[Federal Register Volume 72, Number 78 (Tuesday, April 24, 2007)]
[Notices]
[Pages 20375-20389]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-7534]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a determination by the Commission that such
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from March 30, 2007 to April 12, 2007. The last
biweekly notice was published on April 10, 2007 (72 FR 17944).
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
[[Page 20376]]
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 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
[[Page 20377]]
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of amendment request: March 22, 2007.
Description of amendment request: The proposed amendment would
revise the Technical Specifications to incorporate a revised limit for
the variable low reactor coolant system pressure-temperature core
protection safety limit. The revised limit is associated with the
introduction of AREVA NP's Mark-B-HTP fuel design, which will require
more restrictive Safety Limits and more restrictive Limiting Safety
System Settings for the Reactor Protection System. The proposed limits
are developed in accordance with the method described in the Nuclear
Regulatory Commission (NRC)-approved Topical Report BAW-10179P-A,
``Safety Criteria and Methodology for Acceptable Cycle Reload
Analyses.'' The revised limits will maintain the same magnitude of
departure from nucleate boiling (DNB) protection.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed Technical Specification (TS) limits and reactor
protection system (RPS) trip setpoints are developed in accordance
with the methods and assumptions described in NRC-approved AREVA NP
Topical Reports BAW-10179 P-A, ``Safety Criteria and Methodology for
Acceptable Cycle Reload Analyses'' and BAW-10187 P-A, ``Statistical
Core Design for B&W-Designed 177 FA Plants.'' The core thermal-
hydraulic code (LYNXT) and CHF correlation (BHTP) have been approved
for use with these methods and the Mark-B-HTP fuel type. The
proposed change preserves the design DNB Ratio safety criterion that
there shall be at least a 95% [percent] probability at a 95%
confidence level that the hot fuel rod in the core does not
experience a departure from nucleate boiling during normal operation
or events of moderate frequency. The corresponding core-wide
protection on a pin-by-pin basis is greater than 99.9%. The margin
retained for penalties such as transition core effects, by imposing
a Thermal Design Limit in all DNB analyses supporting the proposed
change, has been shown to be sufficient to offset the mixed core
conditions at TMI Unit 1, where the Mark-B-HTP fuel design will be
co-resident with earlier Mark-B fuel designs. The setpoint
calculation methodology utilized, and the surveillance requirements
established, are in accordance with approved industry standards and
NRC criteria.
The proposed setpoint change does not involve a significant
increase in the consequences of an accident previously evaluated
because the proposed change does not alter any assumptions
previously made in the radiological consequence evaluations, or
affect mitigation of the radiological consequences of an accident
previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed TS limit and reactor protection system (RPS) trip
setpoint provide a core protection safety limit and variable low
pressure trip setpoint developed in accordance with NRC-approved
methods and assumptions. No new accident scenarios, failure
mechanisms or single failures are introduced as a result of the
proposed change. All systems, structures, and components previously
required for the mitigation of an event remain capable of fulfilling
their intended design function.
Therefore, 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 amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed RPS trip setpoint ensures core protection safety
limits will be preserved during power operation. The proposed safety
limit and setpoint are developed in accordance with NRC-approved
methods and assumptions. The margin retained for penalties such as
transition core effects, by imposing a Thermal Design Limit in all
DNB analyses supporting the proposed change, has been shown to be
sufficient to offset the mixed core conditions at TMI Unit 1. The
setpoint calculation methodology utilized, and the surveillance
requirements established, are in accordance with approved industry
standards and NRC criteria.
Therefore, the proposed changes do not involve a significant
reduction in any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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. Brad Fewell, Assistant General Counsel,
Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA
19348.
NRC Branch Chief: Harold K. Chernoff.
Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-317, Calvert
Cliffs Nuclear Power Plant, Unit No. 1, Calvert County, Maryland
Date of amendment request: February 27, 2007.
Description of amendment request: The proposed license amendment
would revise Technical Specification 4.2.1, Fuel Assemblies, to add a
temporary exemption to allow the insertion of up to four lead fuel
assemblies, which contain non-Zircaloy based cladding, into the Unit 1
core for one cycle of operation. These lead fuel assemblies are
currently installed in the Unit 2 core under a previous exemption and
are scheduled to be discharged during the 2007 refueling outage.
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. The licensee has determined
that the proposed change:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
[[Page 20378]]
Calvert Cliffs Technical Specification 4.2.1, Fuel Assemblies,
states that fuel rods are clad with either Zircaloy or
ZIRLOTM. Calvert Cliffs Nuclear Power Plant, Inc.
proposes to re-insert up to four fuel assemblies into Calvert Cliffs
Unit 1 that have some fuel rods clad in zirconium alloys that do not
meet the definition of Zircaloy or ZIRLOTM. A temporary
exemption to the regulations has been requested to allow these fuel
assemblies to be re-inserted into Unit 1. The proposed change to the
Calvert Cliffs Technical Specifications will allow the use of
cladding materials that are not Zircaloy or ZIRLOTM for
one fuel cycle once the temporary exemption is approved. The
proposed change to the Technical Specification is effective only as
long as the temporary exemption is effective. The addition of what
will be an approved temporary exemption for Unit 1 to Technical
Specification 4.2.1 does not change the probability or consequences
of an accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Would not create the possibility of a new or different [kind]
of accident from any accident previously evaluated.
The proposed change does not add any new equipment, modify any
interfaces with existing equipment, change the equipment's function,
or change the method of operating the equipment. The proposed change
does not affect normal plant operations or configuration. Since the
proposed change does not change the design, configuration, or
operation, it could not become an accident initiator.
Therefore, the proposed change does not create the possibility
of a new or different [kind] of accident from any accident
previously evaluated.
3. Would not involve a significant reduction in [a] margin of
safety.
The proposed change will add an approved temporary exemption to
the Calvert Cliffs Technical Specifications allowing the
installation of up to four lead fuel assemblies. The assemblies use
advanced cladding materials that are not specifically permitted by
existing regulations or Calvert Cliffs' Technical Specifications. A
temporary exemption to allow the installation of these assemblies
has been requested. The addition of an approved temporary exemption
to Technical Specification 4.2.1 is an administrative change to
allow the installation of the lead fuel assemblies under the
provisions of the temporary exemption. The license amendment is
effective only as long as the exemption is effective. This amendment
does not change the margin of safety since it only adds a reference
to an approved, temporary exemption to the Technical Specifications.
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: Carey Fleming, Esquire, Senior Counsel--
Nuclear Generation, Constellation Generation Group, LLC, 750 East Pratt
Street, 17th floor, Baltimore, MD 21202.
NRC Acting Branch Chief: John P. Boska.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendment request: February 27, 2007.
Description of amendment request: The proposed license amendment
would revise Technical Specification 5.6.5, Core Operating Limits
Report (COLR), to add the supporting topical report (WCAP-15604-NP,
Revision 2-A, ``Limited Scope High Burnup Lead Test Assemblies,''
September 2003) to the list of references. The topical report provides
guidance for operation with a limited number of lead fuel assemblies to
be irradiated to a higher burnup limit than currently allowed for
Calvert Cliffs fuel assemblies.
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. The licensee has determined
that the proposed change:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed change would modify the Calvert Cliffs Units 1 and
2 Technical Specification 5.6.5.b, Core Operating Limits Report by
adding an approved topical report to the existing list of topical
reports. The topical report provides the technical basis that
supports irradiating a limited number of lead fuel assemblies to a
higher burnup limit than currently approved for Calvert Cliffs. The
proposed change is administrative in nature and has no impact on any
plant configurations or on system performance that is relied upon to
mitigate the consequences of an accident.
In the safety evaluation report approving the requested topical
report (WCAP-15604-NP, Revision 2-A), the Nuclear Regulatory
Commission concluded that it is acceptable for an individual power
licensee to irradiate a limited number of lead fuel assemblies to a
maximum burnup to 75 GWD/MTU [gigawatt days per metric ton of
uranium] provided that certain conditions are met. Calvert Cliffs
meets those required conditions. Because those required conditions
are met and only a limited number of fuel assemblies are included in
this change, the probability or consequences of an accident
previously evaluated are not significantly increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Would not create the possibility of a new or different [kind]
of accident from any accident previously evaluated.
The proposed change does not add any new equipment, modify any
interfaces with existing equipment, change the equipment's function,
or change the method of operating the equipment. The proposed change
does not affect normal plant operations or configuration. Since the
proposed change does not change the plant design, operation, or
configuration, it could not become an accident initiator.
Therefore, the proposed change does not create the possibility
of a new or different [kind] of accident from any accident
previously evaluated.
3. Would not involve a significant reduction in a margin of
safety.
The proposed change will add a reference to an approved topical
report to allow a limited number of lead fuel assemblies to be
irradiated to a higher burnup level than is currently allowed at
Calvert Cliffs. The higher burnup limit has been evaluated and
approved in the topical report being referenced. Calvert Cliffs
conforms to the requirements of the topical report. The addition of
an approved reference to the Technical Specifications is
administrative in nature and has no impact on the margin of safety
for any plant configuration or on system performance that is relied
upon to mitigate the consequences on an accident.
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: Carey Fleming, Esquire, Senior Counsel--
Nuclear Generation, Constellation Generation Group, LLC, 750 East Pratt
Street, 17th floor, Baltimore, MD 21202.
NRC Acting Branch Chief: John P. Boska.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: February 2, 2007.
Description of amendment request: The proposed amendment deletes
requirements from the Technical Specifications (TS) to maintain
hydrogen recombiners and hydrogen monitors. Licensees were generally
required to implement upgrades as described in NUREG-0737,
``Clarification of Three Mile Island
[[Page 20379]]
(TMI) Action Plan Requirements,'' and Regulatory Guide (RG) 1.97,
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess
Plant and Environs Conditions During and Following an Accident.''
Implementation of these upgrades was an outcome of the lessons learned
from the accident that occurred at TMI, Unit 2. Requirements related to
combustible gas control were imposed by Order for many facilities and
were added to or included in the TS for nuclear power reactors
currently licensed to operate. The revised Title 10 of the Code of
Federal Regulations (10 CFR) 50.44, ``Standards for Combustible Gas
Control System in Light-Water-Cooled Power Reactors,'' eliminated the
requirements for hydrogen recombiners and relaxed safety
classifications and licensee commitments to certain design and
qualification criteria for hydrogen and oxygen monitors.
The NRC staff published a notice of opportunity for comment in the
Federal Register on August 2, 2002 (67 FR 50374), on possible
amendments to eliminate requirements regarding containment hydrogen
recombiners and the removal of requirements from TS for containment
hydrogen and oxygen monitors, including a model safety evaluation and
model No Significant Hazards Consideration (NSHC) Determination, in
accordance with 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 September 25, 2003 (68 FR 55416). The licensee affirmed the
applicability of the model NSHC determination in its application dated
February 2, 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 revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen monitors are no longer required to mitigate design-basis
accidents and, therefore, the hydrogen monitors do not meet the
definition of a safety-related component as defined in 10 CFR 50.2.
RG 1.97 Category 1, is intended for key variables that most directly
indicate the accomplishment of a safety function for design-basis
accident events. The hydrogen monitors no longer meet the definition
of Category 1 in RG 1.97. As part of the rulemaking to revise 10 CFR
50.44 the Commission found that Category 3, as defined in RG 1.97,
is an appropriate categorization for the hydrogen monitors because
the monitors are required to diagnose the course of beyond design-
basis accidents.
The regulatory requirements for the hydrogen monitors can be
relaxed without degrading the plant emergency response. The
emergency response, in this sense, refers to the methodologies used
in ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities.
Classification of the hydrogen monitors as Category 3 and removal of
the hydrogen monitors from TS will not prevent an accident
management strategy through the use of the severe accident
management guidelines, the emergency plan, the emergency operating
procedures, and site survey monitoring that support modification of
emergency plan protective action recommendations.
Therefore, the elimination of the hydrogen recombiner
requirements and relaxation of the hydrogen monitor requirements,
including removal of these requirements from TS, does not involve a
significant increase in the probability or the consequences of any
accident previously evaluated.
Criteria 2: The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident From Any Previously
Evaluated.
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TS, will not result in any failure mode
not previously analyzed. The hydrogen recombiner and hydrogen
monitor equipment was intended to mitigate a design-basis hydrogen
release. The hydrogen recombiner and hydrogen monitor equipment are
not considered accident precursors, nor does their existence or
elimination have any adverse impact on the pre-accident state of the
reactor core or post accident confinement of radionuclides within
the containment building.
Therefore, this change does 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 elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TS, in light of existing plant equipment,
instrumentation, procedures, and programs that provide effective
mitigation of and recovery from reactor accidents, results in a
neutral impact to the margin of safety.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The Commission has found that this
hydrogen release is not risk-significant because the design-basis
LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage.
Category 3 hydrogen monitors are adequate to provide rapid
assessment of current reactor core conditions and the direction of
degradation while effectively responding to the event in order to
mitigate the consequences of the accident. The intent of the
requirements established as a result of the TMI, Unit 2 accident can
be adequately met without reliance on safety-related hydrogen
monitors.
Therefore, this change does not involve a significant reduction
in the margin of safety. Removal of hydrogen monitoring from TS will
not result in a significant reduction in their functionality,
reliability, and availability.
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.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: March 19, 2007.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) 3.8.1 entitled ``AC Sources-
Operating'' to change the minimum Emergency Diesel Generator (EDG)
output voltage acceptance criterion from 3740 to 3873 volts.
Specifically, the proposed change would revise the Surveillance
Requirements (SRs) 3.8.1.2, 3.8.1.7, 3.8.1.10, 3.8.1.11, 3.8.1.14, and
3.8.1.17.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
[[Page 20380]]
1. The proposed change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
The increase in the minimum EDG output voltage acceptance
criterion value in TS 3.8.1 surveillance requirements does not
adversely affect any of the parameters in the accident analyses. The
change increases the minimum allowed EDG output voltage acceptance
criterion to ensure that sufficient voltage is available to operate
the required Emergency Safety Feature (ESF) equipment under accident
conditions. The increase in the minimum allowed EDG output voltage
in the TS surveillance requirements ensures that adequate voltage is
available to support the assumptions made in the Design Bases
Accident (DBA) analyses. DBA analyses assume that onsite standby
emergency power will provide an adequate power source to operate
safe shutdown equipment and to mitigate consequences of design bases
accidents. This conservative change of the acceptance criterion
enhances the testing requirements of the onsite emergency diesel
generators and ensures the reliability of this power source.
Changing the acceptance criterion does not affect the probability of
evaluated accidents and it provides better assurance of EDG
reliability in mitigating consequences of accidents. Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from an accident previously evaluated.
The change in the value of the minimum EDG output voltage
acceptance criterion supports the assumptions in the accident
analyses that sufficient voltage will be available to operate ESF
equipment on the Class 1E buses when these buses are powered from
the onsite emergency diesel generators. The maximum EDG output
voltage of 4580 volts is not affected by this change. The change in
the minimum EDG output voltage from 3740 to 3873 volts ensures the
reliability of the onsite emergency power source. Therefore, the
proposed change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
This proposed license amendment involves a change in the minimum
EDG output voltage acceptance criterion in TS 3.8.1 surveillance
requirements. The surveillance frequency and the different test
requirements are unchanged. The change provides a better assurance
that the onsite power source is able to satisfy the design
requirements assumed in the accident analyses to safely shutdown the
reactor and mitigate the consequences of design bases accidents.
Therefore, the proposed change will 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: David G. Pettinari, Legal Department, 688
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279.
NRC Branch Chief: L. Raghavan.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2, New London County, Connecticut
Date of amendment request: November 8, 2006.
Description of amendment request: The proposed amendment would
modify the Technical Specification (TS) Action and Surveillance
Requirements (SRs) for instrumentation identified in TSs 3.3.1 and
3.3.2. In particular, the proposed amendment adds actions to address
the inoperability of one or more automatic bypass removal channels;
revises the terminology used in the notation of TS Tables 2.2-1 and
3.3-1 relative to the implementation and automatic removal of certain
Reactor Protection System (RPS) trip bypasses; revises the frequency
for performing surveillance of the automatic bypass removal function
logic; and incorporates two administrative changes.
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:
Criterion 1: Does the proposed amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response: No.
The proposed changes to Technical Specifications 2.2.1, 3.3.1
and 3.3.2 do not adversely impact structure, system, or component
design or operation in a manner that would result in a change in the
frequency of occurrence of accident initiation. The proposed
technical specification changes do not involve accident initiators,
do not change the configuration or method of operation of any plant
equipment that is used to mitigate the consequences of an accident,
and do not alter any conditions assumed in the plant accident
analyses. The proposed amendment does not change the function or the
manner of operation of the RPS or ESFAS [engineered safety features
actuation system] trip bypass features. Adding actions to be taken
for an inoperable automatic bypass removal function places
additional restriction on plant operation in this condition and does
not alter the setpoint or the logic of the operating bypasses and
automatic bypass removals. Clarifying the frequency of the SR
associated with testing the automatic bypass removal function does
not alter the setpoint or the manner of operation of the operating
bypasses and automatic bypass removals. More accurately reflecting
the input process variable of the operating bypasses and automatic
bypass removals of the affected reactor trips does not alter the
setpoint nor the manner of operation of the operating bypasses and
automatic bypass removals. With respect to the incorporation of the
administrative changes, the proposed changes are spelling
corrections and do not alter any of the requirements of the affected
TS. Therefore, this change does not impact the consequences of any
accident. Based on this discussion, the proposed amendment does not
increase the probability or consequence of an accident previously
evaluated.
Criterion 2: Does the proposed amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response: No.
No new or different accidents result from clarifying actions for
an inoperable automatic bypass removal function, clarifying
surveillance requirements for the automatic bypass removal function,
and more accurately reflecting the parameter being measured for
automatic bypass removal by referring to logarithmic power, the
input process variable. The results of previously performed accident
analyses remain valid. The proposed amendment does not introduce
accident initiators or malfunctions that would cause a new or
different kind of accident. The proposed amendments are
administrative in nature and will not change the physical plant or
the modes of plant operation defined in the facility operating
license. The changes do not involve the addition or modification of
equipment nor do they alter the design or operation of plant
systems. Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Criterion 3: Does the proposed amendment involve a significant
reduction in a margin of safety?
Response: No.
The proposed change does not alter the function or manner of
operation of the operating bypasses and automatic bypass removals of
the affected reactor trips. The proposed changes do not affect any
of the assumptions used in the accident analysis, nor do they affect
any operability requirements for equipment important to plant
safety. Therefore, the proposed amendment does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT
06385.
[[Page 20381]]
NRC Branch Chief: Harold K. Chernoff.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: January 4, 2007.
Description of amendment request: The proposed amendment would
revise the Technical Specification for Limiting Conditions for
Operation (LCOs) and Surveillance Requirements (SRs) for control rod
operability, scram insertion times, and control rod accumulators. 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 changes extend the frequency and revise the
methodology for testing control rod scram times, and identify a new
category of ``slow'' control rods for assessing control rod
operability. The frequency of control rod scram testing is not an
initiator of any accident previously evaluated. The frequency of
surveillance testing does not affect the ability to mitigate any
accident previously evaluated, because the tested component is still
required to be operable. The proposed test methodology is consistent
with industry approved methods and ensures control rod operability
requirements for the number and distribution of operable, slow, and
stuck control rods continue to satisfy scram reactivity rate
assumptions used in plant safety analysis.
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 changes do not involve any physical alteration of
the plant (no new or different type of equipment is being installed)
and do not involve a change in the design, normal configuration, or
basic operation of the plant. The proposed changes do not introduce
any new accident initiators. The proposed changes do not involve
significant changes in the fundamental methods governing normal
plant operation and do not require unusual or uncommon operator
actions. The proposed changes provide assurance that the plant will
not be operated in a mode or condition that violates the assumptions
or initial conditions in the safety analyses and that SSCs
[structures, systems, and components] remain capable of performing
their intended safety functions as assumed in the same analyses.
Consequently, the response of the plant and the plant operator to
postulated events will not be significantly different.
Therefore, the proposed TS change does 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.
Margin of safety is related to confidence in the ability of the
fission product barriers to perform their design functions during
and following an accident situation. The proposed changes address
control rod scram test performance and acceptance criteria as well
as control rod operability requirements. The scam test acceptance
criteria and control rod operability restrictions are based on
industry approved methodology and will continue to ensure control
rod scram design functions and reactivity insertion assumptions used
in safety analyses continue to be protected. The proposed changes
also extend the frequency of testing control rod scram times while
at-power from 120 days to 200 days. The proposed change ensures
scram testing is performed and that test results verify acceptable
operation of the control rods.
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 50.929(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Travis C. McCullough, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White
Plains, NY 10601.
Branch Chief: John P. Boska (Acting).
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2 (ANO-2), Pope County, Arkansas
Date of amendment request: March 15, 2007.
Description of amendment request: The proposed amendment would
revise containment systems surveillance requirements for Technical
Specification (TS) 3/4.6.2, ``Depressurization, Cooling, and pH Control
Systems.'' The proposed amendment would revise the frequency for ANO-2
TS Surveillance Requirement 4.6.2.1.d to require verification that
spray nozzels are unobstructed following maintenance that could result
in a nozzel blockage (loss of foreign material exclusion control)
rather than performing an air or smoke flow test through each spray
header every 5 years.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do[es] the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The Containment Spray System (CSS) is not an initiator of any
analyzed event. The proposed change does not have a detrimental
impact on the integrity of any plant structure, system, or component
that may initiate an analyzed event. The proposed change will not
alter the operation or otherwise increase the failure probability of
any plant equipment that can initiate an analyzed accident. This
change does not affect the plant design. There is no increase in the
likelihood of formation of significant corrosion products. Due to
their location at the top of the containment, introduction of
foreign material into the spray headers is unlikely. Foreign
materials exclusion controls during and following maintenance
provides assurance that the nozzles remain unobstructed.
Consequently, there is no significant increase in the probability of
an accident previously evaluated.
The CSS is designed to address the consequences of a Loss of
Coolant Accident (LOCA) or a Main Steam Line Break (MSLB). The
Containment Spray System is capable of performing its function
effectively with the single failure of any active component in the
system, any of its subsystems, or any of its support systems.
Therefore, the consequences of an accident previously evaluated are
not significantly affected by the proposed change.
2. Do[es] the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change will not physically alter the plant (no new
or different type of equipment will be installed) or change 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. Do[es] the proposed change involve a significant reduction in
a margin of safety?
Response: No.
The system is not susceptible to corrosion-induced obstruction
or obstruction from sources external to the system. Strict controls
are established to ensure the foreign material is not introduced
into the CSS during maintenance or repairs. Maintenance activities
that could introduce significant foreign material into the system
require subsequent system cleanliness verification which would
prevent nozzle blockage. The spray header nozzles are expected to
remain unblocked and available in the event that the safety function
is required. The capacity of the system would remain unaffected.
Therefore, the proposed change does not involve a reduction in a
margin of safety.
[[Page 20382]]
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., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of amendment request: March 1, 2007.
Description of amendment request: The proposed change would revise
the Grand Gulf Nuclear Station, Unit 1 (GGNS) Technical Specifications
(TS) to add a note to the Required Actions of TS 3.6.1.3, ``Primary
Containment Isolation Valves (PCIVs),'' Actions A.1 and B.1. 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 without being deactivated,
to comply with TS 3.6.1.3, Actions A.1 and B.1, 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 1-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.
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.
Primary Containment Isolation Valves (PCIVs) are accident
mitigating features designed to limit releases from the containment
following an accident. The TS specify actions to be taken to
preserve the containment isolation function if a PClV is inoperable.
These actions include isolating the penetration flow path by
specific methods including, closed and de-activated automatic
valves, closed manual valves, blind flanges, and check valves with
flow through the valve secured. The current TS Actions do not
specifically recognize a closed relief valve as an acceptable method
of isolating a penetration flow path. Thus, special measures may
need to be taken to comply with the TS Required Actions, such as
replacing the relief valve with a blind flange or de-activating the
relief valve by installing a gag. While such actions may provide
additional assurance of preserving the containment isolation
function, it may also have adverse safety affects such as disabling
the overpressure protective safety feature, causing additional
safety system unavailability time, and increasing occupational dose.
The proposed change would allow certain relief valves to be used
for isolating the penetration flow path without being de-activated.
The proposed TS changes do not alter the design, operation, or
capability of PCIVs. Relief valves are designed to be normally
closed to preserve the piping boundary integrity yet automatically
open on an abnormal process pressure to protect the piping from
overpressure conditions. Relief valves may also serve as passive
containment isolation devices (i.e., they do not require mechanical
movement to perform the isolation function). The proposed TS changes
preserve both the containment isolation and piping overpressure
protection functions.
The failure of a relief valve to remain closed during or
following an accident is considered a low probability because relief
valves are passive isolation devices that do not require mechanical
movement to perform the isolation function and the relief setpoint
provides sufficient margin to preclude the potential for premature
opening due to containment post-accident pressures. Additional
criteria are established to provide defense-in-depth protection.
Relief valves that are one-inch or smaller provide an additional
physical barrier in that, even in the unlikely event that a relief
valve were to fail to remain fully closed during or following an
accident, the size restriction would limit leakage such that a large
early release would not occur. By definition, penetrations one-inch
and smaller do not contribute to large early releases. Larger relief
valves may be used as isolation devices provided that the
containment penetration flow path through the relief valve would be
contained in a closed system. In the unlikely event that a relief
valve were to fail to remain closed, the leakage would be into a
system which forms a closed loop outside primary containment and any
containment leakage would return to primary containment through this
closed loop.
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 does not introduce any new modes of plant
operation or adversely affect the design function or operation of
safety features. The proposed TS change allows use of existing plant
equipment as compensatory measures to maintain the containment
isolation design intent when the normal isolation features are
inoperable. Since relief valves used for this purpose will not be
disabled by gags or blind flanges, the system piping overpressure
protection design feature will also be preserved.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The safety margin associated with this change is that associated
with preserving the containment integrity. NUREG-0800, the Standard
Review Plan, recognizes that relief valves with relief setpoints
greater than 1.5 times containment design pressure are acceptable as
containment isolation devices. Closed relief valves with relief
setpoints of this margin provide an isolation alternative that is
less susceptible to a single failure (i.e., inadvertent opening) yet
still preserves the overpressure protection that the component was
intended to provide. The failure of a relief valve to remain closed
during or following an accident is considered a low probability
because relief valves are passive isolation devices that do not
require mechanical movement to perform the isolation function and
the relief setpoint provides sufficient margin to preclude the
potential for premature opening due to containment post-accident
pressures. Defense-in-depth containment leakage protection is
provided by additional TS criteria that limit the use of relief
valves to those one-inch or less in size or those where containment
leakage would be into a closed system whose piping pressure rating
exceeds the containment design pressure rating. Relief valves that
are one-inch or smaller provide an additional physical barrier in
that, even in the unlikely event that a relief valve were to fail to
remain closed during or following an accident, the size restriction
would limit leakage such that a large early release would not occur.
In the unlikely event that a relief valve larger than one-inch were
to fail to remain closed, the leakage would be into a system which
forms a closed loop outside primary containment and any containment
leakage would return to primary containment through this closed
loop.
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--
[[Page 20383]]
Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: David Terao.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2),
Beaver County, Pennsylvania
Date of amendment request: February 9, 2007.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) 3.3.2, ``Engineered Safety
Feature Actuation System Instrumentation,'' TS 3.5.2, ``Emergency Core
Cooling System--Operating,'' TS 3.6.5, ``Containment Air Temperature,''
and TS 5.5.12, ``Containment Leakage Rate Testing Program.'' The
revised TSs would be consistent with a proposed change to the
Recirculation Spray System (RSS) pump start signal due to a
modification to the containment sump screens.
The proposed amendment would also replace the use of LOCTIC with
the Modular Accident Analysis Program-Design Basis Accident (MAAP-DBA)
for calculating containment pressure, temperature, and condensation
rates for input to the SWNAUA code. The calculation methodology change
would ultimately change the aerosol removal coefficients used in dose
consequence analysis.
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 RSS pump start signal,
the upper containment temperature technical specification (TS)
limit, the peak containment internal pressure, the nomenclature for
automatic switchover to the containment sump, and the containment
sump screen visual inspection surveillance requirement do not
involve any system or component that are accident initiators. The
RSS is used for accident mitigation only. The Refueling Water
Storage Tank (RWST) level and containment pressure instrumentation
will continue to comply with all applicable regulatory requirements
and design criteria (e.g., train separation, redundancy, single
failure, etc.) following approval of the proposed changes. The
design functions performed by the RSS and the containment are not
changed by this license amendment request.
Delaying the start of the RSS pumps and the change to the upper
containment temperature affect the long-term containment pressure
and temperature profiles. The environmental qualification of safety-
related equipment inside containment will be confirmed to be
acceptable and accident mitigation systems will continue to operate
within design temperatures and pressures. Delaying the RSS pump
start reduces the emergency diesel generator loading in the early
stage of a design basis accident and maintaining the staggered
loading of the RSS pump starts avoids overloading on each emergency
diesel generator at Unit 1. Staggered loading of the emergency
diesel generator is not required for Unit 2.
The methodology change to calculate containment pressure,
temperature and condensation rates for input to the SWNAUA code will
not involve a significant increase in the probability of an accident
previously evaluated because this change in methodology does not
impact accident initiators.
The loss of coolant accident (LOCA) has been evaluated using the
guidance provided in Regulatory Guide 1.183, ``Alternative
Radiological Source Terms for Evaluating Design Basis Accidents at
Nuclear Power Reactors.'' The radiological consequences of the
remaining design basis accidents are not significantly impacted by
the proposed changes. As demonstrated by the supporting analyses,
the estimated dose consequences at the Exclusion Area Boundary
(EAB), Low Population Zone (LPZ), and control room remain within the
acceptance criteria of 10 CFR 50.67 as supplemented by Regulatory
Guide 1.183 and Standard Review Plan Section 15.0.1. In addition,
the supporting analyses also demonstrates that the dose consequences
in the Emergency Response Facility remain compliant with paragraph
IV.E.8 of Appendix E, to 10 CFR part 50, Emergency Planning and
Preparedness for Production and Utilization Facilities, regulatory
guidance provided in Supplement 1 of NUREG-0737. The revised
radiological analyses results in a slight increase in control room
and off-site doses; however, the radiological analyses and
evaluations developed in support of this application demonstrate
that the proposed changes will not impact compliance with applicable
regulatory requirements and will not involve a significant increase
in the consequences of an accident previously evaluated. The slight
increase in control room and off-site doses is more than offset by
the increased assurance of adequate NPSH [net positive suction head]
to the RSS pumps and Emergency Core Cooling System operability.
The safety analysis acceptance criteria will continue to be met
following the proposed changes to the RSS pump start signal, visual
sump inspection, TS containment upper temperature limit, peak
containment internal pressure, nomenclature for automatic switchover
to the containment sump and the change to the control room and off-
site dose consequences analyses.
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. One of the proposed changes alters the RSS pump
start circuitry by initiating the pump start from a coincident
Containment Pressure High-High/[RWST] Level Low signal instead of
from a timer. The RSS pump instrumentation will be included as part
of the Engineered Safety Feature Actuation System (ESFAS)
instrumentation in the TS and will be subject to the ESFAS
surveillance requirements following approval of the proposed
changes. The design of the RSS pump start instrumentation complies
with all applicable regulatory requirements and design criteria. The
failure modes have been analyzed to ensure that the revised RSS pump
start circuitry can withstand a single active failure without
affecting the RSS design functions. The RSS is an accident
mitigation system only, so no new accident initiators are created.
It is not expected that the change in containment temperature
will have a significant impact on equipment qualification. However,
any equipment that must be replaced or re-qualified will be
addressed prior to operation with the proposed change to RSS pump
start. As a result any such equipment will not introduce new failure
modes, accident initiators, or malfunctions that would cause a new
or different kind of accident.
The remaining changes do not change plant equipment design or
function and therefore will not introduce new failure modes,
accident initiators, or malfunctions that would cause a new or
different kind of accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No. The changes to the RSS pump start signal and the
upper containment temperature limit affect the containment response
and the LOCA dose analyses. Analyses demonstrate that containment
design basis limits are satisfied and post-LOCA offsite and control
room dose criteria will continue to be met following approval of the
proposed changes.
The change to the containment sump visual inspection will not
involve a significant reduction in a margin of safety because the
revised surveillance will continue to provide adequate assurance the
sump screens are not blocked with debris and that signs of corrosion
will be detected.
The change to peak containment internal pressure will not result
[in] a significant reduction in a margin of safety because the new
pressure is lower for each of the units.
Although the control room and off-site doses slightly increase
(due to a combination of the change to the start signal and the
proposed methodology change), the increase will not involve a
significant reduction in a margin of safety because operator and
public exposure limits will continue to meet applicable regulatory
requirements.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
[[Page 20384]]
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, FirstEnergy Nuclear
Operating Company, FirstEnergy Corporation, 76 South Main Street,
Akron, OH 44308.
NRC Acting Branch Chief: John P. Boska.
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: March 8, 2007.
Description of amendment request: The proposed amendment would add
Technical Specification (TS) Limiting Condition for Operation (LCO)
3.0.8 to allow a delay time for entering a supported system TS when the
inoperability is due solely to an inoperable snubber, if risk is
assessed and managed. The proposed change is consistent with TS Task
Force (TSTF) change TSTF-372-A, Revision 4, ``Addition of LCO 3.0.8,
Inoperability of Snubbers.''
The NRC staff issued a notice of availability of a model no
significant hazards consideration determination for referencing in
license amendment applications in the Federal Register on November 24,
2004 (69 FR 68412). The licensee affirmed the applicability of the
model 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 allows a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable snubber
if risk is assessed and managed. The postulated seismic event requiring
snubbers is a low-probability occurrence and the overall TS system
safety function would still be available for the vast majority of
anticipated challenges. Therefore, the probability of an accident
previously evaluated is not significantly increased, if at all. The
consequences of an accident while relying on allowance provided by
proposed LCO 3.0.8 are no different than the consequences of an
accident while relying on the TS required actions in effect without the
allowance provided by proposed LCO 3.0.8. Therefore, the consequences
of an accident previously evaluated are not significantly affected by
this change. The addition of a requirement to assess and manage the
risk introduced by this change will further minimize possible concerns.
Therefore, this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of
a New or Different Kind of Accident From Any Previously Evaluated.
The proposed change does not involve a physical alteration of the
plant (no new or different type of equipment will be installed).
Allowing delay times for entering supported system TS when
inoperability is due solely to inoperable snubbers, if risk is assessed
and managed, will not introduce new failure modes or effects and will
not, in the absence of other unrelated failures, lead to an accident
whose consequences exceed the consequences of accidents previously
evaluated. The addition of a requirement to assess and manage the risk
introduced by this change will further minimize possible concerns.
Thus, this change does not create the possibility of a new or different
kind of accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety.
The proposed change allows a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber, if risk is assessed and managed. The postulated seismic event
requiring snubbers is a low-probability occurrence and the overall TS
system safety function would still be available for the vast majority
of anticipated challenges. The risk impact of the proposed TS changes
was assessed following the three-tiered approach recommended in
Regulatory Guide 1.177. A bounding risk assessment was performed to
justify the proposed TS changes. This application of LCO 3.0.8 is
predicated upon the licensee's performance of a risk assessment and the
management of plant risk. The net change to the margin of safety is
insignificant. Therefore, this change does not involve a significant
reduction in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1700 K Street, NW., Washington, DC 20006.
NRC Acting Branch Chief: John P. Boska.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: February 15, 2007.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs) Surveillance Requirement (SR)
3.8.4.2 to correct errors inadvertently introduced by Amendment No.
146. SR 3.8.4.2 currently requires that each battery charger be
verified to supply greater than or equal to 150 amps for 250-volt DC
subsystems, and greater than or equal to 50 amp for 125-volt DC
subsystems. The licensee proposed to correct the errors by
differentiating that the Division 1 battery chargers are verified to
supply greater than or equal to 150 amps and the Division 2 battery
chargers are verified to supply greater than or equal to 110 amps. The
licensee stated that the Division 2 battery charger output current
limiter is field-adjusted to supply 120 to 125 amps in order to stay
within the electrical circuit breaker ratings in the associated
distribution cabinet.
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 (NSHC). The NRC staff reviewed the licensee's analysis,
and has performed its own analysis as follows:
(1) Does the proposed amendment involve a significant increase
in the probability or consequence of an accident previously
evaluated?
No. The proposed amendment would only correct the battery
chargers' DC supply current limits specified by SR 3.8.4.2. The
current limits of the battery chargers were not considered to be a
precursor to, and does not affect the probability of, an accident.
In addition, there is no design or operation change associated with
the proposed amendment. Therefore, the proposed amendment does not
increase the probability of an accident previously evaluated.
The corrected DC supply current limits of the battery chargers
will ensure that the batteries will be charged under as-designed
conditions. The corrected limits will not decrease the functionality
of the Division 1 or Division 2 battery chargers, or the
functionality of the batteries the battery chargers support.
Therefore, the plant systems required to mitigate accidents will
remain capable of performing their design functions. As a result,
the proposed amendment will not lead to a significant change in the
consequences of any accident.
(2) Does the proposed amendment create the possibility of a new
or different kind of
[[Page 20385]]
accident from any accident previously evaluated?
No. The proposed amendment does not involve a physical
alteration of any system, structure, or component (SSC) or a change
in the way any SSC is operated. The proposed amendment does not
involve operation of any SSCs in a manner or configuration different
from those previously recognized or evaluated. No new failure
mechanisms will be introduced by the revised acceptance value. Thus,
the proposed amendment does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
No. The proposed amendment would only change the current supply
limits of the battery chargers. There will be no modification of any
TSs limiting condition for operation, no change to any limit on
previously analyzed accidents, no change to how previously analyzed
accidents or transients would be mitigated, no change in any
methodology used to evaluate consequences of accidents, and no
change in any operating procedure or process. Therefore, the
proposed amendment does not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
the NRC staff's own analysis above, it appears that the three standards
of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to
determine that the proposed amendment 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 Branch Chief: L. Raghavan.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment requests: March 30, 2007.
Description of amendment requests: The proposed change will revise
Technical Specifications (TSs) Surveillance Requirement (SR) 3.3.7.3.b,
``Loss of Voltage Function'' to a narrower voltage band and lower
operating time for channel calibration testing, by replacing the
undervoltage relays with the reset time significantly lower.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the Technical Specifications
Surveillance Requirement 3.3.7.3.b allowable set point values of the
Loss of Voltage Function for the channel calibration testing. This
proposed change will allow Southern California Edison (SCE) to
increase margin and conservatism for the loss of voltage relay
settings and overall loop uncertainties while performing Loss of
Voltage Signal (LOVS) channel calibration testing.
The loss of voltage function is detected by the LOVS relays
installed on the 4.16 kV Safety Related busses. Normally, these
devices are not considered to be accident initiators. The proposed
change narrows the voltage operating band and lowers the allowable
upper limit for this loss of voltage detection by use of the
electronic type Basler BE1-27 under-voltage relays. However, the
reset time of the relay [will be reduced] significantly. [Therefore,
t]he proposed change does not impact 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 [an] accident previously evaluated?
Response: No.
The proposed allowable values for the LOVS relays voltage
settings and the minimum operating voltage of the of[f]site power
will provide acceptable level of protection for the plant equipment.
3. Does the proposed change involve [a] significant reduction in
a margin of safety?
Response: No.
The proposed loss of voltage function is designed to ensure that
plant equipment will not operate beyond its normal operating range
for satisfactory operation of all the safety related equipment. The
proposed loss of voltage function values will not affect the
existing protection criterion for the plant equipment and will not
reduce margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Branch Chief: Thomas G. Hiltz.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: February 16, 2007.
Description of amendment request: The proposed amendment would
permanently revise Technical Specification 2.2.1, Table 2.2-1,
Functional Unit 17.A, Turbine Trip Low Trip System Pressure allowable
value. The proposed revision was previously approved for one operating
cycle at each unit.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the allowable value for reactor trip
as a result of a turbine trip on low trip system pressure. This
change will not alter any plant components, systems, or processes
and will only provide a more appropriate value to assess operability
of the associated pressure switches. Since the plant features and
operating practices are not altered, the possibility of an accident
is not affected. This reactor trip is not directly credited in SQN's
[Sequoyah Nuclear Plant's] accident analysis and is maintained as an
anticipatory trip to enhance the overall reliability of the reactor
trip system. As such, there is not a specific safety limit
associated with this function and the generation of a reactor trip
based on low trip system pressure is above the required actuations
to ensure acceptable mitigation of accidents. As the proposed change
will continue to provide an acceptable anticipatory trip signal, the
offsite dose potential is not affected by this change. Therefore,
the proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
As described above, this change will not alter any plant
equipment or operating practices that have the ability to create a
new potential for accident generation. The proposed change revises
the operability limits for a function that generates a trip signal
when appropriate conditions exist to require accident mitigation
response. This type of function does not have the ability to create
an accident as its purpose and function is to mitigate events.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will revise an allowable value for a reactor
trip initiator that results from a turbine trip condition. This
change will not alter the setpoint, and the calibration of the
associated pressure switches will continue to be set at the current
value. The allowable value change is in response to accuracy aspects
of the
[[Page 20386]]
instrumentation and does not alter the ability of this trip function
to operate when and as needed to mitigate accident conditions.
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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: Thomas H. Boyce.
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].
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: January 19, 2007, as
supplemented by letters dated March 13 and 22, 2007.
Brief description of amendment: The amendment modifies Technical
Specifications 5.5.9 and 5.6.8 to add steam generator alternate repair
criteria and additional steam generator reporting criteria at H. B.
Robinson Steam Electric Plant, Unit No. 2.
Date of issuance: April 9, 2007.
Effective date: This license amendment is effective as of the date
of issuance and shall be implemented within 60 days.
Amendment No.: 214.
Renewed Facility Operating License No. DPR-23. Amendment revises
the Technical Specifications.
Date of initial notice in Federal Register: January 30, 2007 (72 FR
4300). The March 13 and 22, 2007, supplemental letters provided
clarifying information that 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 April 9, 2007.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423,
Millstone Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: March 28, 2006, as supplemented
by letters dated October 26, and December 4, 2006, and January 26,
2007.
Brief description of amendment: The amendment revises Millstone
Power Station, Unit No. 3 Technical Specifications (TS) to delete
redundant surveillance requirements pertaining to post-maintenance/
post-modification testing.
Date of Issuance: March 29, 2007.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 237.
Facility Operating License No. NPF-49: Amendment revised the TS.
Date of initial notice in Federal Register: May 23, 2006 (71 FR
29673). The supplemental letters provided additional information that
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the staff's original proposed
no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 29, 2007.
No significant hazards consideration comments received: No.
Duke Power Company LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: April 11, 2006, as supplemented
October 24, 2006.
Brief description of amendments: The amendments revised the
Technical Specifications requirements related to steam generator tube
integrity consistent with the NRC-approved Revision 4 to Technical
Specification Task Force (TSTF) Standard Technical Specification Change
Traveler TSTF-449, ``Steam Generator Tube Integrity.'' These amendments
also remove license conditions that become outdated with these TS
changes. In addition, the amendments revised the organizational
description in TS 5.2.1, which is solely administrative and unrelated
to steam generator tube integrity.
Date of Issuance: April 2, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 355, 357, 356.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: January 3, 2007 (72 FR
149). The supplement dated October 24, 2006, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated April 2, 2007. No significant hazards
consideration comments received: No.
[[Page 20387]]
Duke Power Company LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: April 11, 2006, as supplemented
by letter dated March 14, 2007.
Brief description of amendments: The amendments added Technical
Specification (TS) Limiting Condition for Operation (LCO) 3.0.8 to
allow a delay time for entering a supported system TS when the
inoperability is due solely to an inoperable snubber, if risk is
assessed and managed with an approved Bases Control Program that is
consistent with the TS Bases Control Program described in Section 5.5
of the applicable vendor's Standard Technical Specifications.
Date of Issuance: April 2, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 356, 358, 357.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the licenses and the Technical Specifications.
Date of initial notice in Federal Register: January 3, 2007 (72 FR
151). The 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 on January 3, 2007 (72 FR 151). The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated April 2, 2007.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will
County, Illinois
Date of application for amendment: November 18, 2005, as
supplemented by letters dated August 18 and September 28, 2006, and
February 15, February 23, and March 7, 2007.
Brief description of amendment: The amendments would revise the
existing steam generator tube surveillance program using Technical
Specification Task Force Traveler No. 449 (TSTF-449), Revision 4,
``Steam Generator Tube Integrity'' as a basis. The amendments would
also revise TS 5.5.9, ``Steam Generator (SG) Tube Surveillance
Program,'' regarding the required SG inspection scope for Byron
Station, Unit No. 2, during outage number 13 and subsequent operating
cycle. A similar approval was granted for Braidwood Station, Unit 2 by
letter from the NRC dated October 24, 2006.
Date of Issuance: March 30, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 150/150, 144/144.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications and License.
Date of initial notice in Federal Register: May 23, 2006 (71 FR
29676). The August 18 and September 28, 2006 and February 15, February
23, and March 7, 2007 supplements, contained clarifying information and
did not change the staff's initial proposed finding of no significant
hazards consideration.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 30, 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: March 23, 2006, as supplemented by
letters dated August 16 and November 28, 2006.
Description of amendment request: The amendment revises the
Seabrook Station, Unit No. 1 Technical Specifications (TSs)
Definitions, TS 3.4.5, ``Steam Generator (SG) Tube Integrity,'' and TS
3.4.6.2, ``Reactor Coolant System Operational Leakage'' consistent with
Technical Specification Task Force (TSTF) Standard Technical
Specification Traveler TSTF-449, ``Steam Generator Tube Integrity,''
Revision 4. Additionally the amendment creates TS 6.7.6.k. ``Steam
Generator (SG) Program'' and TS 6.8.1.7, ``Steam Generator Tube
Inspection Report,'' consistent with TSTF-449, Revision 4.
Date of Issuance: March 28, 2007.
Effective date: As of its date of issuance, and shall be
implemented within 90 days.
Amendment No.: 115.
Facility Operating License No. NPF-86: The amendment revised the
License and Technical Specifications.
Date of initial notice in Federal Register: April 25, 2006 (71 FR
23955). The licensee's August 16 and November 28, 2006, 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 March 28, 2007.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-266, Point Beach Nuclear
Plant, Unit 1, Town of Two Creeks, Manitowoc County, Wisconsin
Date of application for amendments: July 11, 2006, as supplemented
January 19, March 9 and 26, 2007.
Brief description of amendments: The amendment revises Technical
Specification (TS) 5.5.8, ``Steam Generator Program,'' to change the
inspection and repair criteria for portions of the tubes within the
hot-leg region of the tubesheet for a single operating cycle. In
addition, an administrative change corrects a page number in the TS
Table of Contents and deletes two blank pages in TS Section 5.0.
Date of Issuance: April 4, 2007.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment No.: 226.
Renewed Facility Operating License Nos. DPR-24 and DPR-27:
Amendments revised the Technical Specifications/License.
Date of initial notice in Federal Register: August 29, 2006 (71 FR
51230). The supplements dated January 19, March 9 and 26, 2007,
contained clarifying information and did not change the staff's initial
proposed finding of no significant hazards consideration.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 4, 2007.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: December 20, 2006.
Brief description of amendment: The amendment removed annotations
referencing Technical Data Book (TDB)-VIII, ``Equipment Operability
Guidance,'' and annotations referencing Technical Specification
Interpretations (TSIs) from the NRC Authority File of the Technical
Specifications (TSs). These documents are used by Omaha Public Power
District (OPPD) personnel
[[Page 20388]]
for additional guidance in applying certain Limiting Conditions of
Operation requirements to specific equipment and/or situations. OPPD
has annotated references to these documents in the TS copies used at
the Fort Calhoun Station, Unit No.1 (FCS); however, these annotations
were inadvertently included into the NRC Authority File and are not
officially part of the FCS TS. The amendment also corrected a
discrepancy in TS 2.10.4(1)(c).
Date of Issuance: April 3, 2007.
Effective date: As of its date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 249.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Operating License and Technical Specifications.
Date of initial notice in Federal Register: January 30, 2007 (72 FR
4308).
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated April 3, 2007.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-311, Salem Nuclear Generating Station,
Unit No. 2, Salem County, New Jersey
Date of application for amendment: April 6, 2006, as supplemented
by letters dated January 19, and February 27, 2007.
Brief description of amendment: The amendment revises the Technical
Specifications (TSs) related to steam generator tube integrity
consistent with Revision 4 to Technical Specification Task Force (TSTF)
Standard Technical Specification Change Traveler 449 (TSTF-449),
``Steam Generator Tube Integrity.''
Date of Issuance: March 29, 2007.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 262.
Facility Operating License No. DPR-75: The amendment revised the
TSs and the License.
Date of initial notice in Federal Register: July 18, 2006 (71 FR
40753). The letters dated January 19, and February 27, 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 March 29, 2007.
No significant hazards consideration comments received: No.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of application for amendments: June 2, 2006, as supplemented
by letter dated October 19, 2006.
Brief description of amendments: The amendments revised Technical
Specification (TS) 3.8.1, ``AC [alternating current] Sources--
Operating,'' and TS 3.8.3, ``Diesel Fuel Oil, Lube Oil, and Starting
Air,'' to increase the required amount of stored diesel fuel oil to
support a change to Ultra Low Sulfur Diesel fuel from California diesel
fuel presently in use. This change in the type of fuel oil is mandated
by California air pollution control regulations.
Date of Issuance: April 4, 2007.
Effective date: As of its issuance and shall be implemented within
60 days of issuance.
Amendment Nos.: Unit 2--211; Unit 3--203.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: July 18, 2006 (71 FR
40754). The supplemental letter dated October 19, 2006, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 4, 2007.
No significant hazards consideration comments received: No.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: February 21, 2006.
Brief description of amendments: The amendments revised Technical
Specifications 1.1, ``Definitions,'' and 3.4.16, ``RCS [Reactor Coolant
System] Specific Activity.'' The revisions replaced the current
Limiting Condition for Operation (LCO) 3.4.16 limit on RCS
grossspecific activity with limits on RCS Dose Equivalent I-131 (DEI)
and Dose Equivalent Xe-133 (DEX). The conditions and required actions
for LCO 3.4.16 not being met, and surveillance requirements for LCO
3.4.16, are revised. The modes of applicability for LCO 3.4.16 are
extended. TS Figure 3.4.16-1 on the limit for DEI with respect to rated
thermal power is deleted.
Date of issuance: March 29, 2007.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: 137/137.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: February 27, 2007 (72
FR 8805).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 29, 2007.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: August 17, 2006.
Brief description of amendment: The amendment revised Technical
Specifications (TSs) 2.1.1, ``Reactor Core SLs [Safety Limits],''
3.3.1, ``Reactor Trip System (RTS) Instrumentation,'' 3.4.1, RCS
[Reactor Coolant System] Pressure, Temperature, and Flow Departure from
Nucleate Boiling (DNB) Limits,'' and 5.6.5, ``Core Operating Limits
Report (COLR).'' The changes (1) relocated certain operating cycle-
specific core operating limits, including TS Figure 2.1.1-1, ``Reactor
Core Safety Limits,'' from the TSs to the plant COLR, (2) added two new
safety limits for departure from nucleate boiling ratio and peak fuel
centerline temperature, and (3) added topical reports to TS 5.6.5 and
had the reports cited by only the report title and number. TS 5.6.5 was
expanded to include the limits from TSs 2.1.1, 3.3.1, and 3.4.1.
Date of Issuance: April 2, 2007.
Effective date: As of its date of issuance and shall be implemented
within 90 days from the date of issuance. The final TS Bases changes
including the licensee's application dated August 17, 2006, will be
processed under the licensee's program for updates to the TS Bases, in
accordance with TS 5.5.14, at the time this amendment is implemented.
The final changes to the COLR including those in the licensee's
application dated August 17, 2006, will be submitted to the NRC in
accordance with the update process covered by TS 5.6.5.d.
Amendment No.: 183.
Facility Operating License No. NPF-30: The amendment revised the
[[Page 20389]]
Operating License and Technical Specifications.
Date of initial notice in Federal Register: January 16, 2007 (72 FR
1781).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 2, 2007.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, et al., Docket Nos. 50 280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of application for amendments: May 26, 2006, as supplemented
on January 19, 2007.
Brief Description of amendments: These amendments revised the
Technical Specification (TS) requirements related to steam generator
tube integrity and Reactor Coolant System leakage definitions and
requirements. The TSs were revised to implement TS Task Force (TSTF)
Standard TS Change Traveler, TSTF-449, ``Steam Generator Tube
Integrity,'' (TSTF-449, Rev. 4) with minor deviations to be consistent
with Surry's custom TSs.
Date of Issuance: March 29, 2007.
Effective date: As of date of issuance and shall be implemented
within 180 days.
Amendment Nos.: 251, 250.
Renewed Facility Operating License Nos. DPR-32 and DPR-37:
Amendments changed the licenses and the technical specifications.
Date of initial notice in Federal Register: August 15, 2006 (71 FR
46941). The supplement dated January 19, 2007, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated March 29, 2007.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of application for amendments: July 5, 2006, as supplemented
on September 21 and November 20, 2006.
Brief Description of amendments: These amendments revised the main
control room (MCR) and emergency switchgear room (ESGR) air-
conditioning system (ACS) Technical Specifications to reflect the
completion of permanent modifications to the equipment and associated
power supply configuration. The revisions include the addition of
requirements and/or action statements addressing the inoperability of
two or more air handling units (AHUs) on a unit, as well as AHU powered
from an H emergency bus. The proposed change, paralleling requirements
in the Improved Technical Specifications, also adds MCR and ESGR ACS
requirements during refueling operations and irradiated fuel movement
in the fuel building. In addition, the proposed change clarified the
service water requirements for the ACS chillers that serve the MCR and
ESGRs.
Date of Issuance: April 2, 2007.
Effective date: As of date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 252, 251.
Renewed Facility Operating License Nos. DPR-32 and DPR-37:
Amendments changed the licenses and the technical specifications.
Date of initial notice in Federal Register: September 26, 2006 (71
FR 56193). The supplements dated September 21 and November 20, 2006,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the staff's original proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated April 2, 2007.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 16th day of April 2007.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E7-7534 Filed 4-23-07; 8:45 am]
BILLING CODE 7590-01-P