[Federal Register Volume 71, Number 157 (Tuesday, August 15, 2006)]
[Notices]
[Pages 46929-46946]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 06-6921]
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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 July 21, 2006, to August 3, 2006. The last
biweekly notice was published on August 1, 2006 (71 FR 43528).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example, in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for
[[Page 46930]]
leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and
Adjudications Staff; (3) e-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station
(CPS), Unit 1, DeWitt County, Illinois
Date of amendment request: June 30, 2006.
Description of amendment request: The proposed change would revise
the Note preceding Technical Specification (TS) Surveillance
Requirement (SR) 3.4.6.1 to be consistent with the wording in NUREG-
1434, ``Standard Technical Specifications General Electric Plants, BWR/
6,'' Revision 3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment revises the note associated with TS SR
3.4.6.1, which requires verification that the leakage past the
Reactor Coolant System (RCS) Pressure Isolation Valves (PIVs) is
less than a specified limit. The proposed revision provides
clarification that performance of this SR is allowed during plant
shutdown (i.e., a Mode other than Modes 1 and 2).
The proposed change does not require modification to the
facility. The proposed change does not affect the operation of any
facility equipment, the interface between facility systems, or the
reliability of any equipment. In addition, the proposed change does
not alter the requirement to perform the leakage testing of the RCS
PIVs and does not revise the leakage limits associated with this SR.
The function of the RCS PIVs is to separate the high pressure RCS
from an attached low pressure system. Periodic testing of PIVs can
substantially reduce intersystem Loss of Coolant Accident (LOCA)
probability. Since the proposed change does not alter the method or
limits associated with the leak rate testing of the RCS PIVs there
is no significant increase in the probability of a LOCA. Therefore,
the proposed amendment does not involve a significant increase in
the
[[Page 46931]]
probability of an accident previously evaluated.
The consequences of a previously analyzed event are dependent on
the initial conditions assumed in the analysis, the availability and
successful functioning of equipment assumed to operate in response
to the analyzed event, and the setpoints at which these actions are
initiated. The method for performing the leakage testing of the RCS
PIVs and the specified leakage limit for this testing will not
change as a result of the proposed revision and, therefore, there is
no change in the consequences associated with the LOCA analysis. The
radiological consequences remain within applicable regulatory
limits. The proposed change does not alter any system's performance
measures or the ability to perform its accident mitigation
functions. The radiological consequences associated with any
previously evaluated accident do not change as a result of the
proposed revision. Therefore, the proposed change does not involve a
significant increase in the consequences of an accident previously
evaluated.
Based on the above, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change to the wording of the Note to TS SR 3.4.6.1
clarifies the plant conditions for when the surveillance is required
to be performed. The proposed change does not affect the design,
functional performance or operation of the facility. No new
equipment is being introduced and installed equipment is not being
operated in a new or different manner. Similarly, the proposed
change does not affect the design or operation of any structures,
systems or components involved in the mitigation of any accidents,
nor does it affect the design or operation of any component in the
facility such that new equipment failure modes are created. There
are no setpoints at which protective or mitigative actions are
initiated that are affected by this proposed action. No change is
being made to procedures relied upon to respond to an off-normal
event.
As such the proposed amendment will 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.
Margins of safety are established in the design of components,
the configuration of components to meet certain performance
parameters, and in the establishment of setpoints to initiate alarms
or actions. The proposed change revises a note associated with a
surveillance requirement to clarify the plant conditions for when
the surveillance needs to be performed. This change involves an
administrative clarification to reflect the original intent of the
TS. The equipment will continue to be tested in a manner and at a
frequency necessary to provide confidence that the equipment can
perform its intended safety function. There is no change in the
design of the affected systems, no alteration of the setpoints at
which alarms or actions are initiated, and no change in plant
configuration from original design. There is no impact on the plant
safety analyses.
Therefore, operation of CPS in accordance with 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: Mr. Bradley J. Fewell, Assistant General
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett
Square, PA 19348.
NRC Branch Chief: Daniel S. Collins.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: June 14, 2006.
Description of amendment request: The proposed change will delete
Waterford 3 Technical Specification (TS) Surveillance Requirement (SR)
4.8.1.1.2.f. This SR requires that the emergency diesel generator be
subjected to an inspection in accordance with procedures prepared in
conjunction with its manufacturer's recommendations.
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 ability of the emergency diesel generator to perform its
safety function is not proven by the performance of the
manufacturer's recommended inspections. The inspections are not
considered an initiator or mitigating factor in any previously
evaluated accidents.
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 results in the deletion of the SR associated
with the performance of manufacturer's inspections. No modifications
to plant structures, systems, or components, or changes in the
design of the plant structures, systems, or components are required
to support the proposed TS change.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The ability of the emergency diesel generator to perform its
safety function is not proven by the performance of the
manufacturer's recommended inspections. Inspection activities will
continue to be performed.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: N.S. Reynolds, Esquire, Winston & Strawn,
1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: David Terao.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station (DNPS), Units 2 and 3, Grundy County, Illinois
Date of amendment request: June 2, 2006.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) Surveillance Requirement (SR)
3.4.3.1 to increase the allowable as-found main steam safety valve
(MSSV) lift set point tolerance from +/-1 percent to +/-3 percent. The
proposed change would also revise the SR 3.1.7.10 to increase the
enrichment of sodium pentaborate used in the Standby Liquid Control
(SLC) system from greater than or equal to 30 atom percent boron-10 to
greater than or equal to 45 atom percent boron-10.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change increases the allowable as-found MSSV lift
setpoint tolerance, determined by test after the valves have been
removed from service, from +/-1 percent to +/-3 percent. The
proposed change does not alter the TS requirements for
[[Page 46932]]
the number of MSSVs required to be operable, the nominal lift
setpoints, the allowable as-left lift setpoint tolerance, the MSSV
testing frequency, or the manner in which the valves are operated.
Consistent with current TS requirements, the proposed change
continues to require that the MSSVs be adjusted to within +/-1
percent of their nominal lift setpoints following testing. Since the
proposed change does not alter the manner in which the valves are
operated, there is no significant impact on reactor operation.
The proposed change does not involve a physical change to the
valves, nor does it change the safety function of the valves. The
proposed TS revision involves no significant changes to the
operation of any systems or components in normal or accident
operating conditions and no changes to existing structures, systems,
or components, with the exception of the SLC system enrichment
change. The proposed change to increase the enrichment of sodium
pentaborate used in the SLC system by a design modification using a
single SLC pump will ensure that the requirements of 10 CFR 50.62,
``Requirements for reduction of risk from anticipated transients
without scram (ATWS) events for light-water-cooled nuclear power
plants,'' continue to be met. The SLC system is not an initiator to
an accident; rather, the SLC system is used to mitigate a postulated
anticipated transient without scram (ATWS) event. Therefore, these
changes will not increase the probability of an accident previously
evaluated.
Generic considerations related to the change in setpoint
tolerance were addressed in NEDC-31753P, ``BWROG In-Service Pressure
Relief Technical Specification Revision Licensing Topical Report,''
and were reviewed and approved by the NRC in a safety evaluation
dated March 8, 1993. The plant specific evaluations, required by the
NRC's safety evaluation and performed to support this proposed
change, show that there is no change to the design core thermal
limits and adequate margin to the reactor vessel pressure limits
using a +/-3 percent lift setpoint tolerance. These analyses also
show that operation of Emergency Core Cooling Systems is not
affected, and the containment response following a loss-of-coolant
accident is acceptable. The plant systems associated with these
proposed changes are capable of meeting applicable design basis
requirements and retain the capability to mitigate the consequences
of accidents described in the Updated Final Safety Analysis Report.
Therefore, these changes do not involve an increase in the
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. 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 increases the allowable as-found lift
setpoint tolerance for the DNPS MSSVs, and increases the required
enrichment of sodium pentaborate used in the SLC system. The
proposed change to increase the enrichment of sodium pentaborate
used in the SLC system will ensure that the requirements of 10 CFR
50.62 continue to be met.
The proposed change to increase the MSSV tolerance was developed
in accordance with the provisions contained in the NRC safety
evaluation for NEDC-31753P. MSSVs installed in the plant following
testing or refurbishment will continue to meet the current tolerance
as-left acceptance criteria of +/-1 percent of the nominal setpoint.
The proposed change does not affect the manner in which the
overpressure protection system is operated; therefore, there are no
new failure mechanisms for the overpressure protection system.
The proposed change to allow an increase in the MSSV setpoint
tolerance does not alter the nominal MSSV lift setpoints or the
number of MSSVs currently required to be operable by DNPS TS. The
proposed change does not involve physical changes to the valves, nor
does it change the safety function of the valves. There is no
alteration to the parameters within which the plant is normally
operated. As a result, no new failure modes are being introduced.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established through the design of the
plant structures, systems, and components, the parameters within
which the plant is operated, and the establishment of the setpoints
for the actuation of equipment relied upon to respond to an event.
The proposed change does not modify the safety limits or setpoints
at which protective actions are initiated, and does not change the
requirements governing operation or availability of safety equipment
assumed to operate to preserve the margin of safety.
Establishment of the 3 percent MSSV setpoint
tolerance limit does not adversely impact the operation of any
safety-related component or equipment. Evaluations performed in
accordance with the NRC safety evaluation for NEDC-31753P have
concluded that all design limits will continue to be met.
The proposed change to increase the enrichment of sodium
pentaborate used in the SLC system will ensure that the requirements
of 10 CFR 50.62 continue to be met.
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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Assistant General
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett
Square, PA 19348.
NRC Branch Chief: Daniel S. Collins.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station (LSCS), Units 1 and 2, LaSalle County, Illinois
Date of amendment request: March 16, 2006.
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) 3.3.6.1, ``Primary Containment
Isolation Instrumentation,'' Table 3.3.6.1-1 to revise the allowable
values (AVs) for the reactor core isolation cooling (RCIC) temperature-
based leak detection. The proposed change is a result of revising the
setpoint calculation for the subject temperature instruments based on
the current reactor coolant leak detection analytical limit. The
temperature limits correspond to a 25-gallon per minute (gpm) leak as
determined by LSCS calculations. The proposed changes would revise TS
Table 3.3.6.1-1 AVs for the following four RCIC system isolation
functions:
Item 3.e. RCIC Equipment Room Temperature--High
Item 3.f. RCIC Equipment Room Differential Temperature--High
Item 3.g. RCIC Steam Line Tunnel Temperature--High
Item 3.h. RCIC Steam Line Tunnel Differential Temperature--High
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change is a result of revising the setpoint
calculation for the subject temperature instruments based on the
current reactor coolant leak detection calculation analytical limit.
The proposed changes will revise TS Table 3.3.6.1-1 Allowable Values
for the following four RCIC system isolation functions as noted
below.
Increase the Allowable Value for Function 3.e., ``RCIC
Equipment Room Temperature--High,'' from <= 291.0 [deg]F to <= 297.0
[deg]F
Decrease the Allowable Value for Function 3.f., ``RCIC
Equipment Room Differential Temperature--High,'' from <= 189.0
[deg]F to <= 188.0 [deg]F
Decrease the Allowable Value for Function 3.g., ``RCIC
Steam Line Tunnel Temperature--High,'' from <= 277.0 [deg]F to <=
267.0 [deg]F
Increase the Allowable Value for Function 3.h., ``RCIC
Steam Line Tunnel Differential Temperature--High,'' from <= 155.0
[deg]F to <= 163.0 [deg]F
[[Page 46933]]
The function of the instrumentation listed on TS Table 3.3.6.1-
1, in combination with other accident mitigation features, is to
limit fission product release during and following postulated Design
Basis Accidents to within allowable limits. The Allowable Values
specified in TS Table 3.3.6.1-1 provide assurance that the
instrumentation will perform as designed.
The Allowable Values for RCIC system isolation are not a
precursor to any accident previously evaluated. Accidents are
assumed to be initiated by equipment failure. The proposed change
does not alter the initiation conditions or operational parameters
for the system. There is no increase in the failure probability of
the system. As such, the probability of occurrence for a previously
evaluated accident is not increased.
The Allowable Values specified in Table 3.3.6.1-1 provide
assurance that the RCIC system will perform as designed. The
proposed revision to the Allowable Values does not change any of the
RCIC system leak detection isolation actuation setpoints. Thus, the
radiological consequences of any accident previously evaluated are
not increased.
Based on the above information, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change does not affect the control parameters
governing unit operation or the response of plant equipment to
transient conditions. The proposed change does not change or
introduce any new equipment, modes of system operation or failure
mechanisms.
The proposed change is based on revised reactor coolant leak
detection calculation analytical limits determined by the most
current revision to the heat rise calculation. Setpoint calculations
have been performed to determine the nominal trip setpoints and
Allowable Values for the instrumentation associated with the leak
detection function based on the revised analytical limits determined
by the heat rise calculations. The proposed revision to the
Allowable Values does not change any of the RCIC system leak
detection isolation actuation setpoints.
Based on the above information, the proposed change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The proposed change will revise TS Table 3.3.6.1-1 Allowable
Values for the instrument functions associated with RCIC Isolation.
The current Allowable Values for these functions are:
<= 291.0 [deg]F for RCIC Equipment Room Temperature--High
<= 189.0 [deg]F for RCIC Equipment Room Differential Temperature--
High
<= 277.0 [deg]F for the RCIC Steam Line Tunnel Temperature--High
<= 155.0 [deg]F for the RCIC Steam Line Tunnel Differential
Temperature--High
The proposed change revises the Allowable Values to the
following:
<= 297.0 [deg]F for RCIC Equipment Room Temperature--High
<= 188.0 [deg]F for RCIC Equipment Room Differential Temperature--
High
<= 267.0 [deg]F for the RCIC Steam Line Tunnel Temperature--High
<= 163.0 [deg]F for the RCIC Steam Line Tunnel Differential
Temperature--High
The proposed change is a result of revising the setpoint
calculation for the subject temperature instruments based on the
current analytical limit. The proposed changes will revise TS Table
3.3.6.1-1 Allowable Values for the subject four RCIC system
isolation functions and will provide assurance that the RCIC system
will perform as designed. The proposed revision to the Allowable
Values does not change any of the RCIC system leak detection
isolation actuation setpoints.
Margin of safety is established by the design and qualification
of plant equipment, the operation of the plant within analyzed
limits, and the point at which protective or mitigative actions are
being initiated. The proposed change does not alter these
considerations. The proposed allowable values will still ensure that
the results of the accident analysis remain valid.
Based on this information, the proposed changes do not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Assistant General
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett
Square, PA 19348.
NRC Branch Chief: Daniel S. Collins.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: April 4, 2006.
Description of amendment request: The proposed amendment request
will add one NRC approved topical report reference to the list of
analytical methods in Technical Specification (TS) 5.6.5, ``Core
Operating Limits Report (COLR),'' that can be used to determine core
operating limits, and will delete seven obsolete references from the
same TS Section.
The proposed changes are:
1. Add an NRC previously approved Topical Report ANF-1358(P)(A),
Revision 3, ``The Loss of Feedwater Heating Transient in Boiling
Water Reactors,'' (LOFWH), which will list FRA-ANP method for
evaluating the LOFWH transient.
2. Delete seven references describing previously approved Global
Nuclear Fuel (GNF) and FRA-ANP methodologies for the analyses of
ATRIUM-9B and GE9 fuel. Both of these fuel types have been or will
be completely discharged from both Lasalle County Station (LSCS)
reactors after the loading of ATRIUM-10 fuel during the LSCS Unit 2
refuel outage currently scheduled to begin in February 2007 (i.e.,
L2R11).
The proposed changes support the continued irradiation of ATRIUM-10
fuel in the LSCS reactors and the use of the NRC-approved analytical
methodology for evaluation of LOFWH transients.
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.
Technical Specification (TS) 5.6.5 lists NRC-approved analytical
methods used at LaSalle County Station (LSCS) to determine core
operating limits. The proposed changes will add an NRC-approved
topical report reference to the list of administratively controlled
analytical methods in TS 5.6.5, ``Core Operating Limits Report
(COLR),'' that can be used to determine core operating limits, and
delete seven obsolete references.
The addition of a Framatome ANP (FRA-ANP) methodology to
determine overall core operating limits for future LSCS core
configurations was approved by the NRC in Reference 2. LSCS Unit 2
will continue to load Framatome ANP ATRIUM-10 fuel during the Unit 2
Refueling Outage 11 currently scheduled for February 2007. The
proposed change to TS 5.6.5 will add a FRA-ANP methodology as a
reference to determine core operating limits for loss of feedwater
heater (LOFWH) conditions. Thus, the proposed change will allow LSCS
to use the most recent FRA-ANP methodology for analysis of LOFWH
conditions.
The addition and deletion of approved analytical methods in TS
Section 5.6.5 has no effect on any accident initiator or precursor
previously evaluated and does not change the manner in which the
core is operated. The NRC-approved methods ensure that the output
accurately models predicted core behavior, have no effect on the
type or amount of radiation released, and have no effect on
predicted offsite doses in the event of an accident. Additionally,
the NRC-approved methods do not change any key core parameters that
influence any accident consequences. Thus, the proposed changes do
not have any effect on the probability of an accident previously
evaluated.
The methodology conservatively establishes acceptable core
operating limits such that the consequences of previously analyzed
events are not significantly increased.
[[Page 46934]]
The proposed changes in the list of analytical methods do not
affect the ability of LSCS to successfully respond to previously
evaluated accidents and does not affect radiological assumptions
used in the evaluations. Thus, the radiological consequences of any
accident previously evaluated are not increased.
Therefore, the proposed changes do 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 to TS Section 5.6.5 do not affect the
performance of any LSCS structure, system, or component credited
with mitigating any accident previously evaluated. The NRC-approved
analytical methodology for evaluating LOFWH transients will not
affect the control parameters governing unit operation or the
response of plant equipment to transient conditions. The proposed
changes do not introduce any new modes of system operation or
failure mechanism.
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
the margin of safety?
Response: No.
The proposed changes will add a reference to the list of
analytical methods in TS 5.6.5 that can be used to determine core
operating limits and delete seven obsolete references. The proposed
changes do not modify the safety limits or setpoints at which
protective actions are initiated and do not change the requirements
governing operation or availability of safety equipment assumed to
operate to preserve the margin of safety. Therefore, the proposed
changes provide an equivalent level of protection as that currently
provided.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
Based on the above information, EGC concludes that the proposed
amendment presents no significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Daniel S. Collins.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
and Lancaster Counties, Pennsylvania
Date of application for amendments: June 8, 2006.
Description of amendment request: The proposed changes modify
Technical Specifications (TSs) 3.1.3, ``Control Rod OPERABILITY'';
3.1.6, ``Rod Pattern Control''; 3.3.2.1, ``Control Rod Block
Instrumentation''; 3.10.7, ``Control Rod Testing--Operating''; and
3.10.8, ``SHUTDOWN MARGIN (SDM) Test--Refueling'' to replace the
current references to banked position withdrawal sequence (BPWS) with
references to ``the analyzed rod position sequence.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change modifies Technical Specifications (TS)
3.1.3, ``Control Rod OPERABILITY''; TS 3.1.6, ``Rod Pattern
Control''; TS 3.3.2.1, ``Control Rod Block Instrumentation''; TS
3.10.7, ``Control Rod Testing--Operating'', and; TS 3.10.8, SHUTDOWN
MARGIN (SDM) Test--Refueling''. The proposed change would replace
the current references to ``Banked Position Withdrawal Sequence
(BPWS)'' with references to ``the analyzed rod position sequence''.
The use of the ``the analyzed rod position sequence'' will continue
to minimize the consequences of an accident previously evaluated
including the Control Rod Drop Accident (CRDA). Additionally, the
use of the words ``the analyzed rod position sequence'' will provide
an equivalent level of protection during plant startups and
shutdowns and therefore will not increase the consequences of an
accident previously evaluated.
Control rod patterns during startup and shutdown conditions will
continue to be controlled by the operator and the Rod Worth
Minimizer (RWM) (LCO [limiting condition of operation] 3.3.2.1,
``Control Rod Block Instrumentation''), so that only specified
control rod sequences and relative positions are allowed over the
operating range of all control rods inserted to 10% of Rated Thermal
Power. As a result of this change, these sequences will continue to
limit the potential amount of reactivity addition that could occur
in the event of a Control Rod Drop Accident (CRDA).
Accidents are initiated by the malfunction of plant equipment,
or the failure of plant structures, systems, or components. The
proposed change will ensure that analyzed rod position sequences are
developed to minimize incremental control rod reactivity worth in
accordance with the ``General Electric Standard Application for
Reactor Fuel,'' NEDE-24011-P-A-15 (GESTAR-II), and U.S. Supplement,
NEDE-24011-P-A-15-US, September, 2005, NRC approved methodology, and
reviewed and approved in accordance with the 10 CFR 50.59 process.
These analyzed rod position sequences will limit the potential
reactivity increase for a postulated CRDA during reactor startups
and shutdowns below the Low Power Setpoint of 10% of Rated Thermal
Power.
The proposed change will continue to ensure that systems,
structures and components are capable of performing their intended
safety functions.
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 affect the assumed accident
performance of the control rods, nor any plant structure, system, or
component previously evaluated.
The proposed change does not involve the installation of new
equipment, and installed equipment is not being operated in a new or
different manner. The change ensures that control rods remain
capable of performing their safety functions. No set points are
being changed which would alter the dynamic response of plant
equipment. Accordingly, no new failure modes are introduced.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will ensure that analyzed rod position
sequences are developed to minimize incremental control rod
reactivity worth in accordance with the ``General Electric Standard
Application for Reactor Fuel,'' NEDE-24011-P-A-15 (GESTAR-II), and
U.S. Supplement, NEDE-24011-P-A-15-US, September, 2005, NRC approved
methodology, and reviewed and approved in accordance with the 10 CFR
50.59 process. The proposed change will not adversely impact the
plant's response to an accident or transient. All current safety
margins will be maintained. There are no changes proposed which
alter the set points at which protective actions are initiated, and
there is no change to the operability requirements for equipment
assumed to operate for accident mitigation.
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
[[Page 46935]]
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 (Acting): Brooke D. Poole.
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: June 14, 2006.
Description of amendment request: The amendments would incorporate
the results of a new spent fuel pool criticality analysis documented in
WCAP-16518-P/WCAP-16518-NP, ``Beaver Valley Unit 2 Spent Fuel Pool
Criticality Analysis,'' Revision 1, May 2006 for the BVPS-2 spent fuel
storage pool. The revised criticality analysis will permit utilization
of vacant storage locations dictated by the existing Technical
Specification (TS) storage configurations in the BVPS-2 spent fuel
storage pool.
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 relevant accidents previously evaluated are
limited to the fuel handling and criticality accidents.
Administrative controls during fuel fabrication ensure that the
fuel is fabricated to ensure proper loading of fuel in the fuel
assemblies. Administrative and operational controls used to load
fuel assemblies into the spent fuel pool ensure the fuel assemblies
are stored in compliance with the allowed storage configurations.
Fuel handling is performed under administrative controls and
physical limitations. These controls will remain in effect and
continue to protect against criticality and fuel handling accidents
involving new storage configurations dictated by the new analysis.
There is therefore no impact on the probability of fuel handling or
criticality accidents.
The new criticality analysis defines new spent fuel storage
configurations with new enrichment and burnup limits. Integral Fuel
Burnable Absorber (IFBA) limits are used to comply with the 1-out-
of-4 configuration for fresh fuel. The boron dilution evaluation
that supported Amendment [No.] 128 [February 11, 2002, Agencywide
Documents Access and Management System Accession No. ML020020373],
permitting credit for soluble boron at BVPS Unit No. 2 continues to
remain valid. The new analysis demonstrates that keff
remains below unity for the various storage configurations
considered with zero soluble boron, and that keff remains
less than or equal to 0.95 for the entire pool with credit for
soluble boron under non-accident and accident conditions with a 95%
probability at a 95% confidence level (95/95). Potential
consequences of accidents previously analyzed remain unchanged.
The editorial changes made to the table numbers and the LCO
[Limiting Condition for Operation] and Surveillance Requirement
wording do not impact probability or 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 change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No. The relevant types of accidents previously
evaluated are limited to criticality and fuel handling accidents.
Although the new analysis will allow utilization of additional
storage capacity, implementation of fuel loading configurations and
fuel handling activities will continue to be performed under
administrative and operational controls. No new or different
activities are introduced as a result of the proposed changes. The
utilization of additional storage capacity within the allowances of
the revised analysis will introduce no new or other kind of
accident.
The editorial changes made to the table numbers and the LCO and
Surveillance Requirement wording do not impact any previously
evaluated accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No. The margin to safety with respect to analyzed
accidents involves maintaining keff through fuel storage
configurations and boron concentration controls in the spent fuel
pool. The boron dilution evaluation that supported that supported
Amendment [No.] 128 permitting credit for soluble boron at BVPS Unit
No. 2 remains valid. The Amendment [No.] 128 evaluation concluded
that a boron dilution event is not credible for BVPS Unit No. 2. The
new analysis calculates the non-accident soluble boron concentration
to be less than was determined in the Amendment [No.] 128
evaluation. Thus, there is no significant reduction in a margin of
safety because of the new analysis and the conclusions of the
Amendment [No.] 128 dilution evaluation remain valid.
Under accident conditions, the soluble boron needed to maintain
keff below 0.95 with the new storage configurations is
less than what is assumed in current analysis. The proposed change
does not involve a significant reduction in a margin of safety for
accident conditions.
The editorial changes made to the table numbers and the LCO and
Surveillance Requirement wording do not impact a margin of safety.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Branch Chief: Richard J. Laufer.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry
Nuclear Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: June 1, 2006.
Description of amendment request: The proposed amendment would
modify Technical Specification 3.4.10, ``Residual Heat Removal (RHR)
Shutdown Cooling System--Cold Shutdown'' by adding a default Condition
to address situations when an RHR shutdown cooling subsystem becomes
inoperable in MODE 4 and, within the completion time of 1 hour, an
alternate method of decay heat removal can not be verified to be
available.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No. The proposed amendment does not change the design
of any structures, systems or components (SSCs), and does not affect
the manner in which plant systems are operated. It is a change to
the Technical Specifications only, to provide guidance to plant
operators on appropriate actions to take, where no Technical
Specification guidance currently exists. Since the design of plant
SSCs is not changed and plant systems and components are not
operated in a different manner, there is no change to previously
identified accident initiators, and the proposed amendment would not
impact the probability of any of the previously evaluated accidents
in the Updated Safety Analysis Report (USAR).
The USAR event that evaluates the consequences of a loss of RHR
Shutdown Cooling is included in Section 15.2.9 entitled ``Failure of
RHR Shutdown Cooling''. This event examines the consequences of a
loss of not only an RHR shutdown cooling
[[Page 46936]]
subsystem, but also the loss of the suction source from the
recirculation system leading to both RHR Shutdown Cooling
subsystems, and a loss of offsite power. Even with these multiple
failures, this event is not one of the limiting transients. As noted
in Section 15.2.9.5, ``Radiological Consequences,'' there are no
fuel failures, and the consequences of the event are much less than
those for the ``Main Steam Isolation Valve Closure'' transient,
which is evaluated with acceptable results in USAR Section 15.2.4.5.
Since the proposed amendment only involves the addition of a
Required Action where no guidance currently exists, and the design
of plant SSCs is not changed and plant systems and components are
not operated in a different manner, the proposed amendment does not
affect the consequences of the Section 15.2.9 analysis, nor does it
affect the ability of the installed RHR subsystems to perform their
shutdown cooling function. The change adds a default Condition to
provide guidance to the operators in those situations when a
subsystem becomes inoperable with the plant in MODE 4 and an
alternate cannot be verified to be available within an hour, which
does not impact the consequences of the previously evaluated
accidents in the USAR.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No. This change to the required Technical
Specification actions does not involve a change in the design
function or operation of plant SSCs. It does not introduce credible
new failure mechanisms, malfunctions, or accident initiators not
considered in the existing plant design and licensing basis.
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 amendment involve a significant reduction
in a margin of safety?
Response: No. This proposed amendment only involves a change to
the required Technical Specification actions. It does not involve a
change in the evaluation and analysis methods used to demonstrate
compliance with regulatory and licensing requirements, and does not
exceed or alter a design basis or safety limit. The safety margin
before the change remains unchanged after the proposed amendment.
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: David W. Jenkins, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: Daniel S. Collins.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry
Nuclear Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: June 1, 2006.
Description of amendment request: The proposed amendment would
modify Technical Specification 3.4.9, ``Residual Heat Removal (RHR)
Shutdown Cooling System--Hot Shutdown,'' to revise the Required Actions
when both RHR shutdown cooling subsystems are inoperable in MODE 3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No. The proposed amendment does not change the design
of any structures, systems or components (SSCs), and does not affect
the manner in which plant systems are operated. It is a change to
the Technical Specifications only, to provide guidance to plant
operators on appropriate actions to take, when both RHR shutdown
cooling subsystems are inoperable. Since the design of plant SSCs is
not changed and plant systems and components are not operated in a
different manner, there is no change to previously identified
accident initiators, and the proposed amendment would not impact the
probability of any of the previously evaluated accidents in the
Updated Safety Analysis Report (USAR).
The USAR event that evaluates the consequences of a loss of RHR
Shutdown Cooling is included in Section 15.2.9 entitled ``Failure of
RHR Shutdown Cooling.'' This event examines the consequences of a
loss of not only an RHR shutdown cooling subsystem, but also the
loss of the suction source from the recirculation system leading to
both RHR Shutdown Cooling subsystems, and a loss of offsite power.
Even with these multiple failures, this event is not one of the
limiting transients. As noted in Section 15.2.9.5, ``Radiological
Consequences,'' there are no fuel failures, and the consequences of
the event are much less than those for the ``Main Steam Isolation
Valve Closure'' transient, which is evaluated with acceptable
results in USAR Section 15.2.4.5. Since the proposed amendment only
involves the addition of a Required Action where no guidance
currently exists, and the design of plant SSCs is not changed and
plant systems and components are not operated in a different manner,
the proposed amendment does not affect the consequences of the
Section 15.2.9 analysis, nor does it affect the ability of the
installed RHR subsystems to perform their shutdown cooling function.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No. This change to the required Technical
Specification actions does not involve a change in the design
function or operation of plant SSCs. It does not introduce credible
new failure mechanisms, malfunctions, or accident initiators not
considered in the existing plant design and licensing basis.
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 amendment involve a significant reduction
in a margin of safety?
Response: No. This proposed amendment only involves a change to
the required Technical Specification actions. It does not involve a
change in the evaluation and analysis methods used to demonstrate
compliance with regulatory and licensing requirements, and does not
exceed or alter a design basis or safety limit. The safety margin
before the change remains unchanged after the proposed amendment.
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: David W. Jenkins, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: Daniel S. Collins.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of amendment request: April 28, 2006.
Description of amendment request: The proposed amendment would
change the SSES 1 and 2 Technical Specifications (TSs) to modify the
standby liquid control system for single loop pump operation and use of
enriched sodium pentaborate solution.
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:
Does the proposed change involve a significant increase in the
probability or
[[Page 46937]]
consequences of an accident previously evaluated?
Response: No.
The proposed changes revise Technical Specification 3.1.7 for
the Standby Liquid Control (SLC) system to reflect new boron weight-
percent and enrichment requirements. In addition, the change to
single pump operation reduces the required SLC pump flow and
discharge pressure required to satisfy 10 CFR 50.62, thus increasing
the reliability of the system. The changes do not otherwise alter
the design or operation of the SLC system, and the existing design
of the system is sufficient to support operation with the enriched
sodium pentaborate solution. The SLC system is not considered to be
the initiator of any event currently analyzed in the FSAR [Final
Safety Analysis Report]. Therefore, the proposed changes do not
increase the probability of a previously evaluated accident.
The SSES ATWS [anticipated transient without scram] analysis was
performed using standard accepted assumptions, inputs, and codes.
That analysis, which demonstrated that the acceptance criteria for
peak vessel pressure, peak cladding temperature, peak local cladding
oxidation, peak suppression pool temperature, and peak containment
pressure, established the requirements for the proposed boron
weight-percent and concentration, and pump flow rate. The analysis
assumed the use of only a single pump, versus two pumps. The results
of the analysis are that no fission product barriers are adversely
challenged, and the radiological consequences of previously
evaluated accidents (i.e., ATWS) are not increased.
Therefore, the proposed changes do 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 revise Technical Specification 3.1.7 for
the SLC system to reflect new boron weight-percent and enrichment
requirements. In addition, the change to single pump operation
reduces the required SLC pump flow and discharge pressure required
to satisfy 10 CFR 50.62, thus increasing the reliability of the
system. A new Surveillance Requirement (SR 3.1.7.10) is also added
to verify the correct solution enrichment prior to addition of
inventory to the SLC tank. The changes do not otherwise alter the
design or operation of the SLC system, and the existing design of
the system is sufficient to process the enriched sodium pentaborate
solution. With the exception of these changes, no other physical
changes to plant structures or systems are proposed. Thus, the
proposed changes do not create a new initiating event for the
spectrum of events currently postulated in the FSAR.
Therefore, the proposed changes do 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 changes revise Technical Specification 3.1.7 for
the SLC system to reflect new boron weight-percent and enrichment
requirements. In addition, the change to single pump operation
reduces the required SLC pump flow and discharge pressure required
to satisfy 10 CFR 50.62, thus increasing the reliability of the
system. The changes do not otherwise alter the design or operation
of the SLC system, and the existing design of the system is
sufficient to process the enriched sodium pentaborate solution.
The analysis was performed using standard accepted assumptions,
inputs, and codes. That analysis, which demonstrated that ATWS
acceptance criteria are satisfied, established the requirements for
the proposed boron weight-percent and concentration, and pump flow
rate. Further, the analysis assumed only a single pump is in
operation verses two pumps. The evaluation demonstrated that the SLC
system meets this post-LOCA [loss-of-coolant accident] suppression
pool pH control design function.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Branch Chief: Richard J. Laufer.
Tennessee Valley Authority, Docket No. 50-259 , Browns Ferry Nuclear
Plant, Unit 1, Limestone County, Alabama
Date of amendment request: October 12, 2004.
Description of amendment request: As part of Nuclear Regulatory
Commission's (NRC) approval of the Improved Technical Specifications
for Browns Ferry Nuclear Plant, Unit 1, by Amendment No. 234, NRC
imposed License Condition 2.C(4) to ensure that the required analyses
and modifications needed to support the Technical Specification (TS)
changes made by License Amendment No. 234 and any subsequent TS
changes, were completed by licensee prior to entering the mode for
which the TS applies. The proposed amendment would remove this license
condition from the license.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment does not affect any precursors for
accidents described in Chapter 14 of the Browns Ferry Updated Final
Safety Analysis Report (UFSAR). The proposed amendment does not
change the conditions, operating configurations, or minimum amount
of operating equipment assumed in the safety analysis for accident
mitigation. No changes are proposed in plant protection or which
create new modes of plant operation. Therefore, the proposed
amendment does not involve a significant increase in the probability
or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment does not introduce new equipment, which
could create a new or different kind of accident. No new external
threats, release pathways, or equipment failure modes are created.
Therefore, the implementation of the proposed amendment will not
create a possibility for an accident of a new or different type than
those previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment does not impact the redundancy or
availability of equipment credited in the response to accidents
described in Chapter 14 of the UFSAR. For these reasons, 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: L. Raghavan.
Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear
Plant, Unit 1, Limestone County, Alabama
Date of amendment request: May 1, 2006 (TS-455).
Description of amendment request: The proposed amendment would
revise the numeric values of the safety limit minimum critical power
ratio (SLMCPR) in the Technical Specification (TS) Section 2.1.1.2 for
single and two reactor recirculation loop operation to incorporate the
results of the Browns Ferry Nuclear Plant, Unit 1 Cycle 7 SLMCPR
analysis.
[[Page 46938]]
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 Technical Specification change involve a
significant increase in the probability or consequences of an
accident previously evaluated?
Response: No.
The proposed amendment establishes a revised SLMCPR value for
single and two recirculation loop operation. The probability of an
evaluated accident is derived from the probabilities of the
individual precursors to that accident. The proposed SLMCPR values
preserve the existing margin to transition boiling and the
probability of fuel damage is not increased. Since the change does
not require any physical plant modifications or physically affect
any plant components, no individual precursors of an accident are
affected and the probability of an evaluated accident is not
increased by revising the SLMCPR values.
The consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. The revised SLMCPR values have been determined using
NRC-approved methods and procedures. The basis of the MCPR Safety
Limit is to ensure no mechanistic fuel damage is calculated to occur
if the limit is not violated. These calculations do not change the
method of operating the plant and have no effect on the consequences
of an evaluated accident. Therefore, the proposed TS change does not
involve an increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed Technical Specification change create the
possibility of a new or different kind of accident from any accident
previously evaluated?
Response: No.
The proposed license amendment involves a revision of the SLMCPR
value for single and two recirculation loop operation based on the
results of an analysis of the Unit 1 Cycle 7 core. Creation of the
possibility of a new or different kind of accident would require the
creation of one or more new precursors of that accident. New
accident precursors may be created by modifications of the plant
configuration, including changes in the allowable methods of
operating the facility. This proposed license amendment does not
involve any modifications of the plant configuration or changes in
the allowable methods of operation. Therefore, the proposed TS
change does not create the possibility of a new or different kind of
accident previously evaluated.
3. Does the proposed Technical Specification change involve a
significant reduction in a margin of safety?
Response: No.
The margin of safety as defined in the TS bases will remain the
same. The new SLMCPR values were calculated using referenced fuel
vendor methods and procedures, which are in accordance with the fuel
design and licensing criteria. The SLMCPR remains high enough to
ensure that greater than 99.9 percent of all fuel rods in the core
are expected to avoid transition boiling if the limit is not
violated, thereby preserving the fuel cladding integrity. Therefore,
the proposed TS change does not involve a reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC (Acting) Branch Chief: L. Raghavan.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: July 6, 2006 (TS-06-04).
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) for the Sequoyah Nuclear
Plant, Units 1 and 2. Action a.1 of TS 3.1.3.2, ``Position Indication
Systems--Operating,'' requires the verification of rod position by use
of the moveable incore detectors. Tennessee Valley Authority (the
licensee, TVA) is proposing a revision to TS 3.1.3.2 to allow the
position of the control and shutdown rods to be monitored by a means
other than the moveable incore detectors. The amendment will provide a
less burdensome monitoring method should problems with the analog rod
position indication (ARPI) system be experienced. When a recurring
problem in the system requires the monitoring of a rod's position by
the alternate means, TVA plans to continue unit operation and to use
the alternate means until the unit enters Mode 5 and repairs to the
system can safely be implemented.
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 provides an alternative method for the
monitoring of the position of a rod once the position of the rod is
verified using the moveable incore detector system. The proposed
monitoring of rod control system parameters provides a reasonably
similar approach to rod position monitoring as that provided by the
movable incore detector system. In particular, the ability to
immediately detect a rod drop or misalignment is not directly
provided by the movable incore detector system or by the monitoring
of rod control system parameters. Additionally, neither the movable
incore detector system, nor the monitoring of rod control system
parameters, provides the capability to verify rod position following
a reactor trip or shutdown. Therefore, the monitoring of rod control
system parameters, in lieu of the use of the movable incore detector
system, provides an equivalent and acceptable method of monitoring
rod position while a position indicator is inoperable.
The proposed change does not alter plant equipment that is
considered to have the potential to alter the probability of an
accident. The affected components are for monitoring only and do not
actively affect equipment that interacts with the control of the
reactor. Likewise, the affected components are for monitoring and
provide an equivalent level of indication of rod position as the
current action. This maintains an acceptable level of rod position
indication for normal plant operations, as well as post accident
mitigation actions. 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, the proposed change provides only an
alternative method of monitoring the position of a rod. No new
accident initiators are introduced by the proposed alternative
manner of performing rod position monitoring. The proposed change
does not affect the reactor protection system or the reactor control
system. Hence, no new failure modes are created that would cause a
new or different kind of accident from any accident previously
evaluated.
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 rod position indicators are required to determine control
rod positions and thereby ensure compliance with the control rod
alignment and insertion limits. The proposed change does not alter
the requirement to determine rod position but provides an
alternative method for monitoring the position of the affected rod
after the position of the rod is verified using the moveable incore
detector system. As a result, the initial conditions of the accident
analysis are preserved. The components affected by the alternate rod
monitoring will not affect plant setpoints utilized for automatic
mitigation of accident conditions or other equipment necessary for
accident mitigation.
[[Page 46939]]
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: Michael L. Marshall, Jr.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: July 12, 2006 (TS-06-03).
Description of amendment request: The proposed amendment would
revise the limiting condition for operation for the Sequoyah Nuclear
Plant, Units 1 and 2, Technical Specification (TS) Section 3.7.5,
``Ultimate Heat Sink.'' This revision would change the minimum ultimate
heat sink (UHS) water elevation in TS 3.7.5.a from 670 feet to 674
feet. The essential raw cooling water (ERCW) temperature requirement in
TS 3.7.5.b would be increased from 83 degrees Fahrenheit ([deg]F) to 87
[deg]F. The conditional requirements of TS 3.7.5.c would no longer be
required and would be deleted by the proposed change. This change would
also delete a footnote that established a temporary UHS temperature
limit of 87 [deg]F through September 30, 1995. These proposed changes
are supported by a combination of design basis re-analysis, bounding
analysis, and sensitivity analysis of the ERCW system, the UHS, and
supported systems.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to increase the UHS maximum temperature and
the minimum water level does not alter the function, design, or
operating practices for plant systems or components. One exception
is the elimination of non-safety-related station air compressor
loads located in the turbine building. The UHS is utilized to remove
heat loads from plant systems during normal and accident conditions.
This function is not expected or postulated to result in the
generation of any accident and continues to adequately satisfy the
associated safety functions with the proposed changes. Therefore,
the probability of an accident presently evaluated in the safety
analyses will not be increased because the UHS function does not
have the potential to be the source of an accident. The heat loads
that the UHS is designed to accommodate have been evaluated for
functionality with the higher temperature and elevation
requirements. The result of these evaluations is that there is
existing margins associated with the systems that utilize the UHS
for normal and accident conditions. These margins are sufficient to
accommodate the postulated normal and accident heat loads with the
proposed changes to the UHS. Since the safety functions of the UHS
are maintained, the systems that ensure acceptable offsite dose
consequences will continue to operate as designed. 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 UHS function is not an initiator of any accident and only
serves as a heat sink for normal and upset plant conditions. By
allowing the proposed change in the UHS temperature and elevation
requirements, only the parameters for UHS operation are changed
while the safety functions of the UHS and systems that transfer the
heat sink capability continue to be maintained. The UHS function
provides accident mitigation capabilities and does not reflect the
potential for accident generation. Therefore, the possibility for
creating a new or different kind of accident is not created because
the UHS is only utilized for heat removal functions that are not a
potential source for accident generation. 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 has been evaluated for systems that are
needed to support accident mitigation functions as well as normal
operational evolutions. Operational margins were found to exist in
the systems that utilize the UHS capabilities such that these
proposed changes will not result in the loss of any safety function
necessary for normal or accident conditions. The ERCW system has
excess flow margins that will accommodate the increased flows
necessary for the proposed temperature increase. While operating
margins have been reduced by the proposed changes, safety margins
have been maintained as assumed in the accident analyses for
postulated events.
Additionally, the proposed changes do not require the
modification of component setpoints utilized for automatic
mitigation of accident conditions or other equipment necessary for
accident mitigation. Therefore, a significant reduction in the
margin to safety is not created by this proposed change. 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: L. Raghavan.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: June 16, 2006 (WBN-TS-06-04).
Description of amendment request: The proposed amendment change
would revise Technical Specification (TS) 5.7.2.11, ``Inservice Testing
Program,'' to remove ``applicable supports'' from the Inservice Testing
(IST) Program and revise the IST Program for pumps and valves to meet
the requirements of the latest Edition and Addenda of the American
Society of Mechanical Engineers (ASME) Code approved by the NRC for use
on the date 12-months prior to the start of the 10-year IST Interval.
For the Watts Bar Nuclear Plant (WBN), Unit 1, the second 10-year IST
Interval will begin on December 27, 2006. The ASME Code that was
approved in 10 CFR 50.55a(f)(4) for use on December 27, 2005, was ASME
Operations and Maintenance (OM) Code, 2001 Edition, with Addenda
through 2003. The proposed change provides consistency with the
requirements in 10 CFR 50.55a(f)(4) by replacing the reference to ASME
Boiler and Pressure Vessel Code, Section XI, with ASME OM Code. This
proposed change is based on Technical Specification Task Force (TSTF)
Traveler 479, Revision 0, ``Changes to Reflect Revision of 10 CFR
50.55a.'' TSTF 279-A, Revision 0, ``Remove `applicable supports' from
Inservice Testing Program,'' was approved by NRC and incorporated into
Revision 2 of NUREG-1431, ``Standard Technical Specification
Westinghouse Plants.'' In addition, the proposed amendment would add
provisions to TS 5.7.2.11, Item b, to only apply Surveillance
Requirement 3.0.2 to those IST frequencies of 2 years or less.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the
[[Page 46940]]
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 Technical Specification Section
5.7.2.11 for WBN Unit 1 to conform to the requirements of 10 CFR
50.55a(f)(4) regarding the inservice testing of pumps and valves
which are classified as ASME Code Class 1, 2, and 3.
ASME has in the last several years, transitioned the
requirements for inservice testing of pumps and valves out of ASME
Section XI and into a separate, stand alone code entitled the ``Code
for Operation and Maintenance of Nuclear Power Plants,'' (ASME OM
Code). The ASME OM Code has been endorsed by the NRC in 10 CFR
50.55a and is the Code that will be required for inservice testing
of pumps and valves during the WBN Second Inservice Interval. The
proposed change incorporates revisions to the ASME Code that result
in a net improvement in the measures for testing pumps and valves.
The proposed change also deletes the reference to supports from the
Inservice Testing Program as supports are already inspected under
the Inservice Inspection Program.
The proposed changes do not involve any hardware changes, nor do
the changes affect the probability of any event initiators. There
will be no change to normal plant operating parameters, accident
mitigation capabilities, or accident analysis assumptions or inputs.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises the Technical Specifications to
delete the reference to ``applicable supports'' from the Inservice
Testing Program and to incorporate the latest Code requirements in
10 CFR 50.55a(f)(4) for Code Class 1, 2, and 3 pumps and valves for
WBN's next ten year interval. The testing requirements are similar
and reflect the same type testing. Valves are still stroke timed;
remote position indicators are still verified to be accurate; seat
leakage measurements of critical valves are still performed; relief
valves still have their setpoints and seat leakages verified; pumps
are still tested for hydraulic performance and mechanical condition;
check valves are verified to open and close properly; and supports
are still inspected under the appropriate inspection program.
The proposed changes do not involve a modification to the
physical configuration of the plant or change methods governing
normal plant operation. No test methods are added or deleted.
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 revises the TS for consistency with the
Standard Technical Specification and with the requirements in 10 CFR
50.55a(f)(4) regarding the inservice testing of pumps and valves
which are classified as ASME Code Class 1, 2, and 3. This change
incorporates revisions to the ASME Code that result in a net
improvement in the measures of testing. Incorporation of the ASME OM
Code does not alter the limiting values and acceptance criteria used
to judge the continued acceptability of components tested by the
Inservice Testing Program. Deletion of the reference to supports in
the Inservice Testing Program does not alter the support inspection
program as the program is currently under the Inservice Inspection
Program. Since these limits are not altered, the margin of safety is
not altered. Therefore, the proposed changes do not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: L. Raghavan.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: May 30, 2006.
Description of amendment request: The amendment would revise
Surveillance Requirements (SRs) 3.5.2.8 and 3.6.7.1 in the Technical
Specifications (TSs), and delete the footnote to the frequency for SR
3.5.2.5. SR 3.5.2.8 would be revised by replacing the phrase ``trash
racks and screens'' with the word ``strainers.'' This reflects (1) the
replacement of the existing containment recirculation sump suction
inlet trash racks and screens with strainers with significantly greater
effective surface area, and (2) the resulting relocation of the
recirculation fluid pH control system in Refueling Outage 15 schedule
for the spring of 2007. The footnote to SR 3.5.2.5 would be deleted
because it is no longer applicable to the TSs.
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.
None of the changes impact the initiation or probability of
occurrence of any accident [previously evaluated].
The consequences of accidents evaluated in the FSAR [Final
Safety Analysis Report for the Callaway Plant] that could be
affected by this proposed change are those involving the
pressurization of the containment and associated flooding of the
containment and recirculation of this fluid within the Emergency
Core Cooling System (ECCS) or the Containment Spray System (CSS)
(e.g., LOCAs [Loss-of-Coolant Accidents]). [The containment sump
trash racks and screens, and the sump strainers that are replacing
the trash racks and screens are not initiators of accidents.]
Although the configurations of the existing sump screen and the
replacement strainer assemblies are different, they serve the same
fundamental purpose of passively removing debris from the suction of
the supported system pumps. Removal of trash racks does not impact
the adequacy of the pump NPSH [net positive suction head] assumed in
the safety analyses. Likewise the change does not reduce the
reliability of any supported systems or introduce any new system
interactions. The greatly increased surface area of the new strainer
is designed to reduce head loss [at the containment sump] and reduce
the approach velocity at the strainer face significantly, decreasing
the risk of impact from large debris entrained in the sump flow
stream.
The recirculation fluid pH control system storage baskets serve
a passive function to provide a buffering agent to neutralize the
sump solution. The redesign and relocation of the storage baskets
are considered a like kind replacement. The baskets will be located
within the flood plain and will continue to ensure that the
buffering agent is dissolved in the sump fluid to ensure an
equilibrium pH >= 7.1. Failure of a basket would not initiate an
accident. The ECCS and CSS will continue to function in a manner
consistent with the plant design basis.
As such, the proposed change to the Technical Specifications
Surveillance Requirements does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The installed quantity of trisodium phosphate Crystalline will
provide a minimum equilibrium sump pH of 7.1 following dissolution
and mixing. [Deleting the footnote to SR 3.5.2.5 is an
administrative change to remove a one-time required verification
that has already been performed and is no longer a requirement in
the current TSs.] Therefore, there is not a significant increase in
the probability or consequences of an accident previously evaluated.
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 containment recirculation sump strainers and recirculation
fluid pH control
[[Page 46941]]
system are passive systems used for accident mitigation. As such,
they cannot be accident initiators. Therefore, there is no
possibility that this change could create any accident of any kind.
[The containment recirculation sump suction inlet trash racks and
screens are being replaced with a complex strainer design with
significantly larger effective surface area to reduce head loss and
reduce the approach velocity at the strainer face significantly,
decreasing the risk of impact from large debris entrained in the
sump flow stream. This will result in the recirculation fluid pH
control system being relocated.]
No new accident scenarios, transient precursors, or limiting
single failures are introduced as a result of these changes. There
will be no adverse effect[s] or challenges imposed on any safety-
related system as a result of these changes. The quantity of
trisodium phosphate crystalline will provide a minimum equilibrium
sump pH of >= 7.1 following dissolution and mixing. Therefore, the
possibility of a new or different type of accident is not created.
There are no changes which would cause the malfunction of
safety-related equipment, assumed to be operable in the accident
analyses, as a result of the proposed Technical Specification
changes. No new equipment performance burdens are imposed. The
possibility of a malfunction of safety-related equipment with a
different result is not created. [Deleting the footnote to SR
3.5.2.5 is an administrative change to remove a one-time required
verification that has already been performed and is no longer a
requirement in the current TSs.] 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 proposed changes do not adversely affect any plant safety
limits, setpoints, or design parameters. The changes also do not
adversely affect the fuel, fuel cladding, Reactor Coolant System
(RCS), or containment integrity. [The radiological dose consequence
acceptance criteria in the Standard Review Plan for accidents will
continue to be met. Deleting the footnote to SR 3.5.2.5 is an
administrative change to remove a one-time required verification
that has already been performed and is no longer a requirement in
the current TSs.] Therefore, the proposed TS 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: John O'Neill, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: David Terao.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: May 30, 2006, as supplemented by letter
dated June 30, 2006.
Description of amendment request: The proposed amendments would
relocate the American Society for Testing and Materials (ASTM) standard
being used to test the total particulate concentration of the stored
fuel oil to the TS Bases. This proposed change is described in TS Task
force (TSTF) Standard TS Change Traveler TSTF-374-A, Rev. 0, ``Revision
to TS 5.5.13 and Associated TS Bases for Diesel Fuel Oil.'' In
addition, the licensee has proposed to use a ``water and sediment
test'' instead of the ``clear and bright'' test provided in TSTF-374.
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 changes involve a significant increase in the probability
or consequences of an accident previously evaluated?
The proposed change relocates the specific ASTM reference from
the Administrative Controls Section of Technical Specifications (TS)
to a licensee-controlled document. Relocating the specific ASTM
Standard reference from the TS to a licensee-controlled document
will not affect nor degrade the ability of the EDGs [emergency
diesel generators] to perform their specified safety function. Fuel
oil quality will continue to meet the current ASTM requirements for
particulate concentration.
The proposed change is administrative in nature and does not
adversely affect accident initiators or precursors nor alter the
design assumptions, conditions, and configuration of the facility or
the manner in which the plant is operated and maintained. The
proposed change does not alter or prevent the ability of structures,
systems or components from performing their intended function to
mitigate the consequences on an initiating event with the assumed
acceptance limits. The proposed change does not affect the source
term, containment isolation, or radiological release assumptions
used in evaluating the radiological consequences of an accident
previously evaluated. Further, the proposed change does not increase
the types and amounts of radioactive effluent that may be released
offsite, nor significantly increase individual or cumulative
occupational/public radiation exposure.
Therefore, the change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. Do changes create the possibility of a new or different kind
of accident from any accident previously evaluated?
The proposed change relocates the specific ASTM reference from
the Administrative Controls Section of Technical Specifications to a
licensee-controlled document.
The change does not involve a physical alteration of the plant
or a change in the methods governing normal plant conditions. In
addition, the change does not impose any new or different
requirements or eliminate any existing requirements. The change does
not alter assumptions made in the safety analysis and licensing
basis. Therefore, the change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Do changes involve a significant reduction in the margin of
safety?
The proposed change relocates the specific ASTM reference from
the Administrative Controls Section of TS to a licensee-controlled
document. The detail associated with the specific ASTM standard
reference is not required to be in the TS to provide adequate
protection of the public health and safety, since the TS still
retain the requirement for compliance with the applicable ASTM
standard.
The level of safety of facility operation is unaffected by the
proposed change since there is no change in the intent of the TS
requirements of assuring fuel oil is of the appropriate quality for
EDG use. The proposed change provides the flexibility needed to
maintain state-of-the-art technology in fuel oil sampling and
analysis methodology.
The proposed change does not reduce a margin of safety since it
has no impact on any transient or safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Branch Chief: Evangelos C. Marinos.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: May 26, 2006.
Description of amendment request: Item 1: The proposed amendments
would revise the Technical Specification (TS) requirements related to
Reactor Coolant System (RCS) leakage definitions and requirements and
steam generator tube integrity. The licensee requested this change to
implement TS Task Force (TSTF) Standard TS Change Traveler, TSTF-449,
``Steam Generator
[[Page 46942]]
Tube Integrity,'' (TSTF-449, Rev. 4). Item 2: In addition, in its
submittal dated May 26, 2006, the licensee proposed minor deviations
from the TS changes described in TSTF-449, Rev. 4, to provide
consistency with Surry's custom TSs.
Basis for proposed no significant hazards consideration
determination: Item 1: As required by 10 CFR 50.91(a), an analysis of
the issue of no significant hazards consideration is presented below:
1. The Proposed Change Does Not Involve a Significant Increase
in the Probability or Consequences of an Accident Previously
Evaluated.
The proposed change requires a SG Program that includes
performance criteria that will provide reasonable assurance that the
SG tubing will retain integrity over the full range of operating
conditions (including startup, operation in the power range, hot
standby, cooldown and all anticipated transients included in the
design specification). The SG performance criteria are based on tube
structural integrity, accident induced leakage, and operational
leakage.
A SG tube rupture (TR) event is one of the design basis
accidents that are analyzed as part of a plant's licensing basis. In
the analysis of a SGTR event, a bounding primary to secondary
leakage rate equal to the operational leakage rate limits in the
licensing basis plus the leakage rate associated with a double-ended
rupture of a single tube is assumed.
For other design basis accidents such as main steam line break
(MSLB), rod ejection, and reactor coolant pump locked rotor the
tubes are assumed to retain their structural integrity (i.e., they
are assumed not to rupture). These analyses typically assume that
primary to secondary leakage for all SGs is 1 gallon per minute or
increases to 1 gallon per minute as a result of accident induced
stresses. The accident induced leakage criterion introduced by the
proposed changes accounts for tubes that may leak during design
basis accidents. The accident induced leakage criterion limits this
leakage to no more than the value assumed in the accident analysis.
The SG performance criteria proposed change to the TS identify
the standards against which tube integrity is to be measured.
Meeting the performance criteria provides reasonable assurance that
the SG tubing will remain capable of fulfilling its specific safety
function of maintaining reactor coolant pressure boundary integrity
throughout each operating cycle and in the unlikely event of a
design basis accident. The performance criteria are only a part of
the SG Program required by the proposed change to the TS. The
program, defined by NEI 97-06, Steam Generator Program Guidelines,
includes a framework that incorporates a balance of prevention,
inspection, evaluation, repair, and leakage monitoring. The proposed
changes do not, therefore, significantly increase the probability of
an accident previously evaluated.
The consequences of design basis accidents are, in part,
functions of the DOSE EQUIVALENT 1-131 in the primary coolant and
the primary to secondary leakage rates resulting from an accident.
Therefore, limits are included in the plant TS for operational
leakage and for DOSE EQUIVALENT 1-131 in primary coolant to ensure
the plant is operated within its analyzed condition. The typical
analysis of the limiting design basis accident assumes that primary
to secondary leak rate after the accident is 1 gallon per minute
with no more than 500 gallons per day in any one SG, and that the
reactor coolant activity levels of DOSE EQUIVALENT 1-131 are at the
TS values before the accident.
The proposed change does not affect the design of the SGs, their
method of operation, or primary coolant chemistry controls. The
proposed approach updates the current TSs and enhances the
requirements for SG inspections. The proposed change does not
adversely impact any other previously evaluated design basis
accident and is an improvement over the current TSs.
Therefore, the proposed change does not affect the consequences
of a SGTR accident and the probability of such an accident is
reduced. In addition, the proposed changes do not affect the
consequences of an MSLB, rod ejection, or a reactor coolant pump
locked rotor event, or other previously evaluated accident.
2. The Proposed Change Does Not Create the Possibility of a New
or Different Kind of Accident from any Previously Evaluated.
The proposed performance based requirements are an improvement
over the requirements imposed by the current [TS]. Implementation of
the proposed SG Program will not introduce any adverse changes to
the plant design basis or postulated accidents resulting from
potential tube degradation. The result of the implementation of the
SG Program will be an enhancement of SG tube performance. Primary to
secondary leakage that may be experienced during all plant
conditions will be monitored to ensure it remains within current
accident analysis assumptions.
The proposed change does not affect the design of the SGs, their
method of operation, or primary or secondary coolant chemistry
controls. In addition, the proposed change does not impact any other
plant system or component. The change enhances SG inspection
requirements.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
3. The Proposed Change Does Not Involve a Significant Reduction
in the Margin of Safety.
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
isolate the radioactive fission products in the primary coolant from
the secondary system. In summary, the safety function of an SG is
maintained by ensuring the integrity of its tubes.
[SG] tube integrity is a function of the design, environment,
and the physical condition of the tube. The proposed change does not
affect tube design or operating environment. The proposed change is
expected to result in an improvement in the tube integrity by
implementing the SG Program to manage SG tube inspection,
assessment, repair, and plugging. The requirements established by
the SG Program are consistent with those in the applicable design
codes and standards and are an improvement over the requirements in
the current TSs.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed change to the
TS.
The NRC staff has reviewed the licensee's incorporation of the
above analysis by reference and, based on this review, it appears that
the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the
NRC staff proposes to determine that the requested amendments involve
no significant hazards consideration.
Item 2: As required by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below.
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes involve adding a new definition for RCS
[reactor coolant system] leakage and rewording certain [TSs] for
consistency with NUREG-1431, Revision 3. These changes do not
involve any physical plant modifications or changes in plant
operation; consequently, no technical changes are being made to the
existing TS. As such, these changes are administrative in nature and
do not affect initiators of analyzed events or assumed mitigation of
accident or transient events. Therefore, these changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes involve adding a new definition for RCS
leakage and rewording certain [TSs] for consistency with NUREG-1431,
Revision 3. These administrative changes do not involve physical
alteration of the plant (no new or different type of equipment will
be installed) or changes in methods governing normal plant
operation. The changes will not impose any new or different
requirements or eliminate any existing requirements. Therefore,
these changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
[[Page 46943]]
3. Involve a significant reduction in a margin of safety.
The proposed changes involve adding a new definition for RCS
leakage and rewording certain [TS] for consistency with NUREG-1431,
Revision 3. The changes are administrative in nature and will not
involve any technical changes. The changes will not reduce a margin
of safety because they have no impact on any safety analysis
assumptions. Also, since these changes are administrative in nature,
no question of safety is involved. Therefore, the changes do not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Branch Chief: Evangelos C. Marinos.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.
1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: May 26, 2005, as supplemented
by letters dated May 23 and June 20, 2006.
Brief description of amendments: The amendments revised Technical
Specification (TS) 1.1, ``Definitions,'' TS 3.4.14, ``RCS [reactor
coolant system] Operational Leakage,'' TS 5.5.9, ``Steam Generator (SG)
Program,'' and TS 5.6.8, ``Steam Generator Tube Inspection Report,''
and added a new specification, TS 3.4.18, ``Steam Generator (SG) Tube
Integrity.'' The changes are consistent with TS Task Force (TSTF)
Change TSTF-449, Revision 4, ``Steam Generator Tube Integrity.''
Date of issuance: July 27, 2006.
Effective date: As of the date of issuance to be implemented within
150 days from the date of issuance.
Amendment Nos.: Unit 1-161, Unit 2-161, Unit 3-161.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Operating Licenses and the Technical
Specifications for all three units.
Date of initial notice in Federal Register: July 5, 2005 (70 FR
38714). The May 23 and June 20, 2006, supplemental letters provided
additional clarifying information, did not expand the scope of the
application as originally noticed, and did not change the NRC staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 27, 2006.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2 (HBRSEP2), Darlington County, South Carolina
Date of application for amendment: January 21, 2005, as
supplemented by letters dated May 26, 2005, September 19, 2005, and
March 31, 2006.
Brief description of amendment: The amendment approves the
implementation of the alternative source term methodology for a loss-
of-coolant accident at HBRSEP2.
Date of issuance: July 11, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No. 207.
Renewed Facility Operating License No. DPR-23. Amendment does not
revise the Technical Specifications.
Date of initial notice in Federal Register: May 24, 2005 (70 FR
29786). The supplemental letters dated May 26, 2005, September 19,
2005, and March 31, 2006, 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 July 11, 2006.
No significant hazards consideration comments received: No.
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power
Station, Kewaunee County, Wisconsin
Date of application for amendment: January 12, 2006, as
supplemented by letter dated June 2, 2006.
Brief description of amendment: The amendment revises the existing
steam generator (SG) tube surveillance program to be consistent with TS
Task Force (TSTF) Change TSTF-449, Revision 4, ``Steam Generator Tube
Integrity,'' and the model safety evaluation prepared by the Nuclear
Regulatory Commission (NRC) and published in the Federal Register on
March 2, 2005 (70 FR 10298) under the consolidated line item
improvement process (CLIIP).
Date of issuance: July 18, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 188.
Facility Operating License No. DPR-43: Amendment revised the
Facility Operating License and Technical Specifications.
[[Page 46944]]
Date of initial notice in Federal Register: February 14, 2006 (71
FR 7806). The supplement letter contained clarifying information and
did not change the initial no significant hazards consideration
determination, and did not expand the scope of the orginal Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 18, 2006.
No significant hazards consideration comments received: No.
Duke Power Company LLC, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: October 27, 2004.
Brief description of amendments: The amendments revised the
facility operating licenses by removal of license condition 2.F,
``Reporting Requirements'', with regard to maximum power level, Updated
Final Safety Analysis Report, antitrust conditions, fire protection,
and additional conditions.
Date of issuance: July 31, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 230, 226.
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the licenses.
Date of initial notice in Federal Register: July 5, 2005 (70 FR
38717).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 31, 2006.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: April 17, 2006.
Brief description of amendment: The amendment allows a delay time
for entering a supported system Technical Specification (TS) when the
inoperability is due solely to an inoperable snubber, if risk is
assessed and managed consistent with the program in place for complying
with the requirements of 10 CFR 50.65(a)(4). Limiting Condition for
Operation (LCO) 3.0.8 is added to the TS to provide this allowance and
define the requirements and limitations for its use.
This change was proposed by the industry's Technical Specification
Task Force (TSTF) and is designated TSTF-372, Revision 4. The NRC staff
issued a notice of opportunity for comment in the Federal Register on
November 24, 2004 (69 FR 68412), on possible amendments concerning
TSTF-372, including a model safety evaluation and model no significant
hazards consideration (NSHC) determination, using the consolidated line
item improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on May 4, 2005 (70 FR 23252). The
licensee affirmed the applicability of the following NSHC determination
in its application dated April 17, 2006.
Date of issuance: July 11, 2006.
Effective date: As of its date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 198.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 9, 2006 (71 FR
26998).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 11, 2006.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: September 26, 2005, as
supplemented by letter dated April 11, 2006.
Brief description of amendment: The amendment revises the analysis
method used for the large-break loss-of-coolant accident.
Date of issuance: July 24, 2006.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 248.
Facility Operating License No. DPR-26: The amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: November 8, 2005 (70 FR
67747). The April 11, 2006, supplement provided additional information
that clarified the application, did not expand the scope of the
application as originally noticed, and did not change the NRC staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 24, 2006.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant (JAFNPP), Oswego County, New York
Date of application for amendment: January 26, 2006, as
supplemented by letter dated April 12, 2006.
Brief description of amendment: The amendment approves the
implementation of the Boiling Water Reactor Vessel and Internals
Project reactor pressure vessel integrated surveillance program as the
basis for demonstrating the compliance of JAFNPP with the requirements
of Appendix H to Title 10 of the Code of Federal Regulations part 50.
Date of issuance: July 26, 2006.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 285.
Facility Operating License No. DPR-59: The amendment revised the
Updated Final Safety Analysis Report and the License.
Date of initial notice in Federal Register: March 14, 2006 (71 FR
13174). The April 12, 2006, supplement provided additional information
that clarified the application, did not expand the scope of the
application as originally noticed, and did not change the NRC staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 26, 2006.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: September 19, 2005.
Brief description of amendment: The amendment modified ANO-2
Surveillance Requirement TS 3.1.1.4, ``Moderator Temperature
Coefficient,'' and allowed the use of WCAP-16011-P-A, ``Startup Test
Activity Reduction Program.''
Date of issuance: August 2, 2006.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment No.: 265.
Renewed Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications/license.
Date of initial notice in Federal Register: December 6, 2005 (70 FR
72671).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 2, 2006.
[[Page 46945]]
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: September 19, 2005, as
supplemented by letters dated May 11 and June 19, 2006.
Brief description of amendment: The amendment revised the existing
steam generator tube surveillance program to be consistent with the
U.S. Nuclear Regulatory Commission's approved Technical Specification
Task Force Standard Technical Specification Change Traveler, TSTF-449,
``Steam Generator Tube Integrity,'' Revision 4. TSTF-449 is part of the
consolidated line item improvement process.
Date of issuance: August 2, 2006.
Effective date: As of the date of issuance to be implemented within
90 days from the date of issuance.
Amendment No.: 266.
Renewed Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications and Renewed Facility Operating License.
Date of initial notice in Federal Register: January 3, 2006 (71 FR
147). The supplements dated May 11 and June 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 proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 2, 2006.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of application for amendments: January 25, 2006, as
supplemented by letter dated May 17, 2006.
Brief description of amendments: The amendment revised the Quad
Cities licensing basis, as described in the Updated Final Safety
Analysis Report, to allow the use of automatic load tap changers to
operate in automatic mode on the reserve auxiliary transformers to
compensate for potential offsite power voltage fluctuations, in order
to ensure that acceptable voltage is maintained for safety-related
equipment.
Date of issuance: July 24, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 232 and 228.
Renewed Facility Operating License Nos. DPR-29 and DPR-30: The
amendments revised the License.
Date of initial notice in Federal Register: May 23, 2006 (71 FR
29678). The May 17, 2006, supplement contained clarifying information
and did not change the NRC staff's initial proposed finding of no
significant hazards consideration.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 24, 2006.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of application for amendments: August 23, 2005, as
supplemented on April 6, 2006.
Brief description of amendments: The amendments extended the
licensed lives of the Diablo Canyon Power Plant, Unit Nos. 1 and 2
reactors by the amount of time the licensee had expended to perform
low-power testing of the reactors prior to initial startup.
Date of issuance: July 17, 2006.
Effective date: As of its date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: Unit 1-188; Unit 2-190.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Facility Operating Licenses.
Date of initial notice in Federal Register: October 11, 2005 (70 FR
59087). The April 6, 2006, supplemental letter provided additional
information that clarified the application, and did not expand the
scope of the application as originally noticed.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 17, 2006.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: August 4, 2005, as supplemented
by letters dated February 9, July 18, and August 1, 2006.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3.7.1.3, ``Ultimate Heat Sink,'' to permit continued
plant operation if the temperature of the ultimate heat sink (UHS)
exceeds 89 [deg]F, provided the UHS temperature averaged over the
previous 24-hour period is verified at least once per hour to be less
than or equal to 89 [deg]F, and the UHS temperature does not exceed a
maximum value of 91.4 [deg]F.
Date of issuance: August 1, 2006.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 168.
Facility Operating License No. NPF-57: This amendment revised the
TSs.
Date of initial notice in Federal Register: August 30, 2005 (70 FR
51382).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 1, 2006.
No significant hazards consideration comments received: No.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of application for amendment: November 7, 2005, as
supplemented on May 5, 2006.
Brief description of amendment: The amendment revises Technical
Specification 3.9.3, ``Containment Penetrations,'' to allow an
emergency egress door, access door, or roll up door, as associated with
the equipment hatch penetration, to be open, but capable of being
closed, during core alterations or movement of irradiated fuel within
containment.
Date of issuance: July 26, 2006.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 98.
Renewed Facility Operating License No. DPR-18: Amendment revised
the Technical Specifications.
Date of initial notice in Federal Register: January 3, 2006 (71 FR
154). The May 5, 2006, letter provided additional information that
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the staff's original proposed
no significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 26, 2006.
No significant hazards consideration comments received: No.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of application for amendment: November 18, 2005.
Brief description of amendment: The amendment revises the frequency
in Technical Specification Surveillance Requirement 3.6.6.15, which
verifies
[[Page 46946]]
that each containment spray nozzle is unobstructed. The frequency is
changed from ``10 years'' to ``following maintenance which could result
in nozzle blockage.''
Date of issuance: July 31, 2006.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 99.
Renewed Facility Operating License No. DPR-18: Amendment revised
the Technical Specifications and the License.
Date of initial notice in Federal Register: January 3, 2006 (71 FR
154).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 31, 2006.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-259 Browns Ferry Nuclear
Plant, Unit 1, Limestone County, Alabama
Date of application for amendment: December 6, 2004 (TS 428) as
supplemented by letter dated June 16, 2005.
Brief description of amendment: The amendment revised the reactor
vessel Pressure-Temperature curves depicted in the Technical
Specification (TS) Figure 3.4.9-1 and adds a new TS Figure 3.4.9-2.
Date of issuance: July 26, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 256.
Facility Operating License No. DPR-33: Amendment revised the TS.
Date of initial notice in Federal Register: January 18, 2005 (70 FR
2899). The supplement dated June 16, 2005, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 26, 2006.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: December 15, 2005 (TS-05-09), as
supplemented by letter dated June 7, 2006.
Brief description of amendment: The amendment revises the Watts Bar
Nuclear Plant (WBN) Technical Specification Surveillance Requirements
to increase the minimum required average ice basket weight, thus,
increasing the corresponding total weight of the stored ice in the WBN
ice condenser. The changes to the ice basket and total ice weights are
due to the additional energy associated with the Replacement Steam
Generators.
Date of issuance: July 25, 2006.
Effective date: As of the date of issuance and shall be implemented
prior to Mode 4 at startup to begin Cycle 8 fuel cycle.
Amendment No. 62.
Facility Operating License No. NPF-90: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 14, 2006 (71
FR 7814). The supplemental letter provided clarifying information that
was within the scope of the initial notice 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 July 25, 2006.
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: March 28, 2006.
Brief description of amendment: The amendment revised Technical
Specification 5.0, ``Administrative Controls,'' by changing a position
title and department name.
Date of issuance: July 11, 2006.
Effective date: As of its date of issuance, and shall be
implemented within 90 days of the date of issuance.
Amendment No.: 173.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 9, 2006 (71 FR
27005).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 11, 2006.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units 1 and 2, Louisa County, Virginia
Date of application for amendment: July 5, 2005, as supplemented by
letters dated March 30, April 13, and May 11, 2006.
Brief description of amendment: The amendments revised the
Technical Specifications (TSs) to add a reference in TS 5.65.b, ``Core
Operating Limits Report (COLR),'' to permit the use of an alternate
methodology to perform a thermal-hydraulic analysis to predict the
critical heat flux and departure from nucleate boiling ratio for the
AREVA Advanced Mark-BW fuel in the North Anna 1 and 2 cores.
Date of issuance: July 21, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 247, 227.
Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments
changed the Licenses and the TSs.
Date of initial notice in Federal Register: August 16, 2005 (70 FR
48208). The supplements dated March 30, April 13, and May 11, 2006,
contained clarifying information only and did not change the initial no
significant hazards consideration determination or expand the scope of
the initial application.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 21, 2006.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 8th day of August, 2006.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 06-6921 Filed 8-14-06; 8:45 am]
BILLING CODE 7590-01-P