[Federal Register Volume 70, Number 215 (Tuesday, November 8, 2005)]
[Notices]
[Pages 67744-67757]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 05-22002]
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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 October 14, 2005 to October 27, 2005. The
last biweekly notice was published on October 25, 2005 (70 FR 61655).
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
01F21, 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
[[Page 67745]]
Panel, will rule on the request and/or petition; and the Secretary or
the Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1 (HNP), Wake and Chatham Counties,
North Carolina
Date of amendment request: August 18, 2005.
Description of amendment request: The amendment will allow the use
of fire-resistive electrical cable, which has been demonstrated to
provide an equivalent level of protection as would be provided by 3-
hour and 1-hour rated electrical cable raceway fire barriers, for the
protection of safe shutdown electrical cable.
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.
Operation of HNP in accordance with the proposed amendment does
not increase the probability or consequences of accidents previously
evaluated. The Final Safety Analysis Report (FSAR) documents the
analyses of design basis accidents (DBA) at HNP. Any scenario or
previously analyzed accidents that result in offsite dose were
evaluated as part of this analysis. The proposed amendment does not
adversely affect accident initiators nor alter design assumptions,
conditions, or configurations of the facility. The proposed
amendment does not adversely affect the ability of structures,
systems, or components (SSCs) to perform their design function. SSCs
required to safely shut down the reactor and to maintain it in a
safe shutdown condition remain capable of performing their design
functions.
The purpose of this amendment is to assure that redundant trains
of safe shutdown (SSD) control circuits remain protected from damage
in the event of a postulated fire. The proposed amendment revises
the Final Safety Analysis Report (FSAR) to use three-hour fire-
resistive electrical cable, which has been demonstrated to provide
an equivalent level of protection as would be provided by three-hour
and one-hour rated electrical cable raceway fire barriers, for the
protection of
[[Page 67746]]
SSD electrical cables. Based on the above, SSD control circuit
protection is maintained by this amendment.
Therefore, this amendment 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.
Operation of HNP in accordance with the proposed amendment does
not create the possibility of a new or different kind of accident
from any accident previously evaluated. The FSAR documents the
analyses of design basis accidents (DBA) at HNP. Any scenario or
previously analyzed accidents that result in offsite dose were
evaluated as part of this analysis. The proposed amendment does not
change or affect any accident previously evaluated in the FSAR, and
no new or different scenarios are created by the proposed amendment.
The proposed amendment does not adversely affect accident initiators
nor alter design assumptions, conditions, or configurations of the
facility. The proposed amendment does not adversely affect the
ability of SSCs to perform their design function. SSCs required to
safely shut down the reactor and to maintain it in a safe shutdown
condition remain capable of performing their design functions.
The purpose of this amendment is to assure that redundant trains
of Safe Shutdown (SSD) control circuits remain protected from damage
in the event of a postulated fire. The proposed amendment revises
the Final Safety Analysis Report (FSAR) to use three-hour fire-
resistive electrical cable, which has been demonstrated to provide
an equivalent level of protection as would be provided by three-hour
and one-hour rated electrical cable raceway fire barriers, for the
protection of SSD electrical cables. Based on the above, SSD control
circuit protection is maintained by this amendment.
Therefore, this amendment 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.
Operation of HNP in accordance with the proposed amendment does
not involve a significant reduction in a margin of safety. The
proposed amendment does not alter the manner in which safety limits,
limiting safety system settings or limiting conditions for operation
are determined. The safety analysis acceptance criteria are not
affected by this change. The proposed amendment does not adversely
affect existing plant safety margins or the reliability of equipment
assumed to mitigate accidents in the FSAR. The proposed amendment
does not adversely affect the ability of SSCs to perform their
design function. SSCs required to safely shut down the reactor and
to maintain it in a safe shutdown condition remain capable of
performing their design functions.
The purpose of this amendment is to assure that redundant trains
of Safe Shutdown (SSD) control circuits remain protected from damage
in the event of a postulated fire. The proposed amendment revises
the Final Safety Analysis Report (FSAR) to use three-hour fire-
resistive electrical cable, which has been demonstrated to provide
an equivalent level of protection as would be provided by three-hour
and one-hour rated electrical cable raceway fire barriers, for the
protection of SSD electrical cables. Based on the above, SSD control
circuit protection is maintained by this amendment.
Therefore, this 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: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602.
NRC Section Chief: Michael L. Marshall, Jr.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: September 1, 2005.
Description of amendment request: The amendment will add Technical
Specification (TS) 3.7.14, ``Fuel Storage Pool Boron Concentration''
and revise TS 5.6, ``Fuel Storage.'' The proposed changes are related
to requirements for ensuring adequate subcriticality margin in the
spent fuel storage pools. TS 5.6.1 is being revised to include the
design requirements for dry storage of new fuel.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes do not modify the facility. The accident
previously analyzed for the spent fuel pool is a fuel handling
accident. The proposed change applies administrative controls for
maintaining the required boron concentration in the spent fuel
storage pools, revises acceptance criteria and storage arrangements
for fuel storage in PWR [pressurized-water reactor] ``flux trap''
style racks and adds acceptance criteria for dry storage of new fuel
to the Technical Specifications. The controls on spent fuel pool
boron and dry storage of new fuel have previously been implemented
but are being added to the Technical Specifications as requirements.
The proposed change applies new acceptance criteria for criticality
safety of fuel storage in PWR ``flux trap'' style racks in Pools
``A'' and ``B.'' The new acceptance criteria require new
administrative controls on the placement of fuel in Pools ``A'' and
``B.'' Similar administrative controls have previously been placed
on fuel stored in Pools C and D. These changes will eliminate the
dependence on Boraflex in the PWR ``flux trap'' style storage racks.
These changes do not impact the probability of having a fuel
handling accident and do not impact the consequences of a fuel
handling accident.
Therefore, this amendment 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.
No change is being made to the acceptance criteria of the dry
storage of new fuel. These criteria are being added to Technical
Specification Section 5.6.1. Detailed analyses have been performed
to ensure a criticality accident in Pools ``A'' and ``B'' is not a
credible event. The events that could lead to a criticality accident
are not new. These events include a fuel mis-positioning event, a
fuel drop event, and a boron dilution event. The proposed changes do
not impact the probability of any of these events. The detailed
criticality analyses performed demonstrate that criticality would
not occur following any of these events. For the more likely event,
such as a fuel mis-positioning event, the acceptance criteria for
keff remains less than or equal to 0.95. For the unlikely
event that the spent fuel storage pool boron concentration was
reduced to zero, keff remains less than 1.0.
Therefore, a criticality accident remains ``not credible,'' and
this amendment 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.
Incorporation of acceptance criteria for dry storage of new fuel
into TS 5.6.1 does not involve a reduction in the margin of safety.
The new fuel storage condition continues to meet keff <=
0.95 during normal conditions and keff <= 0.98 under
optimal moderation conditions.
The proposed changes for storage of new and irradiated fuel in
Pools ``A'' and ``B'' continue to provide the controls necessary to
ensure a criticality event could not occur in the spent fuel storage
spool. The acceptance criteria are consistent with the acceptance
criteria specified in 10 CFR 50.68, which provide an acceptable
margin of safety with regard to the potential for a criticality
event.
Therefore, this amendment does not involve a significant
reduction in a margin of safety.
[[Page 67747]]
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602.
NRC Section Chief: Michael L. Marshall, Jr.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: September 26, 2005.
Description of amendment request: The proposed amendment will
revise the analysis method used for the large-break loss-of-coolant
accident (LBLOCA) by incorporating the use of a new approach (ASTRUM)
for the treatment of parameter uncertainties. The new approach is
described in Westinghouse Topical Report WCAP-16009-P-A, approved by
the NRC on November 5, 2004.
Changes to the Technical Specifications to reflect the proposed use
of ASTRUM in LBLOCA analysis consist of revisions to the list of
references provided in Technical Specification Section 5.6.5, Core
Operating Limits Report.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change modifies the analysis methodology used to
account for the variation in parameters that are used for the safety
analysis of the LBLOCA. This proposed change has no effect on the
design or operation of plant equipment. Use of the new methodology
will revise the results of the current analysis, but there will be
no change in initiating events for this accident scenario or the
ability of the plant equipment or plant operators to respond.
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 modifications to existing
plant equipment or the installation of any new equipment. The
proposed change only affects the analysis methodology that is used
to evaluate the response of existing plant equipment to the LBLOCA
scenario. Plant operating and emergency procedures that are in place
for the LBLOCA scenario are also not being changed by this proposed
amendment. This proposed change does not create new failure modes or
malfunctions of plant equipment nor is there a new credible failure
mechanism.
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 license amendment revises the analysis methodology
which is used to assess the impact of the LBLOCA scenario with
respect to established acceptance criteria. Margins of safety for
LBLOCA include quantitative limits for fuel performance established
in 10 CFR 50.46. These acceptance criteria and the associated
margins of safety are not being changed. The evaluation of the
LBLOCA scenario, using the proposed new methodology must still meet
the existing established acceptance criteria.
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 Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Section Chief: Richard J. Laufer.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of amendment request: April 4, 2005.
Description of amendment request: The proposed amendments would
revise the maximum and minimum allowable values for the degraded
voltage function of the 4160 volt essential service system (ESS) bus
under-voltage instrumentation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes revise the Technical Specifications (TS)
maximum and minimum allowable values for the degraded voltage
protection function and implement the use of automatic load tap
changers (LTCs) on transformers that provide power to safety-related
equipment. The only accident previously evaluated for which the
probability is potentially affected by these changes is the loss of
offsite power (LOOP). An allowable value for the degraded voltage
protection function that is too high could cause the emergency buses
to transfer to the emergency diesel generators (EDG) and thus
increase the probability of a LOOP. The allowable value for the
degraded voltage protection function has been revised in accordance
with an NRC-approved setpoint methodology and will continue to
ensure that the degraded voltage protection function actuates when
required, but does not actuate prematurely to cause a LOOP.
A failure of an LTC while in automatic operation mode that
results in decreased voltage to the ESS buses could also cause a
LOOP. This could occur in two ways. A failure of the LTC controller
that results in rapidly decreasing the voltage to the emergency
buses is the most severe failure mode. However, a backup controller
is provided with the LTC that makes this failure highly unlikely. A
failure of the LTC controller to respond to decreasing grid voltage
is less severe, since grid voltage changes occur slowly. In both of
the above potential failure modes, operators will take manual
control of the LTC to mitigate the effects of the failure. Thus, the
probability of a LOOP is not significantly increased.
The proposed changes will have no effect on the consequences of
a LOOP, since the EDGs provide power to safety related equipment
following a LOOP. The EDGs are not affected by the proposed changes.
The probability of other accidents previously evaluated is not
affected, since the proposed changes do not affect the way plant
equipment is operated and thus do not contribute to the initiation
of any of the previously evaluated accidents. The only way in which
the consequences of other previously evaluated accidents could be
affected is if a failure of the LTC while in automatic operation
mode caused a sustained high voltage which resulted in damage to
safety related equipment that is used to mitigate an accident.
Damage due to over-voltage is time-dependent. Since the LTC is
equipped with a backup controller, and since operator action is
available to prevent a sustained high voltage condition from
occurring, damage to safety related equipment is extremely unlikely,
and thus the consequences of these accidents are not significantly
increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of
[[Page 67748]]
accident from any accident previously evaluated?
Response: No.
The proposed changes involve functions that provide offsite
power to safety related equipment for accident mitigation. Thus, the
proposed changes potentially affect the consequences of previously
evaluated accidents (as addressed in Question 1), but do not result
in any new mechanisms that could initiate damage to the reactor and
its principal safety barriers (i.e., fuel cladding, reactor coolant
system, or primary containment).
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 changes do not affect the inputs or assumptions of
any of the analyses that demonstrate the integrity of the fuel
cladding, reactor coolant system, or containment during accident
conditions. The allowable values for the degraded voltage protection
function have been revised in accordance with an NRC-approved
setpoint methodology and will continue to ensure that the degraded
voltage protection function actuates when required, but does not
actuate prematurely to cause a LOOP. Automatic operation of the LTC
increases margin by reducing the potential for transferring to the
EDGs during an event.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: Gene Y. Suh.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: September 22, 2005.
Description of amendment request: The proposed amendment would
revise the Seabrook Station, Unit No. 1 operating license and Technical
Specifications to increase the licensed rated power level by 1.7
percent from 3587 megawatts thermal (MWt) to 3648 MWt. 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 will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Seabrook Station performed evaluations of the Nuclear Steam
Supply System (NSSS) and balance of plant systems, components, and
analyses that could be affected by the proposed change. A power
uncertainty calculation was performed, and the effect of increase
core thermal power by 1.7 percent to 3648 MWt on the Seabrook
Station design and licensing basis was evaluated. The result of the
evaluations determined that all systems and components continue to
be capable of performing their design function at the MUR
[measurement uncertainty recapture] core power level of 3648 MWt. An
evaluation of the accident analyses demonstrates that the applicable
analyses acceptance criteria continue to be met. No accident
initiators are affected by the MUR power uprate and no challenges to
any plant safety barriers are created by the proposed change.
The proposed change does not affect the release paths, the
frequency of release, or the analyzed source term for any accidents
previously evaluated in the Seabrook Station Updated Final Safety
Analysis Report (UFSAR). Systems, structures, and components
required to mitigate transients continue to be capable of performing
their design functions, and thus were found acceptable. The reduced
uncertainty in the feedwater flow input to the power calorimetric
measurement ensures that applicable accident analyses acceptance
criteria continue to be met, to support operation at the MUR core
power level of 3648 MWt. Analyses performed to assess the effects of
mass and energy remain valid. The source term used to assess
radiological consequences [has] been reviewed and determined to
bound operation at the MUR core power level.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change will not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
No new accident scenarios, failure mechanisms, or single
failures are introduced as a result of the proposed change. The
installation of the Caldon LEFM CheckPlusTM System has
been analyzed, and failures of the system will have no adverse
effect on any safety-related system or any systems, structures, and
components required for transient mitigation. Systems, structures,
and components previously required for the mitigation of a transient
continue to be capable of fulfilling their intended design
functions. The proposed change has no adverse affect on any safety-
related system or component and does not change the performance or
integrity of any safety-related system.
The proposed change does not adversely affect any current system
interfaces or create any new interfaces that could result in an
accident or malfunction of a different kind than previously
evaluated. Operating at a core power level of 3648 MWt does not
create any new accident initiators or precursors. The reduced
uncertainty in the feedwater flow input to the power calorimetric
measurement ensures that applicable accident analyses acceptance
criteria continue to be met, to support operation at the MUR core
power level of 3648 MWt. Credible malfunctions continue to be
bounded by the current accident analyses of record or evaluations
that demonstrate that applicable criteria continue to be met.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. The proposed change will not involve a significant reduction
in a margin [of] safety.
The margins of safety associated with the MUR are those
pertaining to core thermal power. These include those associated
with the fuel cladding, Reactor Coolant System pressure boundary,
and containment barriers. An engineering evaluation of the 1.7
percent increase in core thermal power from 3587 MWt to 3648 MWt was
performed. The current licensing bases analyzed core power is 3659
MWt. The analyzed core power level of 3659 MWt bounds the NSSS
thermal and hydraulic parameters at the MUR core power level of 3648
MWt. The NSSS systems and components were evaluated at the MUR core
power level and it was determined that the NSSS systems and
components continue to operate satisfactorily at the MUR power
level. The NSSS accident analyses were evaluated at the MUR core
power level of 3648 MWt. In all cases, the accident analyses at the
MUR core power level of 3648 MWt were bounded by the current
licensing bases analyzed core power level of 3659 MWt. As such, the
margins of safety continue to be bounded by the current analyses of
record for this change.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. S. Ross, Florida Power & Light Company,
P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Section Chief: Darrell J. Roberts.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: September 29, 2005.
Description of amendment request: The proposed amendment would
revise the Seabrook Station, Unit No. 1, Technical Specifications (TSs)
to permit a one-time, six-month extension to the currently approved 15-
year test interval for the containment integrated leak rate test.
Basis for proposed no significant hazards consideration
determination:
[[Page 67749]]
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 probability or consequences of accidents previously
evaluated in the UFSAR [updated final safety analysis report] are
unaffected by this proposed change. There is no change to any
equipment response or accident mitigation scenario, and this change
results in no additional challenges to fission product barrier
integrity. The proposed change does not alter the design,
configuration, operation, or function of any plant system,
structure, or component. As a result, the outcomes of previously
evaluated accidents are unaffected. The proposed extension to the
containment integrated leak rate test (ILRT) interval does not
involve a significant increase in consequences because, as discussed
in NUREG 1493, Performance Based Containment Leak Rate Test Program,
Type B and C tests identify the vast majority (greater than 95
percent) of all potential leakage paths. Further, ILRTs identify
only a few potential leakage paths that cannot be identified through
Type B and C testing, and leaks found by Type A testing have been
only marginally greater than existing requirements. In addition,
periodic inspections ensure that any significant containment
degradation will not go undetected. Therefore, the proposed change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed change [does] not create the possibility of a
new or different kind of accident from any [accident] previously
evaluated.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed change.
The proposed change does not challenge the performance or integrity
of any safety-related system. The proposed change neither installs
or removes any plant equipment, nor alters the design, physical
configuration, or mode of operation of any plant structure, system,
or component. No physical changes are being made to the plant, so no
new accident causal mechanisms are being introduced. The proposed
change only changes the frequency of performing the ILRT; however,
the test implementation and acceptance criteria are unchanged.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
3. The proposed change [does] not involve a significant
reduction in a margin of safety.
The margin of safety associated with the acceptance criteria of
any accident is unchanged. The proposed change will have no affect
on the availability, operability, or performance of the safety-
related systems and components. The proposed change does not alter
the design, configuration, operation, or function of any plant
system, structure, or component. The ability of any operable
structure, system, or component to perform its designated safety
function is unaffected by this change. NUREG 1493 concluded that
reducing the frequency of ILRTs to 20 years resulted in an
imperceptible increase in risk. Also, inspections of containment,
required by the ASME code [American Society of Mechanical Engineers
Boiler and Pressure Vessel Code] and the maintenance rule, ensure
that containment will not degrade in a manner that is only
detectable by Type A (ILRT) testing. Therefore, the margin of safety
as defined in the TS is not reduced and the proposed change does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. S. Ross, Florida Power & Light Company,
P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Section Chief: Darrell J. Roberts.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: September 29, 2005.
Description of amendment request: The proposed amendment would
revise the Seabrook Station, Unit No. 1 Technical Specifications to
permit a change in the steam generator tube inspection requirements to
include a sampling of the bulges and over-expansions for portions of
the steam generator tubes within the hot leg tubesheet region.
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 changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The previously analyzed accidents are initiated by the failure
of plant structures, systems, or components. The proposed changes
that alter the steam generator inspection criteria do not have a
detrimental impact on the integrity of any plant structure, system,
or component that initiates an analyzed event. The proposed changes
will not alter the operation of, or otherwise increase the failure
probability of any plant equipment that initiates an analyzed
accident.
Of the applicable accidents previously evaluated, the limiting
transients with consideration to the proposed changes to the steam
generator tube inspection criteria, are the steam generator tube
rupture (SGTR) event and the steam line break (SLB) accident.
During the SGTR event, the required structural integrity margins
of the steam generator tubes will be maintained by the presence of
the steam generator tubesheet area. Tube rupture in tubes with
cracks in the tubesheet is precluded by the constraint provided by
the tubesheet. This constraint results from the hydraulic expansion
process, thermal expansion mismatch between the tube and tubesheet
and from the differential pressure between the primary and secondary
side. Based on this design, the structural margins against burst, as
discussed in Regulatory Guide (RG) 1.121, ``Bases for Plugging
Degraded PWR [pressurized-water reactor] Steam Generator Tubes,''
are maintained for both normal and postulated accident conditions.
At normal operating pressures, leakage from primary water stress
corrosion cracking (PWSCC) below the proposed limited inspection
depth is limited by both the tube-to-tubesheet crevice and the
limited crack opening permitted by the tubesheet constraint.
Consequently, negligible normal operating leakage is expected from
cracks within the tubesheet region. The consequences of an SGTR
event are affected by the primary-to-secondary leakage flow during
the event. Primary-to-secondary leakage flow through a postulated
ruptured tube is not affected by the proposed changes since the
tubesheet enhances the tube integrity in the region of the hydraulic
expansion by precluding tube deformation beyond its initial
hydraulically-expanded outside diameter.
Furthermore, the proposed changes do not affect other systems,
structures, components or operational features. Therefore, the
proposed changes result in no significant increase in the
probability of the occurrence of a SGTR accident.
The probability of a[n] SLB accident is unaffected by the
potential failure of a steam generator tube as this failure is not
an initiator for a[n] SLB accident.
The consequences of a[n] SLB accident are also not significantly
affected by the proposed changes. During a[n] SLB accident, the
reduction in pressure above the tubesheet on the shell side of the
steam generator creates an axially uniformly distributed load on the
tubesheet due to the reactor coolant system pressure on the
underside of the tubesheet. The resulting bending action constrains
the tubes in the tubesheet thereby restricting primary-to-secondary
leakage below the midplane.
Primary-to-secondary leakage from tube degradation in the
tubesheet area during the limiting accident (i.e., a[n] SLB) is
limited by flow restrictions resulting from the crack and tube-to-
tubesheet contact pressures that provide a restricted leakage path
above the indications and also limit the degree of potential crack
face opening as compared to free span indications. The primary-to-
secondary leak rate during postulated SLB accident conditions would
be expected to be less than that during normal operation for
indications near the bottom of the tubesheet (i.e., including
indications in the tube end welds). This conclusion is based on the
[[Page 67750]]
observation that while the driving pressure causing leakage
increases by approximately a factor of (two) 2, the flow resistance
associated with an increase in tube-to-tubesheet contact pressure,
during a[n] SLB accident, increases by approximately a factor of
2.5. While such a leakage decrease is logically expected, the
postulated accident leak rate could be conservatively bounded by
twice the normal operating leak rate even if the increase in contact
pressure is ignored. Since normal operating leakage (spiking) is
limited to less that 0.104 gpm (150 gpd) for continued power
operation per station operating procedure OS 1227.02, ``Steam
Generator Tube Leak,'' the associated accident condition leak rate,
assuming all leakage to be from lower tube sheet indications, would
be bound by 0.208 gpm (twice normal operating leak rate). This value
is well within the assumed accident leakage rate of 0.347 gpm
discussed in the Seabrook Station Updated Safety Analysis Report,
Section 15.1.5 ``Steam System Piping Failure.'' Hence it is
reasonable to omit any consideration of inspection of the tube, tube
end weld, bulges / overexpansions or other anomalies below 17 inches
from the top of the hot leg tubesheet. Therefore, the consequences
of a[n] SLB accident remain unaffected.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any [accident] previously
evaluated.
The proposed changes do not introduce any new equipment, create
new failure modes for existing equipment, or create any new limiting
single failures. Plant operation will not be altered, and all safety
functions will continue to perform as previously assumed in accident
analyses. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed changes do not involve a significant reduction
in the margin of safety.
The proposed changes maintain the required structural margins of
the steam generator tubes for both normal and accident conditions.
Nuclear Energy Institute (NEI) 97-06, ``Steam Generator Program
Guidelines,'' and NRC Regulatory Guide (RG) 1.121, ``Bases for
Plugging Degraded PWR Steam Generator Tubes,'' are used as the bases
in the development of the limited hot leg tubesheet inspection depth
methodology for determining that steam generator tube integrity
considerations are maintained within acceptable limits. RG 1.121
describes a method acceptable to the NRC for meeting General Design
Criteria (GDC) 14, ``Reactor Coolant Pressure Boundary,'' GDC 15,
``Reactor Coolant System Design,'' GDC 31, ``Fracture Prevention of
Reactor Coolant Pressure Boundary,'' and GDC 32, ``Inspection of
Reactor Coolant Pressure Boundary,'' by reducing the probability and
consequences of a SGTR. RG 1.121 concludes that by determining the
limiting safe conditions for tube wall degradation the probability
and consequences of a SGTR are reduced. RG 1.121 uses safety factors
on loads for tube burst that are consistent with the requirements of
Section III of the American Society of Mechanical Engineers (ASME)
Code.
For axially oriented cracking located within the tubesheet, tube
burst is precluded due to the presence of the tubesheet. For
circumferentially oriented cracking, Westinghouse evaluation LTR-
CDME-05-170, ``Limited Inspection of the Steam Generator Tube
Portion Within the Tubesheet at Seabrook Generating Station,''
defines a length of degradation-free expanded tubing that provides
the necessary resistance to tube pullout due to the pressure induced
forces, with applicable safety factors applied. Application of the
limited hot leg tubesheet inspection criteria will preclude
unacceptable primary-to-secondary leakage during all plant
conditions. The methodology for determining leakage provides for
large margins between calculated and actual leakage values in the
proposed limited hot leg tubesheet inspection depth criteria.
Therefore, the proposed changes do not involve a significant
reduction in any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. S. Ross, Florida Power & Light Company,
P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Section Chief: Darrell J. Roberts.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendment request: September 27, 2005.
Description of amendment request: The amendments proposed by
Southern Nuclear Operating Company would revise the Technical
Specifications (TS) to eliminate the Power Range Neutron Flux-High
Negative Rate Reactor Trip function, based on the approved methodology
contained in Westinghouse Topical Report WCAP-11394-P-A, ``Methodology
for the Analysis of the Dropped Rod Event.'' The changes will allow the
elimination of a trip circuitry that is not credited in the Farley
Nuclear Plant safety analysis, and which can result in an unnecessary
reactor trip. These changes will be implemented sequentially,
concurrent with each unit's refueling outage during which the design
change is implemented. Additionally, this amendment request deletes TS
Bases text associated with an unconservative local Departure from
Nucleate Boiling Ratio.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed changes do not significantly increase the
probability or consequences of an accident previously evaluated in
the Updated Final Safety Analysis Report (UFSAR). All of the safety
analyses have been evaluated for impact due to this change. The
elimination of the Power Range Neutron Flux-High Negative Rate
Reactor Trip function and the elimination of text in the TS
[Technical Specifications] Bases for LC0 3.3.1, page B 3.3.1-1 1,
associated with an unconservative local DNBR [departure from
nucleate boiling ratio], does not affect the dropped RCCA [Rod
Cluster Control Assembly] analyses nor any other analyses, since it
is not credited in any of the safety analyses; therefore, the
probability of an accident has not been increased. All dose
consequences have been evaluated with respect to the proposed
changes, there is no impact due to the proposed change, and all
acceptance criteria continue to be met. Therefore, these changes do
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
No. The proposed changes do not create the possibility of a new
or different kind of accident from any accident already evaluated in
the UFSAR. No new accident scenarios, failure mechanisms or limiting
single failures are introduced as result of the proposed changes.
The changes have no adverse effects on any safety-related system.
Therefore, all accident analyses criteria continue to be met and
these changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
No. The proposed changes do not involve a significant reduction
in a margin of safety. The dropped RCCA(s) event does not credit the
Power Range Neutron Flux-High Negative Rate Reactor Trip function.
The conclusion presented in the UFSAR Section 15.2.3.3 that the DNBR
design basis is met for a dropped RCCA(s) event remains valid for
the proposed changes, which are based on the NRC approved
methodology contained in CAP-11394-PA. Additionally, WCAP-11394-P-A
indicates that the analysis for a dropped rod event envelops a
multiple rod drop accident at high power levels, and that such an
accident will not result in an unconservative local DNBR. All
applicable acceptance criteria continue to be met. Therefore, the
proposed changes do not involve a significant reduction in a margin
of safety.
[[Page 67751]]
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201.
NRC Section Chief: Evangelos C. Marinos.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendment request: October 6, 2005.
Description of amendment request: The amendments proposed by
Southern Nuclear Operating Company (SNC) would revise the Technical
Specifications (TS) to support a revision to the Best Estimate Loss of
Coolant Accident (BELOCA) for Farley Nuclear Plant (FNP). The NRC
recently approved a new Westinghouse BELOCA methodology, Automated
Statistical Treatment of Uncertainty Method (ASTRUM). ASTRUM was
submitted in WCAP-16009-P. The NRC issued a Safety Evaluation Report in
a letter dated November 5, 2004. Westinghouse issued WCAP-16009-P-A in
January 2005. SNC has completed the analysis for FNP and the enclosed
proposed amendment is to incorporate a reference to WCAP-16009-P-A in
TS section 5.6.5 Core Operating Limits Report (COLR).
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 physical plant changes are being made as a result of using
the Westinghouse Best Estimate Large Break LOCA [Loss of Coolant
Accident] (BELOCA) analysis methodology. The proposed TS changes
simply involve updating the references in TS 5.6.5.b, Core Operating
Limits Report (COLR), to reference the Westinghouse BELOCA analysis
methodology. The plant conditions assumed in the analysis are
bounded by the design conditions for all equipment in the plant;
therefore, there will be no increase in the probability of a LOCA.
The consequences of a LOCA are not being increased, since the
analysis has shown that the Emergency Core Cooling System (ECCS) is
designed such that its calculated cooling performance conforms to
the criteria contained in 10 CFR 50.46, ``Acceptance criteria for
emergency core cooling systems for light-water nuclear power
reactors.'' No other accident consequence is potentially affected by
this change.
All systems will continue to be operated in accordance with
current design requirements under the new analysis, therefore no new
components or system interactions have been identified that could
lead to an increase in the probability of any accident previously
evaluated in the Updated Final Safety Analysis Report (UFSAR). No
changes were required to the Reactor Protection System (RPS) or
Engineering Safety Features (ESF) setpoints because of the new
analysis methodology.
Therefore, it is concluded that this change does not
significantly increase the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
There are no physical changes being made to the plant as a
result of using the Westinghouse Best Estimate Large Break LOCA
analysis methodology. No new modes of plant operation are being
introduced. The configuration, operation and accident response of
the structures or components are unchanged by utilization of the new
analysis methodology. Analyses of transient events have confirmed
that no transient event results in a new sequence of events that
could lead to a new accident scenario. The parameters assumed in the
analysis are within the design limits of existing plant equipment.
In addition, employing the Westinghouse Best Estimate Large
Break LOCA analysis methodology does not create any new failure
modes that could lead to a different kind of accident. The design of
all systems remains unchanged and no new equipment or systems have
been installed which could potentially introduce new failure modes
or accident sequences. No changes have been made to any RPS or ESF
actuation setpoints.
Based on this review, it is concluded that no new accident
scenarios, failure mechanisms or limiting single failures are
introduced as a result of the proposed changes.
Therefore, the proposed TS changes do not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
It has been shown that the analytic technique used in the
Westinghouse Best Estimate Large Break LOCA analysis methodology
realistically describes the expected behavior of the reactor system
during a postulated LOCA. Uncertainties have been accounted for as
required by 10 CFR 50.46. A sufficient number of LOCAs with
different break sizes, different locations, and other variations in
properties have been considered to provide assurance that the most
severe postulated LOCAs have been evaluated. The analysis has
demonstrated that all acceptance criteria contained in 10 CFR 50.46
paragraph b continue to be satisfied.
Therefore, it is concluded that this change does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201.
NRC Section Chief: Evangelos C. Marino.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of amendment request: January 27, 2005.
Description of amendment request: The proposed amendments would
revise Technical Specifications Limiting Conditions for Operations
3.3.1, 3.3.2, 3.3.6, and 3.3.8, by extending the Surveillance Test
Intervals for the Reactor Protection System.
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?
The proposed changes to the Completion Time, bypass test time,
and Surveillance Frequencies reduce the potential for inadvertent
reactor trips and spurious actuations and, therefore, do not
increase the probability of any accident previously evaluated. The
proposed changes to the allowed Completion Time, bypass test time,
and Surveillance Frequencies do not change the response of the plant
to any accidents and have an insignificant impact on the reliability
of the reactor trip system and engineered safety feature actuation
system (RTS and ESFAS) signals. The RTS and ESFAS will remain highly
reliable, and the proposed changes will not result in a significant
increase in the risk of plant operation. This is demonstrated by
showing that the impact on plant safety as measured by core damage
frequency (CDF) is less than 1.01E-06 per year and the impact on
large early release frequency (LERF) is less than 1.0E-07 per year.
In addition, for the Completion Time change, the incremental
conditional core damage probabilities (ICCDP) and incremental
conditional large early release probabilities (ICLERP) are less than
5.0E-08. These changes meet the
[[Page 67752]]
acceptance criteria in Regulatory Guides 1.174 and 1.177. Therefore,
since the RTS and ESFAS will continue to perform their functions
with high reliability as originally assumed, and the increase in
risk as measured by CDF, LERF, ICCDP, and ICLERP is within the
acceptance criteria of existing regulatory guidance, there will not
be a significant increase in the consequences of any accidents. The
proposed changes do not adversely affect accident initiators or
precursors nor alter the design assumptions, conditions, or
configuration of the facility or the manner in which the plant is
operated and maintained. The proposed changes do not alter or
prevent the ability of structures, systems, and components (SSCs)
from performing their intended function to mitigate the consequences
of an initiating event within the assumed acceptance limits. The
proposed changes do not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of an accident previously evaluated.
Further, the proposed changes do not increase the types or amounts
of radioactive effluent that may be released offsite, nor
significantly increase individual or cumulative occupational/public
radiation exposures. The proposed changes are consistent with the
safety analysis assumptions and resultant consequences. Therefore,
it is concluded that this change does not increase the probability
of occurrence of a malfunction of equipment important to safety.
2. Does the Proposed Change Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated?
The proposed changes do not result in a change in the manner in
which the RTS and ESFAS provide plant protection. The RTS and ESFAS
will continue to have the same setpoints after the proposed changes
are implemented. There are no design changes associated with the
license amendment. The changes to Completion Time, bypass test time,
and Surveillance Frequency do not change any existing accident
scenarios, nor create any new or different accident scenarios. The
changes do not involve a physical alteration of the plant (i.e., no
new or different type of equipment will be installed) or a change in
the methods governing normal plant operation. In addition, the
changes do not impose any new or different requirements or eliminate
any existing requirements. The changes do not alter assumptions made
in the safety analysis. The proposed changes are consistent with the
safety analysis assumptions and current plant operating practice.
Therefore, the possibility of a new or different malfunction of
safety related equipment is not created.
3. Does the Proposed Change Involve a Significant Reduction in
the Margin of Safety?
The proposed changes do not alter the manner in which safety
limits, limiting safety system settings, or limiting conditions for
operation are determined. The safety analysis acceptance criteria
are not impacted by these changes. Redundant RTS and ESFAS trains
are maintained, and diversity with regard to the signals that
provide reactor trip and engineered safety features actuation is
also maintained. All signals credited as primary or secondary and
all operator actions credited in the accident analyses will remain
the same. The proposed changes will not result in plant operation in
a configuration outside the design basis. The calculated impact on
risk is insignificant and meets the acceptance criteria contained in
Regulatory Guides 1.174 and 1.177. Although there was no attempt to
quantify any positive human factors benefit due to increased
Completion Time, bypass test time, and Surveillance Frequencies, it
is expected there would be a net benefit due to a reduced potential
for spurious reactor trips and actuations associated with testing.
Therefore, it is concluded that this change does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Section Chief: Evangelos C. Marinos.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: September 30, 2005 (TS-05-02).
Description of amendment request: The proposed amendment would
revise the SQN Technical Specification (TS) Section 5.0, ``Design
Features,'' to more conform with NUREG-1431 Revision 3, ``Standard
Technical Specifications for Westinghouse Plants.'' The proposed change
included the elimination of exclusion area, low population zone, and
effluent subsections and associated figures referred to in Section 5.1,
``Site''; elimination of Section 5.2, ``Containment''; elimination of
Section 5.4, ``Reactor Coolant System,'' as well as Section 5.5,
``Meteorological Tower Location,'' and its figure. Lastly, a proposed
change to the TS ``Administrative Control'' section to acquire the
component cyclic or transient limits currently located in the ``Design
Features'' section.
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 removal of information and figures featuring the locations
of the site exclusion area, gaseous and liquid effluent boundaries,
low population zone, and the meteorological tower is administrative
in nature. Most, if not, all of this information is located in other
licensee control documents, such as the Final Safety Analysis Report
(FSAR). Congruently, the addition of a site location description
only adds geographical information to the TSs. The relocation and
revision of the component cyclic or transient limits requirement
does not alter the requirement to track and maintain these limits
and thus considered administrative. This proposed amendment involves
no technical changes to the existing TSs and does not impact
initiators of analyzed events. The changes also do not impact the
assumed mitigation of accidents or transient events. 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 change to plant systems,
components, or operating practices that could result in a change in
accident generation potential. The proposed changes do not impose
any new or different requirements or eliminate any existing
requirements. The proposed changes do not alter assumptions made in
the safety analyses and licensing basis. Therefore, the proposed
change does not create the possibility of a new or different kind of
accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The deletion of information and figures featuring the locations
of the site exclusion area, gaseous and liquid effluent boundaries,
low population zone, and the meteorological tower does not affect
operational limits or functional capabilities of plant systems,
structures and components. The addition of a site location
description adds geographical information to the TSs. The relocation
and revision of the component cyclic or transient limits
requirements also does not affect operational limits or functional
capabilities of plant systems, structures and components. These
changes pose no effect on margin of safety. The TS identified
maximum steel containment temperature value is not the current
limiting design value, which is found in the FSAR. Its elimination
is considered administrative in nature and does not result in a
change of margin of safety to the containment design. 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
[[Page 67753]]
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.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit 2, Somervell County, Texas
Date of amendment request: April 27, 2005, as supplemented by
letter dated July 20, 2005.
Brief description of amendments: The amendment revises Technical
Specification (TS) 5.6.5, ``Core Operating Limits Report,'' by adding
topical report WCAP-13060-P-A, ``Westinghouse Fuel Assembly
Reconstitution Evaluation Methodology,'' to the list of approved
methodologies to be used at Comanche Peak Steam Electric Station
(CPSES), Unit 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change is administrative in nature and as such does
not impact the condition or performance of any plant structure,
system or component. The core operating limits are established to
support Technical Specifications 3.1, 3.2, 3.3, 3.4, and 3.9. The
core operating limits ensure that fuel design limits are not
exceeded during any conditions of normal operation or in the event
of any Anticipated Operational Occurrence (AOO). The methods used to
determine the core operating limits for each operating cycle are
based on methods previously found acceptable by the NRC and listed
in TS section 5.6.5.b. Application of these approved methods will
continue to ensure that acceptable operating limits are established
to protect the fuel cladding integrity during normal operation and
AOOs. The requested Technical Specification change does not involve
any plant modifications or operational changes that could affect
system reliability, performance, or possibility of operator error.
The requested change does not affect any postulated accident
precursors, does not affect any accident mitigation systems, and
does not introduce any new accident initiation mechanisms.
As a result, the proposed change to the CPSES Technical
Specifications does not involve any increase in the probability or
the consequences of any accident or malfunction of equipment
important to safety previously evaluated since neither accident
probabilities nor consequences are being affected by this proposed
administrative change.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change is administrative in nature, and therefore
does not involve any change in station operation or physical
modifications to the plant. In addition, no changes are being made
in the methods used to respond to plant transients that have been
previously analyzed. No changes are being made to plant parameters
within which the plant is normally operated or in the setpoints,
which initiate protective or mitigative actions, and no new failure
modes are being introduced.
Therefore, the proposed administrative change to the CPSES
Technical Specifications does not create the possibility of a new or
different kind of accident or malfunction of equipment important to
safety from any accident previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed change is administrative in nature and does not
impact station operation or any plant structure, system or component
that is relied upon for accident mitigation. Furthermore, the margin
of safety assumed in the plant safety analysis is not affected in
any way by the proposed administrative change.
Therefore, the proposed change to the CPSES Technical
Specifications does not involve any reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: David Terao.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: January 24, 2005.
Brief description of amendments: The amendments will revise the
surveillance requirements (SRs) for Technical Specification 3.7.5,
``Auxilary Feed Water (AFW) System.'' Specifically, a note will be
added to SRs 3.7.5.1, 3.7.5.3, and 3.7.5.4 that states, ``AFW train(s)
may be considered OPERABLE during alignment and operation for steam
generator level control, if it is capable of being manually realigned
to the AFW mode of operation.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change has no impact on the probability of any
accident previously evaluated. The consequences of the limiting
transients and accidents (full power operation) are unaffected by
the proposed change. At low power sufficient time is available to
establish auxiliary feedwater injection if needed.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of these changes. There will be no adverse effect or challenges
imposed on any safety-related system as a result of these changes.
There are no changes in the method by which any safety-related plant
system performs its safety function. Overall protection system
performance will remain within the bounds of the previously
performed accident analyses and the protection systems will continue
to function in a manner consistent with the plant design basis. The
proposed changes do not affect the probability of any event
initiators. The proposed changes do not alter any assumptions or
change any mitigation actions in the radiological consequence
evaluations in the Final Safety Analysis Report (FSAR).
Therefore, the proposed change[s] do not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not affect the acceptance criteria for
any analyzed event nor is there a change to any Safety Analysis
Limit (SAL). There will be no effect on the manner in which safety
limits, limiting safety system settings, or limiting conditions for
operation are determined nor will there be any effect on those plant
systems necessary to assure the accomplishment of protection
functions. There will be no impact on the overpower limit, the
Departure from Nucleate Boiling Ratio (DNBR) limits, the Heat Flux
Hot Channel Factor (FQ), the Nuclear Enthalpy Rise Hot Channel
Factor (F'H), the Loss of Coolant Accident Peak Centerline
Temperature (LOCA PCT), peak local power density, or any other
margin of safety. The
[[Page 67754]]
radiological dose consequence acceptance criteria listed in the
Standard Review Plan will continue to be met. Since the limiting
transients and accidents are unaffected, the proposed change[s] do
not involve a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: David Terao.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: August 10, 2005.
Brief description of amendments: The amendments would revise the
Technical Specification (TS) 5.5.13, ``Diesel Fuel Oil Testing
Program,'' to relocate the specific American Society for Testing and
Materials (ASTM) Standard reference from the Administrative Controls
Section of TS to a licensee-controlled document.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes relocate the specific American Society for
Testing and Materials (ASTM) Standard references from the
Administrative Controls of TS to a licensee-controlled document.
Since any change to the licensee-controlled document will be
evaluated pursuant to the requirements of 10 CFR 50.59, ``Changes,
tests and experiments,'' no increase in the probability or
consequences of an accident previously evaluated is involved.
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, or
configuration of the facility or the manner in which the plant is
operated and maintained. The proposed changes do not alter or
prevent the ability of structures, systems, and components (SSCs)
from performing their intended function to mitigate the consequences
of an initiating event within the assumed acceptance limits. The
proposed changes do not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of an accident previously evaluated.
Further, the proposed changes do not increase individual or
cumulative occupational or public radiation exposure.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or change in the methods governing normal plant
operation. In addition, the changes do not alter the assumptions
made in the analysis and licensing basis.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The level of safety of facility operation is unaffected by the
proposed changes since there is no change in the intent of the TS
requirements of assuring fuel oil is of the appropriate quality for
emergency DG [diesel generator] use. The proposed changes provide
the flexibility needed to utilize state-of-the-art technology in
fuel oil sampling and analysis methods.
Therefore the proposed changes do not involve a reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: David Terao.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: August 22, 2005.
Brief description of amendments: The amendments revise Technical
Specification (TS) 3.7.10, ``Control Room Emergency Filtration/
Pressurization System (CREFS) and Control Room Envelope (CRE),'' and
adds new TS 5.5.20, ``Control Room Integrity Program,'' and TS 5.6.11,
``Control Room Report.'' In addition the amendments update the Final
Safety Analysis Report to include new methods and assumptions as
described in Regulatory Guide 1.195 for evaluation of radiological
consequences.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change addresses the Control Room Envelope (CRE),
including updated surveillances for the Control Room Emergency
Filtration/Pressurization System (CREFS) trains and the CRE, a new
TS 5.5.20, ``Control Room Integrity Program,'' and a new TS 5.6.11,
``Control Room Report.'' These changes are consistent with the
guidance in Regulatory Guides 1.196 and 1.197. New methods and
assumptions for evaluating radiological consequences for design
basis accidents are adopted consistent with NRC Regulatory Guide
1.195. The acceptance limits for the Control Room Integrity Program
are based on these revised radiological dose consequences
calculations. The proposed changes do not adversely affect accident
initiators or precursors nor alter the configuration of the
facility. The proposed changes do not alter or prevent the ability
of structures, systems, and components (SSCs) from performing their
intended function to mitigate the consequences of an initiating
event to within the Regulatory Guide 1.195 acceptance limits. This
activity is a revision to the Technical Specifications and the
supporting radiological dose consequences analyses for the control
room ventilation system which is a mitigating system designed to
minimize in-leakage into the control room and to filter the control
room atmosphere to protect the control room operators following
accidents previously analyzed. An important part of the system is
the control room envelope (CRE). The CRE integrity is not an
initiator or precursor to any accident previously evaluated.
Therefore the probability of occurrence of any accident previously
evaluated is not increased. Performing tests and implementing
programs that verify the integrity of the CRE and control room
habitability ensure mitigation features are capable of performing
the assumed function.
The revised radiological consequences analyses, performed using
the assumptions and methodologies presented in Regulatory Guidance
1.195, do not result in significant increases in the radiological
dose consequences to the general public nor to the control room
operators. All calculated dose consequences are within acceptance
limits of Regulatory Guide 1.195.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes will not alter the requirements of the
control room ventilation
[[Page 67755]]
system or its function during accident conditions. No new or
different accidents result from performing the new revised actions
and surveillances or programs required. The changes do not involve a
physical alteration of the plant (i.e., no new or different type of
equipment will be installed) or a change in the methods governing
normal plant operation which could create the possibility of a new
or different kind of accident. The proposed changes are consistent
with the safety analysis assumptions and current plant operating
practices. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The safety analysis acceptance criteria
are not affected by these changes. The proposed changes will not
result in plant operation in a configuration outside the design
basis for an unacceptable period of time without mitigating actions.
The proposed changes do not affect systems that are required to
respond to safely shut down the plant and to maintain the plant in a
safe shutdown condition.
Therefore the proposed change does not involve a reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: David Terao.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: May 27, 2005.
Brief description of amendment: The amendment revised the technical
specification (TS) 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: October 25, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 167.
Facility Operating License No. NPF-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 19, 2005 (70 FR
41443).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 25, 2005.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: May 31, 2005.
Brief description of amendment: The amendment modifies Technical
Specification (TS) requirements to adopt the provisions of Industry/TS
Task Force (TSTF) change TSTF-359, ``Increased Flexibility in Mode
Restraints.''
Date of issuance: October 20, 2005.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 284.
Facility Operating License No. DPR-59: The amendment revised the
TSs.
Date of initial notice in Federal Register: August 16, 2005 (70 FR
48204).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 20, 2005.
No significant hazards consideration comments received: No.
Exelon Generating Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Date of application for amendment: December 17, 2004, as
supplemented by letter dated September 28, 2005.
Brief description of amendment: The amendments revised Appendix B,
Environmental Protection Plan (non-radiological), of the Byron Station
Facility Operating Licenses.
Date of issuance: October 18, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 145.
Facility Operating License Nos. NPF-37 and NPF-66: The amendments
revised the Environmental Protection Plan.
Date of initial notice in Federal Register: April 12, 2005 (70 FR
19115). The supplement dated September 28, 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 amendments is contained
in a
[[Page 67756]]
Safety Evaluation dated October 18, 2005.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: October 25, 2004, as supplement by
letter dated August 1, 2005.
Brief description of amendment: The amendment revises the required
channels per trip system for several instrument functions contained in
Technical Specification Tables 3.3.6.1-1, ``Primary Containment
Isolation Instrumentation,'' 3.3.6.2-1, ``Secondary Containment
Isolation Instrumentation,'' and 3.3.7.1-1 ``Control Room Emergency
Filter System Instrumentation.''
Date of issuance: October 27, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 212.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 4, 2005 (70 FR
402).
The supplement dated August 1, 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 27, 2005.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-272, Salem Nuclear Generating Station
Unit No. 1, Salem County, New Jersey
Date of application for amendment: February 23, 2005, as
supplemented by letters dated August 2, 2005, and September 21, 2005.
Brief description of amendment: The amendments revised Technical
Specifications (TSs) to implement a new steam generator tube
surveillance program that is consistent with the program proposed by
the TS Task Force (TSTF) in TSTF-449.
Date of issuance: October 14, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 268.
Facility Operating License No. DPR-70: The amendments revised the
TSs.
Date of initial notice in Federal Register: May 10, 2005 (70 FR
24655). Supplements dated August 2, 2005, and September 21, 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 amendments is contained
in a Safety Evaluation dated October 14, 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: March 4, 2005, as supplemented
August 2, 2005.
Brief description of amendments: These amendments extend the
completion time from 1 hour to 24 hours for Actions ``a'' and ``b'' of
Salem Nuclear Generating Station, Unit Nos. 1 and 2 Technical
Specification (TS) 3.5.1, ``Accumulators,'' which requires restoration
of an accumulator when it has been declared inoperable for reasons
other than boron concentration in the accumulator not being within the
required range.
Date of issuance: October 14, 2005.
Effective date: As of the date of issuance and to be implemented
within 60 days.
Amendment Nos.: 267 and 249.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the TSs.
Date of initial notice in Federal Register: May 24, 2005 (70 FR
29800). The August 2, 2005, supplement provided clarifying information
only and did not change the scope of the proposed amendment, 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 October 14, 2005.
No significant hazards consideration comments received: No.
Sacramento Municipal Utility District, Docket No. 50-312, Rancho Seco
Nuclear Generating Station, Sacramento County, California
Date of application for amendment: January 24, 2005.
Brief description of amendment: The amendment removes unnecessary
and obsolete information from the facility operating license.
Date of issuance: September 21, 2005.
Effective date: September 21, 2005.
Amendment No.: 132.
Facility Operating License No. DPR-54: The amendment revised the
License.
Date of initial notice in Federal Register: March 29, 2005 (70 FR
15947).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 22, 2005.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: August 12, 2004.
Brief description of amendments: The amendments revised
Surveillance Requirement (SR) 4.7.8.d.3 of the Auxiliary Building Gas
Treatment System (ABGTS) by deleting vacuum relief flow requirements.
The change removes criteria from the SR that is not necessary to verify
the operability of the ABGTS and eliminates confusion regarding the
basis for the vacuum relief flow requirement.
Date of issuance: August 18, 2005.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 303 and 293.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revised the technical specifications.
Date of initial notice in Federal Register: October 12, 2004 (69 FR
60687).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 18, 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: October 27, 2004, as
supplemented by letter dated June 17, 2005.
Brief description of amendment: The amendment (1) deleted
Conditions 2.C.(3), 2.C.(4), 2.C.(6) through 2.C.(14), Section 2.F, and
Attachments 1 and 2, and (2) revised Conditions 2.C.(1) and 2.C.(5), to
the facility operating license, to reflect completed requirements. In
addition, the list of attachments and appendices to the operating
license was revised to reflect the deletion of Attachments 1 and 2. The
proposed
[[Page 67757]]
changes to Technical Specifications Table 5.5.9-2, ``Steam Generator
Tube Inspection,'' and Table 5.5.9-3, ``Steam Generator Repaired Tube
Inspection,'' were also submitted in the licensee's application dated
September 17, 2004 (ULNRC-05056), for the replacement steam generator
project and were approved in Amendment No. 168, which was issued in the
NRC letter dated September 29, 2005.
Date of issuance: October 25, 2005.
Effective date: October 25, 2005, and shall be implemented within
90 days of the date of issuance.
Amendment No.: 169.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 7, 2004 (69 FR
70723). The June 17, 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 25, 2005.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 31st 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. 05-22002 Filed 11-7-05; 8:45 am]
BILLING CODE 7590-01-P