Further, the proposed changes do not increase
the types and the amounts of radioactive effluent that may be
released, nor significantly increase individual or cumulative
occupation/public radiation exposures. Therefore, the changes do 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 [Accident] Previously
Evaluated
The proposed changes add an action statement for two inoperable
control room subsystems. The changes do not involve a physical
altering of the plant (i.e., no new or different type of equipment
will be installed) or a change in methods governing normal plant
operation. The requirements in the TS continue to require
maintaining the control room temperature within the design limits.
Therefore, the changes do not create the possibility of a new or
different kind of accident from any [accident] previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the [a] Margin of Safety
The proposed changes add an action statement for two inoperable
control room subsystems. Instituting the proposed changes will
continue to maintain the control room temperature within design
limits. Changes to the Bases or license[e-]controlled document[s]
are performed in accordance with 10 CFR 50.59. This approach
provides an effective level of regulatory control and ensures that
the control room temperature will be maintained within design
limits. The proposed changes maintain sufficient controls to
preserve the current margins of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1700 K Street, NW., Washington, DC 20006.
NRC Branch Chief: Mark G. Kowal.
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: June 7, 2007.
Description of amendment request: The proposed amendment would
delete the license conditions that require reporting of violations of
other requirements (e.g., conditions listed in Sections 2.C and 2.F for
Unit 1 and Section 2.C for Unit 2) in the operating licenses. This
change is in accordance with Nuclear Regulatory Commission (NRC)-
approved Technical Specification (TS) Task Force (TSTF) change traveler
TSTF-372, Revision 4. The NRC staff issued a notice of availability of
a model no significant hazards consideration (NSHC) determination in
the Federal Register on August 29, 2005 (70 FR 51098). The notice
included a model safety evaluation, a model NSHC determination, and a
model license amendment request. In its application dated June 7, 2007,
the licensee affirmed the applicability of the model NSHC determination
which is presented below.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC adopted by the licensee 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 has reviewed the analysis adopted by the licensee
and, based on this review, it appears that the standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the request for amendments involves NSHC.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRCBranch Chief: Mark G. Kowal.
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: June 8, 2007.
Description of amendment request: The proposed amendment would
revise Limiting Condition for Operation (LCO) 3.10.1, and the
associated Bases, to expand its scope to include provisions for
temperature excursion greater than 200 degrees Fahrenheit ([deg]F) as a
consequence of inservice leak and hydrostatic testing, and as a
consequence of scram time testing initiated in conjunction with an
inservice leak or hydrostatic test, while considering operational
conditions to be in Mode 4 for SSES 1 and 2. This change is in
accordance with Nuclear Regulatory Commission (NRC)-approved Technical
Specification (TS) Task Force (TSTF) change traveler TSTF-484, ``Use of
TS 3.10.1 for Scram Time Testing Activities.'' The NRC staff issued a
notice of opportunity to comment and notice of availability of a model
no significant hazards consideration (NSHC) determination in the
Federal Register on August 21, 2006 (71 FR 48561) and October 27, 2006
(71 FR 63050), respectively. The notices included a model safety
evaluation, a model NSHC determination, and a model license amendment
request. In its application dated June 8, 2007, the licensee affirmed
the applicability of the model NSHC determination which is presented
below.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC adopted by the licensee is presented below:
Criterion 1--The Proposed Change Does not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
Technical Specifications currently allow for operation at
greater than 200 degrees Fahrenheit ([deg]F) while imposing MODE 4
requirements in addition to the secondary containment requirements
required to be met. Extending the activities that can apply this
allowance will not adversely impact the probability or consequences
of an accident previously evaluated. Therefore, the proposed change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Accident Previously
Evaluated
Technical Specifications currently allow for operation at
greater than 200 [deg]F while imposing MODE 4 requirements in
addition to the secondary containment requirements required to be
met. No new operational conditions beyond those currently allowed by
LCO 3.10.1 are introduced. The changes do not involve a physical
alteration of the plant (i.e., no new or different type of equipment
will be installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements or
[[Page 54479]]
eliminate any existing requirements. The changes do not alter
assumptions made in the safety analysis. The proposed changes are
consistent with the safety analysis assumptions and current plant
operating practice. Therefore, the proposed change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety
Technical Specifications currently allow for operation at
greater than 200 [deg]F while imposing MODE 4 requirements in
addition to the secondary containment requirements required to be
met. Extending the activities that can apply this allowance will not
adversely impact any margin of safety. Allowing completion of
inspections and testing and supporting completion of scram time
testing initiated in conjunction with an inservice leak or
hydrostatic test prior to power operation results in enhanced safe
operations by eliminating unnecessary maneuvers to control reactor
temperature and pressure. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the analysis adopted by the licensee
and, based on this review, it appears that the standards of 10 CFR
50.92 (c) are satisfied. Therefore, the NRC staff proposes to determine
that the request for amendments involves NSHC.
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: Mark G. Kowal.
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: August 14, 2007.
Description of amendment request: The proposed amendments would add
a new license condition to the SSES 1 and 2 Operating Licenses to
permit the valves in Title 10 of the Code of Federal Regulations (10
CFR) Part 50, Appendix J leakage test program to be tested at the
higher pressure during the next scheduled test rather than requiring
all of the valves to be tested at the higher pressure prior to the
implementation of the constant pressure power uprate license amendment.
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:
USGPO Galley End:?>1. Does the proposed change involve a
significant increase in the probability or consequences of an
accident previously evaluated?
Response: No.
The proposed License Condition change does not involve any
physical change to structures, systems, or components (SSCs) and
does not alter the method of operation or control of SSCs. The
current assumptions in the safety analysis regarding accident
initiators and mitigation of accidents are unaffected by this
change. No additional failure modes or mechanisms are being
introduced and the likelihood of previously analyzed failures
remains unchanged.
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 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, at which protective or mitigative actions are
initiated, affected by this change. This change will not alter the
manner in which equipment operation is initiated, nor will the
function 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 [final safety analysis report]. As such, no new failure modes
are being introduced. The change does not alter assumptions made in
the safety analysis and licensing basis.
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. The proposed change is acceptable because of the
satisfactory performance of the Primary Containment Integrated Leak
Rate Tests on both Unit 1 and Unit 2 at the new calculated pressure
and the substantial margin to leakage rate acceptance limits based
upon the Integrated Leak Rate Test and the current LLRT [local leak
rate tests] results. Therefore, the plant response to analyzed
events will continue to provide the margin of safety assumed by the
analysis.
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: Mark G. Kowal.
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 amendment request: April 27, 2007.
Description of amendment request: The proposed amendment revises
the Joseph M. Farley Nuclear Plant, Units 1 and 2 Technical
Specifications (TS) for Limiting Condition for Operation 3.9.3
``Containment Penetrations,'' to allow the containment personnel air
locks that provide direct access from the containment atmosphere to the
auxiliary building to be open during refueling activities if
appropriate administrative controls are established.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (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 change would allow the personnel air lock doors,
and emergency air lock doors to remain open during fuel movement and
core alterations. These doors are normally closed during this time
period in order to prevent the release of radioactive material in
the event of a fuel handling accident (FHA) inside containment.
These doors are not initiators of any accident. The probability of a
FHA is unaffected by the operational status of these doors.
The new FHA analysis with open containment personnel air locks
demonstrates that maximum offsite dose is within the acceptance
limits specified in RG [Regulatory Guide] 1.195. The FHA analysis
results in maximum offsite doses of 68.5 rem [roentgen equivalent
man] to the thyroid and 0.2 rem to the whole body. The calculated
control room dose is also within the acceptance criteria specified
in GDC [General Design Criteria] 19. The analysis results in thyroid
and whole body doses to the control room operator of 39.6 rem and <
0.1 rem, respectively.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not involve the addition or
modification of any plant equipment. Also, the proposed change will
not alter the design, configuration, or method of operation of the
plant beyond the standard functional capabilities of the equipment.
The proposed change involves a TS change that
[[Page 54480]]
will allow the air lock doors to be open during core alterations and
fuel movement inside containment. Open doors and penetrations do not
create the possibility of a new accident. Administrative controls
will be implemented to ensure the capability to close the
containment in the event of a FHA.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change has the potential to increase the post-FHA
dose at the Site Boundary, Low Population Zone and in the control
room. However, a revised FHA analysis demonstrates that the dose
consequences at both locations remains within regulatory acceptance
limits and the margin of safety as defined by 10 CFR 100 and GDC 19
has not been significantly reduced. To ensure a bounding
calculation, the revised FHA was performed with conservative
assumptions. For example, it assumes the unfiltered release to the
outside atmosphere of all airborne activity reaching the
containment. Additional margin will be established through
administrative procedures to require that the equipment hatch and at
least one door in each air lock be closed following an evacuation of
containment.
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: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201.
NRC Branch Chief: Evangelos C. Marinos.
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 amendment request: June 5, 2007.
Description of amendment request: The proposed amendments is for a
new technical specification (TS) to address the operation of Engineered
Safety Feature (ESF) Room Coolers required to support ESF TS equipment.
This amendment includes surveillance requirements and will establish a
Completion Time of 72 hours to allow adequate time to complete
maintenance activities on the ESF Room Coolers and thus reduce the need
for unnecessary plant shutdowns.
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 addition of Technical Specification (TS) 3.7.19
creates a Limiting Condition for Operation (LCO) for the Engineering
Safety Feature (ESF) Room Coolers required to support ESF TS
equipment. The Completion Time presented in the new TS is consistent
with other ESF mechanical system Completion Times and is supported
by the inputs used in the current analysis. The possibility of a
loss of off site power (LOSP) is actually reduced by continuing
power operation of the Unit. The radiological consequences of any
associated accidents are not impacted by the proposed amendment.
Therefore, it is concluded that this change does not
significantly increase the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a change in the methods
governing normal operation of the plant. No new accident scenarios,
failure mechanisms or limiting single failures are introduced as
result of the proposed change. The change has no adverse effects on
any safety-related system.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not impact accident offsite dose,
containment pressure or temperature, emergency core cooling system
(ECCS) or reactor protection system (RPS) settings or any other
parameter that could affect a margin of safety.
Therefore, it is concluded that this change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201.
NRC Branch Chief: Evangelos C. Marinos.
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 amendment request: July 17, 2007.
Description of amendment request: The proposed amendments would
revise the current Joseph M. Farley Nuclear Plant, Units 1 and 2
technical specification (TS) requirement for the Plant Manager or the
Operations Manager to hold a Senior Reactor Operator (SRO) license.
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 change to TS 5.2 revises the requirement concerning
the Operations management position that must hold an SRO license. At
least one Operations Superintendent or the Operations Manager will
continue to maintain an SRO license. In addition, a requirement was
added that if not currently licensed, the Operations Manager shall
have previously held an SRO license. The training, qualification and
experience requirements for Operations management personnel will
continue to satisfy the Unit Staff Qualifications as described in
the applicable TS 5.3.1. This change does not impact any accident
initiators or analyzed events. It does not impact any assumed
mitigation capability for any accident or transient event. The
change does not involve the addition or removal of any equipment or
any design changes to the facility. As the proposed change is
administrative in nature, operation of the facility in accordance
with the proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change to TS 5.2 revises the requirement concerning
the Operations management position that must hold an SRO license. At
least one Operations Superintendent or the Operations Manager will
continue to maintain an SRO license. In addition, a requirement was
added that if not currently licensed, the Operations Manager shall
have previously held an SRO license. The training, qualification and
experience requirements for Operations management personnel will
continue to satisfy the Unit Staff Qualifications as described in
the applicable TS 5.3.1. This change does not involve any physical
modifications to plant structures, systems, or components (SSCs), or
[[Page 54481]]
the manner in which SSCs are operated, maintained, modified, tested,
or inspected. In addition, there is no change in the types or
increases in the amounts of effluents that may be released offsite,
and there is no increase in individual or cumulative occupational
radiation exposure. As the proposed change is administrative in
nature, operation of the facility in accordance with 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 change to TS 5.2 revises the requirement concerning
the Operations management position that must hold an SRO license. At
least one Operations Superintendent or the Operations Manager will
continue to maintain an SRO license. The subject Operations
Superintendent will be qualified to fill the Operations Manager
position and have the same management authority over licensed
operators as the Operations Manager.
In addition, a requirement was added that if not currently
licensed, the Operations Manager shall have previously held an SRO
license. Administrative procedures will ensure that there is always
an individual holding a current SRO license within Operations
management. The training, qualification and experience requirements
for Operations management personnel will continue to satisfy the
Unit Staff Qualifications as described in the applicable TS 5.3.1.
This change does not involve any physical modifications to SSCs,
or the manner in which SSCs are operated, maintained, modified,
tested, or inspected. The change does not alter the manner in which
safety limits, limiting safety system settings, or limiting
conditions for operation are determined. The setpoints at which
protective actions are initiated are not altered by the change. As
the proposed change is administrative in nature, operation of the
facility in accordance with the proposed amendment does not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201.
NRC Branch Chief: Evangelos C. Marinos.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2 (HNP), Appling County, Georgia
Date of amendment request: July 17, 2007.
Description of amendment request: The proposed amendments would
revise the current HNP Technical Specification requirement for the
Operations Manager to hold an active or inactive Senior Reactor Opeator
(SRO) license.
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 change to TS 5.2 revises the requirement concerning
the Operations management position that must hold an SRO license. At
least one Operations Superintendent or the Operations Manager will
continue to maintain an SRO license. In addition, a requirement was
added that if not currently licensed, the Operations Manager shall
have previously held an SRO license. The training, qualification and
experience requirements for Operations management personnel will
continue to satisfy the Unit Staff Qualifications as described in
the applicable TS 5.3.1. This change does not impact any accident
initiators or analyzed events. It does not impact any assumed
mitigation capability for any accident or transient event. The
change does not involve the addition or removal of any equipment or
any design changes to the facility. As the proposed change is
administrative in nature, operation of the facility in accordance
with the proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change to TS 5.2 revises the requirement concerning
the Operations management position that must hold an SRO license. At
least one Operations Superintendent or the Operations Manager will
continue to maintain an SRO license. In addition, a requirement was
added that if not currently licensed, the Operations Manager shall
have previously held an SRO license. The training, qualification and
experience requirements for Operations management personnel will
continue to satisfy the Unit Staff Qualifications as described in
the applicable TS 5.3.1. This change does not involve any physical
modifications to plant structures, systems, or components (SSCs), or
the manner in which SSCs are operated, maintained, modified, tested,
or inspected. In addition, there is no change in the types or
increases in the amounts of effluents that may be released offsite,
and there is no increase in individual or cumulative occupational
radiation exposure. As the proposed change is administrative in
nature, operation of the facility in accordance with 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 change to TS 5.2 revises the requirement concerning
the Operations management position that must hold an SRO license. At
least one Operations Superintendent or the Operations Manager will
continue to maintain an SRO license. The subject Operations
Superintendent will be qualified to fill the Operations Manager
position and have the same management authority over licensed
operators as the Operations Manager. In addition, a requirement was
added that if not currently licensed, the Operations Manager shall
have previously held an SRO license. Administrative procedures will
ensure that there is always an individual holding a current SRO
license within Operations management. The training, qualification
and experience requirements for Operations management personnel will
continue to satisfy the Unit Staff Qualifications as described in
the applicable TS 5.3.1.
This change does not involve any physical modifications to SSCs,
or the manner in which SSCs are operated, maintained, modified,
tested, or inspected. The change does not alter the manner in which
safety limits, limiting safety system settings, or limiting
conditions for operation are determined. The setpoints at which
protective actions are initiated are not altered by the change. As
the proposed change is administrative in nature, operation of the
facility in accordance with the proposed amendment does not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Branch Chief: Evangelos C. Marinos.
[[Page 54482]]
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2 (VEGP), Burke
County, Georgia
Date of amendment request: July 17, 2007.
Description of amendment request: The proposed amendments would
revise the current VEGP Technical Specification requirement for the
Operation Manager to hold a Senior Reactor Operator (SRO) license.
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 change to TS 5.2 revises the requirement concerning
the Operations management position that must hold an SRO license. At
least one Operations Superintendent or the Operations Manager will
continue to maintain an SRO license. In addition, a requirement was
added that if not currently licensed, the Operations Manager shall
have previously held an SRO license. The training, qualification and
experience requirements for Operations management personnel will
continue to satisfy the Unit Staff Qualifications as described in
the applicable TS 5.3.1. This change does not impact any accident
initiators or analyzed events. It does not impact any assumed
mitigation capability for any accident or transient event. The
change does not involve the addition or removal of any equipment or
any design changes to the facility. As the proposed change is
administrative in nature, operation of the facility in accordance
with the proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change to TS 5.2 revises the requirement concerning
the Operations management position that must hold an SRO license. At
least one Operations Superintendent or the Operations Manager will
continue to maintain an SRO license. In addition, a requirement was
added that if not currently licensed, the Operations Manager shall
have previously held an SRO license. The training, qualification and
experience requirements for Operations management personnel will
continue to satisfy the Unit Staff Qualifications as described in
the applicable TS 5.3.1. This change does not involve any physical
modifications to plant structures, systems, or components (SSCs), or
the manner in which SSCs are operated, maintained, modified, tested,
or inspected. In addition, there is no change in the types or
increases in the amounts of effluents that may be released offsite,
and there is no increase in individual or cumulative occupational
radiation exposure.
As the proposed change is administrative in nature, operation of
the facility in accordance with 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 change to TS 5.2 revises the requirement concerning
the Operations management position that must hold an SRO license. At
least one Operations Superintendent or the Operations Manager will
continue to maintain an SRO license. The subject Operations
Superintendent will be qualified to fill the Operations Manager
position and have the same management authority over licensed
operators as the Operations Manager. In addition, a requirement was
added that if not currently licensed, the Operations Manager shall
have previously held an SRO license. Administrative procedures will
ensure that there is always an individual holding a current SRO
license within Operations management. The training, qualification
and experience requirements for Operations management personnel will
continue to satisfy the Unit Staff Qualifications as described in
the applicable TS 5.3.1.
This change does not involve any physical modifications to SSCs,
or the manner in which SSCs are operated, maintained, modified,
tested, or inspected. The change does not alter the manner in which
safety limits, limiting safety system settings, or limiting
conditions for operation are determined. The setpoints at which
protective actions are initiated are not altered by the change. As
the proposed change is administrative in nature, operation of the
facility in accordance with the proposed amendment does not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
Nations Bank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Branch Chief: Evangelos C. Marinos.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: August 16, 2007.
Brief description of amendments: The proposed amendments would
revise Technical Specifications (TS) 3.1.4, ``Rod Group Alignment
Limits,'' Table 3.3.1-1, ``Reactor Trip System Instrumentation,'' Table
3.3.2-1, ``Engineered Safety Feature Actuation System
Instrumentation,'' TS 3.4.10, ``Pressurizer Safety Valves,'' TS 3.7.1,
``Main Steam Safety Valves (MSSVs),'' and Table 3.7.1-1, ``Operable
Main Steam Safety Valves Versus Maximum Allowable Power.'' The proposed
change is a request to revise TSs for Comanche Peak Steam Electric
Station, Units 1 and 2, to reflect cycle-specific safety analysis
assumptions and results associated with the adoption of Westinghouse
accident analyses methodologies.
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 only affect the transient and accident
mitigation capability of the plant. The proposed changes to the
pressurizer safety valve set pressure and as-found tolerance do not
overlap with the pressurizer control system operation nor with the
reactor trip setpoint. Therefore, the proposed changes do affect the
probability of an accident previously evaluated.
The revised Reactor Trip System and Engineered Safety Features
Actuation System setpoints have been shown, using NRC-approved
analysis methodologies [the licensee's submittal for incorporating
standard Westinghouse-developed analytical methods at Comanche Peak
Steam Electric Station is under review by NRC], to meet all relevant
event acceptance criteria. Similarly, the change to the nominal set
pressure of the pressurizer safety valve, when evaluated using NRC-
approved analysis methodologies, has been shown to meet the relevant
event acceptance criteria. The proposed reduction to maximum
allowable power level for operation in inoperable MSSVs has been
previously shown to be very conservative. 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 based on analyses and evaluations
performed in accordance with NRC-approved methodologies shown to be
applicable [to] CPNPP [Comanche Peak Nuclear Power Plant] and to be
conservatively applied to CPNPP [Comanche Peak Steam Electric
Station herein referred to as CPNPP]. None of
[[Page 54483]]
the proposed changes can result in plant operation outside the
limits previously considered, nor allow the progression of transient
or accident in a manner different that previously considered.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes are based on analyses and evaluations
performed in accordance with NRC-approved methodologies shown to be
applicable to CPNPP and to be conservatively applied to CPNPP. All
relevant event acceptance criteria were found to be satisfied.
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: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Branch Chief: Thomas G. Hiltz.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendments request: January 22, 2007.
Description of amendments request: The proposed amendments change
the Technical Specifications related to the fuel design description and
the fuel criticality methods to accommodate the transition to AREVA
fuel.
Date of publication of individual notice in the Federal Register:
August 29, 2007 (72 FR 49742).
Expiration dates of individual notice: September 28, 2007 (Public
comments) and October 29, 2007 (Hearing requests).
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of amendment request: July 10, 2007.
Brief description of amendment request: The proposed amendment
would revise the values of the safety limit minimum critical power
ratio in Technical Specification Section 2.1.1, ``Reactor Core SLs.''
Date of publication of individual notice in Federal Register:
September 5, 2007 (72 FR 50986).
Expiration date of individual notice: November 5, 2007.
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].
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, CalvertCliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: November 3, 2005, as
supplemented March 22 and July 17, 2007.
Brief description of amendments: The amendments implement the
alternative source term methodology for analyzing design basis accident
radiological consequences, thereby replacing the existing accident
radiological source term that is described in Technical Information
Document TID-14844, ``Calculation of Distance Factors for Power and
Test Reactor Sites.''
Date of issuance: August 29, 2007.
Effective date: This license amendment is effective as of the date
of its issuance and shall be implemented within 60 days following
completion of the installation and testing of the plant modifications
described in the licensee's letters dated November 3, 2005,March 22 and
July 17, 2007.
Amendment Nos.: 281 and 258.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the License and Technical Specifications.
Date of initial notice in Federal Register: January 17, 2006 (71 FR
2589) The supplements dated March 22 and July 17, 2007, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination.
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated August 29, 2007.
No significant hazards consideration comments received: No.
[[Page 54484]]
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of application for amendment: August 2, 2007 as supplemented
by letters dated March 9 and May 8, 2007.
Brief description of amendment: This amendment revises Technical
Specification 2.2.1 and 3/4.3.2 to modify the statistical summation
error term ``Z'' and one of the allowable values for certain steam
generator water level trip setpoints used in the Reactor Trip system
and Engineered Safety Feature Actuation System instrumentation.
Date of issuance: August 31, 2007.
Effective date: 60 days from the date of issuance.
Amendment No. 126.
Facility Operating License No. NPF-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 27, 2007 (72
CR 8801). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 31, 2007. The
supplemental letters provided clarifying information that did not
expand the scope of the original application or change the initial
proposed no significant hazards consideration determination. No
significant hazards consideration comments received: No.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: October 16, 2006, as supplemented by
letter dated July 30, 2007.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) to add a topical report to the analytical methods
referenced in TS 5.6.5.b, ``Core Operating Limits Report (COLR),''
previously approved by U.S. Nuclear Regulatory Commission. The current
method of performing the loss-of-coolant accident analyses was replaced
by an updated method described in AREVA NP (formerly known as Framatome
or Siemens) topical report, ``EXEM BWR-2000 [Boiling-Water Reactor-
2000] ECCS [Emergency Core Cooling System] Evaluation Model.''
Date of issuance: August 30, 2007.
Effective date: As of the date of issuance and shall be implemented
prior to Cycle 15 operation.
Amendment No.: 153.
Facility Operating License No. NPF-47: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: November 7, 2006 (71 FR
65141). The supplemental letter dated July 30, 2007, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 30, 2007.
No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of application for amendments: October 19, 2006, as
supplemented June 7, 2007.
Brief description of amendments: Amendments revise Technical
Specification 4.6.2.1.d. to change the frequency of air or smoke flow
testing of the containment spray nozzles.
Date of Issuance: September 4, 2007.
Effective Date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 201 and 148.
Renewed Facility Operating License Nos. DPR-67 and NPF-16:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: January 3, 2007 (72 FR
152). The supplement dated June 7, 2007, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards determination as published in
the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 4, 2007.
No significant hazards consideration comments received: No.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of application for amendments: April 26, 2007.
Brief description of amendments: The amendments revised the
technical specifications (TSs) to add new Limiting Condition for
Operation 3.0.6.
Date of issuance: September 5, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos: 235 and 230.
Renewed Facility Operating License Nos. DPR-31 and DPR-41:
Amendments revised the TSs.
Date of initial notice in Federal Register: July 3, 2007 (72 FR
36522).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 5, 2007.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: October 17, 2006, as supplemented by
letters dated February 7, April 17, May 4, and July 26, 2007.
Brief description of amendment: The amendment revised Technical
Specification (TS) 4.3.1.1.c, ``Criticality,'' by adding a new nominal
center-to-center distance between fuel assemblies for two new storage
racks, and TS 4.3.3, ``Capacity,'' by increasing the capacity of the
spent fuel storage pool from 2366 assemblies to 2651 assemblies.
Date of issuance: September 6, 2007.
Effective date: As of the date of issuance and shall be implemented
within 45 days of issuance.
Amendment No.: 227.
Facility Operating License No. DPR-46: Amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: December 5, 2006 (71 FR
70561) and January 19, 2007 (72 FR 2560).
The supplements dated February 7, April 17, May 4, and July 26,
2007, provided additional information that clarified the application,
did not expand the scope of the application as originally noticed, and
did not change the staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 6, 2007.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit No. 2, Oswego County, New York
Date of application for amendment: January 4, 2007, as supplemented
by letters dated April 27, 2007, May 22, 2007, and July 23, 2007.
[[Page 54485]]
Brief description of amendment: The amendment revises Technical
Specification (TS) 3.7.1, ``Service Water (SW) System and Ultimate Heat
Sink (UHS),'' as follows: revises the existing Limiting Condition for
Operation (LCO) statement to require four operable SW pumps to be in
operation when SW subsystem supply header water temperature is <=82
[deg]F; adds a requirement that five operable SW pumps be in operation
when SW subsystem supply header water temperature is >82 [deg]F and
<=84 [deg]F; deletes Condition G and the associated Required Actions
and Completion Times; revises Surveillance Requirement 3.7.1.3 to
increase the maximum allowed SW subsystem supply header water
temperature from 82 [deg]F to 84 [deg]F; and modifies the requirements
for increasing the surveillance frequency as the temperature approaches
the limit.
Date of issuance: September 4, 2007.
Effective date: As of the date of issuance to be implemented within
90 days.
Amendment No.: 119.
Renewed Facility Operating License No. NPF-69: Amendment revises
the License and Technical Specifications.
Date of initial notice in Federal Register: March 13, 2007 (72 FR
11390).
The supplemental letters dated April 27, 2007, May 22, 2007, and
July 23, 2007, provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the Nuclear Regulatory Commission staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 4, 2007.
No significant hazards consideration comments received: No
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of application for amendments: November 16, 2006, as
supplemented on March 29 and July 31, 2007.
Brief Description of amendments: These amendments added a reference
in Technical Specification (TS) Section 6.2.C, ``Core Operating Limits
Report (COLR),'' to permit the use of the Westinghouse Best-Estimate
Large Break Loss-of-Coolant Accident (BE-LBLOCA) analysis methodology
using the Automated Statistical Treatment of Uncertainty Method
(ASTRUM) for the analysis of LBLOCA.
Date of issuance: September 6, 2007.
Effective date: As of date of issuance and shall be implemented at
the completion of Unit 1 fall 2007 refueling outage.
Amendment Nos.: 254 and 253.
Renewed Facility Operating License Nos. DPR-32 and DPR-37:
Amendments changed the licenses and the technical specifications.
Date of initial notice in Federal Register: December 5, 2006 (71 FR
70564). The supplements dated March 29 and July 31, 2007, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 6, 2007.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: March 14, 2007.
Brief description of amendment: The amendment revised Surveillance
Requirements 3.7.2.1 and 3.7.3.1 for the main steam isolation valves
and main feedwater isolation valves, respectively, to replace the
isolation times by the phrase ``within limits.'' The valve closure
times will be stated in the TS Bases, which is controlled by TS 5.5.14,
``Technical Specification (TS) Bases Control Program.'' This amendment
is consistent with the NRC-approved Technical Specification Task Force
Traveler 491, Revision 2, ``Removal of Main Steam and Main Feedwater
Isolation Times.''
There are other proposed changes to the TSs in the application
dated March 14, 2007, that are not being addressed in this amendment.
These will be addressed in future letters to the licensee.
Date of issuance: August 28, 2007.
Effective date: Effective as of its date of issuance and shall be
implemented within 90 days of the date of issuance.
Amendment No.: 174.
Facility Operating License No. NPF-42. The amendment revised the
Operating License and Technical Specifications.
Date of initial notice in Federal Register: June 19, 2007 (72 FR
33785).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 28, 2007.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 17th day of September, 2007.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E7-18634 Filed 9-24-07; 8:45 am]
BILLING CODE 7590-01-P