[Federal Register Volume 72, Number 175 (Tuesday, September 11, 2007)]
[Notices]
[Pages 51852-51869]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-17864]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a
[[Page 51853]]
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 16, 2007 to August 29, 2007. The last
biweekly notice was published on August 28, 2007 (72 FR 49568).
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 at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final
[[Page 51854]]
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 email to [email protected].
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: June 12, 2007.
Description of amendment request: The proposed amendment would
revise Technical Specification 3.7.4 to add an Action Statement for two
inoperable control center air conditioning (AC) subsystems. The
proposed new Action Statement would allow a finite time to restore one
control center AC subsystem to operable status and require verification
that control room temperature remains < 90 [deg]F every 4 hours.
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 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 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 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 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: David G. Pettinari, Legal Department, 688
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279.
NRC Acting Branch Chief: Travis L. Tate.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: July 17, 2007.
[[Page 51855]]
Description of amendment request: The proposed changes would modify
Technical Specification (TS) requirements related to control room
envelope (CRE) habitability in TS 3.7.3, ``Control Room Emergency
Ventilation Air Supply (CREVAS) System'' and adds new TS 5.5.14,
``Control Room Envelope Habitability Program.''
These changes were proposed by the industry's TS Task Force (TSTF)
and is designated TSTF-448. The NRC staff issued a notice of
opportunity for comment in the Federal Register on October 17, 2006 (71
FR 61075), on possible amendments concerning TSTF-448, including a
model safety evaluation and model no significant hazards (NSHC)
determination, using the consolidated line item improvement process
(CLIIP). 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 July 17, 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.
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 a 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.
Based on the above, the NRC staff concludes that the proposed
change presents no significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of
no significant hazards consideration is justified.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Mark G. Kowal.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: July 25, 2007.
Description of amendment request: The proposed amendment would
modify the Technical Specifications (TS) by adding an Action statement
to the Limiting Condition for Operation (LCO) for TS 3.7.4, Control
Room Air Conditioning (AC) System. The new Action statement allows a
finite time to restore one control room AC subsystem to operable status
(72 hours) and requires verification that control room temperature
remains less than 104 [deg]F every 4 hours. The licensing basis control
room air temperature for the James A. FitzPatrick Nuclear Power Plant
(JAFNPP) is 104 [deg]F.
This change was proposed by the industry's TS Task Force (TSTF) and
is designated TSTF-477. The NRC staff issued a notice of opportunity
for comment in the Federal Register on December 18, 2006 (71 FR 75774),
on possible amendments concerning TSTF-477, including a model safety
evaluation and model no significant hazards (NSHC) determination, using
the consolidated line item improvement process (CLIIP). The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on March 26, 2007 (72 FR 14143). The licensee affirmed the
applicability of the following NSHC determination in its application
dated July 25, 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 Changes Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change as 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
[[Page 51856]]
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 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 previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in 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 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.
Based on the above, the NRC staff concludes that the proposed
change presents no significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of
no significant hazards consideration is justified.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Mark G. Kowal.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1 (RBS), West Feliciana Parish, Louisiana
Date of amendment request: July 2, 2007.
Description of amendment request: The proposed amendment would
modify RBS technical specification (TS) requirements for MODE change
limitations in limiting condition for operation (LCO) 3.0.4 and
surveillance requirement (SR) 3.0.4. The proposed TS changes are
consistent with Revision 9 of Nuclear Regulatory Commission (NRC)
approved Industry TS Task Force (TSTF) Standard TS Change Traveler,
TSTF-359, ``Increase Flexibility in MODE Restraints.'' In addition, the
proposed amendment would also change TS section 1.4, Frequency, Example
1.4-1, ``Surveillance Requirements,'' to accurately reflect the changes
made by TSTF-359, which is consistent with NRC-approved TSTF-485,
Revision 0, ``Correct Example 1.4-1.''
The NRC staff issued a notice of opportunity for comment in the
Federal Register on August 2, 2002 (67 FR 50475), as part of the
Consolidated Line Item Improvement Process (CLIIP), on possible
amendments to revise the plant-specific TS to modify requirements for
MODE change limitations in LCO 3.0.4 and SR 3.0.4.
The NRC staff subsequently issued a notice of availability of the
models for Safety Evaluation and No Significant Hazards Consideration
Determination for referencing in license amendment applications in the
Federal Register on April 4, 2003 (68 FR 16579). The licensee affirmed
the applicability of the CLIIP, including the model No Significant
Hazards Consideration Determination, in its application dated February
8, 2007.
The proposed TS changes are consistent with NRC-approved Industry
TSTF Standard TS change, TSTF-359, Revision 8, as modified by 68 FR
16579. TSTF-359, Revision 8, was subsequently revised to incorporate
the modifications discussed in the April 4, 2003, Federal Register
notice and other minor changes. TSTF-359, Revision 9, was subsequently
submitted to the NRC on April 28, 2003, and was approved by the NRC on
May 9, 2003.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the NRC staff's analysis
of the issue of no significant hazards consideration is presented
below:
Criterion 1--The Proposed Changes Do Not Involve a Significant Increase
in the Probability or Consequences of an Accident Previously Evaluated
The proposed changes in TS Section 1.4, Frequency, Example 1.4-
1, would accurately reflect the changes made by TSTF-359 in LCO
3.0.4 and SR 3.0.4, which are consistent with NRC-approved TSTF-485,
Revision 0. These changes are considered administrative in that they
modify the example to demonstrate the proper application of LCO
3.0.4 and SR 3.0.4. The requirements of LCO 3.0.4 and SR 3.0.4 are
clear and are clearly explained in the associated Bases. As a
result, modifying the example will not result in a change in usage
of the TS.
The proposed changes in LCO 3.0.4 and SR 3.0.4 allow entry into
a mode or other specified condition in the applicability of a TS,
while in a TS condition statement and the associated required
actions of the TS. The proposed changes do not adversely affect
accident initiators or precursors, 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, or radiological release assumptions used in
evaluating the radiological consequences of an accident previously
evaluated. Being in a TS condition and the associated required
actions are not an initiator of any accident previously evaluated.
Therefore, the probability of an accident previously evaluated is
not significantly increased. The consequences of an accident while
relying on required actions as allowed by proposed LCO 3.0.4, are no
different than the consequences of an accident while entering and
relying on the required actions while starting in a condition of
applicability of the TS. Therefore, the consequences of an accident
previously evaluated are not significantly affected by these
changes. The addition of a requirement to assess and manage the risk
introduced by these changes will further minimize possible concerns.
Therefore, these changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Changes Do Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
No new or different accidents result from utilizing the proposed
changes. The proposed changes do not involve a physical alteration
of the plant (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 eliminate any existing requirements. The
proposed changes do not alter assumptions made in the safety
analysis and are consistent with the safety analysis assumptions and
current plant operating practice. Entering into a mode or other
specified condition in the applicability of a TS, while in a TS
condition statement and the associated required actions of the TS,
will not introduce new failure modes or effects and will not, in the
absence of other unrelated failures, lead to an accident whose
consequences exceed the consequences of accidents previously
evaluated. The addition of a requirement to assess and manage the
risk introduced by these changes will further minimize possible
concerns. Thus, these changes do not create the possibility of a new
or different kind of accident from an accident previously evaluated.
[[Page 51857]]
Criterion 3--The Proposed Changes Do Not Involve a Significant
Reduction in the Margin of Safety
The proposed changes in TS section 1.4, Example 1.4-1, are
considered administrative and will have no effect on the application
of the TS requirements. Therefore, the margin of safety provided by
the TS requirements is unchanged.
The proposed changes in TS LCO 3.0.4 and SR 3.0.4 allow entry
into a mode or other specified condition in the applicability of a
TS, while in a TS condition statement and the associated required
actions of the TS. The RBS TS allows operation of the plant without
the full complement of equipment through the TS conditions for not
meeting the TS LCO. The risk associated with this allowance is
managed by the imposition of required actions that must be performed
within the prescribed completion times. The net effect of being in a
TS LCO condition on the margin of safety is not considered
significant. The proposed changes do not alter the required actions
or completion times of the TS. The proposed changes allow TS
conditions to be entered, and the associated required actions and
completion times to be used in new circumstances. This use is
predicated upon the licensee's performance of a risk assessment and
the management of plant risk. The changes also eliminate current
allowances for utilizing required actions and completion times in
similar circumstances, without assessing and managing risk. The net
change to the margin of safety is insignificant. Therefore, these
changes do 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: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Thomas G. Hiltz.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1 (RBS), West Feliciana Parish, Louisiana
Date of amendment request: July 16, 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.
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 requirements
regarding control room envelope (CRE) habitability by changing the
action and surveillance requirements associated with the limiting
condition for operability requirements for the CRE emergency
ventilation system. A new TS administrative controls program on CRE
habitability is being added, 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 16, 2007, as supplemented by letter dated August 7, 2007.
Basis for proposed NSHC 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 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 no significant hazards consideration.
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 Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1 (RBS), West Feliciana Parish, Louisiana
Date of amendment request: August 17, 2007.
Description of amendment request: The proposed amendment would
revise the date for performing the ``Type A test'' in the RBS technical
specification (TS) 5.5.13, ``Primary Containment Leak Rate Testing
Program,'' from ``prior to December 14, 2007'' to ``April 14, 2008.''
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
[[Page 51858]]
consequences of an accident previously evaluated?
Response: No.
The proposed amendment to TS 5.5.13 allows a one-time extension
to the current interval for the ILRT [integrated leak rate test].
The current interval of 15 years 4 months, based on past
performance, would be extended on a one-time basis to 15 years and 8
months from the date of the last test. The proposed extension to the
ILRT cannot increase the probability of an accident since there are
no design or operating changes involved and the test is not an
accident initiator. The proposed extension of the test interval does
not involve a significant increase in the consequences since
analysis has shown that, the proposed extension of the ILRT and DWBT
[drywell bypass test] frequency has a minimal impact on plant risk.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed extension to the interval for the ILRT does not
involve any design or operational changes that could lead to a new
or different kind of accident from any accidents previously
evaluated. The tests are not being modified, but are only being
performed after a longer interval. The proposed change does not
involve a physical alteration of the plant (no new or different type
of equipment will be installed) or a change in the methods governing
normal plant operation. 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.
An evaluation of extending the ILRT DWBT surveillance frequency
from once in 10 years to once in 15 years and 8 months has been
performed using methodologies based on the approved ILRT methodologies.
This evaluation assumed that the DWBT frequency was being adjusted in
conjunction with the ILRT frequency. This analysis used realistic, but
still conservative, assumptions with regard to developing the frequency
of leakage classes associated with the ILRT and DWBT. The results from
this conservative analysis indicates that the proposed extension of the
ILRT frequency has a minimal impact on plant risk and 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: 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, St. Charles Parish, Louisiana
Date of amendment request: August 2, 2007.
Description of amendment request: The proposed changes to the
technical specifications (TSs) will add new analytical methods and
modify the containment average air temperature and safety injection
tank level to support the implementation of Combustion Engineering 16 x
16 Next Generation Fuel (NGF) as defined in Westinghouse Topical Report
WCAP-16500-P beginning in Cycle 16 commencing after the spring 2008
refueling outage. The fuel design is intended to provide improved fuel
reliability by reducing grid-to-rod fretting issues, improved fuel
performance for high duty operation, and enhanced operating margin.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Core Operating Limits Report (COLR)
The proposed changes to the COLR TS are administrative in nature
and have no impact on any plant configuration or system performance
relied upon to mitigate the consequences of an accident. Changes to
the calculated core operating limits may only be made using NRC
approved methodologies, must be consistent with all applicable
safety analysis limits, and are controlled by the 10 CFR 50.59
process.
The proposed change will add the following topical reports to
the list of referenced core operating analytical methods.
WCAP-16500-P and Final Safety Evaluation (SE)
Westinghouse topical report WCAP-16500-P describes the methods
and models that will be used to evaluate the acceptability of CE 16
x 16 NGF at CE plants. Entergy has demonstrated that the Limitations
and Conditions associated with the NRC SE will be met. Prior to
implementation of NGF the new core design will be analyzed with
applicable NRC staff approved codes and methods.
WCAP-12610-P-A and CENPD-404-P-A Addendum 1-A
The proposed change allows the use of methods required for the
implementation of Optimized ZIRLOTM clad fuel rods. Entergy has
demonstrated that the Limitations and Conditions associated with the
NRC SE will be met.
WCAP-16523-P and Final Safety Evaluation
This topical report describes the departure from nucleate
boiling correlations that will be used to account for the impact of
the CE 16 x 16 NGF fuel assembly design. Entergy has demonstrated
that the Limitations and Conditions associated with the NRC SE will
be met. Prior to implementation of NGF the new core design will be
analyzed with applicable NRC staff approved codes and methods.
CENPD-387-P-A
The proposed addition of this topical report provides the
departure from nucleate boiling (DNB) correlation that will be used
to evaluate the DNB impact of non-mixing vane grid spans for CE 16 x
16 standard and NGF assemblies Entergy has demonstrated that the
Limitations and Conditions associated with the NRC SE will be met.
CENPD-132, Supplement 4-P-A, Addendum 1-P and Final Safety
Evaluation
The addendum provides an optional steam cooling model that can
be used for Emergency Core Cooling System (ECCS) Performance
analyses to support the implementation of the CE 16 x 16 NGF fuel
assembly design. Entergy has demonstrated that the Limitations and
Conditions associated with the NRC SE will be met.
Assumptions used for accident initiators and/or safety analysis
acceptance criteria are not altered by the addition of these topical
reports.
Safety Injection Tank Water Level and Containment Average Air
Temperature
These values are used as inputs to the LBLOCA and SBLOCA
analyses. The new limits ensure that the analyzed LBLOCA remain
acceptable. The limits have no impact to the SBLOCA analysis
results. The changes do not cause an increase in the probability of
an accident or an increase in the dose consequences associated with
a LBLOCA.
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.
Core Operating Limits Report (COLR)
The proposed change identifies changes in the codes used to
confirm the values of selected cycle-specific reactor physics
parameter limits. The proposed change allows the use of methods
required for the implementation of CE 16 x 16 NGF. The proposed
addition of the referenced topical reports has no impact on any
plant configurations or on system performance that
[[Page 51859]]
is relied upon to mitigate the consequences of an accident. The
change to the COLR is administrative in nature and does not result
in a change to the physical plant or to the modes of operation
defined in the facility license.
WCAP-16500-P and Final Safety Evaluation
The proposed change adds Westinghouse topical report WCAP-16500-
P, which describes the methods and models that will be used to
evaluate the acceptability of CE 16 x 16 NGF at CE plants. Entergy
has demonstrated that the Limitations and Conditions associated with
the NRC SE will be met. Prior to implementation of NGF, the new core
design will be analyzed with applicable NRC staff approved codes and
methods.
WCAP-12610-P-A and CENPD-404-P-A Addendum 1-A
The proposed change allows the use of methods required for the
implementation of Optimized ZIRLOTM clad fuel rods. Entergy has
demonstrated that the Limitations and Conditions associated with the
NRC SE will be met.
WCAP-16523-P and Final Safety Evaluation
This topical report describes the departure from nucleate
boiling correlations that will be used to account for the impact of
the CE 16 x 16 NGF fuel assembly design. Entergy has demonstrated
that the Limitations and Conditions associated with the SE will be
met.
CENPD-387-P-A
The proposed addition of this topical report provides the
departure from nucleate boiling (DNB) correlation that will be used
to evaluate the DNB impact of non-mixing vane grid spans for CE 16 x
16 standard and NGF assemblies. Entergy has demonstrated that the
Limitations and Conditions associated with the NRC SE will be met.
CENPD-132, Supplement 4-P-A, Addendum 1-P and Final Safety
Evaluation
The addendum provides an optional steam cooling model that can
be used for ECCS Performance analyses to support the implementation
of the CE 16 x 16 NGF fuel assembly design. Entergy has demonstrated
that the Limitations and Conditions associated with the NRC SE will
be met.
Safety Injection Tank Water Level and Containment Average Air
Temperature
The safety injection tank (SIT) system provides a passive means
of adding a large quantity of borated water to the reactor core in
the event of a LBLOCA. The SIT system serves no other purpose.
Reducing the maximum volume will not create any new or different
accidents.
The containment average air temperature ensures that the peak
cladding temperature and cladding oxidation remain within limits
during a LBLOCA. The change in the minimum allowable containment
average temperature does not create any new or different accidents.
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.
Core Operating Limits Report (COLR)
The addition of the following topical reports to the list of
analytical methods referenced in the COLR is administrative in
nature:
WCAP-16500-P and Final Safety Evaluation for Westinghouse
Electric Company (Westinghouse) Topical Report (TR) WCAP-16500-P,
Revision 0, ``CE [Combustion Engineering] 16x16 Next Generation Fuel
[(NGF)] Core Reference Report''
WCAP-12610-P-A and CENPD-404-P-A Addendum 1-A
WCAP-16523-P and Final Safety Evaluation for Westinghouse
Electric Company (Westinghouse) Topical Report (TR), WCAP-16523-P,
``Westinghouse Correlations WSSV and WSSV-T for Predicting Critical
Heat Flux in Rod Bundles with Side-Supported Mixing Vanes''
CENPD-387-P-A
CENPD-132, Supplement 4-P-A, Addendum 1-P and Final Safety
Evaluation for Westinghouse Electric Company (Westinghouse) Topical
Report (TR) CENPD-132 Supplement 4-P-A, Addendum 1-P, ``Calculative
Methods for the CE [Combustion Engineering] Nuclear Power Large Break
LOCA Evaluation Model--Improvement to 1999 Large Break LOCA EM Steam
Cooling Model for Less Than 1 in/sec Core Reflood''
Safety Injection Tank Water Level and Containment Average Air
Temperature
The change to the allowable range for these two parameters does
not reduce a margin of safety. The changes add to the margin of
safety and provide assurance that the peak cladding temperature and
cladding oxidation remain within limits.
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: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Thomas G. Hiltz.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will
County, Illinois
Date of amendment request: July 31, 2007.
Description of amendment request: The proposed amendment would
revise Technical Specification 5.5.2, ``Primary Coolant Sources Outside
Containment,'' to clarify the intent of refueling cycle intervals
(i.e., 18 month intervals) with respect to system integrated leak test
requirements and to add a statement that the provisions of Surveillance
Requirement 3.0.2 are applicable.
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 amendment affects only the interval at which
integrated system leak tests are performed, not the effectiveness of
the integrated system leak test requirements. Revising the
integrated system leak test requirements from ``at refueling cycle
interval or less'' to ``at least once per 18 months'' is considered
to be an administrative change because Braidwood Station, Units 1
and 2, and Byron Station, Units 1 and 2, operate on 18-month fuel
cycles. Incorporation of the allowance to extend the 18-month
interval by 25%, as allowed by Surveillance Requirement (SR) 3.0.2,
does not significantly degrade the reliability that results from
performing the Surveillance at its specified Frequency.
Test intervals are not considered as initiators of any accident
previously evaluated. As a result, the probability of any accident
previously evaluated is not significantly increased by the proposed
amendment. Technical Specification (TS) 5.5.2 continues to require
the performance of periodic integrated system leak tests. Therefore,
accident analysis assumptions will still be verified. As a result,
the consequences of any accident previously evaluated are not
significantly increased.
Based on the above discussion, the proposed changes do not
involve an 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 amendment affects only the interval at which
integrated system leak tests are performed; they do not alter the
design
[[Page 51860]]
or physical configuration of the plant. No changes are being made to
Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2,
that would introduce any new accident causal mechanisms.
Based on this evaluation, the proposed change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed amendment does not change the design or function of
plant equipment. The proposed amendment does not significantly
reduce the level of assurance that any plant equipment will be
available to perform its function.
The proposed amendment provides operating flexibility without
significantly affecting plant operation.
Based on this evaluation, 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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Russell Gibbs.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: June 18, 2007.
Description of amendment request: The proposed amendments would
revise Technical Specification 3.7.5, ``Control Room Area Ventilation
Air Conditioning (AC) System,'' to add an Action Statement for two
inoperable control room area ventilation AC subsystems.
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 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
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 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 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 will
continue to maintain the control room temperature within design
limits. Changes to the Bases or license 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.
Based upon the reasoning above, it appears that the three standards
of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to
determine that the requested amendments involve no significant hazards
consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Russell Gibbs.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: June 27, 2007.
Description of amendment request: The proposed amendment would
remove the operability and surveillance requirements for the drywell
air temperature and suppression chamber air temperature instrumentation
from the Limerick Generating Station (LGS) technical specifications.
This will allow a relocation of these requirements to the LGS technical
requirements manual, a licensee controlled document.
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 failure of the drywell air temperature or suppression
chamber air temperature instrumentation is not assumed to be an
initiator of any analyzed event in the UFSAR [Updated Final Safety
Analysis Report]. The proposed changes do not alter the physical
design of this instrumentation or any other plant structure, system,
or component. The proposed changes relocate the drywell air
temperature and suppression chamber air temperature instrumentation
operability and surveillance requirements from the Limerick
Generating Station (LGS) Technical Specifications (TS) to a
licensee-controlled document under the control of 10 CFR 50.59
[Title 10 of the Code of Federal Regulations (10 CFR) Part 50,
Section 50.59].
The proposed changes conform to NRC regulatory requirements
regarding the content of plant TS as identified in 10 CFR 50.36, and
also the guidance as approved by the NRC in NUREG-1433, ``Standard
Technical Specifications-General Electric Plants, BWR/4.''
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 relocate the drywell air temperature and
suppression chamber air temperature instrumentation operability and
surveillance requirements from the LGS TS to a licensee-controlled
document under the control of 10 CFR 50.59. The proposed
[[Page 51861]]
changes do not alter the physical design, safety limits, or safety
analysis assumptions associated with the operation of the plant.
Accordingly, the proposed changes do not introduce any new accident
initiators, nor do they reduce or adversely affect the capabilities
of any plant structure, system, or component in the performance of
their safety function.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The subject instrumentation does not provide primary information
required to permit operators to take specific manually controlled
actions for which no automatic control is provided, and that are
required for safety systems to accomplish their safety functions for
design basis accident events. The instrumentation provides only
drywell air temperature indication and suppression chamber air
temperature indication, and does not provide an input to any
automatic safety function. Operability and surveillance requirements
will be established in a licensee-controlled document to ensure the
reliability of drywell air temperature and suppression chamber air
temperature instrumentation capability. Changes to these
requirements will be subject to the controls of 10 CFR 50.59,
providing the appropriate level of regulatory control.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Bradley Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Harold K. Chernoff.
FirstEnergy Nuclear Operating Company, et. al., Docket No. 50-346,
Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of amendment request: April 12, 2007.
Description of amendment request: The proposed amendment request
would make the operating license and technical specification changes
necessary to allow an increase in the rated thermal power from 2772
megawatts thermal (MWt) to 2817 MWt (approximately 1.63 percent), based
on the use of Caldon, Inc. Leading Edge Flow Meter
CheckPlusTM System instrumentation to improve the accuracy
of the plant power calorimetric measurement.
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.
Under contract to the FirstEnergy Nuclear Operating Company,
AREVA NP Inc. performed evaluations of the Davis-Besse Nuclear Power
Station (DBNPS) Nuclear Steam Supply System (NSSS) and balance of
plant systems, components, and analyses that could be affected by
the proposed change to the licensed power level. A power uncertainty
calculation was performed and the effect of increasing core thermal
power by 1.63 percent to 2817 MWt on the DBNPS design and licensing
basis was evaluated. The evaluations determined that all structures,
systems and components will continue to be capable of performing
their design function at the proposed uprated power level of 2817
MWt. An evaluation of the accident analyses demonstrates that the
applicable analysis acceptance criteria continue to be met with the
proposed changes. No accident initiators are affected by the power
uprate and no challenges to any plant safety barriers are created by
any of the proposed changes.
The proposed change to the licensed power level does not affect
the release paths, the frequency of release, or the analyzed source
term for any accidents previously evaluated in the DBNPS Updated
Final Safety Analysis Report (UFSAR). Systems, structures, and
components required to mitigate transients will continue to be
capable of performing their design functions with the proposed
changes, and thus were found acceptable. The reduced uncertainty in
the power calorimetric measurement ensures that applicable accident
analyses acceptance criteria will continue to be met with operation
at the proposed power level of 2817 MWt. Analyses performed to
assess the effects of mass and energy remain valid. The source term
used to assess radiological consequences has been reviewed and
determined to bound operation at the proposed power level.
The proposed change to the RPS high flux setpoint Allowable
Value does not alter the typical manner in which systems or
components are operated, and, therefore, will not result in an
increase in the probability of an accident. The proposed High Flux
Trip Allowable Values preserve assumptions of current accident
analyses at the higher thermal power allowed by the proposed
amendment, irrespective of the source of Heat Balance calculation
input data. This proposed change does not alter any assumption
previously made in the radiological consequence evaluations, nor
does it affect mitigation of the radiological consequences of an
accident previously evaluated. Therefore, this proposed change does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The addition of references to Note 10 to Functional Unit 2, High
Flux, in Table 4.3-1 is administrative and does not impact the
probability or consequences of an accident previously evaluated
because its inclusion does not involve an accident initiator or
impact any radiological analyses. This change is made to incorporate
NRC guidance in a manner previously determined to be acceptable in
DBNPS License Amendment No. 274.
The proposed change to the volume of the condensate storage
tanks does not alter the typical manner in which the system or
component is operated, and, therefore, will not result in a
significant increase in the probability of an accident. The
condensate storage tanks are not accident initiators. The proposed
change preserves the assumptions previously made in the radiological
consequence evaluations and the radiological consequences of
accidents previously evaluated. Therefore, this proposed change does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes to the Core Operating Limits Report (COLR)
portion of the Administrative Controls Section of the TS are
administrative and do not impact the probability or consequences of
an accident previously evaluated because their inclusion do not
involve accident initiators or impact any radiological analyses.
These changes are made to include the NRC-approved documents
pertaining to the Caldon Leading Edge Flow Meter.
In summary, none of the proposed changes 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.
No new accident scenarios, failure mechanisms, or single
failures are introduced as a result of any of the proposed changes.
Use of the Caldon CheckPlus\TM\ System has been analyzed, and
failures of the system will have no adverse effect on any safety-
related system or any systems, structures, and components required
for transient mitigation. Systems, structures, and components
previously required for the mitigation of a transient continue to be
capable of fulfilling their intended design functions. The proposed
changes have no significant adverse affect on any safety-related
structures, systems or components and do not significantly change
the performance or integrity of any safety-related system.
The proposed changes do not adversely affect any current system
interfaces or create any new interfaces that could result in an
accident or malfunction of a different kind than previously
evaluated. Operating at a core power level of 2817 MWt does not
create any new accident initiators or precursors. The reduced
uncertainty in the power calorimetric measurement ensures that
applicable accident analyses acceptance criteria continue to be met,
to support
[[Page 51862]]
operation at the proposed core power level of 2817 MWt. Credible
malfunctions continue to be bounded by the current accident analyses
of record or recent evaluations that demonstrate that applicable
criteria will continue to be met with the proposed changes.
The proposed change to the RPS high flux setpoint Allowable
Value does not introduce new accident scenarios, failure mechanisms
or single failures. The change does not alter the manner in which
plant systems or components are operated. The proposed High Flux
Trip Allowable Values preserve assumptions of current accident
analyses at the higher thermal power allowed by the proposed
amendment, irrespective of the source of Heat Balance calculation
input data. Therefore, this proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The addition of a reference to Note 10 to Functional Unit 2,
High Flux, in Table 4.3-1 is administrative and will not create the
possibility of a new or different kind of accident from any accident
previously evaluated because its inclusion will not change the
manner in which any equipment is operated. The proposed change to
the volume of the condensate storage tanks does not introduce new
accident scenarios, failure mechanisms or single failures.
Therefore, this proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes to the COLR portion of the Administrative
Controls Section of the TS are administrative and will not create
the possibility of a new or different kind of accident from any
accident previously evaluated because their inclusion will not
change the manner in which any equipment is operated.
In summary, none of the proposed changes will 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 margins of safety associated with the power uprate are those
pertaining to core thermal power. These include those associated
with the fuel cladding, Reactor Coolant System pressure boundary,
and containment barriers. An engineering evaluation of the proposed
1.63 percent increase in core thermal power was performed. The power
uprate required revised NSSS design thermal and hydraulic parameters
to be established to serve as the basis for all of the NSSS analyses
and evaluations. This engineering review identified the design
modifications necessary to accommodate the revised NSSS design
conditions. Evaluations determined that the NSSS systems and
components will continue to operate satisfactorily at the uprated
power level with these modifications and the proposed changes. The
NSSS accident analyses were evaluated at the uprated power level. In
all cases, the evaluations demonstrate that the applicable analyses
acceptance criteria will continue to be met with approval of the
proposed changes. As such, the margins of safety will continue to be
bounded by the analyses for all the changes being proposed.
Therefore, none of the proposed changes will 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: David W. Jenkins, Attorney, FirstEnergy
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: Russell Gibbs.
Florida Power Corporation, et. al., Docket No. 50-302, Crystal River
Unit 3 Nuclear Generating Plant (CR-3), Citrus County, Florida
Date of amendment request: April 25, 2007, as supplemented by
letter dated June 28, 2007.
Description of amendment request: The proposed amendment would
change the operating license and technical specifications to increase
the maximum power level from 2568 megawatts thermal (MWt) to 2609 MWt.
The approximately 1.6 percent increase in power level would be achieved
by use of the Caldon Leading Edge Flowmeter CheckPlus system to
accurately measure power level.
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 not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed change will increase the maximum core power level
from 2568 MWt to 2609 MWt. This increase will only require
adjustments and calibrations of existing plant instrumentation and
control systems. The only equipment upgrades necessary for this
uprate are spool pieces containing multiple ultrasonic flow
instruments, which will be installed in each feedwater line, as well
as more accurate instrumentation for feedwater pressure and steam
pressure and temperature. Indication and control functions will
continue to be performed by the currently installed feedwater
instrumentation.
Nuclear steam supply systems (NSSS) and balance-of-plant (BOP)
systems and components that could be affected by the proposed change
have been evaluated using revised NSSS design parameters based on a
core power level of 2609 MWt. The results of these evaluations,
which used well-defined analysis input assumptions/parameter values
and currently approved analytical techniques, indicate that CR-3
systems and components will continue to function within their design
parameters and remain capable of performing their required safety
functions at 2609 MWt. Since the revised NSSS parameters remain
within the design conditions of the Reactor Coolant System (RCS)
functional specification, the proposed change will not result in any
new design transients or adversely affect the current CR-3 design
transient analyses.
The accidents analyzed in Chapter 14 of the CR-3 Final Safety
Analysis Report (FSAR) have been reviewed for the impact of the
uprate. Based on the power levels assumed in the current safety
analyses, it has been determined that all FSAR and supporting
analyses bound the uprate. This includes the dose calculations for
the design basis radiological accidents, which assume a power level
of 2619 MWt (2568 MWt plus an assumed 2 percent measurement
uncertainty). Since the proposed change relies on less than 0.4%
uncertainty, the assumed power level of 100.4% of 2609 MWt remains
2619 MWt. Therefore, analyses performed at this power remain
bounding.
(2) Does not create the possibility of a new or different kind
of accident from any accident previously evaluated.
As discussed above, the only equipment upgrades necessary for
this uprate are spool pieces containing multiple ultrasonic flow
instruments, which will be installed in each feedwater line, as well
as more accurate instrumentation for feedwater pressure and steam
pressure and temperature. All CR-3 systems and components will
continue to function within their design parameters and remain
capable of performing their required safety functions. The proposed
change does not impact current CR-3 design transients or introduce
any new transients. Equipment failure modes are expected to be the
same as for existing instruments. Protective and control functions
will continue to be performed by the currently installed feedwater
instrumentation. Therefore, the proposed change will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
(3) Does not involve a significant reduction in a margin of
safety
Challenges to the fuel, RCS pressure boundary and containment
were evaluated for uprate conditions. Core analyses show that the
implementation of the power uprate will continue to meet the current
nuclear design basis. Impacts to components associated with RCS
pressure boundary structural integrity, and factors such as
pressure/temperature limits, vessel fluence, and pressurized thermal
shock (PTS) were determined to be bounded by current analyses.
As discussed above, all systems will continue to operate within
their design parameters and remain capable of performing their
intended safety functions following implementation of the proposed
change. Finally, the current CR-3 safety analyses, including the
design basis radiological accident dose calculations, bound the
uprate.
[[Page 51863]]
Therefore, this change does not involve a significant reduction in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602.
NRC Branch Chief: Thomas H. Boyce.
Nine Mile Point Nuclear Station (NMPNS), LLC, Docket No. 50-220, Nine
Mile Point Nuclear Station Unit No. 1 (NMP1), Oswego County, New York
Date of amendment request: July 12, 2007.
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) requirements related to control
room envelope (CRE) habitability in TS 3.4.5, ``Control Room Air
Treatment System,'' and TS 6.5, ``Programs and Manuals.'' The proposed
changes are consistent with TS Task Force (TSTF) change TSTF-448,
Revision 3, ``Control Room Habitability.'' The availability of the TS
improvement was published in the Federal Register on January 17, 2007
(72 FR 2022) 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 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 [a] 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 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. 50-220, Nine
Mile Point Nuclear Station Unit No. 1 (NMP1), Oswego County, New York
Date of amendment request: July 23, 2007.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) section 3.1.1, ``Control Rod
System,'' to incorporate a provision that should the rod worth
minimizer (RWM) become inoperable before a reactor startup is commenced
or before the first 12 control rods have been withdrawn, startup would
be allowed to continue. This provision would rely on the RWM function
being performed manually and would require a double check of compliance
with the control rod program by a second licensed operator or other
qualified member of the technical staff. The use of this allowance
would be limited to one startup in the last calendar year.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change allows plant startup to proceed if the RWM
becomes inoperable prior to withdrawing the first 12 control rods.
The relevant design basis accident is the control rod drop accident
(CRDA), which involves multiple failures to initiate the event. This
change does not increase the probability of occurrence of any of the
failures that are necessary for a CRDA to occur. Use of the RWM or
the alternate use of a second qualified individual to ensure the
correct control rod withdrawal sequence is not in itself an accident
initiator, and adding the new startup allowance does not involve any
plant hardware changes or new operator actions that could serve to
initiate a CRDA. The proposed change will have no adverse effect on
plant operation, or the availability or operation of any accident
mitigation equipment. Also, since the control rod program will
continue to be enforced by either the RWM or verification by a
second qualified individual, the initial conditions of the CRDA
radiological consequence analysis presented in the Updated Final
Safety Analysis Report are not affected. Therefore, there will be no
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
[[Page 51864]]
The proposed change does not introduce any new modes of plant
operation and will not result in a change to the design function or
operation of any structure, system, or component that is used for
accident mitigation. The proposed change allows plant startup to
proceed if the RWM becomes inoperable prior to withdrawing the first
12 control rods, with verification of control rod movement in the
correct sequence performed by a second qualified individual. This
change does not result in any credible new failure mechanisms,
malfunctions, or accident initiators not considered in the design
and licensing basis. This change does not affect the ability of
safety-related systems and components to perform their intended
safety functions. Therefore, the proposed change will not create the
possibility of a new or different kind of accident from any
[accident] previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change allows plant startup to proceed if the RWM
becomes inoperable prior to withdrawing the first 12 control rods.
The proposed change will have no adverse effect on plant operation
or equipment important to safety. The relevant design basis accident
is the [CRDA], which involves multiple failures to initiate the
event. The CRDA analysis consequences and related initial conditions
remain unchanged when invoking the proposed change. The plant
response to the CRDA will not be affected and the accident
mitigation equipment will continue to function as assumed in the
accident analysis. Therefore, there will be no significant reduction
in a margin of safety.
The 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 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 12, 2007.
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) requirements related to control
room envelope (CRE) habitability in TS 3.7.2, ``Control Room Envelope
Filtration (CREF) System,'' and TS 5.5, ``Programs and Manuals.'' The
proposed changes are consistent with TS Task Force (TSTF) change TSTF-
448, Revision 3, ``Control Room Habitability.'' The availability of the
TS improvement was published in the Federal Register on January 17,
2007 (72 FR 2022) 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 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 [a] 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 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
revise the technical specifications (TSs) by changing the testing
frequency for drywell spray nozzles specified in TS Surveillance
Requirement (SR) 3.6.1.6.3 from ``10 years'' to ``following maintenance
that could result in nozzle blockage.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change modifies the surveillance requirement (SR)
to verify that the drywell spray nozzles are unobstructed after
maintenance that could introduce material that could result in
nozzle blockage. The spray nozzles are not assumed to be initiators
of any previously analyzed
[[Page 51865]]
accident. Therefore, the proposed change does not increase the
probability of any accident previously evaluated. The spray nozzles
are used in the accident analyses to mitigate design basis
accidents. The revised SR to verify system operability following
maintenance is considered adequate to ensure operability of the
Residual Heat Removal (RHR) Drywell Spray System.
Since the system will still be able to perform its accident
mitigation function, the consequences of accidents 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.
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 revises the SR to verify that the RHR
Drywell Spray System nozzles are unobstructed after maintenance that
could result in nozzle blockage. The change does not introduce a new
mode of plant operation and does not involve physical modification
to the plant. The change will not introduce new accident initiators
or impact the assumptions made in the safety analysis.
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 revises the frequency for performance of the
SR to verify that the RHR Drywell Spray System nozzles are
unobstructed. The frequency is changed from every 10 years to
following maintenance that could result in nozzle blockage. This
requirement, along with the foreign material exclusion program, the
normal environmental conditions for the system, and the remote
physical location of the spray nozzles, provide assurance that the
spray nozzles will remain unobstructed. As the spray nozzles are
expected to remain unobstructed and able to perform their post-
accident mitigation function, plant safety is not significantly
affected.
Therefore, the proposed change does not involve a significant
reduction in a 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 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.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant (PBNP), Units 1 and 2, Town of Two Rivers,
Manitowoc County, Wisconsin
Date of amendment request: June 29, 2007.
Description of amendment request: The proposed amendments would
modify the Technical Specifications (TSs) 3.7.2, by removing the
specific isolation time for the main steam isolation valves from the
associated TS Surveillance Requirements (SRs) and by replacing it with
the requirement to verify the valve isolation time is within limits.
The changes are consistent with Nuclear Regulatory Commission (NRC)
approved Industry/Technical Specification Task Force (TSTF)-491,
``Removal of the Main Steam and Main Feedwater Valve Isolation Time
from Technical Specifications,'' Revision 2. The proposed amendments
deviate from TSTF-491 in that the current PBNP TS 3.7.3, and associated
SRs do not include the main feedwater valve closure times, and thus
TSTF-491 changes to TS 3.7.3 are not applicable to the PBNP TSs.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on October 5, 2006 (71 FR 58884), on possible
amendments concerning the Consolidated Line Item Improvement Process
(CLIIP), including a model safety evaluation and a model no significant
hazards consideration determination. The NRC staff subsequently issued
a notice of availability of the models for referencing in license
amendment applications in the Federal Register on December 29, 2006 (71
FR 78472) as part of the CLIIP. In its application dated June 29, 2007,
the licensee affirmed the applicability of the following determination.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change allows relocating main steam [ ] valve
isolation times to the Licensee Controlled Document that is
referenced in the Bases. The proposed change is described in
Technical Specification Task Force (TSTF) Standard TS Change
Traveler TSTF-491 related to relocating the main steam [ ] valve
isolation times to the Licensee Controlled Document that is
referenced in the Bases and replacing the isolation time with the
ph[r]ase, ``within limits.''
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 relocate the main steam [ ] isolation valve
times to the Licensee Controlled Document that is referenced in the
Bases. The requirements to perform the testing of these isolation
valves are retained in the TS. 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 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 relocate the main steam [ ] valve isolation
times to the Licensee Controlled Document that is referenced in the
Bases. In addition, the valve isolation times are replaced in the TS
with the ph[r]ase ``within limits.'' 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
testing of the main steam [ ] isolation valves to ensure the proper
functioning of these isolation valves.
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 relocate the main steam [ ] valve isolation
times to the Licensee Controlled Document that is referenced in the
Bases. In addition, the valve isolation times are replaced in the TS
with the ph[r]ase ``within limits.'' Instituting the proposed
changes will continue to ensure the testing of main steam [ ]
isolation valves. Changes to the Bases or license controlled
document are performed in accordance with 10 CFR 50.59. This
approach provides an effective level of regulatory control and
ensures that main steam [ ] isolation valve testing is
[[Page 51866]]
conducted such that there is no significant reduction in the margin
of safety.
The margin of safety provided by the isolation valves is
unaffected by the proposed changes since there continue to be TS
requirements to ensure the testing of main steam [ ] isolation
valves. 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: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Acting Branch Chief: Travis L. Tate.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: August 15, 2007.
Description of amendment request: The proposed amendment would
revise the licensing basis, as described in Appendix 3A of the Salem
Updated Final Safety Analysis Report (UFSAR), regarding the method of
calculating the net positive suction head available (NPSHa) for the
emergency core cooling system (ECCS) and containment heat removal
system pumps. The proposed change relates to issues associated with
Generic Letter 2004-02, ``Potential Impact of Debris Blockage on
Emergency Recirculation During Design Basis Accidents at Pressurized-
Water Reactors.''
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 change in NPSH methodology for ECCS pumps allows the use of
initial containment air pressure in calculating NPSHa. Although this
change is a non-conservative change in the Salem methodology for
calculation of RHR [residual heat removal] pump NPSHa during post
LOCA [loss-of-coolant accident] recirculation (per 10 CFR
50.59(c)(1)(viii) [Title 10 of the Code of Federal Regulations, Part
50, Section 50.59(c)(1)(viii)]), the proposed new methodology is in
accordance with NPSHa calculation methodologies provided in Safety
Guide 1, Regulatory Guides [RG] 1.1, and 1.82, and the guidance of
NEI [Nuclear Energy Institute] 04-07, [``]Pressurized Water Reactor
Sump Performance Evaluation Methodology[,''] (GSI [generic safety
issue]--191) and accompanying SER [safety evaluation report]. The
containment air pressure value used in the NPSHa calculation is
based on the containment conditions prior to the accident only and
does not include any credit for accident pressure conditions, is
conservatively determined based on minimum containment initial
pressure, and maximum temperature and relative humidity conditions.
In addition, the vapor pressure term for the sump water being pumped
is also included in the NPSHa equation, and the value chosen for the
NPSHa calculation is based on the highest temperature of the sump
fluid for the condition being evaluated. This, in conjunction with
the more rigorous GSI-191 analyses, provides assurance that the ECCS
pumps can perform their design function. Consequently, the ECCS
pumps will continue to perform their design function and there is no
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 ECCS pumps take suction from the containment sump during the
recirculation phase of the LOCA to provide long term core cooling.
This system is not utilized during normal operation of the plant.
Therefore, it does not cause initiation of any accident.
However, the ECCS pumps will continue to perform their design
function during the recirculation phase. Crediting initial
containment air pressure in the NPSH methodology does not create any
new or different kind of accident from any accident previously
evaluated. This change removes an additional conservatism built into
the original methodology. By changing the UFSAR described
methodology to credit the containment initial air pressure in the
RHR pump NPSHa calculation, a more realistic methodology is
established. The sole purpose of the additional conservatism was to
ensure credit was not taken for post-LOCA pressure. The revised
methodology continues to meet this requirement.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change removes conservatism from the existing UFSAR
methodology. However, the purpose of the conservatism (equating
containment pressure to sump vapor pressure) was solely to ensure
that no credit was taken for transient (post-LOCA) pressure in the
NPSHa calculation. The purpose was not to deny credit for initial
containment air pressure. Consequently, removing the conservatism
does not alter the basic intent of the NPSH methodology per RG 1.1
requirements, and is consistent with the requirements of RG 1.82,
Revision 1 and NEI 04-07. This change to include a containment air
pressure value establishes a more realistic methodology that still
encompasses adequate conservatisms; no credit is given for the
higher accident pressure conditions, and the value is conservatively
determined based on minimum initial containment air pressure and
maximum temperature and relative humidity conditions. In addition,
the vapor pressure term for the sump water being pumped is also
added to the NPSHa equation, and the value chosen for the NPSHa
calculation is based on the highest temperature of the sump fluid
for the condition being evaluated. Consequently, this change does
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Harold K. Chernoff.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: August 20, 2007.
Description of amendment request: The amendment would increase the
minimum volume of fuel required for the emergency diesel generators
(EDGs) in Technical Specification (TS) 3.8.3, ``Diesel Fuel Oil, Lube
Oil, and Starting Air.''
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 to the minimum required fuel oil volume
required in the EDG storage tanks have no impact on the frequency of
occurrence of any of the accidents evaluated in the FSAR [Final
Safety Analysis Report for Callaway]. Changing the minimum required
fuel oil volume in the EDG fuel oil storage tank has no impact on
the likelihood of occurrence of a loss of coolant accident (LOCA),
line break, plant transient, loss of offsite power, or any such
accident because the precursors for such accidents do not involve
the fuel oil storage tanks.
The EDGs are designed to provide [alternating current]
electrical power to systems required for mitigating the effects of
accidents in the event of a loss of the
[[Page 51867]]
preferred (offsite) power source (i.e., from the grid). However, the
failure or malfunction of an EDG (due, for example, to a loss or
interruption of [the] fuel oil supply) is not itself an initiator of
any accident previously evaluated.
Based on these considerations, the proposed changes have no
impact on the probability of occurrence of any accident evaluated in
the FSAR, and therefore the proposed changes do not involve a
significant increase in the probability of an accident previously
evaluated.
With respect to the consequences of postulated accidents
addressed in [the] FSAR, the support function provided by the EDGs
for accident mitigation is not affected by the proposed TS changes.
[The proposed changes are to provide additional margin for
precluding adverse effects that could result from air entrapment
caused by a vortex condition during fuel oil transfer pump operation
and, thus, to ensure that the EDG has sufficient fuel oil to provide
its support function when needed.] Each of the diesel fuel oil
storage tanks has adequate excess capacity to more than accommodate
a slight increase in the usable volume of fuel oil contained
therein. Thus, even with this increase, the tanks will still be
fully capable of storing the required fuel oil volume needed to
ensure EDG operation throughout the assumed duration of an accident.
At the same time, the proposed changes to TS 3.8.3 will serve to
ensure that the unusable volume in the tanks provides adequate
margin against potentially adverse vortex effects (by precluding the
potential for air ingestion into the fuel oil transfer pumps). On
this basis, the proposed changes have no impact on the capability of
the EDGs to perform their required mitigation/support function for
accidents involving a loss of offsite power. Since the proposed
changes have no impact on accident mitigation capability, they
involve no increase in the consequences of any accident evaluated in
the FSAR.
Based on the above, 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 involve a slight change to the minimum fuel
oil volume required for the EDGs, but they do not involve hardware
changes or changes to EDG operation or testing that would create any
new failure modes for the EDGs or any other [safety-related] system
or component, or that would adversely affect plant operation. The
changes do not involve the addition of any new equipment. No changes
to accident assumptions, including any new limiting single failures,
are involved. With respect to the proposed changes, the plant will
continue to be operated within the envelope of the existing safety
analyses.
Therefore, based on the above, the proposed changes do not
create [the possibility of] a new or different kind of accident
[from any accident] previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is related to the confidence in the ability
of the fission product barriers to perform their design functions
during and following an accident situation. These barriers include
the fuel cladding, the reactor coolant system, and the containment
system. The proposed changes do not directly affect these barriers,
nor do they involve or cause any adverse impact on the EDGs which
serve to support these barriers in the event of an accident
concurrent with a loss of offsite power.
[The margin of safety is also related to the ability of the
safety-related systems to perform their safety function as described
in the safety analyses in the FSAR. The proposed changes are to
provide additional margin for precluding adverse effects that could
result from air entrapment caused by a vortex condition during fuel
oil transfer pump operation and, thus, to ensure that the EDG has
sufficient fuel oil to provide its support function when needed.
Therefore, the proposed changes are to increase margin for the
EDGs.]
The proposed changes do not alter the manner in which safety
limits or limiting safety system settings are determined, nor is
[the] basis of any limiting condition for operation changed or
affected. The safety analysis acceptance criteria are not impacted
by these changes. The proposed changes will not result in plant
operation in a configuration outside the design basis.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR
50.92(copyright) are satisfied. Therefore, the NRC staff proposes to
determine that the amendment request involves no significant hazards
consideration.
Attorney for licensee: John O'Neill, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: Thomas G. Hiltz.
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 51.22(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].
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 application for amendment: March 1, 2007.
Brief description of amendment: The amendment revised the Grand
Gulf Nuclear Station, Unit 1 (GGNS) Technical Specification (TS) to add
a note to the Required Actions of TS 3.6.1.3, ``Primary Containment
Isolation Valves (PCIVs)''. GGNS TS 3.6.1.3 requires specific actions
to be taken for inoperable PCIVs. The TS Required Actions include
isolating the affected penetration by use of a closed and deactivated
automatic valve, closed manual valve, blind flange, or check valve with
flow through the valve secured. The new note would allow a relief valve
to be used to comply with
[[Page 51868]]
TS 3.6.1.3, Actions A.1 and B.1 without being deactivated provided it
has a relief setpoint of at least 1.5 times containment design pressure
(i.e., at least 23 pounds per square inch gauge) and meets one of the
following criteria:
1. The relief valve is one-inch nominal size or less, or
2. The flow path is into a closed system whose piping pressure
rating exceeds the containment design pressure rating.
Date of issuance: August 24, 2007.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment No: 176.
Facility Operating License No. NPF-29: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: April 24, 2007 (72 FR
20382).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 24, 2007.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian
Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New
York
Date of application for amendment: September 25, 2006, as
supplemented March 12, 2007.
Brief description of amendment: Entergy Nuclear Operations, Inc.,
requested an amendment to make editorial changes to the Technical
Specifications of Indian Point Nuclear Generating Unit Nos. 2 and 3.
The editorial changes consist of typographical corrections, update of
references, and deletion of obsolete notes.
Date of issuance: August 16, 2007.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 252 and 234.
Facility Operating License Nos. DPR-26 and DPR-64: The amendment
revised the License and the Technical Specifications.
Date of initial notice in Federal Register: November 7, 2006 (71 FR
65142).
The March 12, 2007, supplement provided additional information that
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the NRC staff's original
proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 16, 2007.
No significant hazards consideration comments received: No.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: August 7, 2006, as supplemented by
letters dated January 22, May 14, and August 7, 2007.
Description of amendment request: The amendment revises the
Seabrook Technical Specifications (TSs) to correct a joint-owner name
in the operating license, remove a license condition from Appendix C to
the FOL, and remove the list of Bases sections from the TS Index.
Additionally, the amendment removes two manual valves from TS table
3.3-9, ``Remote Shutdown System,'' adds the requirement that only one
charging pump is permitted to be aligned for injection into the reactor
coolant system in Modes 4, 5, and 6, removes a 1-hour reporting
requirement for portable makeup pump system storage from TS 3.7.4,
``Service Water System/Ultimate Heat Sink,'' deletes a footnote from TS
3.7.6.2, ``Air Conditioning,'' and modifies TS 6.7.6, ``Radioactive
Effluent Controls Program.''
Date of issuance: August 23, 2007.
Effective date: As of its date of issuance, and shall be
implemented within 90 days.
Amendment No.: 116.
Facility Operating License No. NPF 86: The amendment revised the
License and Technical Specification.
Date of initial notice in Federal Register: June 5, 2007 (72 FR
31101).
The licensee's January 22, May 14, and August 7, 2007, supplements
provided clarifying information that did not change the scope of the
proposed amendment as described in the original notice of proposed
action published in the Federal Register, and did not change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 23, 2007.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of application for amendments: August 14, 2006, as
supplemented by letter dated July 16, 2007.
Brief description of amendments: The amendments make miscellaneous
improvements to the Technical Specifications (TSs) for Prairie Island
Nuclear Generating Plant, Units 1 and 2. The amendments revise the
wording in the section headers in TS 1.3, ``Completion Times''; remove
an unnecessary Note in TS 3.1.4, ``Rod Group Alignment Limits''; remove
applicable modes in TS 3.3.7, ``Spent Fuel Pool Special Ventilation
System (SFPSVS) Actuation Instrumentation''; add reference to a TS
Condition to clarify the requirements of TS 3.7.10, ``Control Room
Special Ventilation System (CRSVS)''; and update a reference in TS 4.0,
``Design Features.''
Date of issuance: August 10, 2007.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 180 & 170.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the TSs.
Date of initial notice in Federal Register: November 21, 2006 (71
FR 67397).
The supplemental letter contained clarifying information and did
not change the initial no significant hazards consideration
determination, and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 10, 2007.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of application for amendments: September 26, 2006, as
supplemented on May 14, 2007.
Brief description of amendments: The amendments revise Technical
Specification (TS) 6.9.1.9, ``Core Operating Limits Report (COLR),'' to
remove the revision numbers and dates from the list of topical reports
that contain the analytical methods used in the COLR. The Salem Unit 2
amendment also adds a new topical report to the list of COLR methods
referenced in TS 6.9.1.9.
Date of issuance: August 23, 2007.
Effective date: The license amendments are effective as of the date
of issuance. The Salem Unit 1 amendment shall be implemented prior to
restart from the 19th refueling outage in fall 2008. The Salem Unit 2
amendment shall be implemented prior to restart from the 16th refueling
outage in spring 2008.
Amendment Nos.: 284 and 267.
[[Page 51869]]
Facility Operating License Nos. DPR 70 and DPR-75: The amendments
revised the TSs and the License.
Date of initial notice in Federal Register: November 7, 2006 (71 FR
65143).
The supplement dated May 14, 2007, provided additional information
that clarified the application, did not expand the scope of the
application as originally noticed, and did not change the NRC staff's
original proposed no significant hazards consideration determination as
published in the Federal Register on November 7, 2006 (71 FR 65143).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 23, 2007.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: November 15, 2006, as
supplemented by letters dated June 21, and August 23, 2007.
Brief description of amendment: The amendment deletes Technical
Specification (TS) Table 3.6.3-1, ``Primary Containment Isolation
Valves,'' and relocates the information to the Hope Creek Generating
Station Technical Requirements Manual (TRM). The amendment also revises
other TS sections that reference TS Table 3.6.3-1.
Date of issuance: August 27, 2007.
Effective date: As of the date of issuance, to be implemented
within 90 days. Implementation shall include the relocation of
information from the TSs to the TRM as described in the licensee's
application dated November 15, 2006, and letters dated June 21, and
August 23, 2007.
Amendment No.: 171.
Facility Operating License No. NPF-57: The amendment revised the
TSs and the License.
Date of initial notice in Federal Register: February 13, 2007 (72
FR 6789).
The supplements dated June 21, and August 23, 2007, provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination or expand the
application beyond the scope of the original Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 27, 2007.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 5th day of September, 2007.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing Office of Nuclear
Reactor Regulation.
[FR Doc. E7-17864 Filed 9-10-07; 8:45 am]
BILLING CODE 7590-01-P