[Federal Register Volume 70, Number 205 (Tuesday, October 25, 2005)]
[Notices]
[Pages 61655-61669]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 05-21180]
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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 September 30 to October 13, 2005. The last
biweekly notice was published on October 11, 2005 (70 FR 59082).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
[[Page 61656]]
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].
Dominion Nuclear Connecticut Inc., et al., Docket No. 50-423, Millstone
Power Station, Unit No. 3, New London County, Connecticut
Date of amendment request: September 13, 2005.
Description of amendment request: The amendment would reduce the
temperature at which shutdown and control rod cluster control
assemblies (RCCA) drop testing is done from greater than or equal to
551 [deg]Fahrenheit (F) to greater than or equal to 500 [deg]F.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1: Does the proposed amendment involve a significant
increase in the probability or consequences of an accident previously
evaluated?
Response: No. DNC [Dominion Nuclear Connecticut, Inc.] is proposing
to change the temperature at which the shutdown and control RCCA drop
tests are performed from ``greater than or equal to 551 [deg]F,'' to
``greater than or equal to 500 [deg]F.'' The proposed change does not
modify any plant equipment and does not impact any failure modes that
could lead to an accident. Additionally, the proposed change has no
effect on the consequence of any analyzed accident since the change
does not affect the function of any equipment credited for accident
mitigation. Based on this discussion, the proposed amendment does not
increase the probability or consequences of an accident previously
evaluated.
Criterion 2: Does the proposed amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
Response: No. The proposed change does not modify any plant
equipment and there is no impact on the capability of existing
equipment to perform its intended functions. No system setpoints are
being modified and no changes are being made to the method in which
plant operations are conducted. No new failure modes are introduced by
the proposed change. The proposed amendment does not introduce accident
initiators or malfunctions that would cause a new or different kind of
accident.
As noted above, the proposed change does not affect the revisions
to plant procedures, which were made to address Westinghouse Nuclear
Safety Advisory Letter, NSAL-00-016 (Rod Withdrawal from Subcritical
Protection in Lower Modes, issued in 2000).
Therefore, the proposed amendment does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
Criterion 3: Does the proposed amendment involve a significant
reduction in a margin of safety?
Response: No. The TS [technical specification] change does not
involve a significant reduction in margin because the acceptance
criterion for the RCCA drop time will not change. The proposed change
will reduce the minimum RCCA drop test temperature from greater than or
equal to 551 [deg]F to greater than or equal to 500 [deg]F. This will
slightly increase the measured test
[[Page 61657]]
RCCA drop time. However, the measured test RCCA drop time is required
to remain within the current TS limit of 2.7 seconds and the 2.19
seconds for surveillance testing acceptance criteria (plant specific
seismic allowance of 0.51 seconds). The proposed change does not affect
any of the assumptions used in the accident analysis, nor does it
affect any operability requirements for equipment important to plant
safety. Therefore, the margin of safety is not impacted by the proposed
amendment.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
NRC Section Chief: Darrell J. Roberts.
Dominion Nuclear Connecticut Inc., et al., Docket No. 50-423, Millstone
Power Station, Unit No. 3, New London County, Connecticut
Date of amendment request: September 13, 2005.
Description of amendment request: The changes revise surveillance
requirements for the recirculation spray system (RSS) to verify proper
initiation of recirculation spray through actuation by the refueling
water storage tank (RWST) low-low level signal instead of actuation by
a timer.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1: Does the proposed amendment involve a significant
increase in the probability or consequences of an accident previously
evaluated?
Response: No. The RSS is only an accident mitigation system. As
such, changes in the operation of RSS cannot have an impact on the
probability of an accident. The delay in the start of the RSS pump is
to assure there is sufficient water in the containment sump for
adequate RSS pump NPSH [net positive suction head] and margin to
suction pipe flashing in light of the debris analysis conducted in
response to GL [Generic Letter] 2004-02. Containment analyses have been
performed to demonstrate that there is no impact on the peak
containment pressure and temperature following a LOCA [loss-of-coolant
accident]. While there are some changes in the predicted post-LOCA
environmental conditions, evaluations have been performed to show that
there is no significant impact on the environmental qualification for
equipment inside containment. The impact to piping and supports has
been demonstrated to be acceptable without modification. Delay in RSS
spray start will result in a reduction in diesel generator loading
since the RSS pumps and the RHS pumps will no longer be running
concurrently. The reduction in iodine removal efficiency during the
delay period is more than offset by elimination of over-conservatisms
in assumptions for long term iodine removal by the RSS system. The net
impact is a reduction in the predicted offsite doses and control room
doses following a design basis LOCA. Based on this discussion, the
proposed amendment does not increase the probability or consequence of
an accident previously evaluated.
Criterion 2: Does the proposed amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
Response: No. The proposed modification alters the RSS pump
circuitry by initiating the start sequence with an existing RWST low-
low level signal instead of a timer. The timer is now used to sequence
pump starts. The pump function is not changed in any way. The proposed
amendment does not introduce failure modes, accident initiators, or
malfunctions that would cause a new or different kind of accident.
Therefore, the proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Criterion 3: Does the proposed amendment involve a significant
reduction in a margin of safety?
Response: No. The proposed change ensures that adequate margin to
suction line flashing and NPSH margin exists for proper operation of
the RSS pumps once the effects of debris are considered as required per
GL 2004-02. Function of the pumps is not affected. Analyses have been
performed that show the containment design basis limits are satisfied
and the post-LOCA offsite and control room doses meet the required
criteria. Therefore, based on the above, the proposed amendment does
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
NRC Section Chief: Darrell J. Roberts.
Duke Energy Corporation, et al., Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina and
Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2,
York County, South Carolina
Date of amendment request: June 29, 2005.
Description of amendment request: The amendment would revise
Technical Specification Bases Section 3.6.11, ``Air Return System
(ARS),'' and the Updated Final Safety Analysis Reports (UFSAR), Section
6.2, ``Containment Systems,'' for McGuire Nuclear Station, Units 1 and
2 and Catawba Nuclear Station, Units 1 and 2. The licensee proposes to
implement an additional manual operator action to respond to NRC
Bulletin 2003-01, ``Potential Impact of Debris Blockage on Emergency
Sump Recirculation at Pressurized-Water Reactors.'' This amendment
would allow plant operators to manually start one air return fan at a
containment pressure of 1 psig prior to the automatic 9 minutes (+ 1
minute) delayed start described in the UFSAR.
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:
First Standard
Does the change involve a significant increase in the probability
or consequences of an accident previously evaluated?
No. The manual start of an Air Return System (ARS) fan will not
result in a significant increase in the probability of an accident
previously evaluated. The starting of an ARS fan is not considered to
be an initiator of any accident or transient. This action is not taken
during normal plant operation, but in response to an accident. The ARS
fans do not operate to provide any normal ventilation requirement. The
Containment Pressure Control System (CPCS) is provided to prevent
excessive depressurization of the containment through inadvertent or
excessive operation of certain engineered safety features. The CPCS
prevents the
[[Page 61658]]
inadvertent actuation of an ARS fan during normal operation.
This change is being requested in order to mitigate the
consequences of a small break loss of coolant accident (SBLOCA) and
help prevent or delay reaching the initiation pressure setpoint for
containment spray, thereby reducing associated problems with possible
sump debris buildup. SBLOCA events are bounded by the consequences of a
design basis large break [loss of coolant accident] LOCA as addressed
in Section 15 of the McGuire and Catawba [Updated Final Safety Analysis
Report] UFSARs. Accordingly, this amendment will not involve a
significant increase in the consequences of an accident previously
evaluated.
Second Standard
Does the change create the possibility of a new or different kind
of accident from any accident previously evaluated?
No. The change proposed in this [license amendment request] LAR
does not involve a physical alteration to the plant (i.e., no new or
different type of equipment will be installed) or a change in the
methods governing any normal plant operation. It does allow for the
early start of one ARS fan during a SBLOCA event with containment
pressure greater than 1 psig and less than 3 psig. This change will not
affect or degrade the ability of the ARS to perform its specified
safety functions.
Accidents of a different type are credible accidents that the
proposed amendment could create that are not bounded by UFSAR evaluated
accidents. This amendment allows for the manual start of an ARS fan
following a SBLOCA within the containment. No new failure modes are
introduced due to the manual start of an ARS fan. The circuit used to
manually start an ARS fan does not interfere with the automatic signal
to start an ARS fan. This change does not require any modifications to
the control circuitry for the ARS. The starting of an ARS fan is not
considered to be an initiator of any accident or transient. This action
(starting of an ARS fan) is not taken during normal operation, but in
response to an accident. Previous accidents considered incredible are
not made more likely by this change. A human performance error, such as
starting the ARS fan too early, too late, or not at all, would not
result in a substantial difference in the calculated differential
pressure across the divider deck. Since no new malfunctions of
equipment with a different result are introduced, all effects of any
malfunctions are bounded by those already evaluated in the UFSAR. Thus
it is concluded that the change contained in this LAR will not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
Third Standard
Does the proposed change involve a significant reduction in a
margin of safety?
No. The early manual start of an ARS fan for SBLOCA events will not
reduce the ability of this system to perform its design functions to
assure the rapid return of air from the upper to the lower containment
compartment after the initial blowdown following a Design Basis
Accident (DBA). The return of this air to the lower compartment and
subsequent recirculation back up through the ice condenser assists in
cooling the containment atmosphere and limiting post accident pressure
and temperature in containment to less than design values. Limiting
pressure and temperature also reduces the release of fission product
radioactivity from containment to the environment in the event of a
DBA. Therefore, there are no adverse dose effects from the early start
of the ARS fan or from the delay of containment spray based on the
current licensing basis.
Analyses have shown that there will be no fan or damper malfunction
due to the early manual start of a fan. The other functions of the
system are not affected by the change proposed in this LAR. The manual
start of the ARS during a SBLOCA will help maintain the margin of
safety by forcing air and steam through the ice condenser with a
subsequent reduction in the rate of pressure increase in the
containment, and a delay in reaching the actuation setpoint for the
containment spray system. The containment spray system will continue to
be initiated at the normal setpoint pressure of the system (-3 psig).
Therefore, the proposed changes listed above 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: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Section Chief: Evangelos C. Marinos.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of amendment request: August 17, 2005.
Description of amendment request: The proposed changes will revise
the Operating License Condition 2.C.(41), Fire Protection Program, to
add a reference to the Nuclear Regulatory Commission (NRC) safety
evaluation that allows the application of National Fire Protection
Agency risk-informed, performance based fire protection methods and
tools that have been approved by the NRC.
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 activity involves the use of a risk-
informed, performance-based method to identify those circuits where a
single fire could damage more than one safe shutdown train. These
circuits would then be provided with one hour rated fire wrap. With the
exception of the fire wrap itself, the proposed activity does not
result in any physical changes to safety-related structures, systems,
or components (SSCs), or the manner in which safety-related SSCs are
operated, maintained, modified, tested, or inspected. The proposed
activity does not degrade the performance or increase the challenges of
any safety-related SSCs assumed to function in the accident analysis.
As a result, the proposed activity does not introduce any new accident
initiators. In addition, fires are not an accident that is previously
evaluated in Chapter 15. Regardless, the proposed activity does not
change the probability of a fire occurring since fire ignition
frequency is independent of the presence of the fire wrap. The
consequences of the proposed activity are bounded by the fire safe
shutdown analysis, which assumes one train is free of fire damage.
Therefore, providing one hour rated fire wrap for those circuits
where a single fire could damage more than one safe shutdown train does
not involve a significant increase in the probability or consequences
of an accident previously evaluated.
[[Page 61659]]
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 activity involves the use of a risk-
informed, performance-based method to identify those circuits where a
single fire could damage more than one safe shutdown train. These
circuits would then be provided with one hour rated fire wrap. With the
exception of the fire wrap itself, the proposed activity does not
result in any physical changes to safety-related structures, systems,
or components (SSCs), or the manner in which safety-related SSCs are
operated, maintained, modified, tested, or inspected. The proposed
activity does not degrade the performance or increase the challenges of
any safety-related SSCs assumed to function in the accident analysis.
As a result, the proposed activity does not introduce nor increase the
number of failure mechanisms of a new or different type than those
previously evaluated. The fire safe shutdown analysis assumes one train
is maintained free of fire damage.
Therefore, providing one hour rated fire wrap for those circuits
where a single fire could damage more than one safe shutdown train 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 activity involves the use of a risk-
informed, performance-based method to identify those circuits where a
single fire could damage more than one safe shutdown train. These
circuits would then be provided with one hour rated fire wrap. With the
exception of the fire wrap itself, the proposed activity does not
result in any physical changes to safety-related structures, systems,
or components (SSCs), or the manner in which safety-related SSCs are
operated, maintained, modified, tested, or inspected. The proposed
activity does not degrade the performance or increase the challenges of
any safety-related SSCs assumed to function in the accident analysis.
The proposed activity does not impact plant safety since the
conclusions of the fire safe shutdown analysis remain unchanged.
Therefore, providing one hour rated fire wrap for those circuits
where a single fire could damage more than one safe shutdown train 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: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1700 K Street, NW., Washington, DC 20006-3817.
NRC Section Chief: David Terao.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: July 21, 2005.
Description of amendment request: The amendment proposes to replace
the existing steam generator tube surveillance program with that being
proposed by the Technical Specification Task Force (TSTF) in TSTF 449,
Revision 4.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No. The proposed change requires a Steam Generator
Program that includes performance criteria that will provide reasonable
assurance that the steam generator (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.
The structural integrity performance criterion is:
Structural integrity performance criterion: All in-service steam
generator tubes shall retain structural integrity over the full range
of normal operating conditions (including startup, operation in the
power range, hot standby, and cool down and all anticipated transients
included in the design specification) and design basis accidents. This
includes retaining a safety factor of 3.0 against burst under normal
steady state full power operation primary to secondary pressure
differential and a safety factor of 1.4 against burst applied to the
design basis accident primary to secondary pressure differentials.
Apart from the above requirements, additional loading conditions
associated with the design basis accidents, or combination of accidents
in accordance with the design and licensing basis, shall also be
evaluated to determine if the associated loads contribute significantly
to burst or collapse. In the assessment of tube integrity, those loads
that do significantly affect burst or collapse shall be determined and
assessed in combination with the loads due to pressure with a safety
factor of 1.2 on the combined primary loads and 1.0 on axial secondary
loads.
The accident induced leakage performance criterion is: The primary
to secondary accident induced leakage rate for any design basis
accidents, other than a SG tube rupture, shall not exceed the leakage
rate assumed in the accident analysis in terms of total leakage rate
for all SGs and leakage rate for an individual SG. Leakage is not to
exceed 540 gallons per day through any one SG, except for specific
types of degradation at specific locations as described in paragraph c
of the Steam Generator Program.
The operational leakage performance criterion is: The RCS
operational primary to secondary leakage through any one SG shall be
limited to <= 75 gallons per day per SG.
A steam generator tube rupture (SGTR) event is one of the design
basis accidents that is 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 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), control element assembly (CEA) ejection, and reactor coolant
pump seized rotor/sheared shaft, the tubes are assumed to retain their
structural integrity (i.e., they are assumed not to rupture). The
accident induced leakage criterion introduced by the proposed changes
account 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 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
[[Page 61660]]
the Steam Generator Program required by the proposed change. The
program, defined by NEI [Nuclear Energy Institute] 97-06, Steam
Generator Program Guidelines, includes a framework that incorporates a
balance of prevention, inspection, evaluation, repair, and leakage
monitoring.
The consequences of design basis accidents are, in part, functions
of the Specific Activity in the primary coolant and the primary to
secondary leakage rates resulting from an accident. Therefore, limits
are included in the plant technical specifications for operational
leakage and for Specific Activity in primary coolant to ensure the
plant is operated within its analyzed condition. For those analyzed
events that do not result in faulted steam generators, greater than or
equal to 75 gpd [gallons per day] primary to secondary leakage per
steam generator is assumed in the analysis. For those analyzed events
that result in a faulted steam generator (e.g., MSLB), 540 gpd primary
to secondary leakage is assumed though the faulted steam generator
while greater than or equal to 75 gpd primary to secondary leakage is
assumed though the intact steam generator.
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 technical specifications 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 Technical
Specifications.
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 other
design basis events.
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated?
Response: No. The proposed performance based requirements are an
improvement over the requirements imposed by the current technical
specifications.
Implementation of the proposed Steam Generator 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 Steam Generator 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. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No. 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 a SG is
maintained by ensuring the integrity of its tubes.
Steam generator 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 Steam Generator Program to manage SG tube
inspection, assessment, repair, and plugging. The requirements
established by the Steam Generator Program are consistent with those in
the applicable design codes and standards and are an improvement over
the requirements in the current technical specifications.
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 Section Chief: David Terao.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of amendment request: July 21, 2005.
Description of amendment request: The requested change will delete
Technical Specification (TS) 6.9.1.2 related to Occupational Radiation
Exposure Reports and TS 6.9.1.5, ``Monthly Operating Reports.''
Basis for proposed no significant hazards consideration
determination: The NRC staff issued a notice of availability of a model
no significant hazards consideration (NSHC) determination for
referencing in license amendment applications in the Federal Register
on June 23, 2004 (69 FR 35067). The licensee affirmed the applicability
of the model NSHC determination in its application dated July 21, 2005.
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 eliminates the Technical
Specifications (TSs) reporting requirements to provide a monthly
operating letter report of shutdown experience and operating statistics
if the equivalent data is submitted using an industry electronic
database. It also eliminates the TS reporting requirement for an annual
occupational radiation exposure report, which provides information
beyond that specified in NRC regulations. The proposed change involves
no changes to plant systems or accident analyses. As such, the change
is administrative in nature and does not affect initiators of analyzed
events or assumed mitigation of accidents or transients. 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 involve a physical
alteration of the plant, add any new equipment, or require any existing
equipment to be operated in a manner different from the present design.
Therefore, the proposed change does not create the possibility of a new
or different kind of accident from any accident previously evaluated.
[[Page 61661]]
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No. This is an administrative change to reporting
requirements of plant operating information and occupational radiation
exposure data, and has no effect on plant equipment, operating
practices or safety analyses assumptions. For these reasons, the
proposed change does not involve a significant reduction in the margin
of safety.
Based upon the reasoning presented above, the requested change does
not involve significance hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Michael L. Marshall, Jr.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: September 1, 2005.
Description of amendment request: The requested change will delete
Technical Specification (TS) 6.9.1.2 related to Occupational Radiation
Exposure Reports and TS 6.9.1.6, ``Monthly Operating Reports.''
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?
No. The proposed change eliminates the Technical Specifications
(TSs) reporting requirements to provide a monthly operating report of
shutdown experience and operating statistics if the equivalent data is
submitted using an industry electronic database. It also eliminates the
TS reporting requirement for an annual occupational radiation exposure
report, which provides information beyond that specified in NRC
regulations. The proposed change involves no changes to plant systems
or accident analyses. As such, the change is administrative in nature
and does not affect initiators of analyzed events or assumed mitigation
of accidents or transients. 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?
No. The proposed change does not involve a physical alteration of
the plant, add any new equipment, or require any existing equipment to
be operated in a manner different from the present design. 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?
No. This is an administrative change to reporting requirements of
plant operating information and occupational radiation exposure data,
and has no effect on plant equipment, operating practices or safety
analyses assumptions. For these reasons, the proposed change does not
involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Michael L. Marshall, Jr.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: July 21, 2005.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) testing frequency for the
surveillance requirement (SR) in TS 3.1.4, ``Control Rod Scram Times.''
Specifically, the proposed change would revise the frequency for SR
3.1.4.2, control rod scram time testing, from ``120 days cumulative
operation in MODE 1'' to ``200 days cumulative operation in MODE 1.''
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in licensing amendment applications in the Federal Register on August
23, 2004 (69 FR 51864). The licensee affirmed the applicability of the
model NSHC determination in its application dated July 21, 2005.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No. The proposed change extends the frequency for testing
control rod scram time testing from every 120 days of cumulative Mode 1
operation to 200 days of cumulative Mode 1 operation. The frequency of
surveillance testing is not an initiator of any accident previously
evaluated. The frequency of surveillance testing does not affect the
ability to mitigate any accident previously evaluated, as the tested
component is still required to be operable. Therefore, the proposed
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No. The proposed change extends the frequency for testing
control rod scram time testing from every 120 days of cumulative Mode 1
operation to 200 days of cumulative Mode 1 operation. The proposed
change does not result in any new or different modes of plant
operation. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No. The proposed change extends the frequency for testing
control rod scram time testing from every 120 days of cumulative Mode 1
operation to 200 days of cumulative Mode 1 operation. The proposed
change continues to test the control rod scram time to ensure the
assumptions in the safety analysis are protected. Therefore, the
proposed change does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John R. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
[[Page 61662]]
NRC Section Chief: David Terao.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of amendment request: November 12, 2004, as supplemented by
letters dated September 2 and September 16, 2005.
Description of amendment request: The proposed amendments would
revise Technical Specifications 3.1.7, ``Standby Liquid Control (SLC)
System,'' for Hatch, Units 1 and 2. The proposed amendments would
update Figures 3.1.7-1 and 3.1.7-2 for Units 1 and 2 TS to reflect the
increased concentration of Boron-10 in the solution. Conforming
revisions to Bases B 3.1.7, ``Standby Liquid Control (SLC) System'' are
also included.
The proposed amendment was previously noticed on February 1, 2005
(70 FR 5249).
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?
This is a proposed change to Figures 3.1.7-1 and 3.1.7-2 of the
Units 1 and 2 TS [Technical Specifications]. Figure 3.1.7-1 is a plot
of the weight percent of Sodium Pentaborate solution in the Standby
Liquid Control (SLC) Tank, as a function of the gross volume of
solution in the tank. Figure 3.1.7-2 is a plot of the Sodium
Pentaborate temperature versus concentration requirements.
Figure 3.1.7-1 is proposed to be changed in order to accommodate an
injection of Sodium Pentaborate solution into the reactor, following an
ATWS [anticipated transient without scram] event, such that the
concentration of Boron-10 atoms in the reactor will be 800 ppm natural
Boron equivalent. This is necessary to accommodate increased cycle
energy requirements for the Hatch Units 1 and 2 cores. Both Figures
3.1.7-1 and 3.1.7-2 are changed to reflect that the boundary between
Region A and B is changing from 6.9% to 7.0%. The proposed change to
the Figures will not increase the probability of an ATWS event because
the curves have nothing to do with the prevention of an ATWS event. The
new requirements will insure that, in the future, the core will have
adequate shutdown margin to mitigate the consequences of an ATWS event.
The minimum concentration of Sodium Pentaborate which also
represents the boundary between Region A and Region B, is changing from
6.9% to 7.0%. This increase in the concentration ensures a conservative
margin to the ATWS equivalency determination required by 10 CFR 50.62.
Also, no systems or components designed to ensure the safe shutdown
of the reactor are being physically changed as a result of this
proposed TS change. In fact, no safety related systems or components
designed for the prevention of previously evaluated events are being
altered by the amendment.
2. The proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
This proposed TS revision results in a change to SLC TS Figures
3.1.7-1 and 3.1.7-2 requirements. However, these changes do not result
in physical changes to the SLC system. SLC pump operation, maintenance
and testing remain the same. Accordingly, no changes to the operation,
maintenance or surveillance procedures will result from this TS
revision request. Therefore, no new modes of operation are introduced
by this TS change.
Since no new modes of operation are introduced, the proposed change
does not create the possibility of a new or different type event from
any previously evaluated.
3. The proposed change does not involve a significant reduction in
the margin of safety.
This proposed TS change is being made to increase the boron
concentration requirements of the sodium pentaborate solution injected
into the reactor vessel following an Anticipated Transient Without
Scram (ATWS) event. The change is necessary due to new fuel designs and
higher energy requirements for fuel cycles. Therefore, the change is
being made to insure that shutdown requirements can be met for the ATWS
event. This will insure the margin of safety with respect to ATWS will
continue to be met.
The increase in the minimum concentration from 6.9% to 7.0% ensures
a conservative margin with respect to the ATWS equivalency
determination. Consequently, this proposed TS change will not result in
a decrease 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: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Section Chief: Evangelos C. Marinos.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: September 1, 2005 (TS-05-04).
Description of amendment request: The proposed amendment would
revise the reactor protection system turbine trip allowable value for
low trip system pressure from greater than or equal to 43 pounds per
square inch gauge (psig) to 39.5 psig. This change would allow the
instrumentation that performs this trip function to be tested and
verified to be operable within the capabilities of the pressure
switches.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No. The proposed change revises the allowable value for
reactor trip as a result of a turbine trip on low trip system pressure.
This change will not alter any plant components, systems, or processes
and will only provide a more appropriate value to assess operability of
the associated pressure switches. Since the plant features and
operating practices are not altered, the possibility of an accident is
not affected. This reactor trip is not directly credited in SQN's
accident analysis and is maintained as an anticipatory trip to enhance
the overall reliability of the reactor trip system. As such, there is
not a specific safety limit associated with this function and the
generation of a reactor trip based on low trip system pressure is above
the required actuations to ensure acceptable mitigation of accidents.
As the proposed change will continue to provide an acceptable
anticipatory trip signal, the offsite dose potential is not affected by
this change. Therefore, the proposed change does not involve a
significant increase in the probability or
[[Page 61663]]
consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No. As described above, this change will not alter any
plant equipment or operating practices that have the ability to create
a new potential for accident generation. The proposed change revises
the operability limits for a function that generates a trip signal when
appropriate conditions exist to require accident mitigation response.
This type of function does not have the ability to create an accident
as its purpose and function is to mitigate events. Therefore, the
proposed change does not create the possibility of a new or different
kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No. The proposed change will revise an allowable value
for a reactor trip initiator that results from a turbine trip
condition. This change will not alter the setpoint, and the calibration
of the associated pressure switches will continue to be set at the
current values. The allowable value change is in response to accuracy
aspects of the instrumentation and does not alter the ability of this
trip function to operate when and as needed to mitigate accident
conditions.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Michael L. Marshall, Jr.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: August 26, 2005.
Description of amendment request: The amendment would authorize
changes to the Updated Safety Analysis Report (USAR) for Wolf Creek
Generating Station (WCGS) that would revise the methodology for the
reactor coolant system (RCS) leak detection instrumentation. This
revision would clarify the requirements of the containment atmosphere
gaseous radioactivity monitor with regard to the RCS leak detection
capability and would justify that the monitor can be considered
operable in compliance with Limiting Condition for Operation 3.4.15, in
Technical Specification (TS) 3.4.15, ``RCS Leakage Detection
Instrumentation,'' during all applicable Modes. There are no proposed
changes to the WCGS 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. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change has been evaluated and determined to not
increase the probability or consequences of an accident previously
evaluated. The proposed change does not make hardware changes and does
not alter the configuration of any plant system, structure, or
component (SSC). The proposed change only clarifies the design and
OPERABILITY requirements for the containment atmosphere gaseous
radioactivity monitors and identifies the capabilities of the monitors
at low RCS [radio]activity levels. The containment atmosphere gaseous
radioactivity monitors are not initiators of any accident; therefore,
the probability of occurrence of an accident is not increased. The USAR
and TSs will continue to require diverse means of [RCS] leakage
detection equipment, thus ensuring that leakage due to cracks [in the
RCS] would continue to be identified prior to propagating to the point
of a[n] [RCS] pipe break. Therefore, the consequences of an accident
[previously evaluated] are not increased.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change does involve the use or installation of new
equipment and the currently installed equipment will not be operated in
a new or different manner. No new or different system interactions are
created and no new processes are introduced. The proposed changes will
not introduce any new failure mechanisms, malfunctions, or accident
initiators not already considered in the design and licensing basis
[for WCGS]. The proposed change does not affect any SSC associated with
an accident initiator. Based on this evaluation, the proposed change
does not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. The proposed change does not involve a significant reduction in
a margin of safety.
The proposed change does not alter any RCS leakage detection
components. The proposed change only clarifies the design and
operability requirements for the containment atmosphere gaseous
radioactivity monitor and identifies the capabilities of the
containment atmosphere gaseous radioactivity monitors at low RCS
[radio]activity levels. This change is required since the level of
radioactivity in the WCGS reactor coolant has become much lower than
what was assumed in the USAR and the gaseous channel [(monitor)] can no
longer promptly detect a small RCS leak under all operating conditions.
The proposed amendment continues to require diverse means of [RCS]
leakage detection equipment with [the] capability to promptly detect
RCS leakage. Although not required by [the] TS[s], additional diverse
means of leakage detection capability are available as described in the
USAR Section 5.2.5. Early detection of [RCS] leakage, as the potential
indicator of a crack(s) in the RCS pressure boundary, will thus
continue to be in place so that such a condition is known and
appropriate actions [are] taken well before any such crack would
propagate to a more severe condition. Based on this evaluation, the
proposed change does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Daniel S. Collins, Acting.
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
[[Page 61664]]
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].
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of application for amendment: October 20, 2004, as
supplemented by letters dated June 30, July 29, August 17, and
September 19, 2005.
Brief description of amendment: The amendment revised the Technical
Specifications to (1) eliminate the existing requirement in Section
3.8.6 regarding maintaining the containment equipment hatch cover in
place with a minimum of four bolts during fuel loading and refueling
operations, and (2) revise or introduce commitments to the Technical
Specifications Bases in support of the change in Section 3.8.6.
Date of issuance: October 13, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 257.
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 7, 2004 (69 FR
70714) The supplements dated June 30, July 29, August 17, and September
19, 2005, provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the NRC staff's original proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated October 13, 2005.
No significant hazards consideration comments received: No.
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 28, 2003, as supplemented
by letters dated January 22 and June 23, 2004, and February 2 and
September 27, 2005.
Brief description of amendments: The amendments revise several
surveillance requirements (SRs) in Technical Specification (TS) 3.8.1
on alternating current power sources and SR 3.8.4.6 for direct current
power sources for plant operation. The revised SRs have notes deleted
or modified to adopt in part the staff-approved TS Task Force 283,
Revision 3, to allow these SRs to be performed, or partially performed,
in reactor modes that previously were not allowed by the TSs.
Date of issuance: September 29, 2005.
Effective date: September 29, 2005, and shall be implemented within
90 days of the date of issuance including the incorporation of the
changes to the TS Bases for TS 3.8.1 and SR 3.8.4.6 as described in the
licensee's letters dated May 28, 2003, January 22 and June 23, 2004,
and February 2 and September 27, 2005.
Amendment Nos.: Unit 1--156, Unit 2--156, Unit 3--156.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: July 8, 2003 (68 FR
40709).
The supplemental letters dated January 22, June 23, 2004, and
February 2 and September 27, 2005, do not expand the scope of the
application as originally noticed and do 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 September 29, 2005.
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: February 3, 2005.
Brief description of amendment: The amendment modified Technical
Specification (TS) 6.16.b.1, ``Radioactive Effluent Controls Program,''
and TS 6.18, ``Off-site Dose Calculation Manual (ODCM),'' to be
consistent with Title 10 of the Code of Federal Regulations (10 CFR)
part 20 and NUREG-1431, ``Standard Technical Specifications
Westinghouse Plants.''
Date of issuance: October 4, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 186.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 29, 2005 (70 FR
15944).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 4, 2005.
No significant hazards consideration comments received: No.
Duke Energy Corporation, 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: November 25, 2002, as
supplemented by letters dated November 13 and December 16, 2003,
September 22, 2004, April 6, June 14, July 8, August 17, and September
8 and September 19, 2005.
Brief description of amendments: The amendments include a full-
scope implementation of an alternative source term for evaluating the
consequences of design basis accidents at Catawba Nuclear Station. The
amendments also revised the Technical Specifications for the
Ventilation Filter Testing Program, Annulus Ventilation System,
Auxiliary Building Filtered Ventilation Exhaust
[[Page 61665]]
System, Fuel Handling Ventilation Exhaust System, and Control Room Area
Ventilation System, and containment penetrations.
Date of issuance: September 30, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 227 and 222.
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: April 15, 2003 (68 FR
18272). This application was renoticed on May 24, 2005 (70 FR 29789).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 30, 2005.
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 19, 2005.
Brief description of amendment: The amendment revised technical
specifications (TSs) testing frequency for the surveillance requirement
(SR) in TS 3.1.4, ``Control Rod Scram Times.'' Specifically, the change
revised the frequency for SR 3.1.4.2, ``Control Rod Scram Time
Testing,'' from ``120 days cumulative operation in MODE 1'' to ``200
days cumulative operation in MODE 1.''
Date of issuance: September 29, 2005.
Effective date: September 29, 2005, and shall be implemented within
60 days from the date of issuance.
Amendment No.: 194.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 7, 2005 (70 FR
33212).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 29, 2005.
No significant hazards consideration comments received: No.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: December 17, 2004, as supplemented by
letters dated June 29, and August 12, 2005.
Brief description of amendment: The amendment revises the Technical
Specification (TS) requirements for direct current (DC) sources. The
current TS only includes ACTION Statements for an inoperable DC Power
subsystems. The change adds a new ACTION Statement to TS 3.8.4, ``DC
Sources--Operating,'' to specifically address an inoperable battery
charger.
Date of issuance: October 7, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 148.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 4, 2005 (70 FR
401). The supplements dated June 29, and August 12, 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 October 7, 2005.
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: December 20, 2004, as
supplemented by letter dated April 26, 2005.
Brief description of amendment: The amendment changed the existing
containment structures and tendon inservice inspection requirements to
be consistent with NUREG-1432, Revision 3, and the American Society of
Mechanical Engineers Boiler and Pressure Vessel Code, Section XI.
Specifically, the amendment modified the Surveillance Requirement of
Technical Specification (TS) 3.6.1.5, added a new Surveillance Program
to TS 6.5.6 and a report to TS 6.5.7, and made two administrative
changes to the TSs.
Date of issuance: September 29, 2005.
Effective date: As of the date of issuance to be implemented within
90 days from the date of issuance.
Amendment No.: 262.
Renewed Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 29, 2005 (70 FR
15943).
The supplement dated April 26, 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 September 29, 2005.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: September 21, 2004, as
supplemented by letters dated March 18, April 7, May 6, August 10, and
September 19, 2005.
Brief description of amendments: The amendments extended the outage
times from 72 hours to 14 days for an inoperable emergency diesel
generator. It also changed formats of the affected technical
specification pages to improve their appearance but not alter any
requirements.
Date of issuance: September 30, 2005.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment Nos.: 291, 273.
Facility Operating License Nos. DPR-58 and DPR-74: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 26, 2004 (69 FR
62476). The supplements provided clarifying information that did not
change the scope of the proposed amendment as described in the original
notice of proposed action published in the Federal Register, and did
not change the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 30, 2005.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: June 30, 2004, as supplemented
by letters dated September 16, 2004, November 5, 2004, March 3, 2005,
July 1, 2005, and September 27, 2005.
Brief description of amendment: The amendment changed the TSs to
support an increase in the length of the fuel cycle from 18 to 24
months at Monticello. In addition, the proposed amendment requested
changes in calibration times of various instruments. These changes will
be evaluated in a separate license amendment.
[[Page 61666]]
Date of issuance: September 30, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 143.
Facility Operating License No. DPR-22: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 18, 2005 (70 FR
2892). The supplements dated September 16, 2004, November 5, 2004,
March 3, 2005, July 1, 2005, and September 27, 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 on January 18, 2005
(70 FR 2892).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 30, 2005.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: November 23, 2004, as supplemented by
letter dated July 8, 2005.
Brief description of amendment: The amendment (1) revised the
descriptive wording of Technical Specifications (TSs) Table 1-1, ``RPS
[reactor protection system] Limiting Safety System Settings,'' for the
reactor trip setpoint for low steam generator water level to relocate
unnecessary detail, and (2) converted TSs Section 4.0, ``Design
Features,'' to the format and content of NUREG-1432, Revision 3,
``Standard Technical Specifications for Combustion Engineering
Plants.''
Date of issuance: October 3, 2005.
Effective date: October 3, 2005, and shall be implemented within 60
days from the date of issuance.
Amendment No.: 236.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: May 24, 2005 (70 FR
29798).
The July 8, 2005, supplemental 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 no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated October 3, 2005.
No significant hazards consideration comments received: No.
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 application for amendments: September 8, 2004, as
supplemented by letters dated July 8 and September 28, 2005.
Brief description of amendments: The amendments changed SSES 1 and
2 Technical Specifications 3.6.4.1, ``Secondary Containment,'' and
3.6.4.3, ``Standby Gas Treatment System (SGTS),'' to extend, on a one-
time basis, the allowable completion time for required actions for
secondary containment inoperable and two SGTS subsystems inoperable, in
mode 1, 2, or 3, from 4 hours to 48 hours.
Date of issuance: October 6, 2005.
Effective date: October 6, 2005.
Amendment Nos.: 226 and 203.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 1, 2005 (70 FR
9994). The supplements dated July 8 and September 28, 2005, contained
clarifying information and did not change the initial no significant
hazards consideration determination or expand the scope of the initial
Federal Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 6, 2005.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of application for amendments: April 15, 2004, as supplemented
by letters dated August 11, 2004, and August 11, 2005. The August 11,
2005, supplement withdrew a portion of the original application from
consideration.
Brief description of amendments: The amendments revised Technical
Specification (TS) 3.6.2.3 Action B, for both units, to correct a non-
conservative action statement.
Date of issuance: September 30, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 266 and 248.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the TSs.
Date of initial notice in Federal Register: October 12, 2004 (69 FR
60684). The licensee's supplement dated August 11, 2005, withdrew a
portion of the original application from consideration and did not
increase the scope of the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 30, 2005.
No significant hazards consideration comments received: No.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of application for amendments: March 24, 2005.
Brief description of amendments: The proposed changes revised
various Technical Specifications (TSs) related to cycle-specific values
and the shutdown margin, and are consistent with the following Nuclear
Regulatory Commission approved Technical Specification Task Force
(TSTF) Standard TS Change Travelers: TSTF-9-A, Revision 1, ``Relocate
Value for Shutdown Margin to COLR;'' TSTF-67-A, Revision 0,
``Correction of Shutdown Margin Definition;'' TSTF-142-A, Revision 0,
``Increase the Completion Time When the Core Reactivity Balance is Not
Within Limit;'' and TSTF-150-A, Revision 0, ``Replace DNBR Power
Decrease Number with Reference to the COLR.''
Date of issuance: October 3, 2005.
Effective date: As of the date of issuance, and shall be
implemented within 60 days from the date of issuance.
Amendment Nos.: 200/191.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 10, 2005 (70 FR
24656).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 3, 2005.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: July 4, 2005.
Brief description of amendments: The amendments change Technical
Specification 4.0.5 to add a reference to the NRC-approved exemption of
selected pumps, valves, and other components from special treatment
requirements. As an editorial change, references to Title 10, Code of
Federal Regulations (10 CFR) part 50, section
[[Page 61667]]
50.55a(f) and 10 CFR part 50, section 50.55a(f)(6)(I) is added to the
paragraph for inservice testing, similar to the existing references for
inservice inspection. In addition, ``inservice testing'' and
``inservice inspection'' are reordered for consistency with the
sequence of the regulations in 10 CFR 50.55a.
Date of issuance: October 4, 2005.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: Unit 1--173; Unit 2--161.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 2, 2005 (70 FR
44403).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 4, 2005.
No significant hazards consideration comments received: No.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: September 9, 2004.
Brief description of amendments: The Amendments revised the
Technical Specification 3.6.6.8 to change the current interval for
surveillance from every 10 years to verification that the nozzles are
unobstructed following a maintenance that could have resulted in nozzle
blockage.
Date of issuance: September 23, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 120 and 120.
Facility Operating License Nos. NPF-87 and NPF-89: The Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 26, 2004 (69 FR
62478).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 23, 2005.
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: September 17, 2004, as
supplemented by letters dated February 11, May 26, June 17 (two
letters), July 15, July 29, August 16, and September 6, 2005.
Brief description of amendment: The amendment supports the
installation of replacement steam generators (SGs) at Callaway during
the refueling outage in the fall of 2005. The amendment affects the
following affected TSs: the reactor core safety limits (TS 2.1.1),
reactor trip system and engineered safety feature actuation system
instrumentation (TSs 3.3.1 and 3.3.2), reactor coolant system (RCS)
limits (TS 3.4.1), RCS loops (TSs 3.4.5, 3.4.6, and 3.4.7), RCS
operational leakage (TS 3.4.13), SG tube integrity (the new TS 3.4.17),
main steam safety valves (TS 3.7.1), SG tube surveillance program (TS
5.5.9), containment integrated leakage rate testing program (TS
5.5.16), and SG tube inspection report (TS 5.6.10).
Date of issuance: September 29, 2005.
Effective date: Effective on the date of issuance, and shall be
implemented before entry into Mode 5 during the restart from the fall
2005 refueling outage when the replacement steam generators are
installed including (1) revising the pressure temperature limits report
to change the cold overpressure mitigation system setpoints to reflect
no reactor coolant pump operation restrictions and (2) incorporating
the TS Bases changes identified in the licensee's letter of September
6, 2005, into the TS Bases.
Amendment No.: 168.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 23, 2004 (69
FR 68185). The supplemental letters dated February 11, May 26, June 17
(two letters), July 15, July 29, August 16, and September 6, 2005,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed and did not
change the NRC staff's original proposed no significant hazards
consideration determination published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 29, 2005.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
[[Page 61668]]
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room (PDR), located at One White Flint North, Public File Area 01F21,
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly
available records will be accessible from the Agencywide Documents
Access and Management System's (ADAMS) Public Electronic Reading Room
on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems
in accessing the documents located in ADAMS, contact the PDR Reference
staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to [email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. Within 60 days after the date
of publication of this notice, the licensee may file a request for a
hearing with respect to issuance of the amendment to the subject
facility operating license and any person whose interest may be
affected by this proceeding and who wishes to participate as a party in
the proceeding must file a written request for a hearing and a petition
for leave to intervene. Requests for a hearing and a petition for leave
to intervene shall be filed in accordance with the Commission's ``Rules
of Practice for Domestic Licensing Proceedings'' in 10 CFR part 2.
Interested persons should consult a current copy of 10 CFR 2.309, which
is available at the Commission's PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland, and electronically on the Internet at the NRC Web site,
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR Reference staff at
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. If a
request for a hearing or petition for leave to intervene is filed by
the above date, the Commission or a presiding officer designated by the
Commission or by the Chief Administrative Judge of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the Chief Administrative Judge of the Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion. The
petition must include sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or
fact.\1\ Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A petitioner/requestor
who fails to satisfy these requirements with respect to at least one
contention will not be permitted to participate as a party.
---------------------------------------------------------------------------
\1\ To the extent that the applications contain attachments and
supporting documents that are not publicly available because they
are asserted to contain safeguards or proprietary information,
petitioners desiring access to this information should contact the
applicant or applicant's counsel and discuss the need for a
protective order.
---------------------------------------------------------------------------
Each contention shall be given a separate numeric or alpha
designation within one of the following groups:
1. Technical: primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the
applications.
2. Environmental: primarily concerns/issues relating to matters
discussed or referenced in the environmental analysis for the
applications.
3. Miscellaneous: does not fall into one of the categories outlined
above.
As specified in 10 CFR 2.309, if two or more petitioners/requestors
seek to co-sponsor a contention, the petitioners/requestors shall
jointly designate a representative who shall have the authority to act
for the petitioners/requestors with respect to that contention. If a
petitioner/requestor seeks to adopt the contention of another
sponsoring petitioner/requestor, the petitioner/requestor who seeks to
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the
petitioners/requestors with respect to that contention.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-
[[Page 61669]]
0001, Attention: Rulemaking and Adjudications Staff; (2) courier,
express mail, and expedited delivery services: Office of the Secretary,
Sixteenth Floor, One White Flint North, 11555 Rockville Pike,
Rockville, Maryland 20852, Attention: Rulemaking and Adjudications
Staff; (3) E-mail addressed to the Office of the Secretary, U.S.
Nuclear Regulatory Commission, [email protected]; or (4) facsimile
transmission addressed to the Office of the Secretary, U.S. Nuclear
Regulatory Commission, Washington, DC, Attention: Rulemakings and
Adjudications Staff at (301) 415-1101, verification number is (301)
415-1966. A copy of the request for hearing and petition for leave to
intervene should also be sent to the Office of the General Counsel,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and it
is requested that copies be transmitted either by means of facsimile
transmission to (301) 415-3725 or by e-mail to [email protected]. A
copy of the request for hearing and petition for leave to intervene
should also be sent to the attorney for the licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer or the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(I)-(viii).
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: August 31, 2005, as supplemented by
letter dated September 13, 2005.
Brief description of amendment: The amendment permitted a one-time
change to Technical Specification Table 3.3.8.1-1 to provide a one-time
relaxation of the Loss of Power instrumentation requirements.
Date of issuance: September 15, 2005.
Effective date: As of the date of issuance to be implemented
immediately.
Amendment No.: 147.
Facility Operating License No. NPF-47: Amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: Yes.
The NRC published a public notice of the proposed amendment, issued
a proposed finding of no significant hazards consideration, and
requested that any comments on the proposed no significant hazards
consideration be provided to the NRC staff by the close of business on
September 9, 2005. The notice was published in The St. Francisville
Democrat (in St. Francisville) on September 8, 2005, and The Advocate
(in Baton Rouge) on September 7, 2005. No public comments were
received.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, consultation with the State of Louisiana, and
final no significant hazards consideration determination are contained
in a Safety Evaluation dated September 15, 2005.
XU Generation Company LP, Docket No. 50-445, Comanche Peak Steam
Electric Station, Unit No. 1, Somervell County, Texas
Date of amendment request: April 27, 2005 as supplemented by letter
dated July 20, 2005.
Description of amendment: The amendments revise the Technical
Specifications to add the topical report WCAP-13060-P-A to the list of
NRC approved methodologies to be used at Comanche Peak Steam Electric
Station, Unit 1.
Date of issuance: October 11, 2005.
Effective date: As of the date of issuance and shall be implemented
immediately.
Amendment No.: 123.
Facility Operating License No. NPF-87: Amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes.
The notice published on September 26, 2005 (70 FR 56191) provided
an opportunity to submit comments on the Commission's proposed NSHC
determination. No comments have been received. The notice also provided
an opportunity to request a hearing within 60 days from the date of
publication, but indicated that if the Commission makes a final NSHC
determination, any such hearing would take place after issuance of the
amendment.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, state consultation, and final NSHC determination
are contained in a safety evaluation dated October 11, 2005.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: David Terao.
Dated at Rockville, Maryland, this 17th day of October, 2005.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 05-21180 Filed 10-24-05; 8:45 am]
BILLING CODE 7590-01-P